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