[Federal Register Volume 69, Number 177 (Tuesday, September 14, 2004)]
[Notices]
[Pages 55466-55478]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: 04-20497]


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NUCLEAR REGULATORY COMMISSION


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

I. Background

    Pursuant to section 189a. (2) of the Atomic Energy Act of 1954, as 
amended (the Act), the U.S. Nuclear Regulatory Commission (the 
Commission or NRC staff) is publishing this regular biweekly notice. 
The Act requires the Commission publish notice of any amendments 
issued, or proposed to be issued and grants the Commission the 
authority to issue and make immediately effective any amendment to an 
operating license upon a determination by the Commission that such 
amendment involves no significant hazards consideration, 
notwithstanding the pendency before the Commission of a request for a 
hearing from any person.
    This biweekly notice includes all notices of amendments issued, or 
proposed to be issued from, August 20, 2004, through September 2, 2004. 
The last biweekly notice was published on August 31, 2004, (69 FR 
53098).

Notice of Consideration of Issuance of Amendments to Facility Operating 
Licenses, Proposed No Significant Hazards Consideration Determination, 
and Opportunity for a Hearing

    The Commission has made a proposed determination that the following 
amendment requests involve no significant hazards consideration. Under 
the Commission's regulations in 10 CFR 50.92, this means that operation 
of the facility in accordance with the proposed amendment would not (1) 
involve a significant increase in the probability or consequences of an 
accident previously evaluated; or (2) create the possibility of a new 
or different kind of accident from any accident previously evaluated; 
or (3) involve a significant reduction in a margin of safety. The basis 
for this proposed determination for each amendment request is shown 
below.
    The Commission is seeking public comments on this proposed 
determination. Any comments received within 30 days after the date of 
publication of this notice will be considered in making any final 
determination. Within 60 days after the date of publication of this 
notice, the licensee may file a request for a hearing with respect to 
issuance of the amendment to the subject facility operating license and 
any person whose interest may be affected by this proceeding and who 
wishes to participate as a party in the proceeding must file a written 
request for a hearing and a petition for leave to intervene.
    Normally, the Commission will not issue the amendment until the 
expiration of 60 days after the date of publication of this notice. The 
Commission may issue the license amendment before expiration of the 60-
day period provided that its final determination is that the amendment 
involves no significant hazards consideration. In addition, the 
Commission may issue the amendment prior to the expiration of the 30-
day comment period should circumstances change during the 30-day 
comment period such that failure to act in a timely way would result, 
for example in derating or shutdown of the facility. Should the 
Commission take action prior to the expiration of either the comment 
period or the notice period, it will publish in the Federal Register a 
notice of issuance. Should the Commission make a final No Significant 
Hazards Consideration Determination, any hearing will take place after 
issuance. The Commission expects that the need to take this action will 
occur very infrequently.
    Written comments may be submitted by mail to the Chief, Rules and 
Directives Branch, Division of Administrative Services, Office of 
Administration, U.S. Nuclear Regulatory Commission, Washington, DC 
20555-0001, and should cite the publication date and page number of 
this Federal Register notice. Written comments may also be delivered to 
Room 6D22, Two White Flint North, 11545 Rockville Pike, Rockville, 
Maryland, from 7:30 a.m. to 4:15 p.m. Federal workdays. Copies of 
written comments received may be examined at the Commission's Public 
Document Room (PDR), located at One White Flint North, Public File Area 
O1F21, 11555 Rockville Pike (first floor), Rockville, Maryland. 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

[[Page 55467]]

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 email 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 email to 
[email protected].

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

    Date of amendment request: August 27, 2004.
    Description of amendment request: The licensee proposed to amend 
the Oyster Creek Nuclear Generating Station (OCNGS) Technical 
Specifications (TSs) regarding the safety limit minimum critical power 
ratio (SLMCPR) to reflect the results of cycle-specific calculations 
performed for the next fuel cycle (i.e., Cycle 20), using Nuclear 
Regulatory Commission (NRC)-approved methodology documented in Topical 
Report NEDE-24011-P-A-14, ``General Electric Standard Application for 
Reactor Fuel'' (GESTAR II), updated to Amendment 25. Specifically, the 
licensee proposed to revise TS Section 2.1.A, changing the SLMCPR 
values from 1.10 to 1.12 for three-recirculation-loop operation, and 
from 1.09 to 1.10 for four- or five-recirculation-loop operation.
    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 has reviewed the licensee's analysis 
against the standards of 10 CFR 50.92(c). The NRC staff's review is 
presented below:
    (1) Does the proposed amendment involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    No. SLMCPR values, and their derivation using NRC-approved methods, 
do not change the design or operating procedures of OCNGS, and have no 
role on the occurrence of an initiating event of an accident or 
transient. The basis of the SLMCPR is to ensure no mechanistic fuel 
damage will occur if the limit is not violated. The new SLMCPR values 
will preserve the existing margin to transition boiling (i.e., in the 
event of an accident or transient, the amount of fuel damaged would not 
be increased as a result of the new SLMCPR values). Furthermore, the 
proposed new SLMCPR values do not lead to, nor do they arise as a 
result of, plant design or procedural changes. Therefore, the proposed 
changes do not involve a significant increase in the probability or 
consequences of an accident previously evaluated.
    (2) Does the proposed amendment create the possibility of a new or 
different kind of accident from any accident previously evaluated?
    No. The new SLMCPR values for OCNGS Cycle 20 core have been 
calculated in accordance with the methods and procedures described in 
an NRC-approved topical report. The proposed new SLMCPR values do not 
lead to, nor do they arise as a result of, plant design or procedural 
changes. The changes do not involve any new method for operating the 
facility and do not involve any facility modifications. As a result, no 
new initiating events or transients could develop from the proposed 
changes. Therefore, the proposed TS 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 margin of safety as defined in OCNGS's licensing basis will 
remain the same. The new cycle-specific SLMCPR values are calculated 
using NRC-approved methods and procedures that are in accordance with 
the current fuel

[[Page 55468]]

design and licensing criteria. The SLMCPR values will remain high 
enough to ensure that greater than 99.9% of all fuel rods in the core 
are expected to avoid transition boiling if the limits are not 
violated, thereby preserving the fuel cladding integrity. Therefore, 
the proposed amendment does not involve a significant reduction in a 
margin of safety.
    Based on the above review, 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.

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

    Date of amendment request: May 27, 2004.
    Description of amendment request: The proposed amendment would 
revise the Technical Specifications (TSs) based on the radiological 
dose analysis margins obtained by using an alternative source term 
consistent with 10 CFR 50.67. Specifically, the amendment would revise 
TS 3/4.7.7, ``Control Room Emergency Air Filtration System,'' 
surveillance requirements and delete TS 3/4.7.8, ``Control Room 
Envelope Pressurization 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. Involve a significant increase in the probability or 
consequences of an accident previously evaluated.
    The proposed amendment does not involve a significant increase 
in the probability or consequence of an accident previously 
analyzed. The Millstone Unit 3 Control Room Emergency Air Filtration 
System only functions following the initiation of a design basis 
radiological accident. Therefore, the change to the value used for 
methyl iodide penetration test acceptance criteria following a 
design basis accident will not increase the probability of any 
previously analyzed accident. The Millstone Unit 3 Control Room 
Envelope Pressurization System is no longer credited in the accident 
analyses described in the Alternative Source Term (AST) 
implementation analyses. In accordance with AST implementation 
analyses, the requirements contained in this Specification do not 
meet any of 10 CFR 50.36(c)(2)(ii) criteria on items for which 
Technical Specifications must be established. Deletion of this 
Technical Specification will not increase the probability of 
occurrence of any previously analyzed accident and does not impact 
the consequences of any evaluated accident since it is no longer 
analytically credited. The Millstone Unit 3 containment and the 
containment systems function to prevent or control the release of 
radioactive fission products following a postulated accident. 
Therefore, the change to the value used for the leakage rate 
acceptance criteria for all penetrations that are secondary 
containment bypass leakage paths following a design basis accident 
will not increase the probability of any previously analyzed 
accident and is limited to ensure it does not increase any accident 
consequence.
    These systems are not initiators of any design bases accident. 
Revised dose calculations, which take into account the changes 
proposed by this amendment and the use of the alternative source 
term, have been performed for the Millstone Unit 3 design basis 
radiological accidents. The results of these revised calculations 
indicate that public and control room doses will not exceed the 
limits specified in 10 CFR 50.67 and Regulatory Guide 1.183. There 
is not a significant increase in predicted dose consequences for any 
of the analyzed accidents. Therefore, the proposed changes do not 
involve a significant increase in the consequences of any previously 
analyzed accident.
    2. Create the possibility of a new or different kind of accident 
from any accident previously evaluated.
    The implementation of the proposed changes does not create the 
possibility of an accident of a different type than was previously 
evaluated in the UFSAR [updated final safety report]. Although the 
proposed changes could affect the operation of the Control Room 
Emergency Air Filtration System, and containment and the containment 
systems following a design basis radiological accident, none of 
these changes can initiate a new or different kind of accident since 
they are only related to system capabilities that provide protection 
from accidents that have already occurred. These changes do not 
alter the nature of events postulated in the UFSAR nor do they 
introduce any unique precursor mechanisms. Therefore, the proposed 
changes do not create the possibility of a new or different kind of 
accident from those previously analyzed.
    3. Involve a significant reduction in the margin of safety.
    The implementation of the proposed changes does not reduce the 
margin of safety. The proposed changes for the Control Room 
Emergency Air Filtration System, and containment and the containment 
systems do not affect the ability of these systems to perform their 
intended safety functions to maintain dose less than the required 
limits during design basis radiological events. The revised dose 
calculations also indicate that the change to the containment 
depressurization times will continue to maintain the dose to the 
public and control room operators less than the required limits. The 
radiological analysis results, when compared with the revised TEDE 
acceptance criteria, meet the applicable limits. These acceptance 
criteria have been developed for application to analyses performed 
with alternative source terms. These acceptance criteria have been 
developed for the purpose of use in design basis accident analyses 
such that meeting the stated limits demonstrates adequate protection 
of public health and safety. It is thus concluded that the margin of 
safety will not be reduced by the implementation of the changes.

    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, et al., Docket Nos. 50-413 and 50-414, Catawba 
Nuclear Station, Units 1 and 2, York County, South Carolina

    Date of amendment request: March 22, 2004.
    Description of amendment request: The amendments would revise the 
Catawba Nuclear Station Facility Operating Licenses and Technical 
Specifications (TSs) to change the surveillance frequency on selected 
Engineered Safety Features Actuation System (ESFAS) slave relays from 
92 days to 18 months.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    (1) The proposed license amendments do not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    This change to the TS does not result in a condition where the 
design, material, and construction standards that were applicable 
prior to the change are altered. Only the slave relay test interval 
is changed. The proposed change will not modify any system interface 
and could not increase the likelihood of an accident since these 
events are independent of this change. The proposed activity will 
not change, degrade, or prevent actions or alter any assumptions 
previously made in evaluating the radiological consequences of an 
accident described in the UFSAR [Updated Final Safety Analysis 
Report]. Therefore, the proposed amendments do not result in any 
increase in the probability or consequences of an accident 
previously evaluated.
    (2) The proposed license amendments do not create the 
possibility of a new or different

[[Page 55469]]

kind of accident from any accident previously evaluated.
    This change does not alter the performance of the affected 
systems. The slave relays will still be tested every 18 months. 
Changing the surveillance frequency for the slave relays will not 
create any new accident initiators or scenarios. Periodic 
surveillance of these instruments will detect significant 
degradation in the channel characteristic. Implementation of the 
proposed amendments does not create the possibility of a new or 
different kind of accident from any accident previously evaluated.
    (3) The proposed license amendments do not involve a significant 
reduction in a margin of safety.
    The surveillance test frequency is relaxed for certain slave 
relays because of demonstrated high reliability of the relay and its 
insensitivity to any short term wear or aging effects. Based on the 
above, it is concluded that the proposed license amendment request 
does not result in a reduction in a margin with respect to plant 
safety.

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

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

    Date of amendment request: April 6, 2004.
    Description of amendment request: The amendments would revise the 
Catawba Nuclear Station Technical Specifications (TSs) to allow a 
diesel generator battery to remain operable with no more than one cell 
less than 1.36 Volts DC on float charge.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    (1) Does the change involve a significant increase in the 
probability or consequences of an accident previously evaluated?
    No. The DC electrical power system provides normal and emergency 
DC electrical power for the diesel generators, emergency 
auxiliaries, and control and switching during all modes of 
operation. This change will not affect or degrade the ability of the 
DC Electrical Power Systems to perform their specified safety 
function.
    The only effect on systems, structures and components (SSCs) by 
this change is that one DG battery with one cell less than 1.36 
volts the system will still be considered operable. With one or more 
DG batteries with one or more battery cell(s) not within limits of 
level or temperature, sufficient capacity to supply the required 
load for the DG is not assumed, and the corresponding DC electrical 
power subsystem must be declared inoperable immediately. With one or 
more DG batteries with two or more battery cells not within limits 
of voltage, sufficient capacity to supply the required load for the 
DG is not assumed, and the corresponding DC electrical power 
subsystem must be declared inoperable immediately.
    Surveillance (SR) 3.8.4.2 is being relocated to TS 3.8.6 as a 
new surveillance and the wording of the Bases section is being 
revised for clarity as follows: ``For this surveillance, a minimum 
of two cells shall be tested every seven days. The cells selected 
for testing shall be rotated on a monthly basis.'' The new SR 
3.8.6.5 will check the DG battery cell voltage on selected cells to 
ensure they are greater than or equal to 1.36 volts on a seven day 
frequency. This test will continue to assure that the batteries are 
available to perform their design functions.
    This amendment will not change any previously evaluated 
accidents such as ``Loss of Non-Emergency AC Power to Station 
Auxiliaries (Blackout)'', ``Loss of Coolant Accident (LOCA),'' and 
``LOCA/Blackout.'' The prevention and mitigation of these accidents 
is also not affected by this change.
    The likelihood of a malfunction of the batteries is not 
increased by this change in the surveillances. The systems will 
continue to be able to perform their design functions of supplying 
emergency power during the evaluated accidents listed above. 
Therefore, the changes do not involve a significant increase in the 
probability or consequences of an accident previously evaluated.
    (2) Does the change create the possibility of a new or different 
kind of accident from any accident previously evaluated?
    No. This change does not involve a physical alteration to the 
plant (i.e., no new or different type of equipment will be 
installed) or a change in the methods governing normal plant 
operation. The change does not alter assumptions made in the safety 
analysis or licensing basis. This change will not affect or degrade 
the ability of the DC Electrical Power Systems to perform their 
specified safety function. Therefore, the change does not create the 
possibility of a new or different kind of credible accident from any 
accident previously evaluated.
    (3) Does the proposed change involve a significant reduction in 
a margin of safety?
    No. Assuming that one cell in a 94-cell battery is at a full-
reverse voltage of -1.80V, the remaining cells would be required to 
supply 106.80V, or 1.1484V/cell, in order to maintain a minimum 
battery terminal voltage of 105.0V. The manufacturer has 
extrapolated new sizing factors for an end-voltage of 1.1484V and 
used the new sizing factors to recalculate the battery capacity 
required to satisfy the design basis requirements. The load profile 
data and sizing methodology was taken from 125 Vdc Diesel Auxiliary 
Power Battery Sizing Calculations. Considering all possible loading 
scenarios, the minimum capacity margin available with one cell 
assumed to be in full reversal (-1.80V) was calculated to be 34%. 
This assumes the battery is at an end-of-life capacity of 80%, the 
electrolyte temperature is at the design-minimum of 60 [deg]F, and 
that no cells are jumpered out.
    Based on the discussion above and the results of the battery 
sizing calculations, a DG battery remains operable and fully capable 
of satisfying its design requirements with one cell < 1.36V on an 
indefinite basis. Therefore, the proposed changes listed above do 
not involve a significant reduction in a margin of safety.

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

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

    Date of amendment request: September 29, 2003.
    Description of amendment request: The proposed amendment would 
revise the Technical Specification (TS) 3.7.15 spent fuel pool (SFP) 
storage criteria based upon fuel type, fuel enrichment, burnup, cooling 
time and partial credit for soluble boron in the SFP. This amendment 
allows for the safe storage of fuel assemblies with a nominal 
enrichment of Uranium-235 up to 5.00 weight percent. In addition, this 
amendment decreases the required soluble boron credit, which provides 
an acceptable margin of subcriticality in the McGuire Nuclear Station 
(McGuire), Units 1 and 2, spent fuel storage pools.
    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

    This license amendment transitions the McGuire SFP from 
conformance with a

[[Page 55470]]

temporary exemption to 10 CFR 70.24 to compliance with 10 CFR 
50.68(b). This regulation requires that the SFP remain subcritical 
if flooded with unborated water and remain 5 percent subcritical 
with credit for soluble boron. The SFP will be maintained with a 
minimum TS required soluble boron concentration that would provide 
substantial margin to criticality. The criticality analysis takes 
into consideration fuel type, fuel enrichment, fuel burnup, spent 
fuel cooling time and partial credit for soluble boron.
    There is no significant increase in the probability or 
consequence of a fuel assembly drop accident in the SFP as a result 
of this amendment. The method of handling fuel assemblies in the SFP 
is not affected by the changes made to the criticality analysis for 
the SFP or by the TS changes. The handling of fuel assemblies during 
normal operation is unchanged, since the same equipment and 
procedures will be used.
    There is no significant increase in the probability or 
consequence of the accidental misloading of spent fuel assemblies. 
Fuel assembly placement and storage will be controlled in accordance 
with approved fuel handling procedures and other approved processes 
to ensure compliance with the TS requirements. Analyses demonstrate 
that the pool will remain subcritical following an accidental 
misloading because the SFP contains an adequate margin of soluble 
boron concentration.
    The mitigating actions as the result of a loss of SFP cooling 
are not changed. The heat up rate in the SFP is a nearly linear 
function of the fuel decay heat load. The fuel decay heat load will 
not be significantly affected since the number of fuel assemblies 
and the fuel burnups are unchanged. In the unlikely event that all 
pool cooling is lost, sufficient time will still be available for 
the operators to provide alternate means of cooling before the onset 
of pool boiling.
    A decrease in pool water temperature from a large emergency 
makeup would cause an increase in water density, increasing fuel 
bundle reactivity. However, the margin provided by the TS required 
minimum boron concentration, above the concentration required to 
maintain 5 percent subcritical, will compensate for the increased 
fuel bundle reactivity which could result from a decrease in SFP 
water temperature.

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

    This license amendment regarding fuel storage requirements, 
nominal fuel enrichment, and the credit for soluble boron in the SFP 
specified by TS 4.3 will have no effect on normal pool operations 
and maintenance. There are no changes in equipment design or in 
plant configuration.
    Criticality and other SFP accidents have been analyzed in the 
McGuire's Updated Final Safety Analysis Report and Criticality 
Analysis reports. Specific accidents considered and evaluated 
include fuel assembly drop, accidental misloading, and significant 
changes in SFP water temperature. Region 1 of the SFP for both units 
had previously been updated with new replacement in-kind fuel racks 
utilizing boral neutron poison. As a result of this amendment no 
credit will be taken for the degrading boraflex neutron poison in 
Region 2 of the SFP.
    Therefore, the proposed amendment will not result in the 
possibility of a new or different kind of accident previously 
evaluated.

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

    The proposed TS changes and the resulting spent fuel storage 
operating limits will provide adequate safety margin to ensure that 
the stored fuel assembly array will always remain subcritical. Those 
limits are based on a plant-specific criticality analysis. This 
methodology takes partial credit for soluble boron in the SFP and 
requires conformance with 10 CFR 50.68(b).
    Therefore, the proposed changes in this license amendment will 
not result in a significant reduction in the facility's margin of 
safety.

    Therefore, the NRC staff proposes to determine that the amendment 
request involves no significant hazards consideration.
    Attorney for licensee: Ms. Lisa F. Vaughn, Duke Energy Corporation, 
422 South Church Street, Charlotte, North Carolina 28201-1006.
    NRC Section Chief: Mary Jane Ross-Lee, Acting.

Entergy Operations, Inc., Docket No. 50-313, Arkansas Nuclear One, Unit 
No. 1, Pope County, Arkansas

    Date of amendment request: June 24, 2004.
    Description of amendment request: The proposed amendment would 
modify the Safety Analysis Report (SAR) by increasing the maximum 
hypothetical accident (MHA) doses to the control room operators, due to 
an increase in the allowable unfiltered in-leakage into the control 
room envelope. However, the new MHA doses would still be within NRC-
approved guidance.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Does the proposed change involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    Response: No.
    The proposed changes adopt new dose acceptance criteria in 
Regulatory Guide 1.195 for calculating radiological consequences of 
design basis accidents. The proposed change increases the allowable 
unfiltered inleakage to 52 scfm [standard cubic feet per minute] 
which increases the licensing basis thyroid doses for ANO [Arkansas 
Nuclear One] operators to 49.9 rem for the ANO-1 [Arkansas Nuclear 
One, Unit 1] Safety Analysis Report MHA. The new MHA doses are 
within NRC approved guidance. The proposed change does not impact 
the probability of an accident previously evaluated in the SAR.
    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 accident analysis performed in establishing [the] new 
control room unfiltered inleakage value of 52 scfm were primarily 
performed using the existing licensing basis for the ANO-1 SAR. 
However, a new thyroid dose acceptance criterion of 50 rem was used 
per Regulatory Guide 1.195 instead of the previous Standard Review 
Plan thyroid dose limit of 30 rem. Dose consequences of non-LOCA 
[non-loss-of-coolant accident] events (except for the Fuel Handling 
Accident) were not historically calculated in the ANO-1 SAR. The 
doses had been assumed to be a fraction of the doses resulting from 
the MHA. New analyses of these control room doses confirmed them to 
be bounded by the revised MHA control room doses.
    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.
    Even though the ANO-1 SAR reported doses for the MHA are being 
increased in the proposed change, they are still within the NRC 
acceptance criteria of Regulatory Guide 1.195. Other assumptions are 
consistent with the current ANO-1 licensing basis or previously NRC 
approved assumptions within the industry. The increase in allowable 
in leakage by the proposed change maintains the operator doses 
within GDC [General Design Criteria] 19 limits with no compensatory 
measures to reduce thyroid uptake.
    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: Nicholas S. Reynolds, Esquire, Winston and 
Strawn, 1400 L Street, NW., Washington, DC 20005-3502.
    NRC Section Chief: Robert A. Gramm.

[[Page 55471]]

FPL Energy Seabrook, LLC, Docket No. 50-443, Seabrook Station, Unit No. 
1, Rockingham County, New Hampshire

    Date of amendment request: June 28, 2004.
    Description of amendment request: The proposed amendment deletes 
the requirements from the technical specifications (TSs) to maintain 
hydrogen recombiners and hydrogen monitors. Licensees were generally 
required to implement upgrades as described in NUREG-0737, 
``Clarification of TMI [Three Mile Island] Action Plan Requirements,'' 
and Regulatory Guide (RG) 1.97, ``Instrumentation for Light-Water-
Cooled Nuclear Power Plants to Assess Plant and Environs Conditions 
During and Following an Accident.'' Implementation of these upgrades 
was an outcome of the lessons learned from the accident that occurred 
at TMI Unit 2. Requirements related to combustible gas control were 
imposed by Order for many facilities and were added to or included in 
the TSs for nuclear power reactors currently licensed to operate. The 
revised 10 CFR 50.44, ``Standards for Combustible Gas Control System in 
Light-Water-Cooled Power Reactors,'' eliminated the requirements for 
hydrogen recombiners and relaxed safety classifications and licensee 
commitments to certain design and qualification criteria for hydrogen 
and oxygen monitors.
    The NRC staff issued a notice of availability of a model no 
significant hazards consideration determination (NSHC) for referencing 
in license amendment applications in the Federal Register on September 
25, 2003 (68 FR 55416). The licensee affirmed the applicability of the 
model NSHC determination in its application dated June 28, 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 revised 10 CFR 50.44 no longer defines a design-basis loss-
of-coolant accident (LOCA) hydrogen release, and eliminates 
requirements for hydrogen control systems to mitigate such a 
release. The installation of hydrogen recombiners and/or vent and 
purge systems required by 10 CFR 50.44(b)(3) was intended to address 
the limited quantity and rate of hydrogen generation that was 
postulated from a design-basis LOCA. The Commission has found that 
this hydrogen release is not risk-significant because the design-
basis LOCA hydrogen release does not contribute to the conditional 
probability of a large release up to approximately 24 hours after 
the onset of core damage. In addition, these systems were 
ineffective at mitigating hydrogen releases from risk-significant 
accident sequences that could threaten containment integrity.
    With the elimination of the design-basis LOCA hydrogen release, 
hydrogen monitors are no longer required to mitigate design-basis 
accidents and, therefore, the hydrogen monitors do not meet the 
definition of a safety-related component as defined in 10 CFR 50.2. 
Category 1 in RG 1.97 is intended for key variables that most 
directly indicate the accomplishment of a safety function for 
design-basis accident events. The hydrogen monitors no longer meet 
the definition of Category 1 in RG 1.97. As part of the rulemaking 
to revise 10 CFR 50.44 the Commission found that Category 3, as 
defined in RG 1.97, is an appropriate categorization for the 
hydrogen monitors because the monitors are required to diagnose the 
course of beyond design-basis accidents.
    The regulatory requirements for the hydrogen monitors can be 
relaxed without degrading the plant emergency response. The 
emergency response, in this sense, refers to the methodologies used 
in ascertaining the condition of the reactor core, mitigating the 
consequences of an accident, assessing and projecting offsite 
releases of radioactivity, and establishing protective action 
recommendations to be communicated to offsite authorities. 
Classification of the hydrogen monitors as Category 3, and removal 
of the hydrogen monitors from TS will not prevent an accident 
management strategy through the use of the severe accident 
management guidelines (SAMGs), the emergency plan (EP), the 
emergency operating procedures (EOP), and site survey monitoring 
that support modification of emergency plan protective action 
recommendations (PARs).
    Therefore, the elimination of the hydrogen recombiner 
requirements and relaxation of the hydrogen monitor requirements, 
including removal of these requirements from TS, does not involve a 
significant increase in the probability or the consequences of any 
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 elimination of the hydrogen recombiner requirements and 
relaxation of the hydrogen monitor requirements, including removal 
of these requirements from TS, will not result in any failure mode 
not previously analyzed. The hydrogen recombiner and hydrogen 
monitor equipment was intended to mitigate a design-basis hydrogen 
release. The hydrogen recombiner and hydrogen monitor equipment are 
not considered accident precursors, nor does their existence or 
elimination have any adverse impact on the pre-accident state of the 
reactor core or post accident confinement of radionuclides within 
the containment building.
    Therefore, this change does not create the possibility of a new 
or different kind of accident from any previously evaluated.

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

    The elimination of the hydrogen recombiner requirements and 
relaxation of the hydrogen monitor requirements, including removal 
of these requirements from TS, in light of existing plant equipment, 
instrumentation, procedures, and programs that provide effective 
mitigation of and recovery from reactor accidents, results in a 
neutral impact to the margin of safety.
    The installation of hydrogen recombiners and/or vent and purge 
systems required by 10 CFR 50.44(b)(3) was intended to address the 
limited quantity and rate of hydrogen generation that was postulated 
from a design-basis LOCA. The Commission has found that this 
hydrogen release is not risk-significant because the design-basis 
LOCA hydrogen release does not contribute to the conditional 
probability of a large release up to approximately 24 hours after 
the onset of core damage.
    Category 3 hydrogen monitors are adequate to provide rapid 
assessment of current reactor core conditions and the direction of 
degradation while effectively responding to the event in order to 
mitigate the consequences of the accident. The intent of the 
requirements established as a result of the TMI Unit 2 accident can 
be adequately met without reliance on safety-related hydrogen 
monitors.
    Therefore, this change does not involve a significant reduction 
in the margin of safety. Removal of hydrogen monitoring from TS will 
not result in a significant reduction in their functionality, 
reliability, and availability.

    The NRC staff proposes to determine that the amendment request 
involves no significant hazards consideration.
    Attorney for licensee: M. S. Ross, Florida Power & Light Company, 
P.O. Box 14000, Juno Beach, FL 33408-0420.
    Acting NRC Section Chief: Daniel S. Collins.

PSEG Nuclear LLC, Docket No. 50-354, Hope Creek Generating Station, 
Salem County, New Jersey

    Date of amendment request: June 7, 2004.
    Description of amendment request: The proposed changes would 
reflect an expanded operating domain resulting from implementation of 
Average Power Range Monitor/Rod Block Monitor/Technical Specifications/
Maximum Extended Load Line Limit Analysis (ARTS/MELLLA). The average 
power range monitor (APRM) flow-biased flux scram setpoint and the APRM 
and rod block monitor (RBM) flow-biased rod block trip setpoints would 
be revised to permit operation in the MELLLA region. In addition, the 
APRM scram and rod

[[Page 55472]]

block trip setdown requirement would be replaced by more direct power 
and flow-dependent thermal limits to reduce the need for APRM gain 
adjustments and to allow more direct thermal limits administration 
during operation at other than rated conditions. The amendment would 
also change the methods used to evaluate annulus pressurization and jet 
loads resulting from the postulated recirculation suction line break.
    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 has reviewed the licensee's analysis 
against the standards of 10 CFR 50.92(c). The staff's review is 
presented below:
    1. Does the proposed change involve a significant increase in the 
probability or consequences of an accident previously evaluated?

Proposed Change of APRM/RBM Setpoints

    The APRM and RBM are not involved in the initiation of any accident 
and the APRM flow-biased simulated thermal power scram and rod block 
functions are not credited in any Hope Creek Generating Station safety 
analyses. The revised evaluation of the rod withdrawal error event will 
continue to demonstrate acceptable results without crediting operation 
of the RBM. Therefore, the proposed change would have no effect on the 
probability of an accident previously evaluated, and the increase in 
consequences of a previously-evaluated accident, if any, would not be 
significant.

Proposed Replacement of APRM Scram and Rod Block Trip Setdown 
Requirements by More Direct Power and Flow Dependent Thermal Limits

    Neither the APRM scram and rod block setdown requirements, nor the 
power and flow-dependent thermal limits have any impact on accident 
initiating mechanisms. Adjustments to the thermal limits will be made 
using NRC-approved methods such that the fuel thermal and mechanical 
design bases will be maintained. Therefore, the proposed change will 
have no effect on the probability of an accident previously evaluated, 
and because the design bases will be maintained, an increase in the 
consequences of a previously-evaluated accident, if any, would not be 
significant.

Proposed Change in the Methods Used To Evaluate Annulus Pressurization 
and Jet Loads Resulting From the Postulated Recirculation Suction Line 
Break

    The proposed change would modify the method of accident analysis 
for selected scenarios, and as such could have no impact on the 
probability of an accident previously evaluated. Since the loads 
resulting from the recirculation suction line break are demonstrated to 
be bounded by the current licensing basis, the increase in consequences 
of a previously-evaluated accident, if any, would not be significant.
    2. Does the proposed change create the possibility of a new or 
different kind of accident from any accident previously evaluated?

Proposed Change of APRM/RBM Setpoints

    Changing the formulation of the flow-biased APRM rod block and 
scram trip setpoints and the RBM flow biased rod block trip setpoint 
would not change their respective functions and manner of operation. 
The change would not introduce a sequence of events or introduce a new 
failure mode that would create a new or different type of accident. 
Operating within the expanded power flow map would not require any 
systems, structures or components to function differently. Therefore, 
the proposed change would not create the possibility of a new or 
different kind of accident from any accident previously evaluated.

Proposed Replacement of APRM Scram and Rod Block Trip Setdown 
Requirements by More Direct Power and Flow Dependent Thermal Limits

    The replacement of the APRM scram and rod block trip setdown 
requirements by power and flow dependent thermal limits will continue 
to maintain the mechanical and thermal fuel design bases. Given that 
these design bases will be maintained, the proposed change would not 
create the possibility of a new or different kind of accident from any 
previously evaluated.

Proposed Change in the Methods Used To Evaluate Annulus Pressurization 
and Jet Loads Resulting From the Postulated Recirculation Suction Line 
Break

    The proposed change to the methods of analysis does not change the 
design function or operation of any plant equipment. Therefore, the 
proposed change would 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?

Proposed Change of APRM/RBM Setpoints

    The minimum critical power ratio (MCPR) and maximum average planar 
linear heat generation rate (MAPLHGR) thermal limits will be developed 
to ensure that the fuel thermal and mechanical design bases shall be 
maintained. Operation in the expanded operating domain would not alter 
the manner in which safety limits, limiting safety system settings, or 
limiting conditions for operation are determined. Given that the 
proposed change will continue to meet the current design basis, any 
reduction in a margin of safety would not be significant.

Proposed Replacement of APRM Scram and Rod Block Trip Setdown 
Requirements by More Direct Power and Flow Dependent Thermal Limits

    Replacement of the APRM setpoint requirements with power- and flow-
dependent adjustments to the MCPR and MAPLHGR or LHGR thermal limits 
will continue to ensure that margins to the fuel cladding safety limit 
are preserved during operation at other than rated conditions. The fuel 
cladding safety limit will continue to be bounding for any anticipated 
operational occurrence. The flow and power dependent adjustments will 
continue to ensure that all fuel thermal and mechanical design bases 
shall remain bounding. The 10 CFR 50.46 acceptance criteria for the 
performance of the emergency core cooling system following postulated 
loss-of-coolant accidents will continue to be met. Therefore, any 
reduction in a margin of safety would not be significant.

Proposed Change in the Methods Used To Evaluate Annulus Pressurization 
and Jet Loads Resulting From the Postulated Recirculation Suction Line 
Break

    The proposed change in methods shows that the loads from a 
postulated recirculation suction line break would be bounded by the 
current design basis loads. Therefore, any reduction in a margin of 
safety would not be significant.
    Based on this review, it appears that the three standards of 10 CFR 
50.92'') are satisfied. Therefore, the NRC staff proposes to determine 
that the amendment request involves no significant hazards 
consideration
    Attorney for licensee: Jeffrie J. Keenan, Esquire, Nuclear Business 
Unit--N21, P.O. Box 236, Hancocks Bridge, NJ 08038.
    NRC Section Chief: James W. Clifford.

[[Page 55473]]

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

    Date of amendment request: July 8, 2004.
    Description of amendment request: The proposed amendment would 
delete requirements from the Technical Specifications (TS) to maintain 
hydrogen and oxygen monitors. A notice of availability for this 
technical specification improvement using the consolidated line item 
improvement process (CLIIP) was published in the Federal Register on 
September 25, 2003 (68 FR 55416). Licensees were generally required to 
implement upgrades as described in NUREG-0737, ``Clarification of TMI 
[Three Mile Island] Action Plan Requirements,'' and Regulatory Guide 
1.97, ``Instrumentation for Light-Water-Cooled Nuclear Power Plants to 
Assess Plant and Environs Conditions During and Following an 
Accident.'' Implementation of these upgrades was an outcome of the 
lessons learned from the accident that occurred at TMI, Unit 2. 
Requirements related to combustible gas control were imposed by Order 
for many facilities and were added to or included in the TS for nuclear 
power reactors currently licensed to operate. The revised 10 CFR 50.44, 
``Standards for combustible gas control system in light-water-cooled 
power reactors,'' eliminated the requirements for hydrogen recombiners 
[not installed at Browns Ferry and, therefore, not addressed by this 
proposed amendment] and relaxed safety classifications and licensee 
commitments to certain design and qualification criteria for hydrogen 
and oxygen monitors.
    The NRC staff issued a notice of availability of a model no 
significant hazards consideration (NSHC) determination for referencing 
in license amendment applications in the Federal Register on September 
25, 2003 (68 FR 55416). The licensee affirmed the applicability of the 
model NSHC determination in its application dated July 8, 2004. Basis 
for proposed no significant hazards consideration determination: As 
required by 10 CFR 50.91(a), an analysis of the issue of no significant 
hazards consideration is presented below:

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

    The revised 10 CFR 50.44 no longer defines a design-basis loss-
of-coolant accident (LOCA) hydrogen release, and eliminates 
requirements for hydrogen control systems to mitigate such a 
release. The installation of hydrogen recombiners and/or vent and 
purge systems required by 10 CFR 50.44(b)(3) was intended to address 
the limited quantity and rate of hydrogen generation that was 
postulated from a design-basis LOCA. The Commission has found that 
this hydrogen release is not risk-significant because the design-
basis LOCA hydrogen release does not contribute to the conditional 
probability of a large release up to approximately 24 hours after 
the onset of core damage. In addition, these systems were 
ineffective at mitigating hydrogen releases from risk-significant 
accident sequences that could threaten containment integrity.
    With the elimination of the design-basis LOCA hydrogen release, 
hydrogen and oxygen monitors are no longer required to mitigate 
design-basis accidents and, therefore, the hydrogen monitors do not 
meet the definition of a safety-related component as defined in 10 
CFR 50.2. RG [Regulatory Guide] 1.97 Category 1, is intended for key 
variables that most directly indicate the accomplishment of a safety 
function for design-basis accident events. The hydrogen and oxygen 
monitors no longer meet the definition of Category 1 in RG 1.97. As 
part of the rulemaking to revise 10 CFR 50.44 the Commission found 
that Category 3, as defined in RG 1.97, is an appropriate 
categorization for the hydrogen monitors because the monitors are 
required to diagnose the course of beyond design-basis accidents. 
Also, as part of the rulemaking to revise 10 CFR 50.44, the 
Commission found that Category 2, as defined in RG 1.97, is an 
appropriate categorization for the oxygen monitors, because the 
monitors are required to verify the status of the inert containment.
    The regulatory requirements for the hydrogen and oxygen monitors 
can be relaxed without degrading the plant emergency response. The 
emergency response, in this sense, refers to the methodologies used 
in ascertaining the condition of the reactor core, mitigating the 
consequences of an accident, assessing and projecting offsite 
releases of radioactivity, and establishing protective action 
recommendations to be communicated to offsite authorities. 
Classification of the hydrogen monitors as Category 3 
[classification of the oxygen monitors as Category 2], and removal 
of the hydrogen and oxygen monitors from TS will not prevent an 
accident management strategy through the use of the severe accident 
management guidelines (SAMGs), the emergency plan (EP), the 
emergency operating procedures (EOPs), and site survey monitoring 
that support modification of emergency plan protective action 
recommendations (PARs).
    Therefore, the relaxation of the hydrogen and oxygen monitor 
requirements, including removal of these requirements from TS, does 
not involve a significant increase in the probability or the 
consequences of any 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 relaxation of the hydrogen and oxygen monitor requirements, 
including removal of these requirements from TS; will not result in 
any failure mode not previously analyzed. The hydrogen and oxygen 
monitor equipment was intended to mitigate a design-basis hydrogen 
release. The hydrogen and oxygen monitor equipment are not 
considered accident precursors, nor does their existence or 
elimination have any adverse impact on the pre-accident state of the 
reactor core or post accident confinement of radionuclides within 
the containment building.
    Therefore, this change does not create the possibility of a new 
or different kind of accident from any previously evaluated.

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

    The relaxation of the hydrogen and oxygen monitor requirements, 
including removal of these requirements from TS; in light of 
existing plant equipment, instrumentation, procedures, and programs 
that provide effective mitigation of and recovery from reactor 
accidents, results in a neutral impact to the margin of safety.
    The installation of hydrogen recombiners and/or vent and purge 
systems required by 10 CFR 50.44(b)(3) was intended to address the 
limited quantity and rate of hydrogen generation that was postulated 
from a design-basis LOCA. The Commission has found that this 
hydrogen release is not risk-significant because the design-basis 
LOCA hydrogen release does not contribute to the conditional 
probability of a large release up to approximately 24 hours after 
the onset of core damage.
    Category 3 hydrogen monitors are adequate to provide rapid 
assessment of current reactor core conditions and the direction of 
degradation while effectively responding to the event in order to 
mitigate the consequences of the accident. The intent of the 
requirements established as a result of the TMI, Unit 2 accident can 
be adequately met without reliance on safety-related hydrogen 
monitors.
    Category 2 oxygen monitors are adequate to verify the status of 
an inerted containment.
    Therefore, this change does not involve a significant reduction 
in the margin of safety. The intent of the requirements established 
as a result of the TMI, Unit 2 accident can be adequately met 
without reliance on safety-related oxygen monitors. Removal of 
hydrogen and oxygen monitoring from TS will not result in a 
significant reduction in their functionality, reliability, and 
availability.

    The NRC staff proposes to determine that the amendment request 
involves no significant hazards consideration.
    Attorney for licensee: General Counsel, Tennessee Valley Authority, 
400 West Summit Hill Drive, ET 11A, Knoxville, Tennessee 37902.
    NRC Acting Section Chief: Michael L. Marshall, Jr.

[[Page 55474]]

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

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

Detroit Edison Company, Docket No. 50-341, Fermi 2, Monroe County, 
Michigan

    Date of amendment request: July 30, 2004.
    Brief description of amendment request: The proposed amendment 
would (1) add License Condition 2.C.(22) requiring an integrated tracer 
gas test of the control room envelope using methods described in 
American Society for Testing and Materials E741-00, ``Standard Test 
Method for Determining Air Change in a Single Zone by Means of a Tracer 
Gas Dilution,'' and (2) delete Surveillance Requirement 3.7.3.6, which 
requires verification that unfiltered inleakage from control room 
emergency filtration system duct work outside the control room envelope 
is within limits.
    Date of publication of individual notice in Federal Register: 
August 13, 2004 (68 FR 50217).
    Expiration date of individual notice: October 12, 2004.

Notice of Issuance of Amendments to Facility Operating Licenses

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

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

    Date of application for amendment: December 23, 2003.
    Brief description of amendment: The amendment clarified the 
requirements for inoperable core spray (CS) system components, rendered 
inoperable CS component verification requirements consistent with each 
other, and modified the location requirement of stored water during 
periods of CS system inoperability.
    Date of Issuance: August 19, 2004.
    Effective date: August 20, 2004, and shall be implemented within 60 
days of issuance.
    Amendment No.: 247.
    Facility Operating License No. DPR-16: Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: January 20, 2004 (69 FR 
2738).
    The Commission's related evaluation of this amendment is contained 
in a Safety Evaluation dated August 19, 2004.
    No significant hazards consideration comments received: No.

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

    Date of application for amendment: February 4, 2004, as 
supplemented by letter dated June 9, 2004.
    Brief description of amendment: This amendment revises Technical 
Specification (TS) Surveillance Requirement 4.4.1.3.2, ``Reactor 
Coolant System Hot Shutdown Surveillance Requirements,'' and Limiting 
Condition for Operation 3.4.1.4.1.b, ``Reactor Coolant System Cold 
Shutdown--Loops Filled Limiting Condition For Operation,'' by 
eliminating a requirement that the wide-range instrumentation be 
inoperable before the narrow-range instrumentation can be used for 
confirmation of the minimum steam generator secondary side water level. 
The amendment also revises the TS Index to restore consistency with 
other sections of the TS.
    Date of issuance: August 16, 2004.
    Effective date: August 16, 2004.
    Amendment No.: 116.
    Facility Operating License No. NPF-63. Amendment revises the 
Technical Specifications.
    Date of initial notice in Federal Register: March 16, 2004 (69 FR 
12365). The June 9, 2004, supplement provided clarifying information 
only and did not change the initial no proposed significant hazards 
consideration determination or expand the scope of the initial 
application.
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated August 16, 2004.
    No significant hazards consideration comments received: No.

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

    Docket Nos. STN 50-456 and STN 50-457, Braidwood Station, Unit Nos. 
1 and 2, Will County, Illinois.
    Date of application for amendments: June 27, 2003, as supplemented 
by letters dated January 29, 2004, March 3,

[[Page 55475]]

2004, June 4, 2004, and August 11, 2004.
    Brief description of amendments: The amendments revise TS 3.4.10, 
``Pressurizer Safety Valves,'' by changing the existing pressurizer 
safety valve lift settings from ``>=2460 psig and <=2510 psig,'' to 
``>=2411 psig and <=2509 psig.''
    Date of issuance: August 26, 2004.
    Effective date: As of the date of issuance and shall be implemented 
within 30 days.
    Amendment Nos.: 138/138, 131/131.
    Facility Operating License Nos. NPF-37, NPF-66, NPF-72 and NPF-77: 
The amendments revised the Technical Specifications.
    Date of initial notice in Federal Register: September 30, 2003 (68 
FR 56343).
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated August 26, 2004.
    No significant hazards consideration comments received: No.

Nebraska Public Power District, Docket No. 50-298, Cooper Nuclear 
Station, Nemaha County, Nebraska

    Date of amendment request: December 29, 2003, as supplemented by 
letters dated March 8 and June 8, 2004.
    Brief description of amendment: The amendment revises the 
following: (1) Incorporates into the Updated Safety Analysis Report 
(USAR) the overall main steam isolation valve leakage pathway 
configuration (including the post-accident manual actions necessary to 
establish that configuration), (2) incorporates into the Cooper Nuclear 
Station licensing basis the loss-of-coolant accident (LOCA) dose 
calculation methodology (previously approved on an interim basis), and 
(3) deletes License Condition 2.C.(6), eliminating the commitment to 
provide potassium iodide to the control room personnel during LOCA 
conditions with core damage.
    Date of issuance: September 1, 2004.
    Effective date: As of the date of issuance and shall be implemented 
within 60 days of issuance.
    Amendment No.: 206.
    Facility Operating License No. DPR-46: Amendment revises the USAR 
and Operating License.
    Date of initial notice in Federal Register: March 2, 2004 (69 FR 
9861).
    The March 8 and June 8, 2004, supplemental letters 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 as published in the Federal Register.
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated September 1, 2004.
    No significant hazards consideration comments received: No.

Nuclear Management Company, LLC, Docket No. 50-282, Prairie Island 
Nuclear Generating Plant, Unit 1, Goodhue County, Minnesota

    Date of application for amendment: August 27, 2003, as supplemented 
December 16, 2003, March 22, 2004, and July 19, 2004.
    Brief description of amendment: The amendment revises Technical 
Specification 5.5.14 to allow the licensee to perform post-modification 
testing of the containment pressure boundary following steam generator 
replacement in accordance with the American Society of Mechanical 
Engineers Boiler and Pressure Vessel Code, Section XI, instead of 10 
CFR Part 50, Appendix J, Option B. The steam generator replacement is 
scheduled for fall 2004.
    Date of issuance: August 20, 2004.
    Effective date: As of the date of issuance and shall be implemented 
within 30 days.
    Amendment No.: 165.
    Facility Operating License No. DPR-42: Amendment revises the 
Technical Specifications.
    Date of initial notice in Federal Register: January 20, 2004 (69 FR 
2744).
    The March 22 and July 19, 2004, supplemental letters provided 
clarifying information that was within the scope of the original 
amendment request and did not change the Nuclear Regulatory Commission 
staff's initial proposed no significant hazards consideration 
determination.
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated August 20, 2004
    No significant hazards consideration comments received: No.

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

    Date of application for amendment: July 23, 2003.
    Brief description of amendment: Revised the near end-of-life 
Moderator Temperature Coefficient (MTC) Surveillance Requirement 
4.1.1.3.b by placing a set of conditions on core operation, which if 
met, would allow exemption from the required MTC measurement. The 
conditional exemption is determined on a cycle-specific basis by 
considering the margin predicted to the surveillance requirement MTC 
limit and the performance of other core parameters, such as beginning 
of life MTC measurements and the critical boron concentration as a 
function of cycle life.
    Date of issuance: July 21, 2004.
    Effective date: As of the date of issuance and shall be implemented 
within 30 days.
    Amendment No.: 169.
    Renewed Facility Operating License No. NPF-12: Amendment revises 
the Technical Specifications.
    Date of initial notice in Federal Register: September 30, 2003 (68 
FR 56346).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated July 21, 2004.
    No significant hazards consideration comments received: No.

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

    Date of amendment request: November 4, 2003, as supplemented by 
letter dated June 29, 2004.
    Brief description of amendments: The amendments revise the South 
Texas Project, Units 1 and 2 Technical Specifications for the Remote 
Shutdown System to reflect requirements consistent with those in NUREG-
1431, ``Standard Technical Specifications--Westinghouse Plants.'' The 
changes increase the allowed outage time for inoperable Remote Shutdown 
System components to a time that is more consistent with their safety 
significance and relocate the description of the required components to 
the Bases where it will be directly controlled by the licensee.
    Date of issuance: August 20, 2004.
    Effective date: As of the date of issuance and shall be implemented 
within 60 days of issuance.
    Amendment Nos.: Unit 1-163; Unit 2-152.
    Facility Operating License Nos. NPF-76 and NPF-80: The amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: November 25, 2003 (68 
FR 66140). The supplement dated June 29, 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 as 
published in the Federal Register.
    The Commission's related evaluation of the amendments is contained 
in a

[[Page 55476]]

Safety Evaluation dated August 20, 2004.
    No significant hazards consideration comments received: No.

Virginia Electric and Power Company, Docket No. 50-338, North Anna 
Power Station, Unit 1, Louisa County, Virginia

    Date of application for amendment: March 28, 2002, as supplemented 
by letters dated May 13, June 19, July 9, July 25, August 2, August 16, 
and November 15, 2002, May 6, May 9, May 27, June 11 (2 letters), July 
18, August 20, August 26, September 4, September 5, September 22, 
September 26 (2 letters), November 10, December 8, and December 17, 
2003, and January 6, January 22 (2 letters), February 12, February 13, 
March 1, June 16, and June 18 (2 letters), 2004. The November 15, 2002, 
submittal replaced the submittals dated July 9, July 25, and August 16, 
2002.
    Brief description of amendment: This amendment revises Improved 
Technical Specification Sections 2.1, 4.2, and 5.6.5 in order to allow 
Virginia Electric and Power Company to implement Framatome ANP Advanced 
Mark-BW fuel at North Anna Power Station, Unit 1.
    Date of issuance: August 20, 2004.
    Effective date: As of the date of issuance and shall be implemented 
prior to the initiation of core onload during Refueling Outage 17 (Fall 
2004).
    Amendment No.: 237.
    Renewed Facility Operating License No. NPF-4: Amendment changes the 
Technical Specifications.
    Date of initial notice in Federal Register: July 22, 2003 (68 FR 
43397). The supplements dated July 18, August 20, August 26, September 
4, September 5, September 22, September 26 (2 letters), November 10, 
December 8, and December 17, 2003, and January 6, January 22 (2 
letters), February 12, February 13, March 1, June 16, and June 18 (2 
letters), 2004, contained clarifying information only and did not 
change the initial no significant hazards consideration determination 
or expand the scope of the initial application.
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated August 20, 2004.
    No significant hazards consideration comments received: No.

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

    During the period since publication of the last biweekly notice, 
the Commission has issued the following amendments. The Commission has 
determined for each of these amendments that the application for the 
amendment complies with the standards and requirements of the Atomic 
Energy Act of 1954, as amended (the Act), and the Commission's rules 
and regulations. The Commission has made appropriate findings as 
required by the Act and the Commission's rules and regulations in 10 
CFR Chapter I, which are set forth in the license amendment.
    Because of exigent or emergency circumstances associated with the 
date the amendment was needed, there was not time for the Commission to 
publish, for public comment before issuance, its usual Notice of 
Consideration of Issuance of Amendment, Proposed No Significant Hazards 
Consideration Determination, and Opportunity for a Hearing.
    For exigent circumstances, the Commission has either issued a 
Federal Register notice providing opportunity for public comment or has 
used local media to provide notice to the public in the area 
surrounding a licensee's facility of the licensee's application and of 
the Commission's proposed determination of no significant hazards 
consideration. The Commission has provided a reasonable opportunity for 
the public to comment, using its best efforts to make available to the 
public means of communication for the public to respond quickly, and in 
the case of telephone comments, the comments have been recorded or 
transcribed as appropriate and the licensee has been informed of the 
public comments.
    In circumstances where failure to act in a timely way would have 
resulted, for example, in derating or shutdown of a nuclear power plant 
or in prevention of either resumption of operation or of increase in 
power output up to the plant's licensed power level, the Commission may 
not have had an opportunity to provide for public comment on its no 
significant hazards consideration determination. In such case, the 
license amendment has been issued without opportunity for comment. If 
there has been some time for public comment but less than 30 days, the 
Commission may provide an opportunity for public comment. If comments 
have been requested, it is so stated. In either event, the State has 
been consulted by telephone whenever possible.
    Under its regulations, the Commission may issue and make an 
amendment immediately effective, notwithstanding the pendency before it 
of a request for a hearing from any person, in advance of the holding 
and completion of any required hearing, where it has determined that no 
significant hazards consideration is involved.
    The Commission has applied the standards of 10 CFR 50.92 and has 
made a final determination that the amendment involves no significant 
hazards consideration. The basis for this determination is contained in 
the documents related to this action. Accordingly, the amendments have 
been issued and made effective as indicated.
    Unless otherwise indicated, the Commission has determined that 
these amendments satisfy the criteria for categorical exclusion in 
accordance with 10 CFR 51.22. Therefore, pursuant to 10 CFR 51.22(b), 
no environmental impact statement or environmental assessment need be 
prepared for these amendments. If the Commission has prepared an 
environmental assessment under the special circumstances provision in 
10 CFR 51.12(b) and has made a determination based on that assessment, 
it is so indicated.
    For further details with respect to the action see (1) the 
application for amendment, (2) the amendment to Facility Operating 
License, and (3) the Commission's related letter, Safety Evaluation 
and/or Environmental Assessment, as indicated. All of these items are 
available for public inspection at the Commission's Public Document 
Room, 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 Public 
Document Room (PDR) Reference staff at 1-800-397-4209, 301-415-4737 or 
by e-mail to [email protected].
    The Commission is also offering an opportunity for a hearing with 
respect to the issuance of the amendment. Within 60 days after the date 
of publication of this notice, the licensee may file a request for a 
hearing with respect to

[[Page 55477]]

issuance of the amendment to the subject facility operating license and 
any person whose interest may be affected by this proceeding and who 
wishes to participate as a party in the proceeding must file a written 
request for a hearing and a petition for leave to intervene. Requests 
for a hearing and a petition for leave to intervene shall be filed in 
accordance with the Commission's ``Rules of Practice for Domestic 
Licensing Proceedings'' in 10 CFR Part 2. Interested persons should 
consult a current copy of 10 CFR 2.309, which is available at the 
Commission's PDR, located at One White Flint North, Public File Area 
01F21, 11555 Rockville Pike (first floor), Rockville, Maryland, and 
electronically on the Internet at the NRC web site, http://www.nrc.gov/reading-rm/doc-collections/cfr/. If there are problems in accessing the 
document, contact the PDR Reference staff at 1-800-397-4209, 301-415-
4737, or by e-mail to [email protected]. If a request for a hearing or 
petition for leave to intervene is filed by the above date, the 
Commission or a presiding officer designated by the Commission or by 
the Chief Administrative Judge of the Atomic Safety and Licensing Board 
Panel, will rule on the request and/or petition; and the Secretary or 
the Chief Administrative Judge of the Atomic Safety and Licensing Board 
will issue a notice of a hearing or an appropriate order.
    As required by 10 CFR 2.309, a petition for leave to intervene 
shall set forth with particularity the interest of the petitioner in 
the proceeding, and how that interest may be affected by the results of 
the proceeding. The petition should specifically explain the reasons 
why intervention should be permitted with particular reference to the 
following general requirements: (1) The name, address and telephone 
number of the requestor or petitioner; (2) the nature of the 
requestor's/petitioner's right under the Act to be made a party to the 
proceeding; (3) the nature and extent of the requestor's/petitioner's 
property, financial, or other interest in the proceeding; and (4) the 
possible effect of any decision or order which may be entered in the 
proceeding on the requestor's/petitioner's interest. The petition must 
also identify the specific contentions which the petitioner/requestor 
seeks to have litigated at the proceeding.
    Each contention must consist of a specific statement of the issue 
of law or fact to be raised or controverted. In addition, the 
petitioner/requestor shall provide a brief explanation of the bases for 
the contention and a concise statement of the alleged facts or expert 
opinion which support the contention and on which the petitioner 
intends to rely in proving the contention at the hearing. The 
petitioner must also provide references to those specific sources and 
documents of which the petitioner is aware and on which the petitioner 
intends to rely to establish those facts or expert opinion. The 
petition must include sufficient information to show that a genuine 
dispute exists with the applicant on a material issue of law or 
fact.\1\ Contentions shall be limited to matters within the scope of 
the amendment under consideration. The contention must be one which, if 
proven, would entitle the petitioner to relief. A petitioner/requestor 
who fails to satisfy these requirements with respect to at least one 
contention will not be permitted to participate as a party.
---------------------------------------------------------------------------

    \1\ To the extent that the applications contain attachments and 
supporting documents that are not publicly available because they 
are asserted to contain safeguards or proprietary information, 
petitioners desiring access to this information should contact the 
applicant or applicant's counsel and discuss the need for a 
protective order.
---------------------------------------------------------------------------

    Each contention shall be given a separate numeric or alpha 
designation within one of the following groups:
    1. Technical--primarily concerns/issues relating to technical and/
or health and safety matters discussed or referenced in the 
applications.
    2. Environmental--primarily concerns/issues relating to matters 
discussed or referenced in the environmental analysis for the 
applications.
    3. Miscellaneous--does not fall into one of the categories outlined 
above.
    As specified in 10 CFR 2.309, if two or more petitioners/requestors 
seek to co-sponsor a contention, the petitioners/requestors shall 
jointly designate a representative who shall have the authority to act 
for the petitioners/requestors with respect to that contention. If a 
petitioner/requestor seeks to adopt the contention of another 
sponsoring petitioner/requestor, the petitioner/requestor who seeks to 
adopt the contention must either agree that the sponsoring petitioner/
requestor shall act as the representative with respect to that 
contention, or jointly designate with the sponsoring petitioner/
requestor a representative who shall have the authority to act for the 
petitioners/requestors with respect to that contention.
    Those permitted to intervene become parties to the proceeding, 
subject to any limitations in the order granting leave to intervene, 
and have the opportunity to participate fully in the conduct of the 
hearing. Since the Commission has made a final determination that the 
amendment involves no significant hazards consideration, if a hearing 
is requested, it will not stay the effectiveness of the amendment. Any 
hearing held would take place while the amendment is in effect.
    A request for a hearing or a petition for leave to intervene must 
be filed by: (1) First class mail addressed to the Office of the 
Secretary of the Commission, U.S. Nuclear Regulatory Commission, 
Washington, DC 20555-0001, Attention: Rulemaking and Adjudications 
Staff; (2) courier, express mail, and expedited delivery services: 
Office of the Secretary, Sixteenth Floor, One White Flint North, 11555 
Rockville Pike, Rockville, Maryland 20852, Attention: Rulemaking and 
Adjudications Staff; (3) E-mail addressed to the Office of the 
Secretary, U.S. Nuclear Regulatory Commission, [email protected]; 
or (4) facsimile transmission addressed to the Office of the Secretary, 
U.S. Nuclear Regulatory Commission, Washington, DC, Attention: 
Rulemakings and Adjudications Staff at (301) 415-1101, verification 
number is (301) 415-1966. A copy of the request for hearing and 
petition for leave to intervene should also be sent to the Office of 
the General Counsel, U.S. Nuclear Regulatory Commission, Washington, DC 
20555-0001, and it is requested that copies be transmitted either by 
means of facsimile transmission to 301-415-3725 or by e-mail to 
[email protected]. A copy of the request for hearing and petition 
for leave to intervene should also be sent to the attorney for the 
licensee.
    Nontimely requests and/or petitions and contentions will not be 
entertained absent a determination by the Commission or the presiding 
officer or the Atomic Safety and Licensing Board that the petition, 
request and/or the contentions should be granted based on a balancing 
of the factors specified in 10 CFR 2.309(a)(1)(i)-(viii).

Florida Power and Light Company, Docket No. 251, Turkey Point Plant, 
Unit 4, Miami-Dade County, Florida

    Date of amendment request: July 28, 2004, as supplemented in a 
letter dated August 5, 2004.
    Description of amendment request: The amendment revised Technical 
Specifications 3/4.1.3.1, 3/4.1.3.2 and 3/4.1.3.5 to allow the use of 
an alternate method of determining rod position for the control rod F-
8, until the end of Cycle 22 or until repairs can be conducted on the 
Analog Rod Indication System at the next outage of sufficient duration, 
whichever comes first.
    Date of issuance: August 20, 2004.

[[Page 55478]]

    Effective date: As of the date of issuance and shall be implemented 
within 60 days of issuance.
    Amendment No.: 221.
    Facility Operating License No. (DPR-41): Amendment revised the 
Technical Specifications.
    Public comments requested as to proposed no significant hazards 
consideration (NSHC): Yes. August 5, 2004 (69 FR 47467). The licensee's 
August 5, 2004 submittal of supplemental information did not affect the 
original no significant hazards consideration determination, and did 
not expand the scope of the request as noticed on August 5, 2004. The 
notice provided an opportunity to submit comments on the Commission's 
proposed NSHC determination. No comments have been received. The notice 
also provided an opportunity to request a hearing by August 19, 2004, 
but indicated that if the Commission makes a final NSHC determination, 
any such hearing would take place after issuance of the amendment.
    The Commission's related evaluation of the amendment, finding of 
exigent circumstances, state consultation, and final NSHC determination 
are contained in a safety evaluation dated August 20, 2004.
    Attorney for licensee: M.S. Ross, Attorney, Florida Power & Light, 
P.O. Box 14000, Juno Beach, Florida 33408-0420.
    NRC Section Chief: Michael L. Marshall, Jr. (Acting).

    Dated at Rockville, Maryland, this 3rd day of September 2004.

    For the Nuclear Regulatory Commission.
Ledyard B. Marsh,
Director, Division of Licensing Project Management, Office of Nuclear 
Reactor Regulation.
[FR Doc. 04-20497 Filed 9-13-04; 8:45 am]
BILLING CODE 7590-01-P