Biweekly Notice; Applications and Amendments to Facility Operating Licenses Involving No Significant Hazards Considerations, 9986-10005 [05-3627]

Agencies

• NUCLEAR REGULATORY COMMISSION

[Federal Register Volume 70, Number 39 (Tuesday, March 1, 2005)]
[Notices]
[Pages 9986-10005]
From the Federal Register Online via the Government Printing 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, https://www.nrc.gov/
reading-rm/doc-collections/cfr/. If a request for a hearing or petition 
for leave to intervene is filed within 60 days, the Commission or a 
presiding officer designated by the Commission or by the Chief 
Administrative Judge of the Atomic Safety and Licensing Board Panel, 
will rule on the request and/or petition; and the Secretary or the 
Chief Administrative Judge of the Atomic Safety and Licensing Board 
will issue a notice of a hearing or an appropriate order.
    As required by 10 CFR 2.309, a petition for leave to intervene 
shall set forth with particularity the interest of the petitioner in 
the proceeding, and how that interest may be affected by the results of 
the proceeding. The petition should specifically explain the reasons 
why intervention should be permitted with particular reference to the 
following general requirements: (1) The name, address, and telephone 
number of the requestor or petitioner; (2) the nature of the 
requestor's/petitioner's right under the Act to be made a party to the 
proceeding; (3) the nature and extent of the requestor's/petitioner's 
property, financial, or other interest in the proceeding; and (4) the 
possible effect of any decision or order which may be entered in the 
proceeding on the requestor's/petitioner's interest. The petition must 
also set forth the specific contentions which the petitioner/requestor 
seeks to have litigated at the proceeding.
    Each contention must consist of a specific statement of the issue 
of law or fact to be raised or controverted. In addition, the 
petitioner/requestor shall provide a brief explanation of the bases for 
the contention and a concise statement of the alleged facts or expert 
opinion which support the contention and on which the petitioner/
requestor intends to rely in proving the contention at the hearing. The 
petitioner/requestor must also provide references to those specific 
sources and documents of which the petitioner is aware and on which the 
petitioner/requestor intends to rely to establish those facts or expert 
opinion. The petition must include sufficient information to show that 
a genuine dispute exists with the applicant on a material issue of law 
or fact. Contentions shall be limited to matters within the scope of 
the amendment under consideration. The contention must be one which, if 
proven, would entitle the petitioner/requestor to relief. A petitioner/
requestor who fails to satisfy these requirements with respect to at 
least one contention will not be permitted to participate as a party.
    Those permitted to intervene become parties to the proceeding, 
subject to any limitations in the order granting leave to intervene, 
and have the opportunity to participate fully in the conduct of the 
hearing.
    If a hearing is requested, and the Commission has not made a final 
determination on the issue of no significant hazards consideration, the 
Commission will make a final determination on the issue of no 
significant hazards consideration. The final determination will serve 
to decide when the hearing is held. If the final determination is that 
the amendment request involves no significant hazards consideration, 
the Commission may issue the amendment and make it immediately 
effective, notwithstanding the request for a hearing. Any hearing held 
would take place after issuance of the amendment. If the final 
determination is that the amendment request involves a significant 
hazards consideration, any hearing held would take place before the 
issuance of any amendment.
    A request for a hearing or a petition for leave to intervene must 
be filed by: (1) First class mail addressed to the Office of the 
Secretary of the Commission, U.S. Nuclear Regulatory Commission, 
Washington, DC 20555-0001, Attention: Rulemaking and Adjudications 
Staff; (2) courier, express mail, and expedited delivery services: 
Office of the Secretary, Sixteenth Floor, One White Flint North, 11555 
Rockville Pike, Rockville, Maryland, 20852, Attention: Rulemaking and 
Adjudications Staff; (3) e-mail addressed to the Office of the 
Secretary, U.S. Nuclear Regulatory Commission, HearingDocket@nrc.gov; 
or (4) facsimile transmission addressed to the Office of the Secretary, 
U.S. Nuclear Regulatory Commission, Washington, DC, Attention: 
Rulemakings and Adjudications Staff at (301) 415-1101, verification 
number is (301) 415-1966. A copy of the request for hearing and 
petition for leave to intervene should also be sent to the Office of 
the General Counsel, U.S. Nuclear Regulatory Commission, Washington, DC 
20555-0001, and it is requested that copies be transmitted either by 
means of facsimile transmission to (301) 415-3725 or by e-mail to 
OGCMailCenter@nrc.gov. A copy of the request for hearing and petition 
for leave to intervene should also be sent to the attorney for the 
licensee.
    Nontimely requests and/or petitions and contentions will not be 
entertained absent a determination by the Commission or the presiding 
officer of the Atomic Safety and Licensing Board that the petition, 
request and/or the contentions should be granted based on a balancing 
of the factors specified in 10 CFR 2.309(a)(1)(I)-(viii).
    For further details with respect to this action, see the 
application for amendment which is available for public inspection at 
the Commission's PDR, located at One White Flint North, Public File 
Area 01F21, 11555 Rockville Pike (first floor), Rockville, Maryland. 
Publicly available records will be accessible from the ADAMS Public 
Electronic Reading Room on the Internet at the NRC Web site, https://
www.nrc.gov/reading-rm/adams.html. If you do not have access to ADAMS 
or if there are problems in accessing the documents located in ADAMS, 
contact the PDR Reference staff at 1 (800) 397-4209, (301) 415-4737 or 
by e-mail to pdr@nrc.gov.

AmerGen Energy Company, LLC, 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 even