[Federal Register Volume 69, Number 128 (Tuesday, July 6, 2004)]
[Notices]
[Pages 40668-40681]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: 04-15061]


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

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 June 11, 2004, through June 23, 2004. The 
last biweekly notice was published on June 22, 2004 (69 FR 34696).

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

[[Page 40669]]

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

    AmerGen Energy Company, LLC, Docket No. 50-461, Clinton Power 
Station, Unit 1, DeWitt County, Illinois
    Date of amendment request: May 20, 2004.
    Description of amendment request: The proposed amendment would 
support conversion from an 18-month to a 24-month fuel cycle. 
Specifically, the proposed amendment would (1) change certain technical 
specification (TS) surveillance requirement (SR)

[[Page 40670]]

frequencies from ``18 months'' to ``24 months'' in accordance with the 
guidance of Generic Letter 91-04, ``Changes in Technical Specification 
Surveillance Intervals to Accommodate a 24-Month Fuel Cycle,'' (2) 
change Administrative Controls Section 5.5.7, ``Ventilation Filter 
Testing Program (VFTP),'' to address changes to 18-month frequencies 
that are specified in Regulatory Guide 1.52, ``Design, Inspection, and 
Testing Criteria for Air Filtration and Adsorption Units of Post-
Accident Engineered-Safety-Feature Atmosphere Cleanup Systems in Light-
Water-Cooled Nuclear Power Plants,'' and (3) change selected allowable 
values for instrumentation setpoints. In addition, two separate 
administrative changes are being proposed to eliminate temporary 
changes that have expired and no longer apply. These include (1) 
removal of TS Table 3.0.2-1, ``Surveillance Intervals Extended to 
November 30, 2000,'' and a reference to it in SR 3.0.2, and (2) removal 
of footnotes (a) and (b) from TS Table 3.3.8.1-1, ``Loss of Power 
Instrumentation.''
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration which is presented below:

    1. Does the proposed amendment involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    Response: No.
    The proposed TS changes involve a change in the surveillance 
testing intervals and allowable values to facilitate a change in the 
operating cycle length. The analytical limit increase for the 
Reactor Vessel Pressure-High function remains conservative with 
respect to considerations for isolating the Residual Heat Removal-
Shut Down Cooling (RHR-SDC) system in the event of a line break and 
for providing overpressure protection to the low pressure RHR-SDC 
system piping. Also included in this application are administrative 
changes to remove Table 3.0.2-1 and the reference to it in SR 3.0.2 
(since this implements an expired one-time TS exception), to 
renumber certain SRs remaining at 18 month frequencies, and to 
remove footnotes (a) and (b) from Table 3.3.8.1-1 that applied 
temporary allowable values until completion of modification to tap 
settings and degraded voltage setpoints. The proposed TS changes do 
not physically impact the plant. The proposed TS changes do not 
degrade the performance of, or increase the challenges to, any 
safety systems assumed to function in the accident analysis. The 
proposed TS changes do not impact the usefulness of the SRs in 
evaluating the operability of required systems and components, or 
the way in which the surveillances are performed. In addition, the 
frequency of surveillance testing is not considered an initiator of 
any analyzed accident, nor does a revision to the frequency 
introduce any accident initiators. The specific value of the 
allowable value is not considered an initiator of any analyzed 
accident. Therefore, the proposed change does not involve a 
significant increase in the probability of an accident previously 
evaluated.
    The consequences of a previously evaluated accident are not 
significantly increased. The proposed change does not affect the 
performance of any equipment credited to mitigate the radiological 
consequences of an accident. Evaluation of the proposed TS changes 
demonstrated that the availability of credited equipment is not 
significantly affected because of other more frequent testing that 
is performed, the availability of redundant systems and equipment, 
and the high reliability of the equipment. Historical review of 
surveillance test results and associated maintenance records did not 
find evidence of failures that would invalidate the above 
conclusions.
    The allowable values have been developed in accordance with RG 
1.105, ``Instrument Setpoints,'' to ensure that the design and 
safety analysis limits are satisfied. The methodology used for the 
development of the allowable values ensures the affected 
instrumentation remains capable of mitigating design basis events as 
described in the safety analyses and that the results and 
radiological consequences described in the safety analyses remain 
bounding. Therefore, the proposed change does not alter the ability 
to detect and mitigate events and, as such, does not involve a 
significant increase in the consequences of an accident previously 
evaluated.
    2. Does the proposed amendment create the possibility of a new 
or different kind of accident from any accident previously 
evaluated?
    Response: No.
    The proposed TS changes involve a change in the surveillance 
testing intervals and allowable values to facilitate a change in the 
operating cycle length. The analytical limit increase for the 
Reactor Vessel Pressure-High function remains conservative with 
respect to considerations for isolating the RHR-SDC system in the 
event of a line break and for providing overpressure protection to 
the low pressure RHR-SDC system piping. Also included in this 
application are administrative changes to remove Table 3.0.2-1 and 
the reference to it in SR 3.0.2 since this implements an expired 
one-time exception, to renumber certain SRs remaining at 18 month 
frequencies, and to remove footnotes (a) and (b) from Table 3.3.8.1-
1 that applied temporary allowable values until completion of 
modification to tap settings and degraded voltage setpoints. The 
proposed TS changes do not introduce any failure mechanisms of a 
different type than those previously evaluated, since there are no 
physical changes being made to the facility. No new or different 
equipment is being installed. No installed equipment is being 
operated in a different manner. As a result, no new failure modes 
are being introduced. The way surveillance tests are performed 
remains unchanged. A historical review of surveillance test results 
and associated maintenance records indicated there was no evidence 
of any failures that would invalidate the above conclusions.
    Therefore, the proposed change does not create the possibility 
of a new or different kind of accident from any previously 
evaluated.
    3. Does the proposed amendment involve a significant reduction 
in a margin of safety?
    Response: No.
    The proposed TS changes involve a change in the surveillance 
testing intervals and allowable values to facilitate a change in the 
operating cycle length. The analytical limit increase for the 
Reactor Vessel Pressure-High function remains conservative with 
respect to considerations for isolating the RHR-SDC system in the 
event of a line break and for providing overpressure protection to 
the low pressure RHR-SDC system piping. Also included in this 
application are administrative changes to remove Table 3.0.2-1 and 
the reference to it in SR 3.0.2 since this implements an expired 
one-time exception, to renumber certain SRs remaining at 18 month 
frequencies, and to remove footnotes (a) and (b) from Table 3.3.8.1-
1 that applied temporary allowable values until completion of 
modification to tap settings and degraded voltage setpoints. The 
impact of these changes on system availability is not significant, 
based on other more frequent testing that is performed, the 
existence of redundant systems and equipment, and overall system 
reliability. Evaluations have shown there is no evidence of time 
dependent failures that would impact the availability of the 
systems. The proposed changes do not significantly impact the 
condition or performance of structures, systems, and components 
relied upon for accident mitigation. The proposed changes in TS 
instrumentation allowable values are the result of application of 
the CPS setpoint methodology using plant specific drift values. The 
revised allowable values more accurately reflect total 
instrumentation loop accuracy including drift while continuing to 
protect any assumed analytical limit. The proposed changes do not 
result in any hardware changes or in any changes to the analytical 
limits assumed in accident analyses. Existing operating margin 
between plant conditions and actual plant setpoints is not 
significantly reduced due to these changes. The proposed changes do 
not significantly impact any safety analysis assumptions or results.
    Therefore, the proposed change does not involve a significant 
reduction in a margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: Mr. Thomas S. O'Neill, Associate General 
Counsel, Exelon Generation Company, LLC, 4300 Winfield Road, 
Warrenville, IL 60666.

[[Page 40671]]

    NRC Section Chief: Anthony J. Mendiola.

    AmerGen Energy Company, LLC, et al., Docket No. 50-219, Oyster 
Creek Nuclear Generating Station, Ocean County, New Jersey
    Date of amendment request: May 20, 2004.
    Description of amendment request: The licensee proposed to revise 
the Technical Specifications (TS), section 3.2.B.4, to clarify the 
application of the action requirements for inoperable control rods. 
Specifically, this involves adding wording to clarify that operable 
control rods that have been taken out of service at the fully inserted 
position (i.e., disarmed) to perform hydraulic control unit maintenance 
are not to be counted as inoperable control rods. Control rods that 
have been fully inserted, and disarmed, fulfill the safety function of 
the control rod since it is in a position of maximum contribution to 
shutdown reactivity. Such clarification is consistent with the intent 
of the current operability requirements, and with the Nuclear 
Regulatory Commission guidance document entitled ``Standard Technical 
Specifications--General Electric Plants, BWR [Boiling Water Reactor]/
4,'' NUREG-1433, Revision 2, where the control rod operability 
requirements explicitly apply to ``inoperable control rods'' and 
``withdrawn stuck control rods.''
    In addition, the licensee proposed to correct a typographical error 
in Table 3.1.1 (page 3.1-12 of the TS), where ``note i'' was 
inadvertently typed as ``note I.''
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration. The NRC staff's analysis is presented below:
    The first standard requires that operation of the unit in 
accordance with the proposed amendment will not involve a significant 
increase in the probability or consequences of an accident previously 
evaluated. The proposed amendment involves clarifying, but not 
changing, the current intent of control rod operability requirements. 
The proposed amendment also corrects a typographical error. These 
changes will not lead to alteration of the physical design or 
operational procedures associated with the control rod system, or any 
other plant structure, system, or component (SSC). All requirements 
needed to assure the operability of the control rod system will remain 
unchanged. Action requirements for control rods were not assumed to be 
precursors of accidents, nor were they assumed to be components in 
previously evaluated accident scenarios. Accordingly, the revised 
specifications will lead to no increase in the consequences of an 
accident previously evaluated, and no increase of the probability of an 
accident previously evaluated.
    The second standard requires that operation of the unit in 
accordance with the proposed amendment will not create the possibility 
of a new or different kind of accident from any accident previously 
evaluated. As stated above, the proposed changes involve clarification 
of control rod operability requirements and correction of a 
typographical error. These changes do not alter the physical design, 
safety limits, or method of operation associated with the operation of 
the plant. Accordingly, the changes do not introduce any new or 
different kind of accident from those previously evaluated.
    The third standard requires that operation of the unit in 
accordance with the proposed amendment will not involve a significant 
reduction in a margin of safety. Since the licensee did not propose to 
exceed or alter a design basis or safety limit, did not propose to 
operate any component in a less conservative manner, and did not 
propose to use a less conservative analysis methodology, the proposed 
amendment will not affect in any way the performance characteristics 
and intended functions of any SSC. Therefore, the proposed amendment 
does not involve a significant reduction in a margin of safety.
    Based on the NRC staff's analysis, it appears that the three 
standards of 10 CFR 50.92(c) are satisfied. Therefore, the NRC staff 
proposes to determine that the proposed amendment involves no 
significant hazards consideration.
    Attorney for licensee: Thomas S. O'Neill, Associate General 
Counsel, Exelon Generation Company, LCC, 4300 Winfield Road, 
Warrenville, IL 60555.
    NRC Section Chief: Richard J. Laufer.

    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: February 4, 2004.
    Description of amendments request: The amendments would revise 
Technical Specification (TS) 3.7.1, ``Main Steam Safety Valves 
(MSSVs),'' to: (1) Permit operation in Mode 3 with 5 to 8 inoperable 
MSSVs (2 to 5 operable MSSVs) per steam generator, (2) increase the 
completion time to reduce the variable overpower trip (VOPT) setpoint 
when 1 to 4 MSSVs per steam generator are inoperable, and (3) make 
associated editorial changes.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. The proposed change does not involve a significant increase 
in the probability or consequences of an accident previously 
evaluated.
    Response: No. Each change is discussed below.
     Revise Technical Specification (TS) 3.7.1 to permit 
operation in Mode 3 when there are five to eight inoperable MSSVs 
(two to five operable MSSVs) per steam generator.
    This proposed change would allow the plant to remain in Mode 3 
with as few as two operable MSSVs per steam generator. Currently, 
the plant must be placed in Mode 4 with fewer than six operable 
MSSVs per steam generator. Two MSSVs have sufficient relieving 
capacity to dissipate core decay heat and reactor coolant pump heat 
in Mode 3 to limit secondary system pressure to less than or equal 
to 110% of design pressure, as required by ASME Code, Section III. A 
minimum of two MSSVs per steam generator (four total) would be 
required to be operable in Mode 3 in case of a single failure of one 
of the valves. Since this proposed change would continue to provide 
over-pressure protection and heat removal capability in Mode 3, this 
change would have no affect on any analyzed accidents. Therefore, 
this proposed change would not involve a significant increase in the 
probability or consequences of an accident previously evaluated.
     Increase the Completion Time for Required Action A.2 of 
TS 3.7.1 (reduce the variable overpower trip [VOPT] setpoint when 
one to four MSSVs per steam generator are inoperable) from 12 hours 
to 36 hours.
    Required Action A.2 of TS 3.7.1 specifies a Completion Time of 
12 hours to reduce the variable overpower trip (VOPT)--high setpoint 
if one or more required MSSVs are inoperable. The proposed increase 
in the Completion Time for Action A.2 from 12 hours to 36 hours is 
consistent with Industry/Technical Specification Task Force TSTF-
235, Revision 1, incorporated in Revision 2 of NUREG-1432, 
Combustion Engineering Standard Technical Specifications. The 
revised TS 3.7.1 Bases associated with TSTF-235, Revision 1, states 
that the Completion Time of 36 hours for Required Action A.2 is 
based on a reasonable time to correct the MSSV inoperability, the 
time required to perform the power reduction, operating experience 
in resetting all channels of a protective function, and on the low 
probability of the occurrence of a transient that could result in 
steam generator overpressure during this period. Increasing the 
Completion Time to reset the VOPT from

[[Page 40672]]

12 hours to 36 hours does not involve a significant increase in the 
probability or consequences of an accident previously evaluated.
     Make associated editorial changes.
    The associated editorial changes do not change any structure, 
system or component (SSC) or affect the operation or maintenance of 
any SSC. They are editorial enhancements to make the TSs easier to 
understand, eliminate potential inconsistencies with other TSs, and 
reduce the potential for human errors. 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 accident from any accident previously 
evaluated.
    Response: No. Each change is discussed below.
     Revise Technical Specification (TS) 3.7.1 to permit 
operation in Mode 3 when there are five to eight inoperable MSSVs 
(two to five operable MSSVs) per steam generator.
    This proposed change would allow the plant to remain in Mode 3 
with as few as two operable MSSVs per steam generator. Currently, 
the plant must be placed in Mode 4 with fewer than six operable 
MSSVs per steam generator. Two MSSVs have sufficient relieving 
capacity to dissipate core decay heat and reactor coolant pump heat 
in Mode 3 to limit secondary system pressure to less than or equal 
to 110% of design pressure, as required by ASME Code, Section III. A 
minimum of two MSSVs per steam generator (four total) would be 
required to be operable in Mode 3 in case of a single failure of one 
of the valves. This proposed change would continue to provide 
overpressure protection and heat removal capability in Mode 3. 
Therefore, this proposed change would not create the possibility of 
a new or different kind of accident from any accident previously 
evaluated.
     Increase the Completion Time for Required Action A.2 of 
TS 3.7.1 (reduce the variable overpower trip [VOPT] setpoint when 
one to four MSSVs per steam generator are inoperable) from 12 hours 
to 36 hours.
    Required Action A.2 of TS 3.7.1 specifies a Completion Time of 
12 hours to reduce the variable overpower trip--high setpoint if one 
or more required MSSVs are inoperable. The proposed increase in the 
Completion Time for Action A.2 from 12 hours to 36 hours is 
consistent with Industry/Technical Specification Task Force TSTF-
235, Revision 1, incorporated in Revision 2 of NUREG-1432, 
Combustion Engineering Standard Technical Specifications. The 
revised TS 3.7.1 Bases associated with TSTF-235, Revision 1, states 
that the Completion Time of 36 hours for Required Action A.2 is 
based on a reasonable time to correct the MSSV inoperability, the 
time required to perform the power reduction, operating experience 
in resetting all channels of a protective function, and on the low 
probability of the occurrence of a transient that could result in 
steam generator overpressure during this period. Therefore, this 
proposed change would not create the possibility of a new or 
different kind of accident from any accident previously evaluated.
     Make associated editorial changes.
    The associated editorial changes do not change any structure, 
system or component (SSC) or affect the operation or maintenance of 
any SSC. They are editorial enhancements to make the TSs easier to 
understand, eliminate potential inconsistencies with other TSs, and 
reduce the potential for human errors. Therefore, the proposed 
change does not create the possibility of a new or different kind of 
accident from any previously evaluated.
    3. The proposed change does not involve a significant reduction 
in a margin of safety.
    Response: No. Each change is discussed below.
     Revise Technical Specification (TS) 3.7.1 permit 
operation in Mode 3 when there are five to eight inoperable MSSVs 
(two to five operable MSSVs) per steam generator.
    This proposed change would allow the plant to remain in Mode 3 
when there are as few as two operable MSSVs per steam generator. 
Currently, the plant must be placed in Mode 4 with fewer than six 
operable MSSVs per steam generator. Two MSSVs have sufficient 
relieving capacity to dissipate core decay heat and reactor coolant 
pump heat in Mode 3 to limit secondary system pressure to less than 
or equal to 110% of design pressure, as required by ASME Code, 
Section III. A minimum of two MSSVs per steam generator (four total) 
would be required to be operable in Mode 3 in case of a single 
failure of one of the valves. Therefore, the proposed change does 
not involve a significant reduction in a margin of safety.
     Increase the Completion Time for Required Action A.2 of 
TS 3.7.1 (reduce the variable overpower trip [VOPT] setpoint when 
one to four MSSVs per steam generator are inoperable) from 12 hours 
to 36 hours.
    Required Action A.2 of TS 3.7.1 specifies a Completion Time of 
12 hours to reduce the variable overpower trip--high setpoint if one 
or more required MSSVs are inoperable. The proposed increase in the 
Completion Time for Action A.2 from 12 hours to 36 hours is 
consistent with Industry/Technical Specification Task Force TSTF-
235, Revision 1, incorporated in Revision 2 of NUREG-1432, 
Combustion Engineering Standard Technical Specifications. The 
revised TS 3.7.1 Bases associated with TSTF-235, Revision 1, states 
that the Completion Time of 36 hours for Required Action A.2 is 
based on a reasonable time to correct the MSSV inoperability, the 
time required to perform the power reduction, operating experience 
in resetting all channels of a protective function, and on the low 
probability of the occurrence of a transient that could result in 
steam generator overpressure during this period. Therefore, the 
proposed change does not involve a significant reduction in a margin 
of safety.
     Make associated editorial changes.
    The associated editorial changes do not change any structure, 
system or component (SSC) or affect the operation or maintenance of 
any SSC. They are editorial enhancements to make the TSs easier to 
understand, eliminate potential inconsistencies with other TSs, and 
reduce the potential for human errors. 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 
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 Section Chief: Stephen Dembek.
    Dominion Nuclear Connecticut Inc., et al., Docket No. 50-423, 
Millstone Power Station, Unit No. 3, New London County, Connecticut
    Date of amendment request: April 15, 2004.
    Description of amendment request: The proposed amendment would 
modify the fire protection license condition to reflect a proposed 
permanent change to the CO2 fire suppression system in the 
cable spreading area (CSA).
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

Criterion 1:

    Does the proposed amendment involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    Response: No.
    The CO2 system is designed to limit the effects of 
fire damage to plant equipment and does not contribute to the 
prevention or initiation of a fire event. The CO2 system 
is not safety-related and is not relied upon to safely shut down the 
reactor, mitigate radiological consequences of any accident, or 
maintain the reactor in a safe shutdown condition. Accordingly, the 
proposed amendment does not affect the inputs or assumptions for any 
accidents previously evaluated nor does it affect initiation of a 
fire event. Modifying the CO2 initiation system to a 
manual mode reduces the possibility of a malfunction leading to an 
inadvertent CO2 discharge. Because the automatic 
initiation feature of the CO2 system would be eliminated 
by the proposed amendment, inadvertent operation would no longer 
need to be a postulated failure for the CO2 system. The 
current analysis for a worst-case fire event allows for complete 
loss of the CSA which is protected by 3-hour fire-rated barriers. 
Alternate safe shutdown methods are available in the event that a 
fire consumes all equipment and cables in the room. The proposed 
amendment does not modify the fire suppression methodology in a way 
that would cause any greater damage than

[[Page 40673]]

complete loss of the CSA. The incipient fire detection system 
offsets the delay time for manual CO2 initiation by 
allowing an earlier response time by the fire brigade. Failure to 
take manual action is bounded by previous failure of the 
CO2 system to operate. Based on this discussion, the 
proposed amendment does not increase the probability or consequence 
of an accident previously evaluated.

Criterion 2:

    Does the proposed amendment create the possibility of a new or 
different kind of accident from any accident previously evaluated?
    Response: No.
    The CO2 system is a mitigating system designed to 
limit the effects of fire damage to plant equipment and is not 
credited for safe shutdown of the plant. The proposed amendment does 
not involve any change that would impact designed CO2 
concentration levels and therefore does not affect the ability of 
the CO2, once delivered, to act as [a] fire extinguishing 
agent. The proposed amendment does not introduce failure modes, 
accident initiators, or malfunctions that would cause a new or 
different kind of accident or fire event. The potential for 
increased water usage due to the proposed change in fire fighting 
methodology for the CSA is within the capability and capacity of the 
existing site fire water system and potential water buildup on the 
CSA floor is bounded by the existing flooding analysis. Therefore, 
the proposed amendment does not create the possibility of a new or 
different kind of accident from any accident previously evaluated.

Criterion 3:

    Does the proposed amendment involve a significant reduction in a 
margin of safety?
    Response: No.
    The evaluated fire event assumes a fire coincident with a loss 
of power, with no additional plant accidents. As stated above, the 
current analysis for a worst-case fire event in the CSA allows for 
complete loss of all cables and equipment in the CSA resulting in 
loss of use of the control room. The proposed amendment changes the 
CO2 system initiation method from automatic to manual and 
impacts the response time of applying CO2 as a fire-
extinguishing agent. This impact is not significant in that any 
potential increase in fire damage does not exceed complete loss of 
all the CSA cables and equipment. In addition, the incipient fire 
detection system offsets the delay time for manual CO2 
initiation by allowing an earlier response time by the fire brigade. 
The proposed amendment does not modify the CSA fire area 3-hour fire 
rated barriers. Therefore, based on the above, the proposed 
amendment does not involve a significant reduction in a margin of 
safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: Lillian M. Cuoco, Senior Nuclear Counsel, 
Dominion Nuclear Connecticut, Inc., Waterford, CT 06141-5127.
    NRC Section Chief: James W. Clifford.
    Duke Energy Corporation, Docket Nos. 50-269, 50-270, and 50-287, 
Oconee Nuclear Station, Units 1, 2, and 3, Oconee County, South 
Carolina
    Date of amendment request: May 25, 2004.
    Description of amendment request: The proposed amendments would 
revise the licensing basis in the Updated Final Safety Analysis Report 
to support installation of a passive low-pressure injection (LPI) cross 
connect inside containment for Unit 3. The proposed changes would 
revise the licensing basis for selected portions of the core flood and 
LPI piping to allow exclusion of the dynamic effects associated with a 
postulated rupture of that piping by application of leak-before-break 
technology. Similar amendments were approved for Unit 1 by NRC letter 
dated September 29, 2003, and for Unit 2 by NRC letter dated February 
5, 2004.
    The proposed amendments would also delete technical specifications 
(TSs) which will no longer apply when the LPI cross connect 
modification has been implemented. 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) Involve a significant increase in the probability or 
consequences of an accident previously evaluated:
    The proposed License Amendment Request (LAR) modifies the Unit 3 
licensing basis to allow the dynamic effects associated with 
postulated pipe rupture of selected portions of the Unit 3 Low 
Pressure Injection (LPI)/Core Flood (CF) piping to be excluded from 
the design basis. The proposed LAR also removes Technical 
Specifications that are no longer applicable due to the completion 
of the LPI cross connect modification on all three Oconee Units. The 
proposed design allowances for these selected portions of piping 
continue to allow the LPI system design to meet General Design 
Criteria (GDC) 4 requirements related to environmental and dynamic 
effects. The proposed LAR will continue to ensure that ONS [Oconee 
Nuclear Station] can meet design basis requirements associated with 
the LPI safety function. The addition of the crossover line will 
enhance the ability of the control room operator to mitigate the 
consequences of specific events for which LPI is credited. 
Therefore, the proposed LAR does not involve a significant increase 
in the probability or consequences of an accident previously 
evaluated.
    (2) Create the possibility of a new or different kind of 
accident from any kind of accident previously evaluated:
    The proposed LAR modifies the Unit 3 licensing basis to allow 
the dynamic effects associated with postulated pipe rupture of 
selected portions of Unit 3 LPI/CF piping to be excluded from the 
design basis and removes TS requirements that are no longer 
applicable due to the completion of the LPI cross connect 
modification on all three Oconee Units. The proposed design 
allowances for these selected portions of piping continue to allow 
the LPI system design to meet GDC 4 requirements related to 
environmental and dynamic effects. The systems affected by the 
changes are used to mitigate the consequences of an accident that 
has already occurred. The proposed licensing basis change does not 
affect the mitigating function of these systems. Consequently, these 
changes do not alter the nature of events postulated in the Safety 
Analysis Report nor do they introduce any unique precursor 
mechanisms. Therefore, the proposed amendment will not create the 
possibility of a new or different kind of accident from any accident 
previously evaluated.
    (3) Involve a significant reduction in a margin of safety.
    The proposed licensing basis and TS changes do not unfavorably 
affect any plant safety limits, set points, or design parameters. 
The changes also do not unfavorably affect the fuel, fuel cladding, 
RCS [Reactor Coolant System], or containment integrity. Therefore, 
the proposed changes, which add new design allowances associated 
with the passive LPI cross connect modification and remove obsolete 
TS requirements, do not involve a significant reduction in the 
margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: Anne W. Cottingham, Winston and Strawn LPP, 
1400 L Street, NW., Washington, DC 20005.
    NRC Section Chief: Stephanie M. Coffin (Acting).
    FirstEnergy Nuclear Operating Company, et al., Docket Nos. 50-334 
and 50-412, Beaver Valley Power Station, Unit Nos. 1 and 2 (BVPS-1 and 
2), Beaver County, Pennsylvania
    Date of amendment request: May 26, 2004.
    Description of amendment request: The proposed amendments would 
revise the current 72-hour allowed outage time (AOT) for the emergency 
diesel generators (EDGs) in Technical Specification (TS) 3.8.1.1 to a 
14-day AOT. Additionally, the proposed amendments delete the 
surveillance requirement in TS 4.8.1.1.2.b.1 which requires an EDG 
inspection, in accordance with the manufacturer's recommendations, 
every 18 months

[[Page 40674]]

during shutdown. The periodic EDG maintenance inspection requirements 
will be relocated to a licensee-controlled maintenance program that is 
referenced in the Updated Final Safety Analysis Report (UFSAR). Future 
changes to the EDG maintenance program would then be controlled 
pursuant to Title 10 of the Code of Federal Regulations (10 CFR), 
Section 50.59. Lastly, the proposed amendments would revise footnote 
(1) of TS 3.8.1.1, which currently provides a 7-day AOT to restore EDG 
fuel oil properties which do not meet the requirements of TS 
4.8.1.1.2.d.2 or TS 4.8.1.1.2.e. The revised footnote wording would 
allow delay of action requirements for up to 7 days when the EDGs are 
inoperable solely as a result of failure to meet TS 4.8.1.1.2.d.2 or TS 
4.8.1.1.2.e surveillance requirements.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Does the proposed change involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    Response: No.
    The proposed change does not affect the design, operational 
characteristics, function or the reliability of the EDGs. The EDGs 
are not initiating conditions for any accident previously evaluated. 
The EDGs mitigate the consequences of previously evaluated accidents 
involving loss of offsite power.
    The consequences of any previously analyzed accident will not be 
significantly affected by extending the AOT for a single EDG, since 
the remaining EDG supporting the redundant Engineered Safety 
Features systems will continue to be available to perform the 
accident mitigation functions. In addition, to fully evaluate the 
effects of the proposed EDG AOT extension, a Probabilistic Risk 
Assessment was performed to quantitatively assess the risk impact of 
the proposed change for each unit. The results of this risk 
assessment concluded that the increase in plant risk is very small 
and consistent with the guidance contained in Regulatory Guide 1.174 
and Regulatory Guide 1.177.
    The deletion of TS surveillance requirement 4.8.1.1.2.b.1 from 
the Technical Specifications will not impact the capability of the 
EDGs to perform their accident mitigation functions. The required 
EDG maintenance inspections will continue to be performed in 
accordance with the licensee EDG maintenance program. The risk of 
performing the maintenance inspections during power operation has 
been considered in the EDG AOT extension supporting risk evaluation 
and determined to be acceptable.
    The proposed change to footnote (1) of TS 3.8.1.1 will also not 
impact the capability of the EDGs to perform their accident 
mitigation functions. Fuel oil properties that are not within the 
specified limits will not have an immediate effect on EDG operation 
and restoring the fuel oil to within limits within 7 days will 
ensure the availability of high grade fuel oil for the EDGs.
    Therefore, the proposed change does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    2. Does the proposed change create the possibility of a new or 
different kind of accident from any accident previously evaluated?
    Response: No.
    The proposed changes do not involve a change in the design, 
configuration, or method of operation of the plant. The changes do 
not involve the addition of new equipment or the modification of 
existing equipment. As such, no new failure modes are introduced by 
these changes.
    Therefore, the proposed change does not create the possibility 
of a new or different kind of accident from any previously 
evaluated.
    3. Does the proposed change involve a significant reduction in a 
margin of safety?
    Response: No.
    The proposed changes do not alter the plant design and do not 
affect any assumptions or inputs to the safety analysis. The 
proposed changes to the EDG allowed outage time have been evaluated 
both deterministically and using a risk informed approach. These 
evaluations demonstrate that power system design defense-in-depth 
capabilities will be maintained and that the risk contribution is 
small.
    In addition, the proposed deletion of the EDG maintenance 
inspection surveillance requirements from the TS[s] and 
modifications to the EDG action requirements associated with the EDG 
fuel oil surveillances will not impact the EDG reliability and their 
capability to perform their accident mitigation function.
    Therefore, the proposed change does not involve a significant 
reduction in a margin of safety.

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

    FirstEnergy Nuclear Operating Company, et al., Docket Nos. 50-334 
and 50-412, Beaver Valley Power Station, Unit Nos. 1 and 2 (BVPS-1 and 
2), Beaver County, Pennsylvania
    Date of amendment request: June 2, 2004.
    Description of amendment request: The proposed amendments would 
revise the Technical Specification (TS) surveillance interval from 
monthly to quarterly for certain reactor trip system and engineered 
safety feature actuation system channel functional tests in accordance 
with the methodology presented in the Nuclear Regulatory Commission-
approved topical report, WCAP-10271, ``Evaluation of Surveillance 
Frequencies and Out of Service Times for the Reactor Protection 
Instrumentation System,'' and supplements thereto.
    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.
    Operation of the Beaver Valley Power Station in accordance with 
the proposed license amendment does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    The proposed change modifies surveillance frequencies. Increases 
in the surveillance test intervals have been established based on 
achieving acceptable levels of equipment reliability. Consequently, 
equipment that is required to operate to mitigate an accident will 
continue to operate as expected and the probability of the 
initiation of any accident previously evaluated will not be 
significantly increased. Implementation of the proposed changes does 
not alter the manner in which protection is afforded. This equipment 
will continue to be tested in a manner and at a frequency to give 
confidence that the equipment can perform its assumed safety 
function. As a result, the proposed surveillance requirement changes 
do not significantly affect the consequences of any accident 
previously evaluated.
    Therefore, the proposed changes do not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    2. Does the proposed change create the possibility of a new or 
different kind of accident from any previously evaluated?
    Response: No.
    The proposed change does not involve any physical changes to the 
plant or the modes of plant operation defined in the Technical 
Specifications. The proposed change does not involve the addition or 
modification of plant equipment nor does it alter the design or 
operation of any plant systems. No new accident scenarios, transient 
precursors or failure mechanisms are introduced as a result of these 
changes.
    There are no changes in this proposal that would cause the 
malfunction of safety-related equipment assumed to be operable in 
accident analyses. No new mode of failure has been created and no 
new equipment performance requirements are imposed. The proposed 
change has no effect on any previously evaluated accident.

[[Page 40675]]

    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 change in surveillance frequencies has been evaluated to 
ensure that it provides an acceptable level of equipment 
reliability. Equipment continues to be tested at a frequency that 
gives confidence that the equipment can perform its assumed safety 
function when required. The proposed changes do not alter the manner 
in which safety limits, limiting safety system setpoints or limiting 
conditions for operations are determined. The impact of reduced 
testing is to allow a longer time interval over which instrument 
uncertainties (e.g. drift) may act. Experience has shown that the 
initial uncertainty assumptions are valid for reduced testing.
    Implementation of the proposed changes is expected to result in 
an overall improvement in safety since plant transients initiated 
from inadvertent safety system actuation should be reduced. Less 
frequent testing will reduce the likelihood for inadvertent reactor 
trips and inadvertent actuation of Engineered Safety Feature 
Actuation System components.
    Therefore, the proposed changes do not involve a significant 
reduction in a margin of safety.

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

    FirstEnergy Nuclear Operating Company, Docket No. 50-440, Perry 
Nuclear Power Plant, Unit 1, Lake County, Ohio
    Date of amendment request: March 31, 2004.
    Description of amendment request: The proposed change allows entry 
into a mode or other specified condition in the applicability of a 
technical specification (TS), while in a condition statement and the 
associated required actions of the TS, provided the licensee performs a 
risk assessment and manages risk consistent with the program in place 
for complying with the requirements of section 50.65(a)(4) of Title 10 
of the Code of Federal Regulations (10 CFR), part 50. Limiting 
Condition for Operation (LCO) 3.0.4 exceptions in individual TS would 
be eliminated, several notes or specific exceptions are revised to 
reflect the related changes to LCO 3.0.4, and Surveillance Requirement 
3.0.4 is revised to reflect the LCO 3.0.4 allowance.
    This change was proposed by the industry's Technical Specification 
Task Force (TSTF) and is designated TSTF-359. The NRC staff issued a 
notice of opportunity for comment in the Federal Register on August 2, 
2002 (67 FR 50475), on possible amendments concerning TSTF-359, 
including a model safety evaluation and model no significant hazards 
consideration (NSHC) determination, using the consolidated line item 
improvement process. The NRC staff subsequently issued a notice of 
availability of the models for referencing in license amendment 
applications in the Federal Register on April 4, 2003 (68 FR 16579). 
The licensee affirmed the applicability of the following NSHC 
determination in its application dated March 1, 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 allows entry into a mode or other specified 
condition in the applicability of a TS, while in a TS condition 
statement and the associated required actions of the TS. Being in a 
TS condition and the associated required actions is not an initiator 
of any accident previously evaluated. Therefore, the probability of 
an accident previously evaluated is not significantly increased. The 
consequences of an accident while relying on required actions as 
allowed by proposed LCO 3.0.4, are no different than the 
consequences of an accident while entering and relying on the 
required actions while starting in a condition of applicability of 
the TS. Therefore, the consequences of an accident previously 
evaluated are not significantly affected by this change. The 
addition of a requirement to assess and manage the risk introduced 
by this change will further minimize possible concerns. 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). 
Entering into a mode or other specified condition in the 
applicability of a TS, while in a TS condition statement and the 
associated required actions of the TS, will not introduce new 
failure modes or effects and will not, in the absence of other 
unrelated failures, lead to an accident whose consequences exceed 
the consequences of accidents previously evaluated. The addition of 
a requirement to assess and manage the risk introduced by this 
change will further minimize possible concerns. Thus, this change 
does not create the possibility of a new or different kind of 
accident from an accident previously evaluated.

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

    The proposed change allows entry into a mode or other specified 
condition in the applicability of a TS, while in a TS condition 
statement and the associated required actions of the TS. The TS 
allow operation of the plant without the full complement of 
equipment through the conditions for not meeting the TS LCO. The 
risk associated with this allowance is managed by the imposition of 
required actions that must be performed within the prescribed 
completion times. The net effect of being in a TS condition on the 
margin of safety is not considered significant. The proposed change 
does not alter the required actions or completion times of the TS. 
The proposed change allows TS conditions to be entered, and the 
associated required actions and completion times to be used in new 
circumstances. This use is predicated upon the licensee's 
performance of a risk assessment and the management of plant risk. 
The change also eliminates current allowances for utilizing required 
actions and completion times in similar circumstances, without 
assessing and managing risk. The net change to the margin of safety 
is insignificant. Therefore, this 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: Mary E. O'Reilly, Attorney, FirstEnergy 
Corporation, 76 South Main Street, Akron, OH 44308.
    NRC Section Chief: Anthony J. Mendiola.

    Nuclear Management Company, LLC, Docket No. 50-305, Kewaunee 
Nuclear Power Plant, Kewaunee County, Wisconsin
    Date of amendment request: May 25, 2004.
    Description of amendment request: The proposed amendment revises 
Technical Specifications (TSs) 3.10.e and 3.10.f to add an allowed 
outage time for the individual rod position indication (IRPI) system of 
24 hours with more than one IRPI group inoperable. Additional changes 
add the demand step counters to the TSs and add a note to allow for a 
soak time subsequent to substantial rod motion for the rods that exceed 
their position limits before invoking the TS requirements. Also, the 
definition of ``immediately'' is added to TS 1.0.

[[Page 40676]]

    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Does the proposed amendment involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    Response: No.
    Rod position indication instrumentation is not an assumed 
accident initiator, providing indication only of the control and 
shutdown rods position. Normal operation, abnormal occurrences and 
accident analyses assume the rods are at certain positions within 
the reactor core. The changes requested herein modify the time the 
existing two rod position indication systems may be inoperable and 
provide appropriate actions to compensate for that inoperability and 
add the second, digital, rod position indication system to the TS. 
Thus, this change does not involve a significant increase in the 
probability of an accident.
    The condition of concern is the alignment of the rods. Operating 
with a rod position indicator inoperable does not change the 
position of the rod; an inoperable rod position indication 
instrument does not make a rod misaligned. An increase in the 
consequences with the rods only comes from a rod being misaligned 
such that an increase in the heat produced in a localized area 
causes the fuel to fail either during operation, during a plant 
transient or post-accident. An inoperable rod position indicator 
does not change the position of the rod. Rod position is 
subsequently verified by other means if the rod is moved by greater 
than a predetermined amount. Indication of rod position by other 
means ensures rod position remains within analytical limits. Thus, 
inoperable rod position indication instrumentation does not involve 
an increase in the consequences of an accident.
    Therefore, the proposed change does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    2. Does the proposed amendment create the possibility of a new 
or different kind of accident from any accident previously 
evaluated?
    Response: No.
    This proposed change does not alter the design, function, or 
operation of any plant component and does not install any new or 
different equipment. The malfunction of safety related equipment, 
assumed operable in the accident analyses, would not be caused 
because of the proposed technical specification change. No new 
failure mode has been created and no new equipment performance 
burdens are imposed. Therefore, the possibility of a new or 
different kind of accident from those previously analyzed has not 
been created.
    Therefore, the proposed change does not create the possibility 
of a new or different kind of accident from any previously 
evaluated.
    3. Does the proposed amendment involve a significant reduction 
in a margin of safety?
    Response: No.
    The rod position indication system is an instrumentation system 
that provides indication to the operators that a control rod may be 
misaligned. Inoperable individual rod position indication 
instrumentation does not by itself harm or affect reactor operation, 
but may impair the ability of the operators to detect a misaligned 
rod. To compensate for this potential impairment of the operators' 
ability to detect a misaligned rod, requirements to verify the 
inoperable rod position indicators position are added. The impact of 
inoperable rod position indication instrumentation is offset by the 
availability of other indications that a rod is misaligned. Excore 
and incore nuclear instrumentation provides indication that reactor 
power, flux density, may have shifted axially or radially. Also, 
thermocouple indication would show that the core temperatures have 
increased in one region of the core and/or decreased in another 
region of the core.
    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: Bradley D. Jackson, Esq., Foley and Lardner, 
P.O. Box 1497, Madison, WI 53701-1497.
    NRC Section Chief: L. Raghavan.

    Nuclear Management Company, LLC, Docket No. 50-305, Kewaunee 
Nuclear Power Plant, Kewaunee County, Wisconsin
    Date of amendment request: June 1, 2004.
    Description of amendment request: The proposed amendment revises 
Technical Specification (TS) 1.0, ``Definitions,'' Table TS 3.5-2, 
``Instrument Operation Conditions for Reactor Trip,'' and Table TS 4.1-
1, ``Minimum Frequencies for Checks, Calibrations and Test of 
Instrument Channels.'' The TS revisions will add a definition for 
``staggered test basis,'' increase surveillance test intervals for 
reactor protection system and engineered safety features actuation 
system analog channels and logic cabinets, and add a completion time 
for the reactor trip breakers.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Does the proposed amendment involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    No. The proposed changes to the STIs [surveillance test 
intervals] and the RTB CT [reactor trip breaker completion time] 
reduce the potential for inadvertent reactor trips and spurious 
actuations, and therefore, do not increase the probability of an 
accident previously evaluated.
    The proposed changes will not result in a significant increase 
in the risk of plant operation as demonstrated in WCAP-15376-P-A. 
The impact of plant safety as measured by core damage frequency 
(CDF) is less than 1.0E-06 per year and the impact of large early 
release frequency (LERF) is less than 1.0E-07 per year. For the 
addition of the RTB CT, the incremental conditional core damage 
probabilities (ICCDP) and incremental conditional large early 
release probabilities (ICLERP) are less than 5.0E-08. These changes 
meet the acceptance criteria in Regulatory Guides 1.174 and 1.177. 
Therefore, there will not be a significant increase in the 
probability of an accident.
    The proposed changes did not include any hardware changes, and 
therefore, all structures, systems, and components will continue to 
perform their intended function to mitigate the consequences of an 
event within the assumed acceptance limits. The proposed changes do 
not affect source term, containment isolation, or the radiological 
release assumptions used in evaluating radiological consequences of 
previously analyzed accidents. Therefore, the proposed changes do 
not increase the consequences of an accident previously evaluated.
    Based on the above paragraphs, it is concluded the proposed 
changes do not involve a significant increase in the probability or 
consequences of an accident previously evaluated.
    2. Does the proposed amendment create the possibility of a new 
or different kind of accident from any accident previously 
evaluated?
    No. The proposed changes do not involve any hardware changes, 
any setpoint changes, any addition of safety related equipment, or 
any changes in the manner in which the systems provide plant 
protection. Additionally, all operator actions credited in accident 
analyses remain the same. There are no new or different accident 
initiators or new accidents scenarios created by the proposed 
changes. Therefore, the proposed changes do not create the 
possibility of a new or different kind of accident from any accident 
previously evaluated.
    3. Does the proposed amendment involve a significant reduction 
in a margin of safety?
    No. The safety analyses acceptance criteria in the Updated 
Safety Analysis Report (USAR) are not impacted by these changes. The 
proposed changes do not alter the manner in which safety limits, 
limiting safety system settings, or limiting conditions for 
operation are determined. All signals and operator actions credited 
in the USAR accident analyses will remain the same. Redundant RPS 
[reactor protection system] and ESFAS [engineered safety features 
actuation system] trains are maintained and diversity with regard to 
the signals that provide reactor trip and engineered safety features 
actuation is also maintained. The calculated impact on risk 
continues to meet

[[Page 40677]]

the acceptance criteria contained in Regulatory Guides 1.174 and 
1.177. Therefore, the proposed changes do not involve a significant 
reduction in the margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: Bradley D. Jackson, Esq., Foley and Lardner, 
P.O. Box 1497, Madison, WI 53701-1497.
    NRC Section Chief: L. Raghavan.

    Nuclear Management Company, LLC, Docket Nos. 50-282 and 50-306, 
Prairie Island Nuclear Generating Plant, Units 1 and 2, Goodhue County, 
Minnesota
    Date of amendment request: May 3, 2004.
    Description of amendment request: The proposed amendments would 
revise the Prairie Island Nuclear Generating Plant (PINGP) licensing 
basis to (1) define a hydraulic analysis methodology for demonstrating 
functionality of the cooling water (CL) system following a design basis 
seismic event and (2) define acceptance criteria from the American 
Society of Mechanical Engineers (ASME) Section III Code, Subsection ND, 
when performing stress analysis of the CL system non-Class I piping 
with design basis seismic loads.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Do the proposed changes involve a significant increase in the 
probability or consequences of an accident previously evaluated?
    Response: No.
    This license amendment proposes to revise the plant licensing 
basis to include: (1) a hydraulic analysis methodology for 
demonstrating functionality of the CL system following a design 
basis seismic event; and (2) American Society of Mechanical 
Engineers, Section III, Subsection ND, ``Class 3 Components,'' 1986 
Edition, Service Level D as the basis for acceptance criteria when 
performing stress analysis of the cooling water system non-Class I 
piping with design basis seismic loads.
    The cooling water system provides a heat sink for removal of 
process and operating heat from safety related components during 
design basis accidents. This system is not an accident initiator and 
thus these proposed licensing basis changes do not increase the 
probability of a previously evaluated accident.
    The proposed plant licensing basis changes will provide the 
basis for evaluating cooling water system capability during and 
following a design basis seismic event. Use of the proposed 
methodology and acceptance criteria will conservatively demonstrate 
that the cooling water system will continue to provide its design 
cooling function. With the cooling water system design heat removal 
capability maintained, accident consequences will not be increased. 
Thus these licensing basis changes do not involve an increase in the 
consequences of an accident previously evaluated.
    Therefore, the proposed changes do not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    2. Do the proposed changes create the possibility of a new or 
different kind of accident from any accident previously evaluated?
    Response: No.
    This license amendment proposes to revise the plant licensing 
basis to include: (1) a hydraulic analysis methodology for 
demonstrating functionality of the CL system following a design 
basis seismic event; and (2) American Society of Mechanical 
Engineers, Section III, Subsection ND, ``Class 3 Components,'' 1986 
Edition, Service Level D as the basis for acceptance criteria when 
performing stress analysis of the cooling water system non-Class I 
piping with design basis seismic loads.
    The proposed licensing basis changes do not involve a change in 
system operation, or procedures involved with the cooling water 
system. The proposed changes provide a conservative basis for 
evaluating cooling water system capability following a design basis 
seismic event. There are no new failure modes or mechanisms created 
through use of the proposed evaluation methodology or pipe stress 
analysis with the proposed acceptance criteria. Use of these 
licensing basis changes with the cooling water system does not 
involve any modification in the operational limits of plant systems. 
There are no new accident precursors generated with use of these 
licensing basis changes.
    Therefore, the proposed changes do not create the possibility of 
a new or different kind of accident from any previously evaluated.
    3. Do the proposed changes involve a significant reduction in a 
margin of safety?
    Response: No.
    This license amendment proposes to revise the plant licensing 
basis to include: (1) A hydraulic analysis methodology for 
demonstrating functionality of the CL system following a design 
basis seismic event; and (2) American Society of Mechanical 
Engineers, Section III, Subsection ND, ``Class 3 Components,'' 1986 
Edition, Service Level D as the basis for acceptance criteria when 
performing stress analysis of the cooling water system non-Class I 
piping with design basis seismic loads.
    The current plant licensing basis does not provide a hydraulic 
analysis methodology for demonstrating functionality of the cooling 
water system following a design basis seismic event and it does not 
provide acceptance criteria for piping stress analysis of the 
cooling water system non-Class I piping with design basis seismic 
loads. The proposed changes provide a conservative basis for 
evaluating cooling water system capability during and following a 
design basis seismic event. The proposed methodology for evaluating 
cooling water system capability is consistent with methods proposed 
by the NRC Staff and current plant methods for evaluating internal 
flooding. The intended use of the proposed acceptance criteria is 
consistent with the intended post-seismic use of the non-Class I 
portions of the cooling water system.
    Therefore, the proposed changes do not involve a significant 
reduction in a margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment requests involve no significant hazards consideration.
    Attorney for licensee: Jonathan Rogoff, Esquire, Vice President, 
Counsel & Secretary, Nuclear Management Company, LLC, 700 First Street, 
Hudson, WI 54016.
    NRC Section Chief: L. Raghavan.

    Omaha Public Power District, Docket No. 50-285, Fort Calhoun 
Station, Unit No. 1, Washington County, Nebraska
    Date of amendment request: May 14, 2004.
    Description of amendment request: The proposed amendment revises 
the Fort Calhoun Station, Unit No. 1 (FCS) Technical Specifications to 
provide a risk-informed alternative to the existing restoration period 
for the high pressure safety injection (HPSI) system. The FCS 
application of the risk-informed change integrates the Westinghouse 
Owners Group recommendations identified in WCAP-15773, ``Joint 
Application Report for the Implementation of a Risk Management 
Technical Specification for the High Pressure Safety Injection (HPSI) 
System.''
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. The proposed change does not involve a significant increase 
in the probability or consequences of an accident previously 
evaluated.
    The proposed change does not require any physical change to any 
plant systems, structures, or components nor does it require any 
change in systems or plant operations; thus the probability of an 
accident previously evaluated occurring remains unchanged. The 
proposed change does not require any change in safety analysis 
methods or results. A single HPSI subsystem inoperability is 
considered

[[Page 40678]]

in existing plant analyses and regulatory criteria with respect to 
single failure criteria and the risk of extended HPSI subsystem 
outages are assessed in accordance with the Maintenance Rule [10 CFR 
50.65]. Because risk is appropriately managed and compensatory 
measures established where necessary, the consequences of an 
accident previously analyzed are not significantly increased. The 
change to establish the extended HPSI CT [completion time] limits is 
justified because operation within the requirements of the 
Maintenance Rule continues to be governed by the current regulation 
and plant programs.
    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 accident from any accident previously 
evaluated.
    HPSI System inoperabilities are assumed in existing analyses 
with respect to single failure criteria and are limited by existing 
regulation. Extending the time in which a HPSI component may remain 
inoperable does not constitute a change that could result in a new 
type of accident initiator than that previously identified. In 
addition, overall plant risk will be managed in accordance with the 
Maintenance Rule to help ensure continued safe plant operation.
    Therefore, the proposed change does not create the possibility 
of a new or different kind of accident from any previously 
evaluated.
    3. The proposed change does not involve a significant reduction 
in a margin of safety.
    The proposed change does not require any change in accident 
analysis methods or results. Overall plant risks will continue to be 
appropriately managed and compensatory measures established when 
appropriate to reduce the overall risk during extended HPSI CT 
periods. In addition, an evaluation of common cause failure and a 
determination of the flow capacity of remaining ECCS [emergency core 
cooling system] components will continue to be performed in relation 
to HPSI System inoperabilities. Although components important to 
safety have an impact on overall plant risk and may impact the 
overall margin to safety, the adverse impacts that are realized due 
to single HPSI subsystem inoperabilities is largely offset by the 
avoidance of unnecessary shutdown transition risks and the 
establishment of compensatory measures and contingency actions where 
appropriate.
    Therefore, the proposed change does not involve a significant 
reduction in a margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: James R. Curtiss, Esq., Winston & Strawn, 
1400 L Street, NW., Washington, DC 20005-3502.
    NRC Section Chief: Stephen Dembek.
    PPL Susquehanna, LLC, Docket Nos. 50-387 and 50-388, Susquehanna 
Steam Electric Station, Units 1 and 2 (SSES 1 and 2), Luzerne County, 
Pennsylvania
    Date of amendment request: May 11, 2004.
    Description of amendment request: The proposed amendment would 
revise the standby liquid control (SLC) pump discharge pressure 
surveillance requirement 3.1.7.7 acceptance criteria from 1224 psig to 
1395 psig in the SSES 1 and 2 Technical Specification 3.1.7.
    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 proposed change establishes the operability requirements 
for the SLC subsystem based on its functional capability to operate 
during an ATWS [anticipated transients without scram] event. This 
proposed change to the surveillance for SLC pump discharge pressure 
does not affect the operation of any other SSES SSC's [structures, 
systems and components]. The SLC system is already being tested on a 
quarterly basis to the proposed new pump discharge pressure to 
demonstrate that the In Service Inspection Program requirements are 
met.
    Consequently, the proposed change has no effect on the 
probability of any accident previously evaluated. Further, the 
consequences of any accident previously evaluated are not affected. 
Therefore, the proposed change does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    2. Does the proposed change create the possibility of a new or 
different kind of accident from any accident previously evaluated?
    No. The proposed change to the surveillance for SLC pump 
discharge pressure does not involve any physical alteration of the 
plant (no new or different type of equipment is installed) or 
changes in methods governing normal plant operation. Since this 
change does not introduce any new accident initiators, 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?
    No. The proposed change to the surveillance for SLC pump 
discharge pressure does not involve any physical alteration of the 
plant (no new or different type of equipment is installed) or 
changes in methods governing normal plant operation. The proposed 
change only affects determination of SLC system Technical 
Specification operability based on the functional capability of the 
SLC subsystems to inject boron during an ATWS event. 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: Bryan A. Snapp, Esquire, Assoc. General 
Counsel, PPL Services Corporation, 2 North Ninth St., GENTW3, 
Allentown, PA 18101-1179.
    NRC Section Chief: Richard J. Laufer.

Notice of Issuance of Amendments to Facility Operating Licenses

    During the period since publication of the last biweekly notice, 
the Commission has issued the following amendments. The Commission has 
determined for each of these amendments that the application complies 
with the standards and requirements of the Atomic Energy Act of 1954, 
as amended (the Act), and the Commission's rules and regulations. The 
Commission has made appropriate findings as required by the Act and the 
Commission's rules and regulations in 10 CFR chapter I, which are set 
forth in the license amendment.
    Notice of Consideration of Issuance of Amendment to Facility 
Operating License, Proposed No Significant Hazards Consideration 
Determination, and Opportunity for A Hearing in connection with these 
actions was published in the Federal Register as indicated.
    Unless otherwise indicated, the Commission has determined that 
these amendments satisfy the criteria for categorical exclusion in 
accordance with 10 CFR 51.22. Therefore, pursuant to 10 CFR 51.22(b), 
no environmental impact statement or environmental assessment need be 
prepared for these amendments. If the Commission has prepared an 
environmental assessment under the special circumstances provision in 
10 CFR 51.12(b) and has made a determination based on that assessment, 
it is so indicated.
    For further details with respect to the action see (1) the 
applications for amendment, (2) the amendment, and (3) the Commission's 
related letter, Safety Evaluation and/or Environmental Assessment as 
indicated. All of these items are available for public inspection

[[Page 40679]]

at the Commission's Public Document Room, located at One White Flint 
North, Public File Area 01F21, 11555 Rockville Pike (first floor), 
Rockville, Maryland. Publicly available records will be accessible from 
the Agencywide Documents Access and Management Systems (ADAMS) Public 
Electronic Reading Room on the internet at the NRC Web site, http://www.nrc.gov/reading-rm/adams.html. If you do not have access to ADAMS 
or if there are problems in accessing the documents located in ADAMS, 
contact the NRC Public Document Room (PDR) Reference staff at 1-800-
397-4209, (301) 415-4737 or by e-mail to [email protected].

    Energy Northwest, Docket No. 50-397, Columbia Generating Station, 
Benton County, Washington
    Date of application for amendment: November 4, 2003.
    Brief description of amendment: The amendment revises technical 
specification (TS) requirements for mode change limitations in Limiting 
Condition for Operation 3.0.4 and Surveillance Requirement 3.0.4 to 
adopt the provisions of Industry TS Task Force (TSTF) change TSTF-359, 
``Increase Flexibility in Mode Restraints.''
    Date of issuance: June 7, 2004.
    Effective date: June 7, 2004, and shall be implemented within 60 
days from the date of issuance.
    Amendment No.: 187.
    Facility Operating License No. NPF-21: The amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: December 9, 2003 (68 FR 
68662).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated June 7, 2004.
    No significant hazards consideration comments received: No.

    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: February 3, 2004.
    Brief description of amendment: The amendment revises Technical 
Specification 3.1.8, ``Scram Discharge Volume (SDV) Vent and Drain 
Valves,'' for the condition of having one or more SDV vent or drain 
lines with one valve inoperable.
    Date of issuance: June 16, 2004.
    Effective date: As of the date of issuance and shall be implemented 
60 days from the date of issuance.
    Amendment No.: 140.
    Facility Operating License No. NPF-47: The amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: March 30, 2004 (69 FR 
16619).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated June 16, 2004.
    No significant hazards consideration comments received: No.

    Entergy Nuclear Operations, Docket No. 50-247, Indian Point Nuclear 
Generating Unit No. 2, Westchester County, New York
    Date of application for amendment: October 21, 2003, as 
supplemented on March 1, 2004.
    Brief description of amendment: The amendment revises Technical 
Specification 5.5.7.b.1 regarding the maximum time interval between 
steam generator (SG) inspections. The amendment permits, on a one-time 
basis, the extension of the SG inspection interval such that the next 
SG inspection, which would have been required to be performed no later 
than November 17, 2004, to be deferred until June 17, 2006. This 
effectively extends the current inspection interval from a maximum of 
24 calendar months to 43 calendar months.
    Date of issuance: June 23, 2004.
    Effective date: As of the date of issuance to be implemented within 
30 days.
    Amendment No.: 239.
    Facility Operating License No. DPR-26: Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: December 9, 2003 (68 FR 
68663).
    The supplement dated March 31, 2004, provided additional 
information that clarified the application, did not expand the scope of 
the application as originally noticed, and did not change the staff's 
original proposed no significant hazards consideration determination. 
The Commission's related evaluation of the amendment is contained in a 
Safety Evaluation dated June 23, 2004.
    No significant hazards consideration comments received: No.

    Entergy Operations, Inc., Docket No. 50-382, Waterford Steam 
Electric Station, Unit 3, St. Charles Parish, Louisiana
    Date of amendment request: March 15, 2004.
    Brief description of amendment: The amendment relocates the 
Waterford Steam Electric Station, Unit 3 Technical Specification 
3.4.8.2, pressurizer heatup and cooldown limits, the associated action 
statements and surveillance requirements to the Technical Requirements 
Manual.
    Date of issuance: June 16, 2004.
    Effective date: As of the date of issuance and shall be implemented 
60 days from the date of issuance.
    Amendment No.: 195.
    Facility Operating License No. NPF-38: The amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: April 13, 2004 (69 FR 
19569).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated June 16, 2004.
    No significant hazards consideration comments received: No.

    Entergy Operations, Inc., Docket No. 50-382, Waterford Steam 
Electric Station, Unit 3, St. Charles Parish, Louisiana
    Date of amendment request: October 22, 2003.
    Brief description of amendment: The licensee is changing the 
existing pressure/temperature limits from 16 effective full power years 
(EFPY) to 32 EFPYs. In addition, the reactor coolant system maximum 
heatup and cooldown temperatures are changed to 60 [deg]F and 100 
[deg]F/hour, respectively. For inservice hydrostatic pressure and leak 
testing, the maximum heatup and cooldown rates are now 60 [deg]F and 
100 [deg]F respectively.
    Date of issuance: June 16, 2004.
    Effective date: As of the date of issuance and shall be implemented 
60 days from the date of issuance.
    Amendment No.: 196.
    Facility Operating License No. NPF-38: The amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: December 9, 2003 (68 FR 
68667).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated June 16, 2004.
    No significant hazards consideration comments received: No.

    Entergy Operations, Inc., System Energy Resources, Inc., South 
Mississippi Electric Power Association, and Entergy Mississippi, Inc., 
Docket No. 50-416, Grand Gulf Nuclear Station, Unit 1, Claiborne 
County, Mississippi
    Date of application for amendment: February 18, 2004, and 
supplemented by letter dated June 8, 2004.
    Brief description of amendment: The amendment deletes the 
requirements from the Technical Specifications to maintain hydrogen 
recombiners and hydrogen analyzers.
    Date of issuance: June 16, 2004.
    Effective date: As of the date of issuance and shall be implemented 
within 120 days of issuance.
    Amendment No: 166.
    Facility Operating License No. NPF-29: The amendment revises the 
Technical Specifications.

[[Page 40680]]

    Date of initial notice in Federal Register: March 16, 2004 (69 FR 
12366).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated June 16, 2004.
    No significant hazards consideration comments received: No.

    Exelon Generation Company, LLC, Docket No. STN 50-454, Byron 
Station, Unit No. 1, Ogle County, Illinois
    Date of application for amendment: December 5, 2003.
    Brief description of amendment: The amendment permits a change in 
the fuel rod-average-burnup limit from 60,000 MWD/MTU to 65,000 MWD/MTU 
for four lead test assemblies during Byron Station, Unit 1, Cycle 13.
    Date of issuance: June 16, 2004.
    Effective date: As of the date of issuance and shall be implemented 
within 30 days.
    Amendment No.: 137.
    Facility Operating License No. NPF-37: The amendment revised the 
License.
    Date of initial notice in Federal Register: January 20, 2004 (69 FR 
2742).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated June 16, 2004.
    No significant hazards consideration comments received: No.

Exelon Generation Company, LLC, and PSEG Nuclear LLC, Docket No. 50-
277, Peach Bottom Atomic Power Station, Unit 2,York and Lancaster 
Counties, Pennsylvania

    Date of application for amendments: February 12, 2004, as 
supplemented March 29, 2004.
    Brief description of amendment: This amendment revised Technical 
Specification (TS) Table 3.3.6.1-1, ``Primary Containment Isolation 
Instrumentation,'' to increase the TS Allowable Value related to the 
setpoint for the Main Steam Tunnel Temperature--High system isolation 
function for those instruments located within the Reactor Building. A 
new Function, 1.f, has been added to represent the Reactor Building 
Main Steam Tunnel Temperature--High. Function 1.e has been renamed to 
clarify that it represents only the Turbine Building Main Steam Tunnel 
Temperature--High.
    Date of issuance: June 16, 2004.
    Effective date: As of the date of issuance, to be implemented 
within 30 days.
    Amendment No.: 250.
    Renewed Facility Operating License No. DPR-44: The amendment 
revised the Technical Specifications.
    Date of initial notice in Federal Register: March 2, 2004 (69 FR 
9860).
    The March 29, 2004, letter provided clarifying information that did 
not change the initial proposed no significant hazards consideration 
determination or expand the application beyond the scope of the 
original Federal Register notice.
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated June 16, 2004.
    No significant hazards consideration comments received: No.

    Nuclear Management Company, LLC, Docket No. 50-331, Duane Arnold 
Energy Center, Linn County, Iowa
    Date of application for amendment: January 30, 2004.
    Brief description of amendment: The amendment deletes the Technical 
Specification requirements associated with the hydrogen and oxygen 
monitors.
    Date of issuance: June 10, 2004.
    Effective date: As of the date of issuance and shall be implemented 
within 120 days.
    Amendment No.: 254.
    Facility Operating License No. DPR-49: The amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: March 2, 2004 (69 FR 
9862).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated June 10, 2004.
    No significant hazards consideration comments received: No.

    Nuclear Management Company, LLC, Docket No. 50-305, Kewaunee 
Nuclear Power Plant, Kewaunee County, Wisconsin
    Date of application for amendment: October 8, 2003, as supplemented 
February 27 and May 3, 2004.
    Brief description of amendment: The amendment revises the Technical 
Specifications with a one-time change to allow a 40-month inspection 
interval after the first (post-replacement) steam generator inservice 
inspection, rather than after two consecutive inspections resulting in 
a C-1 classification.
    Date of issuance: June 18, 2004.
    Effective date: As of the date of issuance and shall be implemented 
within 30 days.
    Amendment No.: 175.
    Facility Operating License No. DPR-43: Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: January 20, 2004 (69 FR 
2743).
    The supplements dated February 27 and May 3, 2004, provided 
additional information that clarified the application, did not expand 
the scope of the application as originally noticed, and did not change 
staff's original proposed no significant hazards consideration 
determination. The Commission's related evaluation of the amendment is 
contained in a Safety Evaluation dated June 18, 2004.
    No significant hazards consideration comments received: No.

    Nuclear Management Company, LLC, Docket No. 50-255, Palisades 
Plant, Van Buren County, Michigan
    Date of application for amendment: January 29, 2004, as 
supplemented on May 14, and June 2, 2004.
    Brief description of amendment: The amendment grants approval to 
update the final safety analysis report (FSAR) to reflect the fuel pool 
building crane (L-3 crane) main hoist upgrade to the new rated capacity 
of 110 tons and reflect the new single-failure-proof design. 
Specifically, the amendment approves the use of the L-3 crane as a 
single-failure-proof crane for below-the-hook loads up to 110 tons.
    Date of issuance: June 16, 2004.
    Effective date: As of the date of issuance and shall be implemented 
within 60 days.
    Amendment No.: 215.
    Facility Operating License No. DPR-20. Amendment updated the FSAR.
    Date of initial notice in Federal Register: March 1, 2004 (69 FR 
9649).
    The supplemental letters contained clarifying information and did 
not change the initial no significant hazards consideration 
determination and did not expand the scope of the original Federal 
Register notice.
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated June 16, 2004.
    No significant hazards consideration comments received: No.
    Southern California Edison Company, et al., Docket Nos. 50-361 and 
50-362, San Onofre Nuclear Generating Station, Units 2 and 3, San Diego 
County, California
    Date of application for amendments: September 15, 2003.
    Brief description of amendments: The amendments revise Technical 
Specification (TS) 2.1.1.2 of TS Section 2.0, ``Safety Limits (SLs).'' 
The amendments replace the peak linear heat rate SL with a peak fuel 
centerline temperature SL so that the SL in TS 2.1.2.2 adequately 
conforms to 10 CFR 50.36(c)(1)(ii)(A) which requires that limiting 
safety system settings prevent a SL from being exceeded.
    Date of issuance: June 10, 2004.
    Effective date: June 10, 2004, to be implemented within 60 days of 
issuance.
    Amendment Nos.: Unit 2-192 ; Unit 3-183.
    Facility Operating License Nos. NPF-10 and NPF-15: The amendments 
revised the Technical Specifications.

[[Page 40681]]

    Date of initial notice in Federal Register: October 14, 2003 (68 FR 
59219).
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated June 10, 2004.
    No significant hazards consideration comments received: No.

    Tennessee Valley Authority, Docket Nos. 50-327 and 50-328, Sequoyah 
Nuclear Plant, Units 1 and 2, Hamilton County, Tennessee
    Date of application for amendment: October 22, 2003 (TS 03-12).
    Brief description of amendment: The amendments extend from 1 hour 
to 24 hours the completion time for Condition B of Technical 
Specification 3.5.1.1, which defines requirements for the restoration 
of an emergency core cooling system accumulator when it has been 
declared inoperable for a reason other than boron concentration.
    Date of issuance: June 18, 2004.
    Effective date: As of the date of issuance and shall be implemented 
within 45 days of issuance.
    Amendment Nos.: 291 and 281.
    Facility Operating License Nos. DPR-77 and DPR-79: Amendments 
revise the technical specifications.
    Date of initial notice in Federal Register: January 6, 2004 (69 FR 
699).
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated June 18, 2004.
    No significant hazards consideration comments received: No.

    Union Electric Company, Docket No. 50-483, Callaway Plant, Unit 1, 
Callaway County, Missouri
    Date of application for amendment: June 6, 2003, as supplemented by 
the letter dated December 19, 2003.
    Brief description of amendment: The amendment revises several 
surveillance requirements (SRs) in Technical Specifications (TSs) 3.8.1 
and 3.8.4 on alternating current and direct current sources, 
respectively, for plant operation. The revised SRs have notes deleted 
or modified to allow the SRs to be performed, or partially performed, 
in reactor modes that previously were not allowed by the TSs. The 
licensee withdrew the changes to SRs 3.8.4.7 and 3.8.4.8 in its letter 
dated April 14, 2004.
    Date of issuance: June 14, 2004.
    Effective date: June 14, 2004, and shall be implemented within 60 
days of the date of issuance including the incorporation of the changes 
to the Technical Specification Bases as described in the licensee's 
letters dated June 6 and December 19, 2003, and April 14, 2004.
    Amendment No.: 162.
    Facility Operating License No. NPF-30: The amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: July 22, 2003 (68 FR 
43394).
    The December 19, 2003, and April 14, 2004, supplemental letters 
provided additional clarifying information, did not expand the scope of 
the application as noticed and did not change the staff's original 
proposed no significant hazards consideration determination.
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated June 14, 2004.
    No significant hazards consideration comments received: No.

    Wolf Creek Nuclear Operating Corporation, Docket No. 50-482, Wolf 
Creek Generating Station, Coffey County, Kansas
    Date of amendment request: February 9, 2004.
    Brief description of amendment: The amendment revises TS 5.5.7, 
``Reactor Coolant Pump Flywheel Inspection Program,'' to increase the 
inspection interval from 10 years to 20 years.
    Date of issuance: June 16, 2004.
    Effective date: June 16, 2004, and shall be implemented within 90 
days from the date of issuance.
    Amendment No.: 153.
    Facility Operating License No. NPF-42. The amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: March 16, 2004 (69 FR 
12373).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated June 16, 2004.
    No significant hazards consideration comments received: No.

    Dated at Rockville, Maryland, this 28th day of June 2004.

    For the Nuclear Regulatory Commission.
Edwin M. Hackett,
Acting Deputy Director, Division of Licensing Project Management, 
Office of Nuclear Reactor Regulation.
[FR Doc. 04-15061 Filed 7-2-04; 8:45 am]
BILLING CODE 7590-01-P