Biweekly Notice; Application and Amendments to Facility Operating Licenses Involving No Significant Hazards Considerations, 38712-38729 [05-12987]

Agencies

• NUCLEAR REGULATORY COMMISSION

[Federal Register Volume 70, Number 127 (Tuesday, July 5, 2005)]
[Notices]
[Pages 38712-38729]
From the Federal Register Online via the Government Printing Office [www.gpo.gov]
[FR Doc No: 05-12987]


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NUCLEAR REGULATORY COMMISSION


Biweekly Notice; Application 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 June 10, 2005 to June 23, 2005. The last 
biweekly notice was published on June 21, 2005 (70 FR 35735).

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.

[[Page 38713]]

    The Commission is seeking public comments on this proposed 
determination. Any comments received within 30 days after the date of 
publication of this notice will be considered in making any final 
determination. Within 60 days after the date of publication of this 
notice, the licensee may file a request for a hearing with respect to 
issuance of the amendment to the subject facility operating license and 
any person whose interest may be affected by this proceeding and who 
wishes to participate as a party in the proceeding must file a written 
request for a hearing and a petition for leave to intervene.
    Normally, the Commission will not issue the amendment until the 
expiration of 60 days after the date of publication of this notice. The 
Commission may issue the license amendment before expiration of the 60-
day period provided that its final determination is that the amendment 
involves no significant hazards consideration. In addition, the 
Commission may issue the amendment prior to the expiration of the 30-
day comment period should circumstances change during the 30-day 
comment period such that failure to act in a timely way would result, 
for example in derating or shutdown of the facility. Should the 
Commission take action prior to the expiration of either the comment 
period or the notice period, it will publish in the Federal Register a 
notice of issuance. Should the Commission make a final No Significant 
Hazards Consideration Determination, any hearing will take place after 
issuance. The Commission expects that the need to take this action will 
occur very infrequently.
    Written comments may be submitted by mail to the Chief, Rules and 
Directives Branch, Division of Administrative Services, Office of 
Administration, U.S. Nuclear Regulatory Commission, Washington, DC 
20555-0001, and should cite the publication date and page number of 
this Federal Register notice. Written comments may also be delivered to 
Room 6D22, Two White Flint North, 11545 Rockville Pike, Rockville, 
Maryland, from 7:30 a.m. to 4:15 p.m. Federal workdays. Copies of 
written comments received may be examined at the Commission's Public 
Document Room (PDR), located at One White Flint North, Public File Area 
O1F21, 11555 Rockville Pike (first floor), Rockville, Maryland. The 
filing of requests for a hearing and petitions for leave to intervene 
is discussed below.
    Within 60 days after the date of publication of this notice, the 
licensee may file a request for a hearing with respect to issuance of 
the amendment to the subject facility operating license and any person 
whose interest may be affected by this proceeding and who wishes to 
participate as a party in the proceeding must file a written request 
for a hearing and a petition for leave to intervene. Requests for a 
hearing and a petition for leave to intervene shall be filed in 
accordance with the Commission's ``Rules of Practice for Domestic 
Licensing Proceedings'' in 10 CFR Part 2. Interested persons should 
consult a current copy of 10 CFR 2.309, which is available at the 
Commission's PDR, located at One White Flint North, Public File Area 
01F21, 11555 Rockville Pike (first floor), Rockville, Maryland. 
Publicly available records will be accessible from the Agencywide 
Documents Access and Management System's (ADAMS) Public Electronic 
Reading Room on the Internet at the NRC Web site, 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

[[Page 38714]]

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.

Arizona Public Service Company, et al., Docket Nos. STN 50-528, STN 50-
529, and STN 50-530, Palo Verde Nuclear Generating Station (PVNGS), 
Units 1, 2, and 3, Maricopa County, Arizona

    Date of amendments request: May 26, 2005.
    Description of amendments request: The amendments would revise the 
Technical Specification (TS) requirements related to steam generator 
(SG) tube integrity, consistent with those in NRC-approved Revision 4 
to Technical Specification Task Force (TSTF) Standard Technical 
Specification Change Traveler, TSTF-449, ``Steam Generator Tube 
Integrity.'' The proposed amendment also includes changes to the 
revised SG program in TS Section 5.5.9 to specify the SG tube 
inspection length through the SG tubesheet and establish plugging 
criteria in the inspected tubesheet region for the remaining original 
SGs containing Alloy 600 mill annealed (MA) tubes. This change is being 
proposed to establish conformance with the NRC position identified in 
Generic Letter (GL) 2004-01, ``Requirements for Steam Generator Tube 
Inspections.''
    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 analysis that established the inspection length through the 
SG tube sheet for the PVNGS Alloy 600 MA-tube SGs took into account 
the reinforcing effect the tubesheet has on the external surface of 
an expanded SG tube. Tube-bundle integrity will not be adversely 
affected by the implementation of the revised tube inspection scope. 
SG tube burst or collapse cannot occur within the confines of the 
tubesheet; therefore, the tube burst and collapse criteria of draft 
Regulatory Guide (RG) 1.121, ``Bases for Plugging Degraded PWR Steam 
Generator Tubes,'' are inherently met. Any degradation below the 
inspection length is shown by analyses and test results to be 
acceptable, thereby precluding an event with consequences similar to 
a postulated tube rupture event.
    Tube burst is precluded for cracks within the tubesheet by the 
constraint provided by the tubesheet. Thus, structural integrity is 
maintained by the tubesheet constraint. However, a 360-degree 
circumferential crack or many axially oriented cracks could permit 
severing of the tube and tube pullout from the tubesheet under the 
axial forces on the tube from primary to secondary pressure 
differentials. Analysis and testing was performed to define the 
length of non-degraded tubing that is sufficient to compensate for 
the axial forces on the tube and thus prevent pullout. That length 
is bounded by the inspection length proposed in this change.
    In conclusion, incorporation of the revised inspection scope 
into PVNGS TS maintains existing design limits and 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 performance based requirements are an improvement 
over the requirements imposed by the current TS.
    Implementation of the proposed Steam Generator Program will not 
introduce any adverse changes to the plant design basis or 
postulated accidents resulting from potential tube degradation. The 
result of the implementation of the Steam Generator Program will be 
an enhancement of SG tube performance. Primary to secondary leakage 
that may be experienced during all plant conditions will be 
monitored to ensure it remains within current accident analysis 
assumptions.
    The proposed change does not affect the design of the SGs, their 
method of operation, or primary or secondary coolant chemistry 
controls. In addition, the proposed change does not impact any other 
plant system or component. The change enhances SG inspection 
requirements.
    Tube-bundle integrity is expected to be maintained during all 
plant conditions upon implementation of the proposed tube inspection 
scope. Use of this scope does not introduce a new mechanism that 
would result in a different kind of accident from those previously 
analyzed. Even with the limiting circumstances of a complete 
circumferential separation of a tube occurring below the inspection 
length into the tubesheet, SG tube pullout is precluded and leakage 
is predicted to be maintained within the Updated Final Safety 
Analysis Report limits during all plant conditions.
    Therefore, the proposed change does not create the possibility 
of a new or different type of accident from any accident previously 
evaluated.
    3. Does the proposed change involve a significant reduction in a 
margin of safety.
    Response: No.
    The SG tubes in pressurized water reactors are an integral part 
of the reactor coolant pressure boundary and, as such, are relied 
upon to maintain the primary system's pressure and inventory. As 
part of the reactor coolant pressure boundary, the SG tubes are 
unique in that they are also relied upon as a heat transfer surface 
between the primary and secondary systems such that residual heat 
can be removed from the primary system. In addition, the SG tubes 
also isolate the radioactive fission products in the primary coolant 
from the secondary system. In summary, the safety function of a SG 
is maintained by ensuring the integrity of its tubes.
    Steam generator tube integrity is a function of the design, 
environment, and the physical condition of the tube. The proposed 
change does not affect tube design or operating environment. The 
proposed change is expected to result in an improvement in the tube 
integrity by implementing the Steam Generator Program to manage SG 
tube inspection, assessment, repair, and plugging. The requirements 
established by the Steam Generator Program are consistent with those 
in the applicable design codes and standards and are an improvement 
over the requirements in the current TS.
    Upon implementation of the revised inspection scope, operation 
with potential cracking below the Inspection Extent length in the 
expansion region of the SG tubing will meet the margin of safety as 
defined by Regulatory Guide (RG) 1.83 [Inservice Inspection of 
Pressurized Water Reactor Steam Generator Tubes], draft RG 1.121 
[Bases for Plugging Degraded PWRSteam Generator Tubes], and the 
requirements of General Design Criteria 14, 15, 31, and 32 of 
Appendix A to 10 CFR 50.


[[Page 38715]]


    The NRC staff has reviewed the licensee's analysis and, based on 
that review, it appears that the three 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.
    Attorney for licensee: Kenneth C. Manne, Senior Attorney, Arizona 
Public Service Company, P.O. Box 52034, Mail Station 7636, Phoenix, 
Arizona 85072-2034. NRC Acting Section Chief: Daniel S. Collins.

Arizona Public Service Company, et al., Docket Nos. STN 50-528, STN 50-
529, and STN 50-530, Palo Verde Nuclear Generating Station, Units 1, 2, 
and 3, Maricopa County, Arizona

    Date of amendments request: June 3, 2005.
    Description of amendments request: The proposed amendments would 
revise the Updated Final Safety Analysis Report (UFSAR) for Palo Verde 
Nuclear Generating Station (PVNGS), Units 1, 2 and 3. The proposed 
amendments would reflect a modification performed by the licensee that 
replaced the automatic water makeup function for the emergency diesel 
generator jacket water cooling system with that of manual operator 
actions.
    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 emergency diesel generator (EDG) is a system that must 
function in response to an accident that has been evaluated in 
either Chapter 6 or 15 of the PVNGS UFSAR. It is designed to respond 
to certain described accident scenarios. None of the accidents 
evaluated are initiated within the EDG system. Therefore, this 
request to allow the replacement of the automatic makeup feature(s) 
with a manual feature can not increase the probability of an 
accident previously postulated in the UFSAR.
    None of the accidents evaluated which credit operation of the 
EDG system require automatic fill of the DGCWS [Diesel Generator 
Cooling Water System] in order to mitigate the consequences of the 
accident. The fill system, whether automatic in nature as originally 
designed or manual, simply maintains the EDG in the ready state.
    Therefore, the proposed change does not involve a significant 
increase in the 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 EDG is a piece of equipment important to safety. This 
modification replaces the automatic water makeup function for the 
EDG jacket water cooling system with that of manual operator 
actions. The jacket water makeup is needed for normal leakage and 
possible evaporation. Area walkdowns occur twice daily when the 
diesel generator is in a standby mode (not running) and more 
frequently (thirty minutes after initial loading and every two hours 
while loaded) when the EDG is being tested or has responded to an 
emergency event. The area operator walkdown procedures instruct the 
operators to log the standpipe level and ensure it is in the normal 
operating range. If the level is not, operators are required to 
restore level and conduct further investigation of the condition and 
notify appropriate personnel. This ensures that enough water remains 
in the jacket water system to allow the diesel to remain operational 
and evaluations are performed in order to detect any abnormal 
leakrates. Therefore, the normal area operator walkdowns and 
frequencies are adequate to ensure that sufficient jacket water 
standpipe inventory is maintained.
    With this modification, the EDG is still maintained and 
monitored for proper conditions in a standby status to ensure that 
it will respond to emergencies when called upon. Once the EDG 
responds to an emergency signal and is loaded, its jacket water 
system is required to be monitored every two hours to help ensure 
that all parameters are observed and maintained for proper 
operation, including its jacket water standpipe level.
    So, with these measures in place it can be expected that the EDG 
will be maintained capable of performing as designed to any 
emergency safety signal. The [E]DG safety system and its support 
jacket water cooling system do not initiate any accident events.
    Therefore, the modification of this non-safety support system 
cannot 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 PVNGS UFSAR states that the design basis function for the 
emergency diesel generators is to provide a standby source of onsite 
Class 1E AC power for the two trains of engineered safety features 
equipment for safe plant shutdown and decay heat removal in the 
event of loss of preferred (off-site) power. Supporting this design 
basis function of supplying emergency power is the function of the 
emergency diesel generator jacket cooling water system, which is to 
remove rejected heat from each diesel engine at the rated design 
load of the emergency diesel generator. The UFSAR further describes 
the emergency diesel generator jacket cooling water surge tank 
(standpipe), stating that the surge tank is sized to provide an 
adequate reservoir to compensate for any minor leaks. The UFSAR also 
described makeup to the jacket cooling water system as being 
automatically actuated and provided from the safety-grade condensate 
transfer system or manually from the demineralized water systems. 
The subject modification replaced the automatic features with manual 
operator action--the sources of the makeup water have not changed.
    The PVNGS engineering analyses and the safety analyses that 
demonstrate the functional goals and the design basis of the 
emergency diesel generator system do not credit any makeup water 
supply to the jacket cooling water system of the emergency diesel 
generator for an initial 25 hours into an event. Operator monitoring 
and manual makeup provides adequate control for maintaining the 
DGCWS standpipe level, both for standby and loaded conditions. An 
automatically actuated makeup water supply is not essential to the 
safe and continued operation of the emergency diesel generator. 
Makeup water is provided as a convenient source of water to 
compensate for anticipated normal system losses and evaporation. It 
is not provided to serve as an emergency source of makeup water to 
the jacket cooling water system in the event of a major failure or 
leak occurring within the jacket cooling water system.
    Makeup to the system is required to compensate for normal 
expected system losses, minor leaks, and evaporation. In addition, 
an engineering calculation has been performed to address 10 CFR 50, 
Appendix R concerns, which demonstrates that no operator action is 
required or credited during the first twenty-five hours of emergency 
diesel generator loaded operation provided that the initial water 
level is at the specified minimum level. This twenty-five hour 
period before operator intervention, which is assumed to occur, 
sufficiently bounds the thirty minutes of no operator action that is 
normally assumed in most of the accident analyses.
    In addition, the area operator walkdown procedures instruct the 
operators to log the standpipe level and ensure it is in the normal 
operating range. If the level is not, operators are required to 
restore level and conduct further investigation of the condition and 
notify appropriate personnel. This ensures that enough water remains 
in the jacket water system to allow the diesel to remain operational 
and evaluations are performed in order to detect any abnormal 
leakrates.
    Therefore, APS has concluded that the proposed license amendment 
request does not involve a significant reduction in a margin of 
safety.
    Based on the above, Arizona Public Service Company (APS) 
concludes that the proposed amendment presents no significant 
hazards consideration under the standards set forth in 10 CFR 
50.92(c), and, accordingly, a finding of ``no significant hazards 
consideration'' is justified.

    The NRC staff has reviewed the licensee's analysis and, based on 
that review, it appears that the three 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 38716]]

    Attorney for licensee: Kenneth C. Manne, Senior Attorney, Arizona 
Public Service Company, P.O. Box 52034, Mail Station 7636, Phoenix, 
Arizona 85072-2034.
    NRC Section Chief: Daniel S. Collins.

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 amendments request: June 7, 2005.
    Description of amendments request: The proposed amendment would 
revise Technical Specification (TS) 3.1.1, ``Shutdown Margin,'' to 
modify Required Action B.1 restricting a positive reactivity addition. 
The proposed amendment would also correct an administrative error 
regarding an incorrect TS reference in TS 3.4.17, ``Special Test 
Exception RCS [reactor coolant system] Loops--Modes 4 and 5.''
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Would not involve a significant increase in the probability 
or consequences of an accident previously evaluated.
    The intent of this change is to clarify a Technical 
Specification involving positive reactivity additions to the 
shutdown reactor so that small, controlled, safe insertions of 
positive reactivity will be allowed where they are now categorically 
prohibited, posing a potential conflict between two required 
actions. These controlled activities could result in a slight change 
in the probability of an event occurring as a RCS manipulation that 
is currently prohibited would now be allowed. However, RCS 
manipulations are rigidly controlled to minimize the possibility of 
a significant reactivity increase.
    In addition, there is sufficient shutdown margin available in 
this condition to allow for slight reactivity changes without 
significantly increasing the probability of an accident previously 
evaluated.
    The proposed change involving positive reactivity additions does 
not permit the shutdown margin required by the Technical 
Specifications to be reduced. While the proposed change will permit 
changes in the discretionary boron concentration above the Technical 
Specification requirements, this excess concentration is not 
credited in the Updated Final Safety Analysis Report safety 
analysis. Because the initial conditions assumed in the safety 
analysis are preserved, no increase in the consequence of an 
accident previously evaluated would occur. These small changes are 
within the required shutdown margin, therefore, there is no increase 
in the consequence of an accident previously evaluated.
    The administrative error was in the marked up Technical 
Specification pages submitted with a proposed change. The correct 
Technical Specification number was provided in the proposal letter 
and was used by the staff in the discussion for accepting the 
proposed change. Correcting this administrative error does not 
change the significant hazards discussion previously submitted.
    Therefore, the proposed change does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    2. Would not create the possibility of a new or different kind 
of accident from any accident previously evaluated.
    This proposed change involving positive reactivity addition 
allows for a minor plant operational adjustment without adversely 
impacting the safety analysis required shutdown margin. It does not 
involve any change to plant equipment or the shutdown margin 
requirements in the Technical Specifications.
    The administrative error was in the marked up Technical 
Specification pages submitted with a proposed change. The correct 
Technical Specification number was provided in the proposal letter 
and was used by the staff in the discussion for accepting the 
proposed change. Correcting this administrative error does not 
change the significant hazards discussion previously submitted.
    Therefore, the proposed change will not create the possibility 
of a new or different kind of accident from any accident previously 
evaluated.
    3. Would not involve a significant reduction in [a] margin of 
safety.
    The margin of safety in Modes 3, 4 and 5 is preserved by the 
calculated shutdown margin which prevents an inadvertent 
criticality. The proposed change involving positive reactivity 
addition will permit reductions in discretionary shutdown margin 
that is beyond Technical Specification requirements. However, the 
shutdown margin required by the Technical Specifications is not 
changed. By not impacting the shutdown margin, the margin of safety 
is not affected.
    The administrative error was in the marked up Technical 
Specification pages submitted with a proposed change. The correct 
Technical Specification number was provided in the proposal letter 
and was used by the staff in the discussion for accepting the 
proposed change. Correcting this administrative error does not 
change the significant hazards discussion previously submitted.
    Therefore, the proposed change will 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 
amendments request involves no significant hazards consideration.
    Attorney for licensee: Carey Fleming, Sr. Counsel--Nuclear 
Generation, Constellation Generation Group, LLC, 750 East Pratt Street, 
17th floor, Baltimore, MD 21202.
    NRC Section Chief: Richard J. Laufer.

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

    Date of amendments request: June 7, 2005.
    Description of amendments request: The proposed amendment would 
revise the Technical Specifications (TSs) to eliminate the use of the 
defined term Core Alterations. The proposed amendment would incorporate 
the changes reflected in TS Task Force (TSTF) Travelers 471-T (TSTF-
471-T) and TSTF-51-A. In addition, the proposed amendment would revise 
TS 3.9.2, ``Nuclear Instrumentation,'' by replacing ``Core 
Alterations'' with ``positive reactivity additions'' in the required 
action for an inoperable source range monitor during refueling 
operations. The limiting conditions for operation in TS 3.9.4, 
``Shutdown Cooling (SDC) and Coolant Recirculation--High Water Level,'' 
would also be revised by replacing ``core alterations'' with ``movement 
of fuel assemblies within containment.''
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Operation of the facility in accordance with the proposed 
amendment would not involve a significant increase in the 
probability or consequences of an accident previously evaluated.
    The proposed change eliminates the use of the defined term CORE 
ALTERATIONS from the Technical Specifications. Core alterations are 
not an initiator of any accident previously evaluated except a fuel 
handling accident. Those revised Technical Specifications that 
protect the initial conditions of a fuel handling accident also 
require the suspension of movement of irradiated fuel assemblies, 
which protects the initial condition of a fuel handling accident. 
Therefore, suspension of CORE ALTERATIONS do not affect the 
initiators of the accidents previously evaluated and suspension of 
CORE ALTERATIONS does not affect the mitigation of the accidents 
previously evaluated.
    Therefore, the proposed change does not involve a significant 
increase in the probability or consequences of an accident 
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 accident previously evaluated.
    No new or different accidents result from utilizing the proposed 
change. The changes

[[Page 38717]]

do not involve a physical modification of the plant (i.e., no new or 
different type of equipment will be installed) or a significant 
change in the methods governing normal plant operation. In addition, 
the changes do not impose any new or different requirements. The 
changes do not alter assumptions made in the safety analysis. The 
proposed changes are consistent with the safety analysis 
assumptions.
    Therefore, the proposed change does not create the possibility 
of a new or different kind of accident from any previously 
evaluated.
    3. Operation of the facility in accordance with the proposed 
amendment would not involve a significant reduction in a margin of 
safety.
    Only two accidents are postulated to occur during plant 
conditions where CORE ALTERATIONS may be made: A fuel handling 
accident and a boron dilution accident. Suspending movement of 
irradiated fuel assemblies prevents a fuel handling accident. Also 
requiring the suspension of CORE ALTERATIONS is redundant to 
suspending movement of irradiated fuel assemblies and does not 
increase the margin of safety. CORE ALTERATIONS have no effect on a 
boron dilution accident. Core components are not involved in the 
initiation or mitigation of a boron dilution accident. Therefore, 
CORE ALTERATIONS have no effect on the margin of safety related to a 
boron dilution accident.
    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 
amendments request involves no significant hazards consideration.
    Attorney for licensee: Carey Fleming, Sr. Counsel--Nuclear 
Generation, Constellation Generation Group, LLC, 750 East Pratt Street, 
17th floor, Baltimore, MD 21202.
    NRC Section Chief: Richard J. Laufer.

Duke Energy Corporation, et al., Docket Nos. 50-413 and 50-414, Catawba 
Nuclear Station, Units 1 and 2, York County, South Carolina and Docket 
Nos. 50-369 and 50-370, McGuire Nuclear Station, Units 1 and 2, 
Mecklenburg County, North Carolina

    Date of amendment request: October 27, 2004.
    Description of amendment request: The amendments would revise the 
facility operating licenses (FOLs) to remove a license condition that 
limits the maximum rod average burnup for any rod to 60 GWd/mtU. This 
deletion would allow the 62 GWd/mtU limit, approved by the NRC, as 
documented in Duke Topical Report DPC-2009-P-A, to become the burnup 
limit. The amendments would also revise both of the station's Updated 
Final Safety Analysis Reports (Section 4.0) to include a new discussion 
of the fuel burnup limit. Additionally, approval would allow Duke to 
make an administrative revision to Duke Topical Report DPC-NE-2009-P-A, 
Revision 2, to reference the approval of these amendments and to 
reflect removal of the current license condition. Furthermore, the 
amendments would remove the McGuire FOL Section 2.E, that lists 
reporting requirements with regard to Maximum Power Level, Fire 
Protection, Protection of the Environment (Unit 2 FOL only), and 
Physical Protection. It would also remove the Catawba FOL Section 2.F, 
that lists reporting requirements with regard to Maximum Power Level, 
Updated Final Safety Analysis Report, Antitrust Conditions, Fire 
Protection, and Additional Conditions.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Would implementation of the changes proposed in this LAR 
[License Amendment Request] involve a significant increase in the 
probability or consequences of an accident previously evaluated?
    No, deletion of the fuel burnup limit currently stated as an 
additional license condition in the McGuire and Catawba Facility 
Operating Licenses has no impact on accident probabilities. Further, 
as determined in the NRC's environmental assessment which supports 
the increased burnup limit (NUREG/CR-6703, Environmental Effects of 
Extending Fuel Burnup Above 60 GWd/mtU), the potential environmental 
consequences of postulated accidents are not expected to increase 
significantly with increased burnup. Duke concurs with this 
assessment conclusion for the burnup range in this LAR.
    The deletion of the reporting requirements from the FOLs is 
solely administrative. No plant equipment or accident analyses will 
be affected by this deletion.
    2. Would implementation of the changes proposed in this LAR 
create the possibility of a new or different kind of accident from 
any accident previously evaluated?
    No, implementation of this amendment would not create the 
possibility of a new or different kind of accident from any accident 
previously evaluated. No new accident causal mechanisms will be 
created as a result of the NRC approval of this LAR. No changes are 
being made to the plant which will introduce any new accident causal 
mechanisms. This amendment does not otherwise impact any plant 
structures, systems, or components that are accident initiators; 
therefore, no new accident types are being created.
    3. Would implementation of the changes proposed in this LAR 
involve a significant reduction in a margin of safety?
    No, margin of safety is related to the confidence in the ability 
of the fission product barriers to perform their design functions 
during and following an accident situation. These barriers include 
the fuel cladding, the reactor coolant system, and the containment 
system. These barriers are not significantly affected by the changes 
proposed in this LAR. The effect of the increased burnup on fuel 
cladding was considered in the NRC's environmental assessment 
supporting the increase in the fuel burnup limit. Further, the 
proposed limit is equal to that approved for the fuel rod cladding 
at McGuire and Catawba.
    The deletion of the reporting requirements from the FOLs is 
solely administrative in nature. No plant equipment or accident 
analyses will be affected by this deletion.
    The margin of safety is established through the design of the 
plant structures, systems, components, the parameters within which 
the plant is operated, and the establishment of the setpoints for 
the actuation of equipment relied upon to respond to an event, and 
thereby protect the fission product barriers. The proposed changes 
have no significant impact on any of these considerations in regard 
to the physical plant or the manner in which it is operated.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: Ms. Lisa F. Vaughn, Legal Department 
(PB05E), Duke Energy Corporation, 422 South Church Street, Charlotte, 
North Carolina 28201-1006.
    NRC Section Chief: Evangelos C. Marinos

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

    Date of amendment request: October 11, 2004.
    Description of amendment request: The proposed amendments apply to 
Technical Specifications 3.8.1, ``AC Sources--Operating,'' and 3.8.9, 
``Distribution Systems--Operating.'' They would extend several 
completion times and would modify several Surveillance Requirement (SR) 
Notes. Additionally, they would correct a recently identified non-
conservative situation that currently exists with SR 3.8.1.4.
    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

[[Page 38718]]

consideration, which is presented below:

First Standard

    Will implementation of the changes proposed in this license 
amendment request involve a significant increase in the probability 
or consequences of an accident previously evaluated?
    No. The changes proposed in this license amendment request 
increase the Technical Specifications Completion Times for the 
emergency diesel generators and electrical power and distribution 
systems. Increasing these Completion Times will not cause a 
significant increase in the probability or consequences of an 
accident which has been previously evaluated. This license amendment 
request is supported by an extensive risk-informed study performed 
by the nuclear industry and documented in a topical report and 
Technical Specifications Task Force travelers that have been 
submitted for NRC review and approval. Within this study, the risk 
impacts of increasing the Completion Times were calculated and 
compared against the acceptability guidelines contained in the 
applicable regulatory guides and found to be acceptable. The 
emergency diesel generators and electrical power and distribution 
systems and equipment affected by this license amendment request 
will remain highly reliable. Thus there will be no significant 
increase in the probability or consequences of an accident which has 
been previously evaluated.
    The proposed changes that modify Surveillance Requirement notes 
are consistent with an NRC [Nuclear Regulatory Commission]-approved 
industry initiative. Implementation of these changes will require 
that the plant's risk be managed. Thus there will be no significant 
increase in the probability or consequences of an accident which has 
been previously evaluated.
    The proposed change that corrects the non-conservative 
Surveillance Requirement only increases a Technical Specifications 
parameter value in the conservative direction. Thus this change will 
not contribute to any increase in the probability or consequences of 
an accident which has been previously evaluated.

Second Standard

    Will implementation of the changes proposed in this license 
amendment request create the possibility of a new or different kind 
of accident from any accident previously evaluated?
    No. The proposed changes would create no new accidents since no 
changes are being made that introduce any new accident casual 
mechanisms. The deterministic evaluation that supports this license 
amendment request consisted of a review of plant systems and safety 
functions impacted by entry into the expanded Completion Times, the 
performance of testing in previously prohibited operating modes, or 
increasing a Technical Specification mandated parameter in the 
conservative direction. The emergency diesel generators and 
electrical power and distribution systems were quantitatively and 
qualitatively assessed. It was determined that no new accidents or 
transients would be introduced by the proposed changes.

Third Standard

    Will implementation of the changes proposed in this license 
amendment request involve a significant reduction in a margin of 
safety?
    No. The impact of the proposed changes on the safety margins was 
considered in the deterministic evaluations that support this 
license amendment request. Extending the Completion Times, 
performing testing activities to confirm operability, or 
conservatively increasing a Technical Specification controlled 
parameter does not adversely impact any assumptions or inputs in the 
transient analyses contained in the McGuire Updated Final Safety 
Analysis Report (UFSAR). The proposed changes have no negative 
impact upon the ability of the fission product barriers (fuel 
cladding, the reactor coolant system, and the containment system) to 
perform their design functions during and following an accident 
situation. Additionally, the proposed changes have no adverse impact 
on setpoints or limits established or assumed within the UFSAR.

    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: Evangelos C. Marinos.

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

    Date of amendment request: May 24, 2005.
    Description of amendment request: The proposed amendment would 
revise the steam generator (SG) tube inspection scope for Byron 
Station, Unit 2 for Refueling Outage 12 and the subsequent operating 
cycle. The proposed changes modify the inspection requirements for 
portions of SG tubes within the hot leg tubesheet region of the SGs.
    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 previously analyzed accidents are initiated by the failure 
of plant structures, systems, or components. The proposed changes 
that alter the SG inspection criteria do not have a detrimental 
impact on the integrity of any plant structure, system, or component 
that initiates an analyzed event. The proposed changes will not 
alter the operation of, or otherwise increase the failure 
probability of any plant equipment that initiates an analyzed 
accident. Therefore, the proposed change does not involve a 
significant increase in the probability of an accident previously 
evaluated.
    Of the applicable accidents previously evaluated, the limiting 
transients with consideration to the proposed changes to the SG tube 
inspection criteria, are the SG tube rupture (SGTR) event and the 
steam line break (SLB) accident.
    During the SGTR event, the required structural integrity margins 
of the SG tubes will be maintained by the presence of the SG 
tubesheet. SG tubes are hydraulically expanded in the tubesheet 
area. Tube rupture in tubes with cracks in the tubesheet is 
precluded by the constraint provided by the tubesheet. This 
constraint results from the hydraulic expansion process, thermal 
expansion mismatch between the tube and tubesheet and from the 
differential pressure between the primary and secondary side. Based 
on this design, the structural margins against burst, discussed in 
Regulatory Guide (RG) 1.121, ``Bases for Plugging Degraded PWR 
[pressurized water reactor] SG Tubes,'' are maintained for both 
normal and postulated accident conditions.
    The proposed changes do not affect other systems, structures, 
components or operational features. Therefore, the proposed changes 
result in no significant increase in the probability of the 
occurrence of a SGTR accident.
    At normal operating pressures, leakage from primary water stress 
corrosion cracking (PWSCC) below the proposed limited inspection 
depth is limited by both the tube-to-tubesheet crevice and the 
limited crack opening permitted by the tubesheet constraint. 
Consequently, negligible normal operating leakage is expected from 
cracks within the tubesheet region. The consequences of an SGTR 
event are affected by the primary-to-secondary leakage flow during 
the event. Primary-to-secondary leakage flow through a postulated 
broken tube is not affected by the proposed change since the 
tubesheet enhances the tube integrity in the region of the hydraulic 
expansion by precluding tube deformation beyond its initial 
hydraulically expanded outside diameter.
    The probability of a SLB is unaffected by the potential failure 
of a SG tube as this failure is not an initiator for a SLB.
    The consequences of a SLB are also not significantly affected by 
the proposed changes. During a SLB accident, the reduction in 
pressure above the tubesheet on the shell side of the SG creates an 
axially uniformly distributed load on the tubesheet due to the 
reactor coolant system pressure on the underside of the tubesheet. 
The resulting bending action constrains the tubes in the tubesheet 
thereby restricting primary-to-secondary leakage below the midplane.
    Primary-to-secondary leakage from tube degradation in the 
tubesheet area during the

[[Page 38719]]

limiting accident (i.e., SLB) is limited by flow restrictions 
resulting from the crack and tube-to-tubesheet contact pressures 
that provide a restricted leakage path above the indications and 
also limit the degree of potential crack face opening as compared to 
free span indications. The primary-to-secondary leak rate during 
postulated SLB accident conditions would be expected to be less than 
that during normal operation for indications near the bottom of the 
tubesheet (i.e., including indications in the tube end welds). This 
conclusion is based on the observation that while the driving 
pressure causing leakage increases by approximately a factor of two, 
the flow resistance associated with an increase in the tube-to-
tubesheet contact pressure, during a SLB, increases by up to 
approximately a factor of three. While such a leakage decrease is 
logically expected, the postulated accident leak rate could be 
conservatively bounded by twice the normal operating leak rate if 
the increase in contact pressure is ignored. Since normal operating 
leakage is limited to less than 0.104 gpm (150 gpd) per TS 3.4.13, 
``RCS Operational Leakage,'' the associated accident condition leak 
rate, assuming all leakage to be from lower tubesheet indications, 
would be bounded by approximately 0.2 gpm. This value is well within 
the assumed accident leakage rate of 0.5 gpm discussed in Updated 
Final Safety Analysis Table 15.1-3, ``Parameters Used in Steam Line 
Break Analyses.'' Hence it is reasonable to omit any consideration 
of inspection of the tube, tube end weld, bulges/overexpansions or 
other anomalies below 17 inches from the top of the hot leg 
tubesheet. Therefore, the consequences of a SLB accident remain 
unaffected.
    Based on the above discussion, the proposed changes do not 
involve an increase in the 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 the use or installation of 
new equipment and the currently installed equipment will not be 
operated in a new or different manner. No new or different system 
interactions are created and no new processes are introduced. The 
proposed changes will not introduce any new failure mechanisms, 
malfunctions, or accident initiators not already considered in the 
design and licensing bases.
    Based on this evaluation, the proposed change does not create 
the possibility of a new or different kind of accident from any 
accident previously evaluated.
    3. Does the proposed change involve significant reduction in a 
margin of safety?
    Response: No.
    The proposed changes maintain the required structural margins of 
the SG tubes for both normal and accident conditions. Nuclear Energy 
Institute (NEI) 97-06, ``Steam Generator Program Guidelines,'' 
Revision 1 and Regulatory Guide (RG) 1.121, ``Bases for Plugging 
Degraded PWR Steam Generator Tubes,'' are used as the bases in the 
development of the limited hot leg tubesheet inspection depth 
methodology for determining that SG tube integrity considerations 
are maintained within acceptable limits. RG 1.121 describes a method 
acceptable to the NRC for meeting General Design Criteria (GDC) 14, 
``Reactor coolant pressure boundary,'' GDC 15, ``Reactor coolant 
system design,'' GDC 31, ``Fracture prevention of reactor coolant 
pressure boundary,'' and GDC 32, ``Inspection of reactor coolant 
pressure boundary,'' by reducing the probability and consequences of 
a SGTR. RG 1.121 concludes that by determining the limiting safe 
conditions for tube wall degradation the probability and 
consequences of a SGTR are reduced. This RG uses safety factors on 
loads for tube burst that are consistent with the requirements of 
Section III of the American Society of Mechanical Engineers (ASME) 
Code.
    For axially oriented cracking located within the tubesheet, tube 
burst is precluded due to the presence of the tubesheet. For 
circumferentially oriented cracking, Westinghouse letter LTR-CDME-
05-32-P, ``Limited Inspection of the Steam Generator Tube Portion 
Within the Tubesheet at Byron Unit 2 and Braidwood Unit 2,'' 
Revision 1, dated May 2005, defines a length of degradation free 
expanded tubing that provides the necessary resistance to tube 
pullout due to the pressure induced forces, with applicable safety 
factors applied. Application of the limited hot leg tubesheet 
inspection depth criteria will preclude unacceptable primary-to-
secondary leakage during all plant conditions. The methodology for 
determining leakage provides for large margins between calculated 
and actual leakage values in the proposed limited hot leg tubesheet 
inspection depth criteria.
    Therefore, the proposed changes do not involve a significant 
hazards consideration under the criteria set forth in 10 CFR 
50.92(c).

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

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

    Date of amendment request: May 11, 2005. The proposed amendment 
supercedes, in its entirety, a previous amendment request dated April 
29, 2004, published in the Federal Register on May 25, 2004 (69 FR 
29766).
    Description of amendment request: The proposed amendment would 
revise technical specification (TS) 3/4.4.10, ``Reactor Coolant 
System--Structural Integrity, ASME Code Class 1, 2, and 3 Components,'' 
to allow a one-time extension of the surveillance interval for the 
reactor vessel internals vent valves from September 2005 to March 2006.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Does the proposed change involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    Response: No.
    The proposed one-time surveillance interval exception does not 
alter the design, operation, or testing method of any structure, 
system, or component. Therefore, the proposed change does not 
involve a significant increase in the probability of an accident 
previously evaluated. In addition, no accident initiators are 
affected and no previously analyzed accident scenario is changed. 
Initiating conditions and assumptions remain as previously analyzed. 
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 one-time surveillance interval exception does not 
alter the design, operation, or testing method of any structure, 
system, or component. The proposed change does not introduce any new 
or different accident initiators. Therefore, the proposed change 
does not create the possibility of a new or different kind of 
accident from any previously evaluated.
    3. Does the proposed change involve a significant reduction in a 
margin of safety?
    Response: No.
    The proposed one-time surveillance interval exception does not 
affect the capabilities of the Reactor Vessel Internals Vent Valves. 
Therefore, the proposed change will not involve a margin of safety.

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

[[Page 38720]]

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

    Date of amendment request: May 22, 2005.
    Description of amendment request: The proposed amendment would 
adopt a qualified alternate repair criteria (ARC) for axial tube end 
cracking (TEC) indications in the Davis-Besse Nuclear Power Station, 
Unit 1 once-through steam generator tubes. Specifically, the proposed 
amendment would revise the technical specification surveillance 
requirements for steam generator tube inservice inspection to include 
the TEC ARC. The technical basis for the ARC is provided in Babcock & 
Wilcox Owners Group Topical Report BAW-2346P, ``Alternate Repair 
Criteria for Tube End Cracking in the Tube-to-Tubesheet Roll Joint of 
Once-Through Steam Generators,'' dated April 1999.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Does the proposed change involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    Response: No.
    The proposed amendment does not increase the probability of any 
accident. Steam generator tube failure is an initiating condition 
for the steam generator tube rupture (SGTR) accident. The proposed 
TEC ARC does not affect the probability of an SGTR because the TEC 
ARC is limited to crack indications that are precluded from burst 
due to the presence of the tubesheet. Therefore, the proposed change 
does not involve a significant increase in the probability of an 
accident previously evaluated.
    The proposed amendment does not increase the consequences of any 
previously evaluated accident. Primary-to-secondary leakage affects 
the radiological consequences of accidents evaluated in the Updated 
Safety Analysis Report. The proposed amendment may result in an 
increase in post-accident primary-to-secondary leakage. Analyses 
have been performed to determine the expected post-accident leakage 
from each TEC left in service. The proposed amendment would impose 
inservice inspection and leakage assessment requirements that would 
ensure that the expected post-accident primary-to-secondary leakage 
through TECs and all other sources is maintained below the value 
assumed in the accident analyses. Therefore, the proposed change 
does not involve a significant increase in the 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 TEC ARC does not introduce any new failure modes or 
accident scenarios. Analyses have demonstrated that structural and 
leakage integrity is maintained for normal operating and accident 
conditions. Any failure of a tube from a TEC would be bounded by the 
SGTR analysis. Therefore, the proposed change does not create the 
possibility of a new or different kind of accident from any 
previously evaluated.
    3. Does the proposed change involve a significant reduction in a 
margin of safety?
    Response: No.
    The proposed amendment does not reduce the structural margin of 
the steam generator tubes. Structural integrity of the tube is 
maintained since the TEC ARC is limited to crack indications that 
are precluded from burst due to the presence of the tubesheet. The 
proposed amendment would impose inservice inspection and leakage 
assessment requirements that will ensure that the expected post-
accident primary-to-secondary leakage through TECs and all other 
sources is maintained below the value assumed in the accident 
analyses. Therefore, the proposed change does not involve a 
significant reduction in a margin of safety.

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

Florida Power and Light Company, Docket Nos. 50-335 and 50-389, St. 
Lucie Nuclear Plant, Units 1 and 2, St. Lucie County, Florida

    Date of amendment request: April 21, 2005.
    Description of amendment request: The submittal requests revision 
to several Technical Specifications (TSs) using seven TS Task Force 
(TSTF) generic changes. The seven TSTFs (nos. 5, 65, 101, 258, 299, 
308, and 361) delete redundant safety limit violation notification 
requirements; adopt use of generic titles for utility positions; change 
the auxiliary feedwate