Federal Register, Volume 70 Issue 39 (Tuesday, March 1, 2005)

[Federal Register Volume 70, Number 39 (Tuesday, March 1, 2005)]
[Notices]
[Pages 9986-10005]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: 05-3627]


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NUCLEAR REGULATORY COMMISSION


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

I. Background

    Pursuant to section 189a. (2) of the Atomic Energy Act of 1954, as 
amended (the Act), the U.S. Nuclear Regulatory Commission (the 
Commission or NRC staff) is publishing this regular biweekly notice. 
The Act requires the Commission publish notice of any amendments 
issued, or proposed to be issued and grants the Commission the 
authority to issue and make immediately effective any amendment to an 
operating license upon a determination by the Commission that such 
amendment involves no significant hazards consideration, 
notwithstanding the pendency before the Commission of a request for a 
hearing from any person.
    This biweekly notice includes all notices of amendments issued, or 
proposed to be issued from February 4, 2005, through February 17, 2005. 
The last biweekly notice was published on February 15, 2005 (70 FR 
7762).

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 
01F21, 11555 Rockville Pike (first floor), Rockville, Maryland. The 
filing of requests for a hearing and petitions for leave to intervene 
is discussed below.
    Within 60 days after the date of publication of this notice, the 
licensee may file a request for a hearing with respect to issuance of 
the amendment to the subject facility operating license and any person 
whose interest may be affected by this proceeding and who wishes to 
participate as a party in the proceeding must file a written request

[[Page 9987]]

for a hearing and a petition for leave to intervene. Requests for a 
hearing and a petition for leave to intervene shall be filed in 
accordance with the Commission's ``Rules of Practice for Domestic 
Licensing Proceedings'' in 10 CFR part 2. Interested persons should 
consult a current copy of 10 CFR 2.309, which is available at the 
Commission's PDR, located at One White Flint North, Public File Area 
01F21, 11555 Rockville Pike (first floor), Rockville, Maryland. 
Publicly available records will be accessible from the Agencywide 
Documents Access and Management System's (ADAMS) Public Electronic 
Reading Room on the Internet at the NRC Web site, http://www.nrc.gov/reading-rm/doc-collections/cfr/. If a request for a hearing or petition 
for leave to intervene is filed within 60 days, the Commission or a 
presiding officer designated by the Commission or by the Chief 
Administrative Judge of the Atomic Safety and Licensing Board Panel, 
will rule on the request and/or petition; and the Secretary or the 
Chief Administrative Judge of the Atomic Safety and Licensing Board 
will issue a notice of a hearing or an appropriate order.
    As required by 10 CFR 2.309, a petition for leave to intervene 
shall set forth with particularity the interest of the petitioner in 
the proceeding, and how that interest may be affected by the results of 
the proceeding. The petition should specifically explain the reasons 
why intervention should be permitted with particular reference to the 
following general requirements: (1) The name, address, and telephone 
number of the requestor or petitioner; (2) the nature of the 
requestor's/petitioner's right under the Act to be made a party to the 
proceeding; (3) the nature and extent of the requestor's/petitioner's 
property, financial, or other interest in the proceeding; and (4) the 
possible effect of any decision or order which may be entered in the 
proceeding on the requestor's/petitioner's interest. The petition must 
also set forth the specific contentions which the petitioner/requestor 
seeks to have litigated at the proceeding.
    Each contention must consist of a specific statement of the issue 
of law or fact to be raised or controverted. In addition, the 
petitioner/requestor shall provide a brief explanation of the bases for 
the contention and a concise statement of the alleged facts or expert 
opinion which support the contention and on which the petitioner/
requestor intends to rely in proving the contention at the hearing. The 
petitioner/requestor must also provide references to those specific 
sources and documents of which the petitioner is aware and on which the 
petitioner/requestor intends to rely to establish those facts or expert 
opinion. The petition must include sufficient information to show that 
a genuine dispute exists with the applicant on a material issue of law 
or fact. Contentions shall be limited to matters within the scope of 
the amendment under consideration. The contention must be one which, if 
proven, would entitle the petitioner/requestor to relief. A petitioner/
requestor who fails to satisfy these requirements with respect to at 
least one contention will not be permitted to participate as a party.
    Those permitted to intervene become parties to the proceeding, 
subject to any limitations in the order granting leave to intervene, 
and have the opportunity to participate fully in the conduct of the 
hearing.
    If a hearing is requested, and the Commission has not made a final 
determination on the issue of no significant hazards consideration, the 
Commission will make a final determination on the issue of no 
significant hazards consideration. The final determination will serve 
to decide when the hearing is held. If the final determination is that 
the amendment request involves no significant hazards consideration, 
the Commission may issue the amendment and make it immediately 
effective, notwithstanding the request for a hearing. Any hearing held 
would take place after issuance of the amendment. If the final 
determination is that the amendment request involves a significant 
hazards consideration, any hearing held would take place before the 
issuance of any amendment.
    A request for a hearing or a petition for leave to intervene must 
be filed by: (1) First class mail addressed to the Office of the 
Secretary of the Commission, U.S. Nuclear Regulatory Commission, 
Washington, DC 20555-0001, Attention: Rulemaking and Adjudications 
Staff; (2) courier, express mail, and expedited delivery services: 
Office of the Secretary, Sixteenth Floor, One White Flint North, 11555 
Rockville Pike, Rockville, Maryland, 20852, Attention: Rulemaking and 
Adjudications Staff; (3) e-mail addressed to the Office of the 
Secretary, U.S. Nuclear Regulatory Commission, [email protected]; 
or (4) facsimile transmission addressed to the Office of the Secretary, 
U.S. Nuclear Regulatory Commission, Washington, DC, Attention: 
Rulemakings and Adjudications Staff at (301) 415-1101, verification 
number is (301) 415-1966. A copy of the request for hearing and 
petition for leave to intervene should also be sent to the Office of 
the General Counsel, U.S. Nuclear Regulatory Commission, Washington, DC 
20555-0001, and it is requested that copies be transmitted either by 
means of facsimile transmission to (301) 415-3725 or by e-mail to 
[email protected]. A copy of the request for hearing and petition 
for leave to intervene should also be sent to the attorney for the 
licensee.
    Nontimely requests and/or petitions and contentions will not be 
entertained absent a determination by the Commission or the presiding 
officer of the Atomic Safety and Licensing Board that the petition, 
request and/or the contentions should be granted based on a balancing 
of the factors specified in 10 CFR 2.309(a)(1)(I)-(viii).
    For further details with respect to this action, see the 
application for amendment which is available for public inspection at 
the Commission's PDR, located at One White Flint North, Public File 
Area 01F21, 11555 Rockville Pike (first floor), Rockville, Maryland. 
Publicly available records will be accessible from the ADAMS Public 
Electronic Reading Room on the Internet at the NRC Web site, http://www.nrc.gov/reading-rm/adams.html. If you do not have access to ADAMS 
or if there are problems in accessing the documents located in ADAMS, 
contact the PDR Reference staff at 1 (800) 397-4209, (301) 415-4737 or 
by e-mail to [email protected].

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

    Date of amendment request: June 24, 2004.
    Description of amendment request: The proposed amendment would 
revise Surveillance Requirement (SR) 4.0.2 to extend the delay period, 
before entering a Limiting Condition for Operation, following a missed 
surveillance. The delay period would be extended from the current limit 
of `` * * * up to 24 hours or up to the limit of the specified 
Frequency, whichever is less'' to `` * * * up to 24 hours or up to the 
limit of the specified Frequency, whichever is greater.'' In addition, 
the following requirement would be added to SR 4.0.2: ``A risk 
evaluation shall be performed for any Surveillance delayed greater than 
24 hours and the risk impact shall be managed.'' In addition, a 
Technical Specifications (TSs) Bases Control Program would be adopted 
as new TS 6.18.
    Basis for proposed no significant hazards consideration 
determination: The NRC staff issued a notice of

[[Page 9988]]

opportunity for comment in the Federal Register on June 14, 2001 (66 FR 
32400), on possible amendments concerning missed surveillances, 
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 September 28, 2001 (66 FR 
49714). The licensee affirmed the applicability of the following NSHC 
determination in its application dated June 24, 2004.
    As required by 10 CFR 50.91(a), an analysis of the issue of no 
significant hazards consideration is presented below:

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

    The proposed change relaxes the time allowed to perform a missed 
surveillance and adds a Bases Control Program. The time between 
surveillances is not an initiator of any accident previously 
evaluated. Consequently, the probability of an accident previously 
evaluated is not significantly increased. The equipment being tested 
is still required to be operable and capable of performing the 
accident mitigation functions assumed in the accident analysis. As a 
result, the consequences of any accident previously evaluated are 
not significantly affected. Any reduction in confidence that a 
standby system might fail to perform its safety function due to a 
missed surveillance is small and would not, in the absence of other 
unrelated failures, lead to an increase in consequences beyond those 
estimated by existing analyses. The addition of a requirement to 
assess and manage the risk introduced by the missed surveillance 
will further minimize possible concerns. The addition of a new 
Section 6.18 to add a Bases Control Program has no effect on the 
operation or testing of any plant equipment and would not affect any 
accident initiator. The addition of a Bases Control Program is 
administrative in nature, and would not affect the probability or 
consequences of an accident. Therefore, this change does not involve 
a significant increase in the probability or consequences of an 
accident previously evaluated.

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

    The proposed change does not involve a physical alteration of 
the plant (no new or different type of equipment will be installed) 
or a change in the methods governing normal plant operation. A 
missed surveillance will not, in and of itself, introduce new 
failure modes or effects and any increased chance that a standby 
system might fail to perform its safety function due to a missed 
surveillance would not, in the absence of other unrelated failures, 
lead to an accident beyond those previously evaluated. The addition 
of a requirement to assess and manage the risk introduced by the 
missed surveillance will further minimize possible concerns. The 
addition of a Bases Control Program is administrative in nature, and 
will not create any new accident initiators. Thus, this change does 
not create the possibility of a new or different kind of accident 
from any accident previously evaluated.

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

    The extended time allowed to perform a missed surveillance does 
not result in a significant reduction in the margin of safety. As 
supported by the historical data, the likely outcome of any 
surveillance is verification that the LCO [Limiting Condition for 
Operation] is met. Failure to perform a surveillance within the 
prescribed frequency does not cause equipment to become inoperable. 
The only effect of the additional time allowed to perform a missed 
surveillance on the margin of safety is the extension of the time 
until inoperable equipment is discovered to be inoperable by the 
missed surveillance. However, given the rare occurrence of 
inoperable equipment, and the rare occurrence of a missed 
surveillance, a missed surveillance on inoperable equipment would be 
very unlikely. This must be balanced against the real risk of 
manipulating the plant equipment or condition to perform the missed 
surveillance. In addition, parallel trains and alternate equipment 
are typically available to perform the safety function of the 
equipment not tested. Thus, there is confidence that the equipment 
can perform its assumed safety function. The addition of a Bases 
Control Program is administrative in nature, serves to ensure that 
changes to the Bases are made in accordance with approved criteria, 
and will not have a significant affect on the margin of safety.

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

Carolina Power & Light Company, et al., Docket No. 50-400, Shearon 
Harris Nuclear Power Plant, Unit 1, Wake and Chatham Counties, North 
Carolina

    Date of amendment request: October 15, 2004.
    Description of amendment request: The proposed amendment revises 
surveillance requirements related to the reactor coolant pump flywheel 
inspections to extend the allowable inspection interval to 20 years.
    The NRC staff issued a notice of availability of a model safety 
evaluation and model no significant hazards consideration (NSHC) 
determination for referencing in license amendment applications in the 
Federal Register on October 22, 2003 (68 FR 60422). The licensee 
affirmed the applicability of the model NSHC determination in its 
application dated October 15, 2004.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), an analysis of the issue 
of no significant hazards consideration is presented below:

    (1) The proposed change does not involve a significant increase 
in the probability or consequences of an accident previously 
evaluated.
    The proposed change to the RCP flywheel examination frequency 
does not change the response of the plant to any accidents. The RCP 
will remain highly reliable and the proposed change will not result 
in a significant increase in the risk of plant operation. Given the 
extremely low failure probabilities for the RCP motor flywheel 
during normal and accident conditions, the extremely low probability 
of a loss-of-coolant accident (LOCA) with loss of offsite power 
(LOOP), and assuming a conditional core damage probability (CCDP) of 
1.0 (complete failure of safety systems), the core damage frequency 
(CDF) and change in risk would still not exceed the NRC's acceptance 
guidelines contained in Regulatory Guide (RG) 1.174 (<1.0E-6 per 
year). Moreover, considering the uncertainties involved in this 
evaluation, the risk associated with the postulated failure of an 
RCP motor flywheel is significantly low. Even if all four RCP motor 
flywheels are considered in the bounding plant configuration case, 
the risk is still acceptably low.
    The proposed change does 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; alter or prevent the ability 
of structures, systems, components (SSCs) from performing their 
intended function to mitigate the consequences of an initiating 
event within the assumed acceptance limits; or affect the source 
term, containment isolation, or radiological release assumptions 
used in evaluating the radiological consequences of an accident 
previously evaluated. Further, the proposed change does not increase 
the type or amount of radioactive effluent that may be released 
offsite, nor significantly increase individual or cumulative 
occupational/public radiation exposure. The proposed change is 
consistent with the safety analysis assumptions and resultant 
consequences. Therefore, the proposed change does not involve a 
significant increase in the probability or consequences of an 
accident previously evaluated.
    (2) The proposed change does not create the possibility of a new 
or different kind of

[[Page 9989]]

accident from any accident previously evaluated.
    The proposed change in flywheel inspection frequency does not 
involve any change in the design or operation of the RCP. Nor does 
the change to examination frequency affect any existing accident 
scenarios, or create any new or different accident scenarios. 
Further, the change does not involve a physical alteration of the 
plant (i.e., no new or different type of equipment will be 
installed) or alter the methods governing normal plant operation. In 
addition, the change does not impose any new or different 
requirements or eliminate any existing requirements, and does not 
alter any assumptions made in the safety analysis. The proposed 
change is consistent with the safety analysis assumptions and 
current plant operating practice. Therefore, the proposed change 
does not create the possibility of a new or different kind of 
accident from any accident previously evaluated.
    (3) The proposed change does not involve a significant reduction 
in a margin of safety.
    The proposed change does not alter the manner in which safety 
limits, limiting safety system settings, or limiting conditions for 
operation are determined. The safety analysis acceptance criteria 
are not impacted by this change. The proposed change will not result 
in plant operation in a configuration outside of the design basis. 
The calculated impact on risk is insignificant and meets the 
acceptance criteria contained in RG 1.174. There are no significant 
mechanisms for inservice degradation of the RCP flywheel. Therefore, 
the proposed change does not involve a significant reduction in a 
margin of safety.

    The NRC staff proposes to determine that the amendment request 
involves no significant hazards consideration.
    Attorney for licensee: David T. Conley, Associate General Counsel 
II--Legal Department, Progress Energy Service Company, LLC, Post Office 
Box 1551, Raleigh, North Carolina 27602.
    NRC Section Chief: Michael L. Marshall, Jr.

Connecticut Yankee Atomic Power Company, Docket No. 50-213, Haddam Neck 
Plant, Middlesex County, Connecticut

    Date of amendment request: December 1, 2004.
    Description of amendment requests: The Haddam Neck Plant (HNP) is 
currently undergoing active decommissioning. The proposed amendment 
would revise the License Termination Plan (LTP) to revise the buried 
debris dose model and surface contamination release limits for various 
piping sizes. Specifically CYAPCO proposes to:
    1. Modify the dose model for volumetrically contaminated concrete, 
rebar (hereafter referred to as simply ``concrete''), the containment 
liner and embedded piping in basements that are to remain in place at 
the HNP site. The revised approach results in the offsite disposal of a 
larger percentage of the concrete structures (approximately 75% of that 
which would remain under the current approach). The overall effect 
results in a smaller amount of radioactivity contained in concrete to 
remain on-site than is allowed by the current LTP. The method of 
calculating the future groundwater pathway dose using the concrete 
debris model is being revised to an inventory based approach which will 
include activity inventories from the containment liner, embedded 
piping inside surfaces and radioactivity released from volumetrically 
contaminated concrete (which is controlled by diffusion rate through 
basement walls and flowable fill). The concrete that will remain is in 
the containment lower walls and floor mat, the in-core instrumentation 
sump, and the lower walls and floor of the spent fuel pool in the fuel 
building. The Basement Fill Model will also be used for other basements 
and footings that will remain on site using the results of future 
characterization surveys.
    2. Additionally, CYAPCO proposes to include surface contamination 
release levels for other pipe diameters that may be encountered during 
the decommissioning beyond that currently included in the LTP for 4 
inch piping.
    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:

    In accordance with 10 CFR 50.92, CYAPCO has reviewed the 
amendment request and concluded that the amendment request does not 
involve a Significant Hazards Consideration (SHC). The basis for 
this conclusion is that the three criteria of 10 CFR 50.92(c) are 
not compromised. The amendment request does not involve an SHC 
because the amendment request would not:
    A. Involve a significant increase in the probability or 
consequences of an accident previously evaluated.
    The activities included in the amendment request are within the 
bounds of those contained in the HNP Updated Final Safety Analysis 
Report (UFSAR). The HNP UFSAR Chapter 15 provides a discussion of 
the radiological events postulated to occur as a result of 
decommissioning activities with bounding consequences resulting from 
a resin container accident. This accident is expected to contain 
more potential airborne activity than can be released from other 
decommissioning events. The radionuclide distribution assumed for 
the resin container has a greater inventory of transuranics 
radionuclides (major dose contributor) than the distribution of 
plant derived radionuclides in the components involved in other 
decommissioning activities. The HNP UFSAR also discusses a fuel 
handling accident in the fuel building, involving the drop of a 
spent fuel assembly onto the fuel racks. The postulated drop assumes 
the rupture of all fuel rods in the associated assembly. The 
probability or consequences of this accident would not be increased 
during any future fuel operations in the spent fuel building related 
to decommissioning. Transfer of the spent fuel to canisters for dry 
cask storage involves additional restrictions contained in the cask 
certificate of compliance in order to maintain decommissioning 
activities within the assumptions of and consequences of the fuel 
handling accident. No systems, structures, or components that could 
initiate or be required to mitigate consequences of an accident are 
affected by the amendment request in any way not previously 
evaluated in the HNP UFSAR. Therefore, the amendment request does 
not involve any increase in the probability or consequences of any 
accident previously evaluated.
    B. Create the possibility of a new or different kind of accident 
from any accident previously evaluated.
    Accident analyses related to decommissioning activities are 
addressed in the HNP UFSAR. The activities included in the amendment 
request are within the bounds of those considered in the HNP UFSAR. 
Thus, the amendment request does not affect plant systems, 
structures, or components in any way previously evaluated in the HNP 
UFSAR. The amendment request does not introduce any new failure 
modes. Therefore, the amendment request will not create the 
possibility of a new or different kind of accident from any 
previously evaluated.
    C. Involve a significant reduction in a margin of safety.
    The HNP LTP is a plan for demonstrating compliance with 
radiological criteria for license termination as provided in 10 CFR 
20.1402. The margin of safety defined in the statements of 
consideration for the final rule on the Radiological Criteria for 
License Termination is described as the margin between 100 mrem/yr 
public dose limit established in 10 CFR 20.1301 for licensed 
operation and the 25 mrem/yr dose limit to the average member of the 
critical group at a site considered acceptable for unrestricted use 
(one of the criteria of 10 CFR 20.1402). This margin of safety 
accounts for the potential effects of multiple sources of radiation 
exposure to the critical group. Since the HNP LTP was designed to 
comply with the radiological criteria for license termination for 
unrestricted use, this license amendment request supports this 
margin of safety. Also, as previously discussed, the bounding 
accident for decommissioning is the resin container accident. Since 
the bounding decommissioning accident results in more airborne 
radioactivity than can be released from the other decommissioning 
events, the margin of safety associated with consequences of 
decommissioning accidents is not reduced by this amendment request. 
Thus, the amendment request does not involve a significant reduction 
in the margin of safety.


[[Page 9990]]


    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.
    NRC Section Chief: Claudia Craig.

Duke Power Corporation (DPC), Docket Nos. 50-369 and 50-370, McGuire 
Nuclear Station (McGuire), Units 1 and 2, Mecklenburg County, North 
Carolina

    Date of amendment request: January 19, 2005.
    Description of amendment request: The proposed amendments would 
revise the McGuire, Units 1 and 2, Technical Specification (TS) 5.6.5.b 
to add an NRC-approved Topical Report to the list of analytical methods 
used to determine core operating limits.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

Criterion 1--Does this LAR Involve a Significant Increase in the 
Probability or Consequences of an Accident Previously Evaluated?

    No. This LAR makes an administrative change to Technical 
Specification (TS) 5.6.5.b, ``Core Operating Limits Report (COLR).'' 
This TS contains a listing of documents (analytical methods) that 
are used to determine core operating limits. These documents are 
separately and individually reviewed and approved by the NRC. The 
current LAR adds a new document, DPC-NE-1005P-A, ``Duke Power 
Nuclear Design Methodology Using CASMO-4/SIMULATE-3 MOX,'' (DPC 
Proprietary), to the list in TS 5.6.5.b. Topical Report ``DPC-NE-
1005P-A'' has been previously reviewed by the NRC and determined to 
be appropriate for use at McGuire. The NRC's determination was 
documented in a safety evaluation report dated August 20, 2004. 
Based on these considerations, it has been determined that the 
proposed administrative change has no impact on any accident 
probabilities or consequences.

Criterion 2--Does This LAR Create the Possibility of a New or Different 
Kind of Accident From Any Accident Previously Evaluated?

    No. This LAR is solely administrative in nature since it only 
adds an NRC-approved licensing basis document to the TS. No new 
accident causal mechanisms will be created as a result of the NRC 
approval of this LAR.

Criterion 3--Does This LAR Involve a Significant Reduction in a Margin 
of Safety?

    No. This LAR is solely administrative in nature. The analytical 
methodologies used to generate the core operating limits are 
separately and individually reviewed and approved by the NRC, and 
are unchanged by this LAR. The change contained in this LAR merely 
revises the McGuire TS in an administrative manner in order to 
conform with a Duke licensing action that has been previously 
approved by the NRC. Therefore the change proposed in this amendment 
request has no impact on margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: Ms. Lisa F. Vaughn, Duke Energy Corporation, 
422 South Church Street, Charlotte, North Carolina 28201-1006.
    NRC Section Chief: John A. Nakoski.

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

    Date of amendment request: December 20, 2004.
    Description of amendment request: The requested change will delete 
Technical Specification (TS) 5.5.1, ``Occupational Radiation Exposure 
Report,'' and TS 5.6.4, ``Monthly Operating Reports.''
    The NRC staff issued a notice of availability of a model no 
significant hazards consideration (NSHC) determination for referencing 
in license amendment applications in the Federal Register on June 23, 
2004 (69 FR 35067). The licensee affirmed the applicability of the 
model NSHC determination in its application dated December 20, 2004.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), an analysis of the issue 
of no significant hazards consideration is presented below:

    1. Does the proposed change involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    Response: No.
    The proposed change eliminates the Technical Specifications 
(TSs) reporting requirements to provide a monthly operating report 
of shutdown experience and operating statistics if the equivalent 
data is submitted using an industry electronic database. It also 
eliminates the TS reporting requirement for an annual occupational 
radiation exposure report, which provides information beyond that 
specified in NRC regulations. The proposed change involves no 
changes to plant systems or accident analyses. As such, the change 
is administrative in nature and does not affect initiators of 
analyzed events or assumed mitigation of accidents or transients. 
Therefore, the proposed change does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    2. Does the proposed change create the possibility of a new or 
different kind of accident from any accident previously evaluated?
    Response: No.
    The proposed change does not involve a physical alteration of 
the plant, add any new equipment, or require any existing equipment 
to be operated in a manner different from the present design. 
Therefore, the proposed change does not create the possibility of a 
new or different kind of accident from any accident previously 
evaluated.
    3. Does the proposed change involve a significant reduction in a 
margin of safety?
    Response: No.
    This is an administrative change to reporting requirements of 
plant operating information and occupational radiation exposure 
data, and has no effect on plant equipment, operating practices or 
safety analyses assumptions. For these reasons, the proposed change 
does not involve a significant reduction in the margin of safety.

    The NRC staff proposes to determine that the amendment request 
involves no significant hazards consideration.
    Attorney for licensee: Mark Wetterhahn, Esq., Winston & Strawn, 
1400 L Street, NW., Washington, DC 20005.
    NRC Section Chief: Allen G. Howe.

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

    Date of amendment request: December 14, 2004.
    Description of amendment request: The proposed amendment would 
eliminate certain administrative requirements for safety limit 
violations that are adequately addressed in 10 CFR 50.36(c)(1)(i)(A), 
10 CFR 50.72, 10 CFR 50.73, and by procedures; replace plant-specific 
titles with generic titles; remove the remaining responsibilities of 
the Operations Review Committee; replace descriptive details specified 
in Technical Specification (TS) 3.13.A.1 associated with 10 CFR 
50.55a(f), ``Inservice Testing Requirements,'' with reference to the 
``Inservice Code Testing Program''; make administrative changes to TS 
5.5.4, ``Radioactive Effluent Controls Program''; and make editorial 
corrections and clarifications.
    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:

    Entergy has evaluated whether or not a significant hazards 
consideration is involved with the proposed amendment(s) by focusing

[[Page 9991]]

on the three standards set forth in 10 CFR 50.92, ``Issuance of 
amendment,'' as discussed 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 is administrative in nature 
and does not involve the modification of any plant equipment or 
affect basic plant operation. There is no impact to any 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 any physical 
alteration of plant equipment and does not change the method by 
which any safety-related system performs its function. As such, no 
new or different types of equipment will be installed, and the basic 
operation of installed equipment is unchanged. The methods governing 
plant operation and testing remain consistent with current safety 
analysis assumptions.
    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 represents the relocation of 
specific Technical Specification requirements, based on regulatory 
guidance and previously approved changes for other stations or 
deletion of detail redundant to regulations or no longer applicable 
(i.e., expired one-time exceptions). The proposed change is 
administrative in nature, does not negate or revise any existing 
requirement, and does not adversely affect existing plant safety 
margins or the reliability of the equipment assumed to operate in 
the safety analysis. As such, there are no changes being made to 
safety analysis assumptions, safety limits or safety system settings 
that would adversely affect plant safety as a result of the proposed 
change. Margins of safety are unaffected by requirements that are 
retained, but relocated from the Technical Specifications. 
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 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: J.M. Fulton, Esquire, Assistant General 
Counsel, Pilgrim Nuclear Power Station, 600 Rocky Hill Road, Plymouth, 
Massachusetts, 02360-5599.
    NRC Section Chief: Darrell Roberts.

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

    Date of amendment request: December 14, 2004.
    Description of amendment request: The proposed amendment would 
remove the additional requirement to perform functional testing of the 
Average Power Range Monitor (APRM) and Anticipated Transient Without 
Scram Recirculation Pump Trip Alternate Rod Insertion instrumentation 
on each startup, even when the nominally required quarterly testing is 
current. Additionally, performance of the APRM High Flux heat balance 
calibration is modified to apply only after 12 hours at >25% power. 
Additional editorial clarifications related to Table 4.2.A through 
4.2.G, Note 2 and associated Table references are also proposed.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Does the proposed change involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    Response: No. The proposed changes to eliminate startup-related 
functional testing, even when the nominally required quarterly 
testing is current, will not result in a significant increase in the 
probability or consequences of an accident previously evaluated 
because there is no change to the requirement that the instrument 
channels remain operable and are periodically tested throughout the 
time that the associated function is required. The surveillance 
continues to be performed at the normal frequency and the normal 
surveillance frequency has been shown, based on operating 
experience, to be adequate for assuring that required conditions are 
established and maintained.
    Delaying the APRM [Average Power Range Monitor] heat balance 
calibration until conditions allow for accurate results will not 
result in a significant increase in the probability or consequences 
of an accident previously evaluated because there is no change to 
the requirement that the instrument channels remain operable. The 
ability of the APRMs to adequately respond to power excursions from 
< 25% that assume an APRM trip at 120% is not significantly impacted 
by deferring the APRM-to-heat balance calibration from the currently 
required 15% power, until the proposed 12 hours after >= 25% power. 
Additional editorial changes have no technical or operational 
impact.
    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 any physical 
alteration of plant equipment and does not change the method by 
which any safety-related system performs its function. As such, no 
new or different types of equipment will be installed, and the basic 
operation of installed equipment is unchanged. The methods governing 
plant operation and testing remain consistent with current safety 
analysis assumptions.
    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 changes do not negate any existing 
equipment or system performance requirements, and do not adversely 
affect existing plant safety margins or the reliability of the 
equipment assumed to operate in the safety analysis. As such, there 
are no changes being made to safety analysis assumptions, safety 
limits or safety system settings that would adversely affect plant 
safety as a result of the proposed change.
    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 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: J. M. Fulton, Esquire, Assistant General 
Counsel, Pilgrim Nuclear Power Station, 600 Rocky Hill Road, Plymouth, 
Massachusetts, 02360-5599.
    NRC Section Chief: Darrell Roberts.

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

    Date of amendment request: December 14, 2004.
    Description of amendment request: The proposed amendment would 
relocate various requirements from the Technical Specification (TS) to 
the Final Safety Analysis Report (FSAR) or TS Bases.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Does the proposed change involve a significant increase in 
the probability or

[[Page 9992]]

consequences of an accident previously evaluated?
    Response: No. The proposed relocations are administrative in 
nature and do not involve the modification of any plant equipment or 
affect basic plant operation. The associated instrumentation and 
inspections are not assumed to be an initiator of any analyzed 
event, nor are these limits assumed in the mitigation of 
consequences of accidents.
    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 any physical 
alteration of plant equipment and does not change the method by 
which any safety-related system performs its function. As such, no 
new or different types of equipment will be installed, and the basic 
operation of installed equipment is unchanged. The methods governing 
plant operation and testing remain consistent with current safety 
analysis assumptions. 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 changes to relocate current TS 
requirements to the FSAR, consistent with regulatory guidance and 
previously approved changes for other stations, are administrative 
in nature. These changes do not negate any existing requirement, and 
do not adversely affect existing plant safety margins or the 
reliability of the equipment assumed to operate in the safety 
analysis. As such, there are no changes being made to safety 
analysis assumptions, safety limits or safety system settings that 
would adversely affect plant safety as a result of the proposed 
change. Margins of safety are unaffected by requirements that are 
retained, but relocated from the Technical Specifications to the 
FSAR. 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 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: J.M. Fulton, Esquire, Assistant General 
Counsel, Pilgrim Nuclear Power Station, 600 Rocky Hill Road, Plymouth, 
Massachusetts, 02360-5599.
    NRC Section Chief: Darrell Roberts.

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

    Date of amendment request: December 17, 2004.
    Description of amendment request: The proposed amendment would 
delete the Technical Specification (TS) requirements to submit monthly 
operating reports and occupational radiation exposure reports.
    The NRC staff issued a notice of availability of a model no 
significant hazards consideration (NSHC) determination for referencing 
in licensing amendment applications in the Federal Register on June 23, 
2004 (69 FR 35067). The licensee affirmed the applicability of the 
model NSHC determination in its application dated December 17, 2004.
    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 NSHC, 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 eliminates the Technical Specifications 
(TSs) reporting requirements to provide a monthly operating report 
of shutdown experience and operating statistics if the equivalent 
data is submitted using an industry electronic database. It also 
eliminates the TS reporting requirement for an annual occupational 
radiation exposure report, which provides information beyond that 
specified in NRC regulations. The proposed change involves no 
changes to plant systems or accident analyses. As such, the change 
is administrative in nature and does not affect initiators of 
analyzed events or assumed mitigation of accidents or transients. 
Therefore, the proposed change does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    2. Does the proposed change create the possibility of a new or 
different kind of accident from any accident previously evaluated?
    Response: No
    The proposed change does not involve a physical alteration of 
the plant, add any new equipment, or require any existing equipment 
to be operated in a manner different from the present design. 
Therefore, the proposed change does not create the possibility of a 
new or different kind of accident from any accident previously 
evaluated.
    3. Does the proposed change involve a significant reduction in a 
margin of safety?
    Response: No
    This is an administrative change to reporting requirements of 
plant operating information and occupational radiation exposure 
data, and has no effect on plant equipment, operating practices or 
safety analyses assumptions. For these reasons, the proposed change 
does not involve a significant reduction in the margin of safety.

    The NRC staff proposes to determine that the amendment request 
involves no significant hazards consideration.
    Attorney for licensee: Nicholas S. Reynolds, Esquire, Winston and 
Strawn, 1400 L Street, NW., 12th Floor, Washington, DC 20005-3502
    NRC Section Chief: Allen G. Howe.

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

    Date of amendment request: October 15, 2004.
    Description of amendment request: The proposed amendment revises 
surveillance requirements related to the reactor coolant pump (RCP) 
flywheel inspections to extend the allowable inspection interval to 20 
years.
    The NRC staff issued a model safety evaluation and model no 
significant hazards consideration (NSHC) determination for referencing 
in license amendment applications in the Federal Register on June 24, 
2003 (68 FR 37590). The notice of availability of the model application 
was issued on October 22, 2003 (68 FR 60422). The licensee affirmed the 
applicability of the model NSHC determination in its application dated 
October 15, 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

    The proposed change to the RCP flywheel examination frequency 
does not change the response of the plant to any accidents. The RCP 
will remain highly reliable and the proposed change will not result 
in a significant increase in the risk of plant operation. Given the 
extremely low failure probabilities for the RCP motor flywheel 
during normal and accident conditions, the extremely low probability 
of a loss-of-coolant accident (LOCA) with loss of offsite power 
(LOOP), and assuming a conditional core damage probability (CCDP) of 
1.0 (complete failure of safety systems), the core damage frequency 
(CDF) and change in risk would still not exceed the NRC's acceptance 
guidelines [contained] in RG [Regulatory Guide] 1.174 (<1.0E-6 per 
year). Moreover, considering the uncertainties involved in this 
evaluation, the risk associated with the postulated failure of an 
RCP motor flywheel

[[Page 9993]]

is significantly low. Even if all four RCP motor flywheels are 
considered in the bounding plant configuration case, the risk is 
still acceptably low.
    The proposed change does 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; alter or prevent the ability 
of structures, systems, components (SSCs) from performing their 
intended function to mitigate the consequences of an initiating 
event within the assumed acceptance limits; or affect the source 
term, containment isolation, or radiological release assumptions 
used in evaluating the radiological consequences of an accident 
previously evaluated. Further, the proposed change does not increase 
the type or amount of radioactive effluent that may be released 
offsite, nor significantly increase individual or cumulative 
occupational/public radiation exposure. The proposed change is 
consistent with the safety analysis assumptions and resultant 
consequences. Therefore, the proposed change does not involve a 
significant increase in the probability or consequences of an 
accident previously evaluated.

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

    The proposed change in flywheel inspection frequency does not 
involve any change in the design or operation of the RCP. Nor does 
the change to examination frequency affect any existing accident 
scenarios, or create any new or different accident scenarios. 
Further, the change does not involve a physical alteration of the 
plant (i.e., no new or different type of equipment will be 
installed) or alter the methods governing normal plant operation. In 
addition, the change does not impose any new or different 
requirements or eliminate any existing requirements, and does not 
alter any assumptions made in the safety analysis. The proposed 
change is consistent with the safety analysis assumptions and 
current plant operating practice. Therefore, the proposed change 
does not create the possibility of a new or different kind of 
accident from any accident previously evaluated.

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

    The proposed change does not alter the manner in which safety 
limits, limiting safety system settings, or limiting conditions for 
operation are determined. The safety analysis acceptance criteria 
are not impacted by this change. The proposed change will not result 
in plant operation in a configuration outside of the design basis. 
The calculated impact on risk is insignificant and meets the 
acceptance criteria contained in RG 1.174. There are no significant 
mechanisms for inservice degradation of the RCP flywheel. Therefore, 
the proposed change does not involve a significant reduction in a 
margin of safety.

    The NRC staff proposes to determine that the amendment request 
involves no significant hazards consideration.
    Attorney for licensee: David T. Conley, Associate General Counsel 
II--Legal Department, Progress Energy Service Company, LLC, Post Office 
Box 1551, Raleigh, North Carolina 27602.
    NRC Section Chief: Michael L. Marshall, Jr.

Florida Power and Light Company, Docket No. 50-389, St. Lucie Plant, 
Unit No. 1, St. Lucie County, Florida

    Date of amendment request: December 20, 2004.
    Description of amendment request: The proposed license amendment 
would extend the effectiveness of the current Technical Specification 
pressure/temperature (P/T) limit curves, also called the heatup and 
cooldown curves, from 23.6 to 35 effective full power years (EFPY). The 
low temperature overpressure protection requirements, which are based 
on the P/T limits, would also be extended to 35 EFPY. The proposed 
amendment would revise Technical Specification Figures 3.1-1b, 3.4-2a, 
3.4-2b, and 3.4-3.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    (1) Operation of the facility in accordance with the proposed 
amendment would not involve a significant increase in the 
probability or consequences of an accident previously evaluated.
    The pressure/temperature (P/T) limit curves in the Technical 
Specifications are conservatively generated in accordance with the 
fracture toughness requirements of 10 CFR 50, Appendix G, as 
supplemented by the ASME [American Society of Mechanical Engineers] 
Code [Boiler and Pressure Vessel Code], Section Xl, Appendix G 
recommendations. The adjusted reference temperature (ART) values are 
based on the Regulatory Guide 1.99, Revision 2, shift prediction and 
attenuation formula and have been validated by a credible reactor 
vessel surveillance program. There are no changes to the limit 
curve, only a change in the period of applicability based on more 
recent fluence predictions and new best estimate chemistry 
information. Based on the current fluence projections, analysis has 
demonstrated that the current P/T limit curves will remain 
conservative for up to 35 EFPY.
    In conjunction with extending the effectiveness of the existing 
P/T limit curves, the low temperature overpressure protection (LTOP) 
analysis for 23.6 EFPY is also extended to 35 EFPY. The LTOP 
analysis confirms that the current setpoints for the power operated 
relief valves (PORVs) will provide the appropriate overpressure 
protection at low reactor coolant system (RCS) temperatures. Because 
the P/T limit curves have not changed, the existing LTOP values have 
not changed, which include the PORV setpoints.
    The P/T limit curves and LTOP analysis have not changed; 
therefore, the proposed amendment does not represent a change in the 
configuration or operation of the plant. The results of the existing 
LTOP analysis have not changed, and the limiting pressures for given 
temperatures will not be exceeded for the postulated transients. 
Therefore, assurance is provided that reactor vessel integrity will 
be maintained. Thus, the proposed amendment does not involve an 
increase in the probability or consequences of accidents previously 
evaluated.
    (2) Operation of the facility in accordance with the proposed 
amendment would not create the possibility of a new or different 
kind of accident from any previously evaluated.
    The requirements for P/T limit curves and LTOP have been in 
place since the beginning of plant operation. The only changes in 
these curves are the extension of the period of applicability 
(EFPY), which is based on new fluence data and the operating time 
(EFPY) required to reach the same limiting adjusted reference 
temperature projection used for the current 23.6 EFPY P/T limit 
curves. Since there is no change in the configuration or operation 
of the facility as a result of the proposed amendment, the proposed 
amendment does not create the possibility of a new or different kind 
of accident from any accident previously evaluated.
    (3) Operation of the facility in accordance with the proposed 
amendment would not involve a significant reduction in a margin of 
safety.
    Analysis has demonstrated that the fracture toughness 
requirements of 10 CFR 50, Appendix G, are satisfied and that 
conservative operating restrictions are maintained for the purpose 
of low temperature overpressure protection. The P/T limit curves 
will provide assurance that the RCS pressure boundary will behave in 
ductile manner and that the probability of a rapidly propagating 
fracture is acceptably low. Therefore, operation 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 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: M.S. Ross, Attorney, Florida Power & Light, 
P.O. Box 14000, Juno Beach, Florida 33408-0420.
    NRC Section Chief: Michael L. Marshall, Jr.

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

    Date of amendment request: January 6, 2005.
    Description of amendment request: The proposed amendment revises

[[Page 9994]]

Technical Specification Section 3/4.4.5, Steam Generators, to allow 
repair of steam generator tubes by installing Westinghouse Electric LLC 
Alloy 800 leak limiting sleeves.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    (1) Does the proposed change involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    No, the leak limiting Alloy 800 tube sleeves are designed using 
the applicable ASME [American Society of Mechanical Engineers] 
Boiler and Pressure Vessel Code and meet the design objectives of 
the original steam generator tubing. The applied stresses and 
fatigue usage factors for the sleeves are bounded by the limits 
established in the ASME Code. Mechanical testing has shown that the 
structural strength of leak limiting sleeves under normal, upset, 
emergency, and faulted conditions provides margin to the acceptance 
limits. These acceptance limits bound the most limiting burst margin 
of three times the normal operating pressure differential as 
recommended by NRC [U.S. Nuclear Regulatory Commission] Regulatory 
Guide 1.121. Burst testing of sleeved-tube assemblies has confirmed 
the analytical results and demonstrated that levels of primary-to-
secondary leakage are not expected to exceed acceptable levels 
during any anticipated plant operating condition.
    The leak limiting Alloy 800 sleeve depth-based structural limit 
is determined using NRC guidance and the pressure-stress equation of 
the ASME Code, Section III with margin added to account for the 
configuration of long axial cracks. An Alloy 800 sleeved tube will 
be plugged on detection of an imperfection in the sleeve or in the 
pressure boundary portion of the original tube wall.
    An evaluation of repaired steam generator tubes, plus testing, 
and analysis indicates that unacceptable detrimental effects on the 
leak limiting Alloy 800 sleeve or of a sleeved tube are not expected 
from the reactor coolant system flow, primary or secondary coolant 
chemistries, thermal conditions or transients, or pressure 
conditions as may be experienced at St. Lucie Unit 2. Corrosion 
testing and historical performance of sleeved steam generator tubes 
indicates no evidence of sleeve or tube corrosion considered 
detrimental under anticipated service conditions. The implementation 
of the proposed tube sleeving has no significant effect on either 
the configuration of the plant or the manner in which it is 
operated.
    The consequences of a hypothetical failure of a leak limiting 
Alloy 800 sleeved tube is bounded by the current steam generator 
tube rupture analysis described in the St. Lucie Unit 2 Updated 
Final Safety Analysis Report. Due to the slight reduction in the 
inside diameter caused by the sleeve wall thickness, primary coolant 
release rates through the parent tube during a tube rupture event 
would be slightly less than that assumed for the steam generator 
tube rupture analysis and therefore, would result in lower total 
primary fluid mass release to the secondary system. A main steam 
line break or feedwater line break will not cause a steam generator 
tube rupture since the sleeves are analyzed for a maximum accident 
differential pressure greater than that predicted in the St. Lucie 
Unit 2 safety analysis.
    Fluid leakage from a sleeved tube during plant operation would 
be minimal and is well within the allowable Technical Specification 
leakage limits. Therefore, the proposed tube sleeving 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, the leak limiting Alloy 800 sleeves are designed using the 
applicable ASME Code as guidance, and therefore, meet the objectives 
of the original steam generator tubing. As a result, the function of 
the steam generator will not be significantly affected by the 
installation of the proposed sleeves. The proposed sleeves do not 
interact with any other plant systems. Any accident that would 
result from potential tube or sleeve degradation in the repaired 
portion of the tube is bounded by the existing steam generator tube 
rupture accident analysis, thus the potential for a new type of 
accident is not created. The continued integrity of the sleeved tube 
is periodically verified by surveillance inspections performed in 
compliance with Technical Specification requirements. A sleeved tube 
will be plugged on detection of any service induced imperfection, 
degradation, or defect in the sleeve and/or pressure boundary 
portion of the original tube wall in the sleeve/tube assembly (i.e., 
the sleeve-to-tube joint).
    Implementation of the proposed change has no significant effect 
on either the configuration of the plant or the manner in which it 
is operated. Therefore, the proposed change does not create the 
possibility of a new or different accident from any accident 
previously evaluated.
    (3) Does the proposed change involve a significant reduction in 
a margin of safety?
    No, the repair of degraded steam generator tubes with leak 
limiting Alloy 800 sleeves restores the structural integrity of the 
degraded tube under normal operating and postulated accident 
conditions. The reduction in core cooling margin due to the addition 
of Alloy 800 sleeves is not significant because the cumulative 
effect of all sleeved and plugged tubes will continue to be less 
than the currently-allowed core cooling margin threshold established 
by the total steam generator tube plugging level. Design safety 
factors utilized for the sleeves are consistent with the safety 
factors in the ASME Boiler and Pressure Vessel Code used in the 
original steam generator design. Each tube and portions of the tube 
with an installed sleeve that constitute the reactor coolant 
pressure boundary will be monitored; a sleeved tube will be plugged 
on detection of any service induced imperfection, degradation, or 
defect in the sleeve and/or pressure boundary portion of the 
original tube wall in the sleeve/tube assembly (i.e., the sleeve-to-
tube joint). Use of the previously-identified design criteria and 
design verification testing assures that the margin to safety is not 
significantly different from that of the original steam generator 
tubes. Therefore, the proposed repairs employing leak limiting Alloy 
800 tube sleeves 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 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: M.S. Ross, Attorney, Florida Power & Light, 
P.O. Box 14000, Juno Beach, Florida 33408-0420.
    NRC Section Chief: Michael L. Marshall, Jr.

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

    Date of amendment request: September 8, 2004.
    Description of amendment request: The proposed amendments would 
change SSES 1 and 2 Technical Specifications 3.6.4.1, ``Secondary 
Containment,'' and 3.6.4.3, ``Standby Gas Treatment System (SGTS),'' to 
extend, on a one-time basis, the allowable completion time for required 
actions for secondary containment inoperable and two SGTS subsystems 
inoperable, in mode 1, 2, or 3, from 4 hours to 48 hours.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Does the proposed change involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    Response: No.
    The proposed change does not involve a significant increase in 
the probability of an accident previously evaluated because neither 
Secondary Containment nor the Standby Gas Treatment System is an 
initiator of an accident. Both mitigate accident consequences.
    The consequences of a Design Basis Analysis-Loss of Coolant 
Accident (DBA-LOCA) have been evaluated in the FSAR [Final Safety 
Analysis Report]. Increasing the completion time for Secondary 
Containment and two SGTS subsystems inoperable from 4

[[Page 9995]]

hours to 48 does not result in a significant increase in the 
consequences of a DBA-LOCA event nor change the evaluation of DBA-
LOCA events as stated in the FSAR evaluation. The radiological 
evaluation of DBA-LOCA doses, including doses offsite, Control Room 
habitability, and exposures for personnel access demonstrates that 
there would be no significant impact. Movement of irradiated fuel 
within Secondary Containment will be prohibited during the extended 
LCO period, to preclude a fuel handling accident, which might lead 
to a radiological consequence.
    Therefore, this change does not involve a significant increase 
in the probability or consequences of an accident previously 
evaluated.
    2. Does the proposed change create the possibility of a new or 
different kind of accident from any accident previously evaluated?
    Response: No.
    The proposed changes do not involve a physical alteration of the 
plant. No new or different type of equipment will be installed 
(damper motors will be replaced) nor will there be changes in 
methods governing normal plant operation.
    The accident analyses affected by this extension are the 
radiological events that are discussed in the FSAR. The potential 
for the loss of other plant systems or equipment to mitigate the 
effects of an accident is not altered.
    The proposed changes do not require any new operator response or 
introduce any new opportunities for operator error not previously 
considered.
    Thus, this change does not create the possibility of a new or 
different kind of accident from any accident previously evaluated.
    3. Does the proposed change involve a significant reduction in a 
margin of safety?
    Response: No.
    The increase in completion time for Standby Gas Treatment does 
not result in any effect on the margin of safety. There is no 
increase in Core Damage Frequency (CDF) or Large Early Release 
Frequency (LERF). A recovery plan will be in place to restore the 
SGTS and Secondary Containment to functional, if a DBA-LOCA accident 
should occur. Implementation of the compensatory measures minimizes 
the probability that an accident will be initiated, maximizes the 
probability that accident mitigation equipment will be available and 
ensures that SGTS and Secondary Containment will be able to be 
restored in a timely manner. Thus the potential impact of extending 
the Completion Time is small. Therefore, this one-time extension 
will not involve a significant reduction in safety margin.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: Bryan A. Snapp, Esquire, Assoc. General 
Counsel, PPL Services Corporation, 2 North Ninth St., GENTW3, 
Allentown, PA 18101-1179.
    NRC Section Chief: Richard J. Laufer.

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

    Date of amendment request: September 8, 2004.
    Description of amendment request: The proposed amendment would 
revise the SSES 1 and 2 Technical Specifications Surveillance 
Requirement 3.6.1.3.6 to reduce the frequency of performing leakage 
rate testing for each primary containment purge valve with resilient 
seals from 184 days to 24 months.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Does the proposed change involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    Response: No.
    The proposal would change the Technical Specification 
Surveillance Requirement for containment purge valves with resilient 
seals. The proposed change does not involve a significant increase 
in the probability or consequences of an accident previously 
evaluated because the extensive industry operating experience 
derived from test results has demonstrated that the resilient seal 
material does not degrade and cause containment isolation valves to 
leak. Further, these valves are not accident initiators. Thus, the 
valves will perform as assumed in the accident analyses. Therefore, 
this change does not involve a significant increase in the 
probability or consequences of an accident previously evaluated.
    2. Does the proposed change create the possibility of a new or 
different kind of accident from any accident previously evaluated?
    Response: No.
    The proposal would change the Technical Specifications 
Surveillance Requirement for containment purge valves with resilient 
seals. The proposed change does not involve a physical alteration of 
the plant (no new or different type of equipment will be installed 
nor changes in methods governing normal plant operation). In 
particular, it does not require the valves to function in any manner 
other than that which is currently required. Thus, this change does 
not create the possibility of a new or different kind of accident 
from any accident previously evaluated.
    3. Does the proposed change involve a significant reduction in a 
margin of safety?
    Response: No.
    The proposal would change the Technical Specifications 
Surveillance Requirement for containment purge valves with resilient 
seals. The proposed change does not involve a significant reduction 
in margin of safety because it has no effect on any safety analysis 
bases or assumptions. It does not change the leakage acceptance 
criteria. Sufficient data has been collected to demonstrate that 
resilient seals do not degrade. Testing at the same frequency as 
other containment isolation valves will not reduce the margin of 
safety provide by Technical Specifications.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: 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.

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

    Date of amendment request: May 21, 2004.
    Description of amendment request: The proposed amendment revises 
the Reactor Coolant Pump (RCP) Flywheel Inspection Program to extend 
the allowable inspection interval to 20 years.
    The NRC staff issued a notice of availability of a model safety 
evaluation and model no significant hazards consideration (NSHC) 
determination for referencing in license amendment applications in the 
Federal Register on October 22, 2003 (68 FR 60422). The licensee 
affirmed the applicability of the model NSHC determination in its 
application dated May 21, 2004.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), an analysis of the issue 
of no significant hazards consideration is presented below:

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

    The proposed change to the RCP flywheel examination frequency 
does not change the response of the plant to any accidents. The RCP 
will remain highly reliable and the proposed change will not result 
in a significant increase in the risk of plant operation. Given the 
extremely low failure probabilities for the RCP motor flywheel 
during normal and accident conditions, the extremely low probability 
of a loss-of-coolant accident (LOCA) with loss of offsite power 
(LOOP), and assuming a conditional core damage probability (CCDP) of 
1.0 (complete

[[Page 9996]]

failure of safety systems), the core damage frequency (CDF) and 
change in risk would still not exceed the NRC's acceptance 
guidelines contained in Regulatory Guide (RG) 1.174 (<1.0E-6 per 
year). Moreover, considering the uncertainties involved in this 
evaluation, the risk associated with the postulated failure of an 
RCP motor flywheel is significantly low. Even if all four RCP motor 
flywheels are considered in the bounding plant configuration case, 
the risk is still acceptably low.
    The proposed change does 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; alter or prevent the ability 
of structures, systems, components (SSCs) from performing their 
intended function to mitigate the consequences of an initiating 
event within the assumed acceptance limits; or affect the source 
term, containment isolation, or radiological release assumptions 
used in evaluating the radiological consequences of an accident 
previously evaluated. Further, the proposed change does not increase 
the type or amount of radioactive effluent that may be released 
offsite, nor significantly increase individual or cumulative 
occupational/public radiation exposure. The proposed change is 
consistent with the safety analysis assumptions and resultant 
consequences. Therefore, the proposed change does not involve a 
significant increase in the probability or consequences of an 
accident previously evaluated.

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

    The proposed change in flywheel inspection frequency does not 
involve any change in the design or operation of the RCP. Nor does 
the change to examination frequency affect any existing accident 
scenarios, or create any new or different accident scenarios. 
Further, the change does not involve a physical alteration of the 
plant (i.e., no new or different type of equipment will be 
installed) or alter the methods governing normal plant operation. In 
addition, the change does not impose any new or different 
requirements or eliminate any existing requirements, and does not 
alter any assumptions made in the safety analysis. The proposed 
change is consistent with the safety analysis assumptions and 
current plant operating practice. Therefore, the proposed change 
does not create the possibility of a new or different kind of 
accident from any accident previously evaluated.

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

    The proposed change does not alter the manner in which safety 
limits, limiting safety system settings, or limiting conditions for 
operation are determined. The safety analysis acceptance criteria 
are not impacted by this change. The proposed change will not result 
in plant operation in a configuration outside of the design basis. 
The calculated impact on risk is insignificant and meets the 
acceptance criteria contained in RG 1.174. There are no significant 
mechanisms for inservice degradation of the RCP flywheel. Therefore, 
the proposed change does not involve a significant reduction in a 
margin of safety.

    The NRC staff proposes to determine that the amendment request 
involves no significant hazards consideration.
    Attorney for licensee: Thomas G. Eppink, South Carolina Electric & 
Gas Company, Post Office Box 764, Columbia, South Carolina 29218.
    NRC Section Chief: John A. Nakoski.

Southern California Edison Company (SCE), et al., Docket Nos. 50-361, 
San Onofre Nuclear Generating Station, Unit 2, San Diego County, 
California

    Date of amendment requests: January 28, 2005.
    Description of amendment requests: The proposed change would revise 
Technical Specifications (TSs) 1.1 ``Definitions,'' 3.4 ``Reactor 
Coolant System [RCS],'' and 5.7 ``Reporting Requirements'' to relocate 
the RCS pressure-temperature curves and limits from the TSs to a 
licensee-controlled document identified as the Pressure and Temperature 
Limit Report.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Does the proposed change involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    Response: No.
    This proposed change revises the Technical Specifications by 
relocating the reactor coolant system (RCS) Pressure and Temperature 
Limits, Heatup and Cooldown Curves and Low Temperature Overpressure 
Protection (LTOP) enable temperatures from the Technical 
Specifications to a RCS Pressure and Temperature Limits Report 
(PTLR). Relocation of this information will not impact the activity 
to update the RCS pressure and temperature curves and limits in 
accordance with the requirements of 10 CFR 50 Appendix G and H to 
ensure the reactor coolant system's pressure boundary integrity will 
be protected until end of life (EOL). Consequently, this proposed 
change is determined to not contribute to the probability of or the 
initiation of accidents. There is no change to the safety analysis.
    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.
    This proposed change revises the Technical Specifications by 
relocating the RCS Pressure and Temperature Limits, Heatup and 
Cooldown Curves and LTOP enable temperatures from the Technical 
Specifications to a RCS PTLR to document removal, testing and 
analyzing the surveillance capsule. This document will be updated by 
SCE to reflect the testing and analysis of specimens. Removal, 
testing and analyzing the surveillance capsule resulted in changes 
to the RCS pressure and temperature limits. These changes are 
required to maintain the RCS pressure boundary integrity until EOL. 
Changes to the RCS pressure and temperature curves and limits will 
not create a new or different kind of accident. There is no change 
to the safety 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 change involve a significant reduction in a 
margin of safety?
    Response: No.
    Pressure and temperature curves and limits are provided as 
limits to plant operation for ensuring RCS pressure boundary 
integrity and maintained until EOL. Changes to the RCS pressure and 
temperature curves and limits, resulting from the removal, testing 
and analyzing of a surveillance capsule, are only made within the 
acceptable margin limits maintaining the required margin of safety. 
There is no change to the safety analysis.
    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 requests involve no significant hazards consideration.
    Attorney for licensee: Douglas K. Porter, Esquire, Southern 
California Edison Company, 2244 Walnut Grove Avenue, Rosemead, 
California 91770.
    NRC Section Chief: Robert A. Gramm.

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

    Date of amendment requests: February 3, 2005.
    Description of amendment requests: The proposed change would revise 
Technical Specification 3.6.3, ``Containment Isolation Valves,'' 
Surveillance Requirements 3.6.3.3 and 3.6.3.4 for Containment Isolation 
Valves and Blind Flanges (ClVs) by adding a provision to exempt CIVs 
that are locked, sealed, or otherwise secured from the position 
verification surveillance requirements.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the

[[Page 9997]]

licensee has provided its analysis of the issue of no significant 
hazards consideration, which is presented below:

    1. Do the proposed changes involve a significant increase in the 
probability or consequences of an accident previously evaluated?
    The proposed change does not affect the CIV design or function. 
In addition, mis-positioned or failed ClVs are not the initiator of 
any event. The position of a locked, sealed, or otherwise secured 
valve and blind flange is verified at the time it is locked, sealed, 
or secured, and these ClVs are administratively controlled to remain 
in the required position. Further, since the change impacts only the 
re-verification of the blind flange and valve position as a 
Technical Specification Surveillance, it does not result in any 
change in the response of the equipment to an accident.
    Based on the above, SCE concludes that deleting the re-
verification of the position of a locked, sealed, or secured CIV as 
a Technical Specification Surveillance does not affect the 
probability or consequences of an accident previously evaluated.
    2. Do the proposed changes create the possibility of a new kind 
of accident from any accident previously evaluated?
    This change does not add any new equipment or result in any 
changes to equipment design or capabilities. This change also does 
not result in any changes to the operation of the plant. The 
position of a locked, sealed, or otherwise secured blind flange and 
valve is verified at the time it is locked, sealed, or secured, and 
these ClVs are administratively controlled to remain in the required 
position. Further, since the change impacts only the re-verification 
of the blind flange and valve position as a Technical Specification 
Surveillance, it does not result in any change in the response of 
the equipment to an accident.
    Based on the above, SCE concludes that deleting the re-
verification of the position of a locked, sealed, or secured CIV as 
a Technical Specification Surveillance does not create the 
possibility of a new or different kind of accident from any 
previously evaluated.
    3. Do the proposed changes involve a significant reduction in a 
margin of safety?
    The CIVs are administratively controlled and their operation is 
a nonroutine event. The position of a locked, sealed, or otherwise 
secured blind flange and valve is verified at the time it is locked, 
sealed, or secured. Also, no CIVs were found to be out of position 
from a review of all the San Onofre Units 2 and 3 surveillance data 
from January 2000 through December 2004. Since the change only 
deletes the re-verification of the blind flange and valve position 
as a Technical Specification Surveillance and the administrative 
controls are in place, the proposed change will provide a similar 
level of assurance of correct CIV position as the current 
verifications.
    Based on the above, SCE concludes that deleting the re-
verification of the position of a locked, sealed, or secured CIV as 
a Technical Specification Surveillance 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 requests involve no significant hazards consideration.
    Attorney for licensee: Douglas K. Porter, Esquire, Southern 
California Edison Company, 2244 Walnut Grove Avenue, Rosemead, 
California 91770.
    NRC Section Chief: Robert A. Gramm.

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

    Date of amendment request: January 20, 2005.
    Description of amendment request: The proposed amendments would 
change Technical Specification (TS) 3/4.8.2.1, ``DC Sources--
Operating,'' and TS 3/4.8.2.2, ``DC Sources--Shutdown,'' with addition 
of a new TS 3/4.8.2.3, ``Battery Parameters'', to incorporate actions 
for responding to ``out-of-limit'' conditions, and surveillances for 
verification of battery parameters. The proposed changes would revise 
allowed outage times for battery chargers as well as battery charger 
testing criteria. The proposed changes would also relocate a number of 
battery surveillance requirements to a licensee-controlled Battery 
Monitoring and Maintenance Program.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Does the proposed change involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    Response: No.
    The proposed change rearranges the Technical Specifications for 
the direct current [DC] electrical power system, and adds new 
Conditions and required actions with revised completion times to 
allow for battery charger inoperability. Neither the direct current 
electrical power subsystem nor associated battery chargers are 
initiators of an accident sequence previously evaluated. Performance 
of plant operational activities in accordance with the proposed 
Technical Specification changes ensures that the direct current 
electrical power subsystem is capable of performing its function as 
previously described. Therefore, the mitigating functions supported 
by the subject subsystem will continue to provide the protection 
assumed by the safety analysis.
    Relocation of preventive maintenance surveillances and certain 
operating limits and actions to a ``Battery Monitoring and 
Maintenance Program'' will not challenge the ability of the subject 
subsystem to perform its design function. Maintenance and monitoring 
currently required will continue to be performed. In addition, the 
direct current electrical power subsystem is within the scope of 10 
CFR 50.65, ``Requirements for monitoring the effectiveness of 
maintenance at nuclear power plants,'' which will ensure continued 
control of maintenance activities associated with the subject 
subsystem.
    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 accident from any accident previously evaluated?
    Response: No.
    The proposed change does not involve any physical alteration of 
the units. No new equipment is introduced, and installed equipment 
is not operated in a new or different manner. The proposed changes 
do not affect setpoints for initiation of protective or mitigating 
actions.
    Operability of the DC electrical power subsystems in accordance 
with the proposed technical specifications is consistent with the 
initial assumptions of the accident analyses and is based upon 
meeting the design basis of the plant.
    The proposed changes will not alter the manner in which 
equipment operation is initiated, nor will the functional demands on 
credited equipment be changed. No alteration in the operating 
procedures is proposed, and no change is being made to procedures 
relied upon in response to an off-normal event. No new failure modes 
are being introduced, and the proposed change does not alter 
assumptions made in the safety analyses.
    Therefore, the proposed change does not create the possibility 
of a new or different 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 will not adversely affect operation of plant 
equipment and will not result in a change to the setpoints at which 
protective actions are initiated. Sufficient DC capacity to support 
operation of mitigation equipment is ensured. The provisions of the 
Battery Monitoring and Maintenance Program will ensure that the 
station batteries are maintained in a highly reliable manner. The 
equipment fed by the DC electrical system will continue to provide 
adequate power to safety-related loads in accordance with analysis 
assumptions.
    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 standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
request for amendments involves no significant hazards consideration.

[[Page 9998]]

    Attorney for licensee: A. H. Gutterman, Esq., Morgan, Lewis & 
Bockius, 1111 Pennsylvania Avenue, NW., Washington, DC 20004.
    NRC Section Chief: Allen G. Howe.

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

    Date of amendment request: September 30, 2004.
    Brief description of amendments: The proposed amendment revises TS 
5.5.7, ``Reactor Coolant Pump [RCP] Flywheel Inspection Program,'' to 
extend the allowable inspection interval to 20 years.
    The NRC staff issued a notice of availability of a model safety 
evaluation and model no significant hazards consideration (NSHC) 
determination for referencing in license amendment applications in the 
Federal Register on October 22, 2003 (68 FR 60422). The licensee 
affirmed the applicability of the model NSHC determination in its 
application dated September 30, 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

    The proposed change to the RCP flywheel examination frequency 
does not change the response of the plant to any accidents. The RCP 
will remain highly reliable and the proposed change will not result 
in a significant increase in the risk of plant operation. Given the 
extremely low failure probabilities for the RCP motor flywheel 
during normal and accident conditions, the extremely low probability 
of a loss-of-coolant accident (LOCA) with loss of offsite power 
(LOOP), and assuming a conditional core damage probability (CCDP) of 
1.0 (complete failure of safety systems), the core damage frequency 
(CDF) and change in risk would still not exceed the NRC's acceptance 
guidelines contained in Regulatory Guide (RG) 1.174 (<1.0E-6 per 
year). Moreover, considering the uncertainties involved in this 
evaluation, the risk associated with the postulated failure of an 
RCP motor flywheel is significantly low. Even if all four RCP motor 
flywheels are considered in the bounding plant configuration case, 
the risk is still acceptably low.
    The proposed change does 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; alter or prevent the ability 
of structures, systems, components (SSCs) from performing their 
intended function to mitigate the consequences of an initiating 
event within the assumed acceptance limits; or affect the source 
term, containment isolation, or radiological release assumptions 
used in evaluating the radiological consequences of an accident 
previously evaluated. Further, the proposed change does not increase 
the type or amount of radioactive effluent that may be released 
offsite, nor significantly increase individual or cumulative 
occupational/public radiation exposure. The proposed change is 
consistent with the safety analysis assumptions and resultant 
consequences. Therefore, the proposed change does not involve a 
significant increase in the probability or consequences of an 
accident previously evaluated.

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

    The proposed change in flywheel inspection frequency does not 
involve any change in the design or operation of the RCP. Nor does 
the change to examination frequency affect any existing accident 
scenarios, or create any new or different accident scenarios. 
Further, the change does not involve a physical alteration of the 
plant (i.e., no new or different type of equipment will be 
installed) or alter the methods governing normal plant operation. In 
addition, the change does not impose any new or different 
requirements or eliminate any existing requirements, and does not 
alter any assumptions made in the safety analysis. The proposed 
change is consistent with the safety analysis assumptions and 
current plant operating practice. Therefore, the proposed change 
does not create the possibility of a new or different kind of 
accident from any accident previously evaluated.

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

    The proposed change does not alter the manner in which safety 
limits, limiting safety system settings, or limiting conditions for 
operation are determined. The safety analysis acceptance criteria 
are not impacted by this change. The proposed change will not result 
in plant operation in a configuration outside of the design basis. 
The calculated impact on risk is insignificant and meets the 
acceptance criteria contained in RG 1.174. There are no significant 
mechanisms for inservice degradation of the RCP flywheel. Therefore, 
the proposed change does not involve a significant reduction in a 
margin of safety.

    The NRC staff proposes to determine that the amendment request 
involves no significant hazards consideration.
    Attorney for licensee: George L. Edgar, Esq., Morgan, Lewis and 
Bockius, 1800 M Street, NW., Washington, DC 20036.
    NRC Section Chief: Allen G. Howe.

Virginia Electric and Power Company, Docket Nos. 50-338 and 50-339, 
North Anna Power Station, Units No. 1 and No. 2, Louisa County, 
Virginia

    Date of amendment request: December 17, 2004.
    Description of amendment request: The proposed changes to the 
Technical Specifications would increase the completion times from 72 
hours to 7 days for the following systems: Low-Head Safety Injection 
(LHSI) Emergency Core Cooling System (ECCS), Auxiliary Feedwater (AFW) 
System, Quench Spray (QS) System, and Chemical Addition System (CAS).
    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 license amendment involve a significant 
increase in the probability or consequences of an accident 
previously evaluated?
    The proposed changes do not alter any plant equipment or 
operating practices in such a manner that the probability of an 
accident is increased. The proposed changes will not alter 
assumptions relative to the mitigation of an accident or transient 
event.
    The CDF [core damage frequency] impact and the LERF [large early 
release frequency] impact, as well as the ICCDP [incremental 
conditional core damage probability] and ICLERP [incremental 
conditional large early release probability], associated with the 
proposed completion time changes meet the acceptance criteria in RG 
[Regulatory Guide] 1.174 and RG 1.177 for the proposed changes. The 
cumulative CDF and LERF impact for the proposed completion time 
changes also meet the acceptance criteria in RG 1.174 for the 
proposed changes.
    Therefore, the proposed changes do not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    2. Does the proposed license amendment create the possibility of 
a new or different kind of accident from any accident previously 
evaluated?
    The proposed changes do 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. 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 amendment involve a significant reduction 
in a margin of safety?
    The overall margin of safety is not decreased due to the 
increased completion times for the LHSI ECCS, QS including the CAS, 
and AFW since the systems design and operation are not altered by 
the proposed increase in completion times. The risk impacts of the 
changes are also consistent with the acceptance criteria in RG 1.174 
and RG 1.177.
    For the Chemical Addition System, which is not modeled in the 
PRA [probabilistic risk assessment] due to its limited capability to 
mitigate severe accidents, the proposed completion time change takes 
into account the ability of the spray systems to remove iodine at a 
reduced capability and the low probability of the worst case DBA 
[design-basis accident] occurring during this period.

[[Page 9999]]

    The codes and standards or their alternatives approved for use 
by the NRC continue to be met. In addition, the safety analysis 
acceptance criteria in the licensing basis (e.g., FSAR [final safety 
analysis report], supporting analyses) continue to be met.
    Therefore, the proposed change does not involve a significant 
reduction in the margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 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: John A. Nakoski.

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: December 17, 2004.
    Description of amendment request: The proposed Technical 
Specifications (TS) change would revise the reactor coolant system 
(RCS) pressure temperature (P/T) operating limits, the Low-Temperature 
Overpressure Protection System (LTOPS) setpoint, and the LTOPS enable 
temperature (Tenable) basis for cumulative core burnups up to 47.6 
effective full-power years (EFPY) and 48.1 EFPY for Surry Power 
Station, Units 1 and 2, respectively.
    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 change involve a significant increase in the 
probability or consequences of an accident previously evaluated?
    The proposed change modifies the Surry Units 1 and 2 RCS P/T 
limit curves, LTOPS setpoint, and LTOPS Tenable value and extends 
the cumulative core burnup applicability limits for these 
parameters. The allowable operating pressures and temperatures under 
the proposed RCS P/T limit curves are not significantly different 
from those allowed under the existing Technical Specification P/T 
limits. The revisions in the values for the LTOPS setpoint and LTOPS 
Tenable do not significantly change the plant operating space. No 
changes to plant systems, structures or components are proposed, and 
no new operating modes are established. The P/T limits, LTOPS 
setpoint, and Tenable value do not contribute to the probability of 
occurrence or consequences of accidents previously analyzed. The 
revised licensing basis analyses utilize acceptable analytical 
methods, and continue to demonstrate that established accident 
analysis acceptance criteria are met. Therefore, there is no 
increase in the probability or consequences of any accident 
previously evaluated.
    2. Does the change create the possibility of a new or different 
kind of accident from any accident previously evaluated?
    The proposed change modifies the Surry Units 1 and 2 RCS P/T 
limit curves, LTOPS setpoint, and LTOPS Tenable value and extends 
the cumulative core burnup applicability limits for these 
parameters. The allowable operating pressures and temperatures under 
the proposed RCS P/T limit curves are not significantly different 
from those allowed under the existing Technical Specification P/T 
limits. No changes to plant systems, structures or components are 
proposed, and no new operating modes are established. Therefore, the 
proposed changes do not create the possibility of any accident or 
malfunction of a different type previously evaluated.
    3. Does the change involve a significant reduction in the margin 
of safety?
    The proposed revised RCS P/T limit curves, LTOPS setpoint, and 
LTOPS Tenable value analysis bases do not involve a significant 
reduction in the margin of safety for these parameters. The proposed 
revised RCS P/T limit curves are valid to cumulative core burnups of 
47.6 EFPY and 48.1 EFPY for Surry Units 1 and 2, respectively. The 
proposed revised LTOPS setpoint and Tenable analyses support these 
same cumulative core burnup limits. The proposed revised RCS P/T 
limit curves utilize ASME [American Society of Mechanical Engineers] 
Code Section XI, which supports use of a conservative but less 
restrictive stress intensity formulation (K1c). The proposed 
extension of the cumulative core burnup applicability limits along 
with a small increase in the LTOPS PORV [power-operated relief 
valve] setpoint is accommodated by the margin provided by ASME Code 
Section XI. The analyses demonstrate that established analysis 
acceptance criteria continue to be met. Specifically, the proposed 
P/T limit curves, LTOPS setpoint and LTOPS Tenable value provide 
acceptable margin to vessel fracture under both normal operation and 
LTOPS design basis (mass addition and heat addition) accident 
conditions. Therefore, the proposed change does not result in a 
significant reduction in 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: John A. Nakoski.

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.

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

    Date of application for amendment: July 15, 2004.
    Brief description of amendment: The amendment added references to 
the list of approved core operating limits analytical methods in 
Technical Specification 5.6.5.b for Calvert Cliffs, Unit Nos. 1 and 2.
    Date of publication of individual notice in Federal Register: 
December 29, 2004 (69 FR 78056).
    Expiration date of individual notice: February 28, 2005.

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

    Date of amendment request: November 5, 2003, as supplemented by 
letter dated April 22, 2004.
    Brief description of amendment request: The proposed amendment 
would revise the Point Beach Nuclear Plant (PBNP), Units 1 and 2, 
Updated Final Safety Analysis Report to reflect the Commission staff's 
approval of the WCAP-14439-P, Revision 2 analysis entitled, ``Technical 
Justification for Eliminating Large Primary Loop Pipe Rupture as the 
Structural Design Basis for the Point Beach Nuclear Plant Units 1 and 2 
for the Power Uprate and License Renewal Program.''

[[Page 10000]]

    Date of publication of individual notice in Federal Register: 
February 7, 2005 (70 FR 6466).
    Expiration date of individual notice: April 8, 2005.

Notice of Issuance of Amendments to Facility Operating Licenses

    During the period since publication of the last biweekly notice, 
the Commission has issued the following amendments. The Commission has 
determined for each of these amendments that the application complies 
with the standards and requirements of the Atomic Energy Act of 1954, 
as amended (the Act), and the Commission's rules and regulations. The 
Commission has made appropriate findings as required by the Act and the 
Commission's rules and regulations in 10 CFR Chapter I, which are set 
forth in the license amendment.
    Notice of Consideration of Issuance of Amendment to Facility 
Operating License, Proposed No Significant Hazards Consideration 
Determination, and Opportunity for A Hearing in connection with these 
actions was published in the Federal Register as indicated.
    Unless otherwise indicated, the Commission has determined that 
these amendments satisfy the criteria for categorical exclusion in 
accordance with 10 CFR 51.22. Therefore, pursuant to 10 CFR 51.22(b), 
no environmental impact statement or environmental assessment need be 
prepared for these amendments. If the Commission has prepared an 
environmental assessment under the special circumstances provision in 
10 CFR 51.12(b) and has made a determination based on that assessment, 
it is so indicated.
    For further details with respect to the action see (1) the 
applications for amendment, (2) the amendment, and (3) the Commission's 
related letter, Safety Evaluation and/or Environmental Assessment as 
indicated. All of these items are available for public inspection at 
the Commission's Public Document Room (PDR), located at One White Flint 
North, Public File Area 01F21, 11555 Rockville Pike (first floor), 
Rockville, Maryland. Publicly available records will be accessible from 
the Agencywide Documents Access and Management Systems (ADAMS) Public 
Electronic Reading Room on the Internet at the NRC Web site, http://www.nrc.gov/reading-rm/adams.html. If you do not have access to ADAMS 
or if there are problems in accessing the documents located in ADAMS, 
contact the PDR Reference staff at 1 (800) 397-4209, (301) 415-4737 or 
by e-mail to [email protected].

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

    Date of application for amendment: June 22, 2004.
    Brief description of amendment: The proposed amendment revises 
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.
    Date of issuance: February 10, 2005.
    Effective date: As of the date of issuance and shall be implemented 
within 60 days.
    Amendment No.: 162.
    Facility Operating License No. NPF-62: The amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: August 31, 2004 (68 FR 
53099).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated February 10, 2005.
    No significant hazards consideration comments received: No.

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

    Date of application for amendments: December 9, 2003, as 
supplemented May 19 and August 3, 2004.
    Brief description of amendments: The amendments revise Technical 
Specification 3.7.1, ``Main Steam Safety Valves (MSSVs),'' to increase 
the maximum allowable lift setting on two MSSVs on each unit. In 
addition, the amendments increase the completion time for reducing the 
Power Level-High Trip setpoint.
    Date of issuance: February 10, 2005.
    Effective date: As of the date of issuance to be implemented within 
30 days.
    Amendment Nos.: 270 and 247.
    Renewed Facility Operating License Nos. DPR-53 and DPR-69: 
Amendments revised the Technical Specifications.
    Date of initial notice in Federal Register: October 26, 2004 (69 FR 
62470).
    The supplemental letters dated May 19 and August 3, 2004, provided 
additional information that clarified the application, did not expand 
the scope of the application as originally noticed, and did not change 
the staff's original proposed no significant hazards consideration 
determination.
    The Commission's related evaluation of these amendments is 
contained in a Safety Evaluation dated February 10, 2005.
    No significant hazards consideration comments received: No.

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

    Date of application for amendment: August 19, 2004, as supplemented 
December 2, 2004.
    Brief description of amendment: The amendment revises the reactor 
coolant system pressure and temperature limits by replacing Technical 
Specification Section 3.4.3, ``RCS Pressure and Temperature (P/T) 
Limits,'' Figures 3.4.3-1 and 3.4.3-2, with figures that are applicable 
up to 35 effective full-power years.
    Date of issuance: February 7, 2005.
    Effective date: February 7, 2005.
    Amendment No.: 202.
    Renewed Facility Operating License No. DPR-23: Amendment revises 
the Technical Specifications.
    Date of initial notice in Federal Register: September 28, 2004 (69 
FR 57981). The December 2, 2004, supplement contained clarifying 
information only that did not change the initial proposed no 
significant hazards consideration determination or expand the scope of 
the initial application.
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated February 7, 2005.
    No significant hazards consideration comments received: No.

Entergy Nuclear Operations, Inc., Docket No. 50-333, James A. 
FitzPatrick Nuclear Power Plant, Oswego County, New York

    Date of application for amendment: June 4, 2004, as supplemented on 
July 27, September 27, and December 14, 2004.
    Brief description of amendment: The amendment revises the safety 
limit values in Technical Specifications 2.1.1.2 for the minimum 
critical power ratio for both single and two recirculation loop 
operation.
    Date of issuance: February 3, 2005.
    Effective date: As of the date of issuance to be implemented within 
30 days.
    Amendment No.: 281.
    Facility Operating License No. DPR-59: Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: July 20, 2004 (69 FR 
43459).
    The July 27, September 27, and December 14, 2004, letters provided 
information that clarified the

[[Page 10001]]

application, did not expand the scope of the application as originally 
noticed, and did not change the staff's original proposed no 
significant hazards consideration determination as published in the 
Federal Register.
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated February 3, 2005.
    No significant hazards consideration comments received: No.

Entergy Nuclear Operations, Inc., Docket No. 50-333, James A. 
FitzPatrick Nuclear Power Plant, Oswego County, New York

    Date of application for amendment: September 1, 2004.
    Brief description of amendment: The amendment eliminates the 
Technical Specification requirements to submit monthly operating 
reports and annual occupational radiation exposure reports.
    Date of issuance: February 3, 2005.
    Effective date: As of the date of issuance to be implemented within 
60 days.
    Amendment No.: 282.
    Facility Operating License No. DPR-59: Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: September 28, 2004 (69 
FR 57984).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated February 3, 2005.
    No significant hazards consideration comments received: No.

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

    Date of application for amendment: August 19, 2003, as supplemented 
on March 12, 2004.
    Brief description of amendment: The amendment revised Pilgrim 
Nuclear Power Station (Pilgrim) Technical Specification (TS) Table 
3.2.C-1 by changing the rod block monitor (RBM) low power setpoint 
(LPSP) allowable value from 29% to 25.9%. The amendment corrected the 
RBM LPSP (currently <=29%) that was incorrectly inserted into Note 5 
for TS Table 3.2.C-1 under License Amendment No. 138, dated July 1, 
1991. Pilgrim plant procedures and the Core Operating Limits Report 
have enforced the correct setpoint value of <=25.9% since issuance of 
License Amendment No. 138.
    Date of issuance: February 2, 2005.
    Effective date: As of the date of issuance, and shall be 
implemented within 60 days.
    Amendment No.: 210.
    Facility Operating License No. DPR-35: The amendment revised the 
TSs.
    Date of initial notice in Federal Register: February 17, 2004 (69 
FR 7521).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated February 2, 2005.
    No significant hazards consideration comments received: No.

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

    Date of application for amendment: April 5, 2004, as supplemented 
by letters dated June 22 and December 6, 2004.
    Brief description of amendment: This amendment modifies the 
existing minimum critical power ratio (MCPR) safety limit contained in 
Technical Specification 2.1.1.2. Specifically, the change modifies the 
MCPR safety limit values, as calculated by Global Nuclear Fuel (GNF), 
by decreasing the limit for two recirculation loop operation from 1.10 
to 1.08, and decreasing the limit for single recirculation loop 
operation from 1.11 to 1.10.
    Date of issuance: February 3, 2005.
    Effective date: As of the date of issuance and shall be implemented 
within 90 days.
    Amendment No.: 132.
    Facility Operating License No. NPF-58: This amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: May 11, 2004 (69 FR 
26189).
    The supplements dated June 22 and December 6, 2004, provided 
clarifying information that did not change the scope of the April 5, 
2004, application nor the initial proposed no significant hazards 
consideration determination.
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated February 3, 2005.
    No significant hazards consideration comments received: No.

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

    Date of application for amendment: March 31, 2004.
    Brief description of amendment: This amendment modified the 
technical specification (TS) surveillance requirements (SRs) for manual 
actuation of certain main steam safety/relief valves (S/RVs), including 
those valves that provide an automatic depressurization system (ADS) 
and low-low set (LLS) valve function. The specific TS changes revised 
SR 3.4.4.3 for S/RVs, SR 3.5.1.7 for ADS valves, and SR 3.6.1.6.1 for 
LLS valves. The changes removed the requirement for the S/RV disks to 
be lifted from their seats when manually actuated.
    The revised SRs specify that the actuator is to stroke when 
manually actuated, without physically lifting the disks off their seats 
at power.
    Date of issuance: February 10, 2005.
    Effective date: As of the date of issuance and shall be implemented 
within 30 days.
    Amendment No.: 133.
    Facility Operating License No. NPF-58: This amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: May 11, 2004 (69 FR 
26188).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated February 10, 2005.
    No significant hazards consideration comments received: No.

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

    Date of application for amendment: October 14, 2004.
    Brief description of amendment: The amendment corrects errors in 
Technical Specifications 3.10.i and 6.9.a.4.A.
    Date of issuance: February 15, 2005.
    Effective date: As of the date of issuance and shall be implemented 
within 60 days.
    Amendment No.: 180.
    Facility Operating License No. DPR-43: Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: December 7, 2004 (69 FR 
70720).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated February 15, 2005.
    No significant hazards consideration comments received: No.

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

    Date of application for amendments: October 5, 2004.
    Brief description of amendments: The amendments deleted technical 
specification (TS) 5.6.1, ``Occupational Radiation Exposure Reports,'' 
and TS 5.6.3, ``Monthly Operating Reports,'' as described in the Notice 
of Availability

[[Page 10002]]

published in the Federal Register on June 23, 2004 (69 FR 35067).
    Date of issuance: February 7, 2005.
    Effective date: As of the date of issuance and shall be implemented 
within 90 days.
    Amendment Nos.: 216, 221.
    Facility Operating License Nos. DPR-24 and DPR-27: Amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: November 9, 2004 (69 FR 
64989).
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated February 7, 2005.
    No significant hazards consideration comments received: No.

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

    Date of application for amendment: May 21, 2004.
    Brief description of amendment: This amendment deletes the 
Technical Specification requirements associated with hydrogen 
recombiners and hydrogen monitors.
    Date of issuance: February 3, 2005.
    Effective date: As of the date of issuance and shall be implemented 
within 60 days.
    Amendment No.: 170.
    Renewed Facility Operating License No. NPF-12: Amendment revises 
the Technical Specifications.
    Date of initial notice in Federal Register: September 28, 2004 (69 
FR 57990).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated February 3, 2005.
    No significant hazards consideration comments received: No.

Tennessee Valley Authority, Docket Nos. 50-259, 50-260, and 50-296, 
Browns Ferry Nuclear Plant, Units 1, 2, and 3, Limestone County, 
Alabama

    Date of application for amendments: July 2, 2004.
    Description of amendment request: The amendments eliminated the 
requirements for the licensee to submit monthly operating reports and 
occupational radiation exposure reports.
    Date of issuance: January 25, 2005.
    Effective date: Date of issuance, to be implemented within 60 days.
    Amendment Nos.: 252, 291 and 250.
    Facility Operating License Nos. DPR-33, DPR-52, and DPR-68. 
Amendments revised the Technical Specifications.
    Date of initial notice in Federal Register: October 12, 2004 (69 FR 
60687).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated January 25, 2005.
    No significant hazards consideration comments received: No.

Tennessee Valley Authority, Docket Nos. 50-259, 50-260, and 50-296, 
Browns Ferry Nuclear Plant, Units 1, 2, and 3, Limestone County, 
Alabama

    Date of application for amendments: July 8, 2004.
    Description of amendment request: The amendments revised Technical 
Specifications by eliminating the requirements associated with hydrogen 
monitors.
    Date of issuance: February 14, 2005.
    Effective date: Date of issuance, to be implemented within 60 days.
    Amendment Nos.: 253, 292 and 251.
    Facility Operating License Nos. DPR-33, DPR-52, and DPR-68. 
Amendments revised the Technical Specifications.
    Date of initial notice in Federal Register: September 14, 2004 (69 
FR 55473).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated February 14, 2005.
    No significant hazards consideration comments received: No.

Tennessee Valley Authority, Docket Nos. 50-327 and 50-328, Sequoyah 
Nuclear Plant, Units 1 and 2, Hamilton County, Tennessee

    Date of application for amendments: March 3, 2004.
    Brief description of amendments: The amendments revised the Updated 
Final Safety Analysis Report (UFSAR) by modifying the licensing basis 
for the seismic qualification of round flexible ducting, triangular 
ducting, and associated air bars installed as part of the suspended 
ceiling air delivery system in the main control room.
    Date of issuance: January 31, 2005.
    Effective date: As of the date of issuance and shall be implemented 
as part of the next UFSAR update made in accordance with 10 CFR 
50.71(e).
    Amendment Nos.: 298 and 287.
    Facility Operating License Nos. DPR-77 and DPR-79: Amendments 
revised the UFSAR.
    Date of initial notice in Federal Register: April 27, 2004 (69 FR 
22883).
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated January 31, 2005.
    No significant hazards consideration comments received: No.

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

    Date of application for amendment: July 8, 2004.
    Brief description of amendment: The amendment revises Technical 
Specification (TS) Section 3.8.4, ``DC Sources-Operating.'' 
Specifically, the amendment removes the term ``inter-rack'' and 
associated wording from TS Surveillance Requirements 3.8.4.6 and 
3.8.4.10 for the 125 Volt Direct Current electrical power subsystems of 
the emergency diesel generators.
    Date of issuance: February 7, 2005.
    Effective date: As of the date of issuance and shall be implemented 
within 60 days of issuance.
    Amendment No.: 54.
    Facility Operating License No. NPF-90: Amendment revises the 
Technical Specifications.
    Date of initial notice in Federal Register: August 3, 2004 (69 FR 
46593).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated February 7, 2005.
    No significant hazards consideration comments received: No.

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

    Date of amendment request: January 21, 2004, as supplemented by 
letters dated November 18 and December 3, 2004.
    Brief description of amendments: The amendments revise Technical 
Specifications (TSs) 3.3.1, ``Reactor Trip System (RTS) 
Instrumentation,'' 3.3.2, ``Engineered Safety Feature Actuation System 
(ESFAS) Instrumentation,'' and 3.3.6, ``Containment Ventilation 
Isolation Instrumentation,'' to adopt the completion time, test bypass 
time, and surveillance frequency time changes approved by the NRC in 
Topical Reports WCAP-14333-P-A, ``Probabilistic Risk Analysis of the 
RPS [reactor protection system] and ESFAS Test Times and Completion 
Times,'' and WCAP-15376-P-A, ``Risk-Informed Assessment of the RTS and 
ESFAS Surveillance Test Intervals and Reactor Trip Breaker Test and 
Completion Times.'' The amendments revise the required actions for 
certain action conditions; increase the completion times for several 
required actions (including some notes); delete notes in certain 
required actions; and increase frequency time intervals (including 
certain notes) in several surveillance requirements.
    Date of issuance: January 31, 2005.

[[Page 10003]]

    Effective date: As of the date of issuance and shall be implemented 
within 90 days from the date of issuance.
    Amendment Nos.: 114, 114.
    Facility Operating License Nos. NPF-87 and NPF-89: The amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: March 2, 2004 (69 FR 
9866).
    The supplemental letters dated November 18 and December 3, 2004, 
provided clarifying information that did not change the scope of the 
original application as noticed or the NRC staff's original proposed no 
significant hazards consideration determination.
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated January 31, 2005.
    No significant hazards consideration comments received: No.

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

    Date of amendment request: July 23, 2004.
    Brief description of amendment: The amendment eliminates the 
requirements in the technical specifications associated with hydrogen 
recombiners and hydrogen monitors.
    Date of issuance: January 31, 2005.
    Effective date: January 31, 2005, and shall be implemented within 
90 days from the date of issuance.
    Amendment No.: 157.
    Facility Operating License No. NPF-42. The amendment revises the 
Technical Specifications.
    Date of initial notice in Federal Register: August 31, 2004 (69 FR 
53115).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated January 31, 2005.
    No significant hazards consideration comments received: No.

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

    Date of amendment request: July 23, 2004.
    Brief description of amendment: The amendment revises the technical 
specifications by eliminating the requirements to provide the NRC 
monthly operating reports and annual occupational radiation exposure 
reports.
    Date of issuance: January 31, 2005.
    Effective date: January 31, 2005, and shall be implemented within 
90 days from the date of issuance.
    Amendment No.: 158.
    Facility Operating License No. NPF-42. The amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: August 31, 2004 (69 FR 
53116).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated January 31, 2005.
    No significant hazards consideration comments received: No.

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

    During the period since publication of the last biweekly notice, 
the Commission has issued the following amendments. The Commission has 
determined for each of these amendments that the application for the 
amendment complies with the standards and requirements of the Atomic 
Energy Act of 1954, as amended (the Act), and the Commission's rules 
and regulations. The Commission has made appropriate findings as 
required by the Act and the Commission's rules and regulations in 10 
CFR Chapter I, which are set forth in the license amendment.
    Because of exigent or emergency circumstances associated with the 
date the amendment was needed, there was not time for the Commission to 
publish, for public comment before issuance, its usual Notice of 
Consideration of Issuance of Amendment, Proposed No Significant Hazards 
Consideration Determination, and Opportunity for a Hearing.
    For exigent circumstances, the Commission has either issued a 
Federal Register notice providing opportunity for public comment or has 
used local media to provide notice to the public in the area 
surrounding a licensee's facility of the licensee's application and of 
the Commission's proposed determination of no significant hazards 
consideration. The Commission has provided a reasonable opportunity for 
the public to comment, using its best efforts to make available to the 
public means of communication for the public to respond quickly, and in 
the case of telephone comments, the comments have been recorded or 
transcribed as appropriate and the licensee has been informed of the 
public comments.
    In circumstances where failure to act in a timely way would have 
resulted, for example, in derating or shutdown of a nuclear power plant 
or in prevention of either resumption of operation or of increase in 
power output up to the plant's licensed power level, the Commission may 
not have had an opportunity to provide for public comment on its no 
significant hazards consideration determination. In such case, the 
license amendment has been issued without opportunity for comment. If 
there has been some time for public comment but less than 30 days, the 
Commission may provide an opportunity for public comment. If comments 
have been requested, it is so stated. In either event, the State has 
been consulted by telephone whenever possible.
    Under its regulations, the Commission may issue and make an 
amendment immediately effective, notwithstanding the pendency before it 
of a request for a hearing from any person, in advance of the holding 
and completion of any required hearing, where it has determined that no 
significant hazards consideration is involved.
    The Commission has applied the standards of 10 CFR 50.92 and has 
made a final determination that the amendment involves no significant 
hazards consideration. The basis for this determination is contained in 
the documents related to this action. Accordingly, the amendments have 
been issued and made effective as indicated.
    Unless otherwise indicated, the Commission has determined that 
these amendments satisfy the criteria for categorical exclusion in 
accordance with 10 CFR 51.22. Therefore, pursuant to 10 CFR 51.22(b), 
no environmental impact statement or environmental assessment need be 
prepared for these amendments. If the Commission has prepared an 
environmental assessment under the special circumstances provision in 
10 CFR 51.12(b) and has made a determination based on that assessment, 
it is so indicated.
    For further details with respect to the action see (1) the 
application for amendment, (2) the amendment to Facility Operating 
License, and (3) the Commission's related letter, Safety Evaluation 
and/or Environmental Assessment, as indicated. All of these items are 
available for public inspection at the Commission's Public Document 
Room (PDR), located at One White Flint North, Public File Area 01F21, 
11555 Rockville Pike (first floor), Rockville, Maryland. Publicly 
available records will be accessible from the Agencywide Documents 
Access and Management System's (ADAMS) Public Electronic Reading Room 
on the Internet at the NRC Web site, http://www.nrc.gov/

[[Page 10004]]

reading-rm/adams.html. If you do not have access to ADAMS or if there 
are problems in accessing the documents located in ADAMS, contact the 
PDR Reference staff at 1 (800) 397-4209, (301) 415-4737 or by e-mail to 
[email protected].
    The Commission is also offering an opportunity for a hearing with 
respect to the issuance of the amendment. Within 60 days after the date 
of publication of this notice, the licensee may file a request for a 
hearing with respect to issuance of the amendment to the subject 
facility operating license and any person whose interest may be 
affected by this proceeding and who wishes to participate as a party in 
the proceeding must file a written request for a hearing and a petition 
for leave to intervene. Requests for a hearing and a petition for leave 
to intervene shall be filed in accordance with the Commission's ``Rules 
of Practice for Domestic Licensing Proceedings'' in 10 CFR Part 2. 
Interested persons should consult a current copy of 10 CFR 2.309, which 
is available at the Commission's PDR, located at One White Flint North, 
Public File Area 01F21, 11555 Rockville Pike (first floor), Rockville, 
Maryland, and electronically on the Internet at the NRC Web site, 
http://www.nrc.gov/reading-rm/doc-collections/cfr/. If there are 
problems in accessing the document, contact the PDR Reference staff at 
1 (800) 397-4209, (301) 415-4737, or by e-mail to [email protected]. If a 
request for a hearing or petition for leave to intervene is filed by 
the above date, the Commission or a presiding officer designated by the 
Commission or by the Chief Administrative Judge of the Atomic Safety 
and Licensing Board Panel, will rule on the request and/or petition; 
and the Secretary or the Chief Administrative Judge of the Atomic 
Safety and Licensing Board will issue a notice of a hearing or an 
appropriate order.
    As required by 10 CFR 2.309, a petition for leave to intervene 
shall set forth with particularity the interest of the petitioner in 
the proceeding, and how that interest may be affected by the results of 
the proceeding. The petition should specifically explain the reasons 
why intervention should be permitted with particular reference to the 
following general requirements: (1) The name, address, and telephone 
number of the requestor or petitioner; (2) the nature of the 
requestor's/petitioner's right under the Act to be made a party to the 
proceeding; (3) the nature and extent of the requestor's/petitioner's 
property, financial, or other interest in the proceeding; and (4) the 
possible effect of any decision or order which may be entered in the 
proceeding on the requestor's/petitioner's interest. The petition must 
also identify the specific contentions which the petitioner/requestor 
seeks to have litigated at the proceeding.
    Each contention must consist of a specific statement of the issue 
of law or fact to be raised or controverted. In addition, the 
petitioner/requestor shall provide a brief explanation of the bases for 
the contention and a concise statement of the alleged facts or expert 
opinion which support the contention and on which the petitioner 
intends to rely in proving the contention at the hearing. The 
petitioner must also provide references to those specific sources and 
documents of which the petitioner is aware and on which the petitioner 
intends to rely to establish those facts or expert opinion. The 
petition must include sufficient information to show that a genuine 
dispute exists with the applicant on a material issue of law or 
fact.\1\ Contentions shall be limited to matters within the scope of 
the amendment under consideration. The contention must be one which, if 
proven, would entitle the petitioner to relief. A petitioner/requestor 
who fails to satisfy these requirements with respect to at least one 
contention will not be permitted to participate as a party.
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    \1\ To the extent that the applications contain attachments and 
supporting documents that are not publicly available because they 
are asserted to contain safeguards or proprietary information, 
petitioners desiring access to this information should contact the 
applicant or applicant's counsel and discuss the need for a 
protective order.
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    Each contention shall be given a separate numeric or alpha 
designation within one of the following groups:
    1. Technical--primarily concerns/issues relating to technical and/
or health and safety matters discussed or referenced in the 
applications.
    2. Environmental--primarily concerns/issues relating to matters 
discussed or referenced in the environmental analysis for the 
applications.
    3. Miscellaneous--does not fall into one of the categories outlined 
above.
    As specified in 10 CFR 2.309, if two or more petitioners/requestors 
seek to co-sponsor a contention, the petitioners/requestors shall 
jointly designate a representative who shall have the authority to act 
for the petitioners/requestors with respect to that contention. If a 
petitioner/requestor seeks to adopt the contention of another 
sponsoring petitioner/requestor, the petitioner/requestor who seeks to 
adopt the contention must either agree that the sponsoring petitioner/
requestor shall act as the representative with respect to that 
contention, or jointly designate with the sponsoring petitioner/
requestor a representative who shall have the authority to act for the 
petitioners/requestors with respect to that contention.
    Those permitted to intervene become parties to the proceeding, 
subject to any limitations in the order granting leave to intervene, 
and have the opportunity to participate fully in the conduct of the 
hearing. Since the Commission has made a final determination that the 
amendment involves no significant hazards consideration, if a hearing 
is requested, it will not stay the effectiveness of the amendment. Any 
hearing held would take place while the amendment is in effect.
    A request for a hearing or a petition for leave to intervene must 
be filed by: (1) First class mail addressed to the Office of the 
Secretary of the Commission, U.S. Nuclear Regulatory Commission, 
Washington, DC 20555-0001, Attention: Rulemaking and Adjudications 
Staff; (2) courier, express mail, and expedited delivery services: 
Office of the Secretary, Sixteenth Floor, One White Flint North, 11555 
Rockville Pike, Rockville, Maryland 20852, Attention: Rulemaking and 
Adjudications Staff; (3) e-mail addressed to the Office of the 
Secretary, U.S. Nuclear Regulatory Commission, [email protected]; 
or (4) facsimile transmission addressed to the Office of the Secretary, 
U.S. Nuclear Regulatory Commission, Washington, DC, Attention: 
Rulemakings and Adjudications Staff at (301) 415-1101, verification 
number is (301) 415-1966. A copy of the request for hearing and 
petition for leave to intervene should also be sent to the Office of 
the General Counsel, U.S. Nuclear Regulatory Commission, Washington, DC 
20555-0001, and it is requested that copies be transmitted either by 
means of facsimile transmission to (301) 415-3725 or by e-mail to 
[email protected]. A copy of the request for hearing and petition 
for leave to intervene should also be sent to the attorney for the 
licensee.
    Nontimely requests and/or petitions and contentions will not be 
entertained absent a determination by the Commission or the presiding 
officer or the Atomic Safety and Licensing Board that the petition, 
request and/or the contentions should be granted based on a balancing 
of the factors specified in 10 CFR 2.309(a)(1)(i)-(viii).

[[Page 10005]]

Duke Energy Corporation, et al., Docket No. 50-414, Catawba Nuclear 
Station Unit 2, York County, South Carolina

    Date of amendment request: February 5, 2005, as supplemented by 
letter dated February 7, 2005.
    Description of amendment request: The amendment revises the system 
bypass leakage acceptance criterion for the charcoal adsorber in the 2B 
Auxiliary Building Filtered Ventilation Exhaust System train as listed 
in Technical Specification 5.5.11, ``Ventilation Filter Testing 
Program.''
    Date of issuance: February 7, 2005.
    Effective date: As of the date of issuance and shall be implemented 
within 30 days from the date of issuance.
    Amendment No.: 213.
    Renewed Facility Operating License No. NPF-52: Amendments revised 
the Technical Specifications.
    Public comments requested as to proposed no significant hazards 
consideration (NSHC): No.
    The Commission's related evaluation of the amendment, finding of 
emergency circumstances, state consultation, and final NSHC 
determination are contained in a safety evaluation dated February 7, 
2005.
    Attorney for licensee: Ms. Anne Cottingham, Esquire.
    NRC Section Chief: John A. Nakoski.

    Dated in Rockville, Maryland, this 17th day of February 2005.

    For the Nuclear Regulatory Commission.
Ledyard B. Marsh,
Director, Division of Licensing Project Management, Office of Nuclear 
Reactor Regulation.
[FR Doc. 05-3627 Filed 2-28-05; 8:45 am]
BILLING CODE 7590-01-P