[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]


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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