[Federal Register Volume 64, Number 115 (Wednesday, June 16, 1999)]
[Notices]
[Pages 32280-32284]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: 99-15244]
-----------------------------------------------------------------------
NUCLEAR REGULATORY COMMISSION
[Docket Nos. STN 50-454, STN 50-455, STN 50-456 and STN 50-457]
Commonwealth Edison Company; Notice of Consideration of Issuance
of Amendments to Facility Operating Licenses, Proposed No Significant
Hazards Consideration Determination and Opportunity for a Hearing
The U.S. Nuclear Regulatory Commission (the Commission) is
considering issuance of amendments to Facility Operating License Nos.
NPF-37 and NPF-66 issued to the Commonwealth Edison Company (ComEd, the
licensee) for operation of Byron Station, Unit Nos. 1 and 2,
respectively, located in Ogle County, Illinois, and Facility Operating
License Nos. NPF-72 and NPF-77 issued to ComEd for the operation of
Braidwood Station, Unit Nos. 1 and 2, respectively, located in Will
County, Illinois.
The proposed amendments would change the Technical Specifications
to support a plant modification to install new storage racks for fuel
in the spent fuel pools (SFP). As part of the modification, the total
capacity of the SFP at each station is being increased from 2,870
assemblies to 2,984 assemblies.
Before issuance of the proposed license amendments, the Commission
will have made findings as required by the Atomic Energy Act of 1954,
as amended (the Act) and the Commission's regulations.
The Commission has made a proposed determination that the
amendments requested 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 amendments
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. 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:
The proposed Technical Specifications (TS) changes do not
involve a significant increase in the probability or consequences of
an accident previously evaluated.
During the installation of the new Holtec spent fuel pool
storage racks, both Holtec and the existing Joseph Oat spent fuel
pool storage racks will be in the spent fuel pool at the same time.
This interim arrangement will not increase the probability or
consequences of an accident previously evaluated. The criticality
analysis for the Joseph Oat spent fuel pool storage racks states
that should a spent fuel pool water temperature change accident or a
fuel assembly misload accident occur in the Region 1, Region 2, or
failed fuel storage cells, keff will be maintained less
than or equal to 0.95 due to the presence of at least 550 ppm (no
fuel handling) or 1650 ppm (during fuel handling) of soluble boron
in the spent fuel pool water. These assumptions are more
conservative than the requirements stated in the criticality
analysis for the Holtec spent fuel pool storage racks which only
requires 220 ppm boron to maintain keff less than or
equal to 0.95 during the worst case fuel assembly misload accident.
The new Holtec racks have a superior neutron attenuation capability
due to their improved design. The requirement of 2000 ppm boron will
be maintained during the entire change out process, therefore,
ensuring that keff will remain less than or equal to
0.95. At the completion of installation, only Holtec spent fuel pool
storage racks will be in the spent fuel pool.
The previously evaluated Byron and Braidwood Stations accidents
relative to spent fuel storage are discussed in the Updated Final
Safety Analysis Report (UFSAR) Section 15.7.4, ``Fuel Handling
Accidents,'' and UFSAR Section 15.7.5, ``Spent Fuel Cask Drop
Accident.'' These accidents were considered for the new Holtec spent
fuel pool racks and are listed below.
a. Spent fuel assembly dropped onto the spent fuel pool floor.
[[Page 32281]]
b. Spent fuel assembly dropped between racks.
c. Spent fuel assembly dropped between a rack and the spent fuel
pool wall.
d. Spent fuel assembly loaded contrary to placement
restrictions.
e. Spent fuel assembly dropped onto to [sic] a rack.
f. Spent fuel cask drop.
g. Change in spent fuel pool water temperature.
Spent Fuel Assembly Dropped Onto the Spent Fuel Pool Floor
The probability and consequences of dropping a spent fuel
assembly onto the spent fuel pool liner have been evaluated and
shown to be bounded by the existing design basis as described in the
Byron and Braidwood Stations UFSAR. The maximum drop distance for a
fuel assembly will not change as a result of this design change and,
therefore, the consequences of this fuel handling accident remain
unchanged. The probability of this fuel handling accident is not
changed by the installation of new Holtec spent fuel pool storage
racks or by the small increase (approximately 4.0%) in spent fuel
storage capacity as the spent fuel handling procedures and equipment
are unaffected by the change. Also, the number of spent fuel
assemblies is not an input to the initial conditions of this
accident evaluation.
Spent Fuel Assembly Dropped Between Racks
The probability and consequences of dropping a fuel assembly
between rack modules was previously evaluated under UFSAR Section
9.1.2.3.9, ``Accident/Abnormal Storage Conditions in Spent Fuel Pool
Racks,'' which supports TS Limiting Condition for Operation (LCO)
3.7.15 and was shown to have no effect on reactivity. This is
considered a bounding analysis and is applicable to this design
change since the new Holtec rack layout still precludes a reactivity
increase due to this fuel handling accident. The probability of this
event is unaffected due to the similarity between the new Holtec
spent fuel pool rack layout and the existing Joseph Oat spent fuel
pool rack layout.
Spent Fuel Assembly Dropped Between a Rack and the Spent Fuel Pool
Wall
The probability and consequences of dropping a spent fuel
assembly between a rack module and the spent fuel wall has been
evaluated for the new Holtec spent fuel pool racks. The worst case
scenario consists of a fresh fuel assembly, of the highest allowed
enrichment, accidentally placed in a cut out area between a rack and
the new fuel elevator or tool bracket. The consequences of this
event remain within the design basis criticality limit of less than
or equal to 0.95 keff, assuming a minimum soluble boron
concentration of 220 ppm in the spent fuel pool water. The
probability of this event is unaffected due to the similarity
between the new Holtec spent fuel pool rack layout and the existing
Joseph Oat spent fuel pool rack layout. This event is bounded by the
analysis of misloading an assembly into a Region 2 rack, discussed
below.
Spent Fuel Assembly Loaded Contrary to Placement Restrictions
The probability and consequences of loading a fuel assembly
contrary to placement restrictions has been evaluated for the Holtec
racks. A worst case scenario of placing a fuel assembly of the
highest enrichment (i.e., 5.0 weight percent U-235) into a Region 2
rack cell was shown to remain within the design basis criticality
limit of 0.95 keff, assuming a minimum soluble boron
concentration of 220 ppm in the spent fuel pool water. The current
required soluble boron concentration in the spent fuel pool is 2000
ppm. The minimum soluble boron concentration, proposed in
conjunction with this design change, is 300 ppm for conservatism.
The probability of this event is unaffected by this design change
since the existing pool already includes a two region layout,
similar to the new Holtec racks. Further, the possibility of a
misloaded fuel assembly is minimized by an independent verification
of the Nuclear Component Transfer List that prescribes the exact
location of each fuel assembly. After an assembly is placed in a
spent fuel pool storage cell, station personnel once again
independently verify it.
Spent Fuel Assembly Dropped onto to [sic] a Rack
The probability and consequences of dropping a spent fuel
assembly onto a spent fuel storage rack have been evaluated for the
Holtec racks. The consequences are shown to meet all existing design
basis requirements as described in the Byron and Braidwood Station
UFSAR. Analyses of the spent fuel drop accidents onto the top of a
spent fuel pool storage rack (shallow drop), and a deep drop into
the bottom of a cell, resulting in impact at the bottom of the rack
cell, were performed to demonstrate that the spent fuel rack retains
its structural integrity and capability to safely store spent fuel
in adjacent cells. The damage due to a perfectly vertical drop, on
the top of a rack, bounds an inclined fuel assembly drop because the
impact energy is focused on a single cell wall, which results in
maximum cell blockage. The radiological consequences of the drop
onto the spent fuel pool liner, shallow drop onto to [sic] the top
of the rack, and deep drop into the bottom of a rack cell, are
bounded by the existing UFSAR assumptions that 314 fuel rods
rupture. The UFSAR design basis dose is shown to be much less than
the 10 CFR 100 off-site dose limits of 300 rem to the thyroid and 25
rem to the whole body. The probability of these fuel handling
accidents occurring is unaffected by the installation of new spent
fuel storage racks. The spent fuel handling procedures and equipment
are unaffected by this change and therefore there is no increase in
the probability of these fuel handling accidents.
Spent Fuel Cask Drop
The probability and consequences of a cask drop were evaluated
and shown to be unaffected by the replacement of the existing Joseph
Oat spent fuel pool storage racks with Holtec racks. There are no
changes to any of the systems, structures, components or equipment
associated with the movement of a spent fuel cask. The cask is shown
by the Byron and Braidwood Stations UFSAR to be isolated from the
spent fuel pool by the combination of guard walls, which are
designed to withstand the impact of a cask drop, and both
administrative and physical controls. These controls are designed to
preclude the fuel handling building crane from traveling over the
spent fuel pool. There are also trolley stops on the crane bridge
which physically prevent the main hook of the crane from traveling
into the spent fuel pool storage area when handling a spent fuel
cask. Spent fuel pool rack installation activities and cask handling
will not be performed simultaneously, thus minimizing the
possibility of improper movement of the cask. This practice is
consistent with the Byron and Braidwood Stations UFSAR assumptions
relative to new fuel operations. Since there will be no changes to
any of the equipment, procedures or operations relative to spent
fuel cask handling that are associated with this design change,
there is no increase in the probability or consequences of this fuel
handling accident.
Change in Spent Fuel Pool Water Temperature
The probability and consequences of a change in the temperature
of the spent fuel pool water was evaluated for the potential for an
increase in reactivity. The new Holtec rack analysis was performed
assuming a spent fuel pool water temperature of 4 deg.C (39
deg.F), which is well below the lowest normal operating temperature
of 50 deg.F. Because the reactivity temperature coefficient in the
spent fuel pool is negative, temperatures greater than 4 deg.C will
result in a decrease in reactivity. The probability of this event is
unaffected by the spent fuel pool rack replacement because there are
no features of this design change affecting the spent fuel pool
cooling system or that would prompt a spent fuel pool water
temperature decrease.
Rack Installation
Holtec International personnel will execute the construction
phases of the Byron and Braidwood Stations rack installations. All
construction work will be performed in compliance with Byron and
Braidwood Stations' commitments to NUREG-0612 and site-specific
procedures. Holtec International and Commonwealth Edison are
developing a complete set of operating procedures which cover the
entire gamut of operations pertaining to the rack installation
effort. Similar procedures have been utilized and successfully
implemented by Holtec International on previous rack installation
projects. These procedures assure that ALARA practices are followed
and provide detailed requirements to assure equipment, personnel,
and plant safety.
Crane and fuel bridge operators will be adequately trained in
the operation of load handling machines per the station specific
training program. The lifting device designed for handling and
installation of the new racks at Byron and Braidwood Stations is in
compliance with the provisions of NUREG-0612, including compliance
with the primary stress criteria, load testing with a multiplier
[[Page 32282]]
for maximum working load, and nondestructive examination of critical
welds.
An intensive surveillance and inspection program shall be
maintained throughout the rack installation phase of the project. A
set of inspection points has been established based on experience in
numerous previous rack installation campaigns. These inspections
have proven to eliminate incidence of rework or erroneous
installation.
Based on the review of the accidents previously analyzed in the
UFSAR, and considering the rigorous controls in place for
installation of the new spent fuel pool storage racks, it is
concluded that there will not be a significant increase in the
probability or consequences of an accident previously evaluated.
The proposed TS changes do not create the possibility of a new
or different kind of accident from any accident previously
evaluated.
The replacement of the existing Byron and Braidwood spent fuel
pool storage racks, having a capacity of 2870 cells, with new racks
having a capacity of 2984 cells, was evaluated for the possibility
of creating a new or different accident. The following cases were
reviewed:
a. An accidental drop of a rack into the spent fuel pool, and
b. Additional heat load resulting from the additional storage
capacity.
A construction accident resulting in a rack drop is an extremely
unlikely event. Operability of the cranes will be checked prior to
use. Lift equipment and rigging will also be inspected prior to use.
Operators of lift equipment and cranes will be trained prior to use.
Safe load paths will be followed and Byron and Braidwood Stations'
commitments to the provisions of NUREG-0612 will be implemented by
use of written procedures that have been utilized for numerous other
similar rack installation projects. The Technical Requirements
Manual requires that Fuel Handling Building Crane loads be limited
to 2000 pounds when traveling over fuel assemblies. This limitation
will be adhered to during the entire course of rack installation. In
the unlikely event of a rack drop, a leak chase system located
beneath the spent fuel pool liner is capable of collecting and
isolating the leakage. A rack drop would present limited structural
damage to the spent fuel pool slab on grade, due to the slab being
founded on rock and soil. Local concrete crushing and possible liner
puncture could occur. Failure of the liner would not result in a
significant loss of water and no safety related equipment would be
affected by the leakage. Make up water is available from 3 separate
sources. There are two 500,000 gallon Refueling Water Storage Tanks,
non-category 1 back up water sources, and the unborated Safety
Category 1 fire protection system, available for spent fuel pool
water make up. A rack drop, therefore, does not create the
possibility of creating a new or different kind of accident.
The additional heat load resulting from the additional storage
capacity of 114 cells (i.e., approximately 4%) has been evaluated
for the possibility of creating a new or different kind of accident.
The existing spent fuel pool cooling system has been shown to be
capable of removing the decay heat generated by the additional spent
fuel assemblies utilizing the standard Byron and Braidwood Stations
operating procedures. Since it is shown that the spent fuel pool
cooling system will maintain the spent fuel pool water temperature
within the existing design basis, as detailed in the Byron and
Braidwood UFSAR, it is concluded that the proposed changes do not
create a new or different kind of accident.
Replacing the existing 23 Joseph Oat Boraflex racks with 24 new
Holtec racks containing Boral, and increasing the spent fuel storage
capacity in each of the spent fuel pools at Byron and Braidwood
Stations to 2984 assemblies, will not create the possibility of an
accident of a different type. The fuel pool rack and fuel
configurations have been analyzed considering criticality, thermal
hydraulic, and structural effects. The increase in storage capacity
is achieved by the installation of additional racks of similar, but
improved design, which are passive components. No new operating
schemes or active equipment types will be required to store
additional fuel assemblies in the fuel pools. The possibility of a
different type of accident occurring is not created since the new
racks meet or exceed the requirements applicable to the existing
racks.
Therefore, implementation of the proposed TS changes do not
create the possibility of a new or different kind of accident from
any previously evaluated.
The proposed TS changes do not involve a significant reduction
in a margin of safety.
The function of the spent fuel pool is to store fuel assemblies
in a subcritical and coolable configuration throughout all
environmental and abnormal loadings, such as earthquakes, dropped
fuel assemblies, or loss of spent fuel pool cooling. The new spent
fuel storage racks are designed to meet all applicable requirements
for safe storage of spent fuel and are functionally compatible with
the spent fuel pool.
The Holtec Licensing Report has analyzed the consequences of
this reracking project by area. In each area, (i.e., criticality,
seismic, structural, thermal hydraulics, and radiological exposure),
design basis margins of safety will be maintained. Installation
controls specified in Byron and Braidwood Stations' commitments to
NUREG-0612 preserve the margins of safety with regard to heavy load
restrictions. Compliance with the Byron and Braidwood Station design
basis limits and procedure adherence will preclude reducing margins
of safety.
The margin of safety is not reduced as demonstrated by analysis
of the seismic, structural, thermal hydraulic, criticality, and
radiological aspects of this design change. The Byron and Braidwood
Station design basis spent fuel pool maximum bulk temperature
acceptance limit of 140 deg. F has been demonstrated to be preserved
by analysis. Criticality calculations show that keff will
be maintained at less than or equal to 0.95. The new Holtec spent
fuel pool storage racks have been designed in accordance with the
Byron and Braidwood Station design bases requirements and the NRC OT
position paper.
Since all aspects of the design change have been demonstrated to
be within the existing design bases for Byron and Braidwood Stations
and the NRC requirements applicable to spent fuel storage, the
proposed changes do not involve a significant reduction in the
margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendments requested involve no significant hazards consideration.
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.
Normally, the Commission will not issue the amendments until the
expiration of the 30-day notice period. However, should circumstances
change during the notice period such that failure to act in a timely
way would result, for example, in derating or shutdown of the facility,
the Commission may issue the license amendments before the expiration
of the 30-day notice period, provided that its final determination is
that the amendments involve no significant hazards consideration. The
final determination will consider all public and State comments
received. Should the Commission take this action, it will publish in
the Federal Register a notice of issuance and provide for opportunity
for a hearing 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 6D59, 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 NRC Public Document
Room, the Gelman Building, 2120 L Street, NW., Washington, DC.
The filing of requests for hearing and petitions for leave to
intervene is discussed below.
By July 16, 1999, the licensee may file a request for a hearing
with respect to issuance of the amendments to the subject facility
operating license and any person whose interest may be affected by this
proceeding and who
[[Page 32283]]
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 2.714 which is available at the
Commission's Public Document Room, the Gelman Building, 2120 L Street,
NW., Washington, DC, and at the local public document room located at
the Byron Public Library District, 109 N. Franklin, P.O. Box 434,
Byron, Illinois 61010 for Byron Station, and the Wilmington Public
Library, 201 S. Kankakee Street, Wilmington, Illinois 60481 for
Braidwood Station. If a request for a hearing or petition for leave to
intervene is filed by the above date, the Commission or an Atomic
Safety and Licensing Board, designated by the Commission or by the
Chairman of the Atomic Safety and Licensing Board Panel, will rule on
the request and/or petition; and the Secretary or the designated Atomic
Safety and Licensing Board will issue a notice of hearing or an
appropriate order.
As required by 10 CFR 2.714, 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 factors: (1) the nature of the petitioner's right under the
Act to be made party to the proceeding; (2) the nature and extent of
the petitioner's property, financial, or other interest in the
proceeding; and (3) the possible effect of any order which may be
entered in the proceeding on the petitioner's interest. The petition
should also identify the specific aspect(s) of the subject matter of
the proceeding as to which petitioner wishes to intervene. Any person
who has filed a petition for leave to intervene or who has been
admitted as a party may amend the petition without requesting leave of
the Board up to 15 days prior to the first prehearing conference
scheduled in the proceeding, but such an amended petition must satisfy
the specificity requirements described above.
Not later than 15 days prior to the first prehearing conference
scheduled in the proceeding, a petitioner shall file a supplement to
the petition to intervene which must include a list of the contentions
which are sought to be litigated in the matter. Each contention must
consist of a specific statement of the issue of law or fact to be
raised or controverted. In addition, the petitioner shall provide a
brief explanation of the bases of 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. Petitioner must provide 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 amendments under consideration. The contention must be
one which, if proven, would entitle the petitioner to relief. A
petitioner who fails to file such a supplement which satisfies 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, including the opportunity to present evidence and cross-
examine witnesses.
If a hearing is requested, 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 amendments requested involve
no significant hazards consideration, the Commission may issue the
amendments and make them immediately effective, notwithstanding the
request for a hearing. Any hearing held would take place after issuance
of the amendments.
If the final determination is that the amendments requested involve
a significant hazards consideration, any hearing held would take place
before the issuance of any amendments.
A request for a hearing or a petition for leave to intervene must
be filed with the Secretary of the Commission, U.S. Nuclear Regulatory
Commission, Washington, DC 20555-0001, Attention: Rulemakings and
Adjudications Staff, or may be delivered to the Commission's Public
Document Room, the Gelman Building, 2120 L Street, NW., Washington, DC,
by the above date. A copy of the petition should also be sent to the
Office of the General Counsel, U.S. Nuclear Regulatory Commission,
Washington, DC 20555-0001, and to Ms. Pamela B. Stroebel, Senior Vice
President and General Counsel, Commonwealth Edison Company, P.O. Box
767, Chicago, Illinois 60690-0767, attorney for the licensee.
Nontimely filings of petitions for leave to intervene, amended
petitions, supplemental petitions and/or requests for hearing will not
be entertained absent a determination by the Commission, the presiding
officer or the presiding Atomic Safety and Licensing Board that the
petition and/or request should be granted based upon a balancing of the
factors specified in 10 CFR 2.714(a)(1)(i)-(v) and 2.714(d).
The Commission hereby provides notice that this is a proceeding on
an application for license amendments falling within the scope of
section 134 of the Nuclear Waste Policy Act of 1982 (NWPA), 42 U.S.C.
10154. Under section 134 of the NWPA, the Commission, at the request of
any party to the proceeding, must use hybrid hearing procedures with
respect to ``any matter which the Commission determines to be in
controversy among the parties.''
The hybrid procedures in section 134 provide for oral argument on
matters in controversy, preceded by discovery under the Commission's
rules and the designation, following argument of only those factual
issues that involve a genuine and substantial dispute, together with
any remaining questions of law, to be resolved in an adjudicatory
hearing. Actual adjudicatory hearings are to be held on only those
issues found to meet the criteria of section 134 and set for hearing
after oral argument.
The Commission's rules implementing section 134 of the NWPA are
found in 10 CFR Part 2, Subpart K, ``Hybrid Hearing Procedures for
Expansion of Spent Fuel Storage Capacity at Civilian Nuclear Power
Reactors'' (published at 50 FR 41662 dated October 15, 1985). Under
those rules, any party to the proceeding may invoke the hybrid hearing
procedures by filing with the presiding officer a written request for
oral argument under 10 CFR 2.1109. To be timely, the request must be
filed within ten (10) days of an order granting a request for hearing
or petition to intervene. The presiding officer must grant a timely
request for oral argument. The presiding officer may grant an untimely
request for oral argument only upon a showing of good cause by the
requesting party for the failure to file on time and after providing
the other parties an opportunity to respond to the untimely request. If
the presiding officer grants a request for oral argument, any hearing
held on the application must be
[[Page 32284]]
conducted in accordance with the hybrid hearing procedures. In essence,
those procedures limit the time available for discovery and require
that an oral argument be held to determine whether any contentions must
be resolved in an adjudicatory hearing. If no party to the proceeding
timely requests oral argument, and if all untimely requests for oral
argument are denied, then the usual procedures in 10 CFR part 2,
Subpart G apply.
For further details with respect to this action, see the
application for amendments dated March 23, 1999, which is available for
public inspection at the Commission's Public Document Room, the Gelman
Building, 2120 L Street, NW., Washington, DC, and at the local public
document room located at the Byron Public Library District, 109 N.
Franklin, P.O. Box 434, Byron, Illinois 61010 for Byron Station, and
the Wilmington Public Library, 201 S. Kankakee Street, Wilmington,
Illinois 60481 for Braidwood Station.
Dated at Rockville, Maryland, this 10th day of June 1999.
For the Nuclear Regulatory Commission.
Stewart N. Bailey,
Project Manager, Section 2, Project Directorate 3, Division of
Licensing Project Management, Office of Nuclear Reactor Regulation.
[FR Doc. 99-15244 Filed 6-15-99; 8:45 am]
BILLING CODE 7590-01-P
