ML20195F964
| ML20195F964 | |
| Person / Time | |
|---|---|
| Site: | Byron, Braidwood  |
| Issue date: | 06/10/1999 |
| From: | Stewart Bailey NRC (Affiliation Not Assigned) |
| To: | |
| Shared Package | |
| ML20195F969 | List: |
| References | |
| NUDOCS 9906150150 | |
| Download: ML20195F964 (13) | |
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7590-01-P UNITED STATES NUCLEAR REGULATORY COMMISSION COMMONWEALTH EDISON COMPANY DOCKET NOS. STN 50-454. STN 50-455. STN 50-456 AND STN 50-457 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 4
previously evaluated; or (2) create the possibility of a new or different kind of accident from 9906150150 990610 yDR ADOCK 05000454 PDR
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/, 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 mistoad accident occur in the Region 1, Region 2, or failed fuel storage cells, k,, 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 k,,less than or equal to 0.95 during the worst case fuel assembly mistoad 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 k., 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, 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.
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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
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' 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.
' Coent Fuel Assembly Dropped Between Rocks The probabii!ty 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 i
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 k,, assuming a minimum soluble boron 1
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 k,.
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, simliar to the new Holtec racks. Further, the possibility of a mistoaded 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.
4 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 Holtee 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 s
pool storage ack (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 lntegrity 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 hahdling 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 book 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*C (39'F),
which is well below the lowest normal operating temperature of 50*F. Because the
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- reactivity temperature coefficient in the spent fuel pool is negative, temperatures greater
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than 4*C will result in a decrease in reactivity. The probaulity of this event is unaffected i
by the spent fuel pool rack replacement because there are no features of this design
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change affecting the spent fuel pool cooling system or that would prompt a spent fuel pool water ternperature decrease.
I RackInstallation ;
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 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 j
inspections have proven to eliminate incidence of rework or erroneous installation.
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. 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 j
storage racks, it is concluded that there will not be a significant increase in the i
probability or consequences of an accident previously evaluated.
I 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 Additional heat load resultinC rom the additional storage capacity.
f b.
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'A construction accident resulting in a rack drop is an enremely unlikely event.
Operability of the cranes will be checked prior to use. i.ift equipmeM and rigging will also be inspected prior to use. Operators of lift equipment and crano. 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 ever fugl assemblies. This limitation will be adhered to during the entire course of rack installation. In the unlike!y 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 affecte'l 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 kid 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 proceduresi 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 changas do not create a new or different kind of accident.
Replacing the existing 23 Joseph Ost Bcmflex racks with 24 new Holtec racks 1
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 l
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 caoacity 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 fue; 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.
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7 The function of the spent fuel poolis to store fuel assemblies in a suberitical 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 af icable requirements for safe H
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, structumi, thermal hydraulics, and radiological exposure), design basis margins of safety will be maintained. 'nstallation 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 Byrcq 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*F has been demonstrated to be preserved by analysis. Criticality calculations show that k,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.
I 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 publicstion of this hotice will be considered h 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
o frecility, 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
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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 6.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 uterest 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 intervane.
Requests for a hearing and a petition for leave to intervene shall be filed in accordance with the i
Commission's " Rules of Practice for Domestic Licensing Proceedings"in 10 CFR Part 2.
j 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,.
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DC, and at the local public document room located at the Byron Public Library District,109 N.
Fran' lin, P.O. Box 434, Byron, Illinois 61010 for Byron Gtation, ed the Wilmington Public k
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Library,201 S. Kankakee Street, Wilmington, Illinois 60481 for Braidwood Station. If a request 1
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 pet;tionj and the Secretary or the designated Atomic Safety and Licensing Board willissue 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 siffected 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 i
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 prc.eeding, but such an amended petition must satisfy the specificity requirements described above.
l Not later than 15 days prior to the first prehearing conference scheduled in the proceeding, a petitioner shall file a suppiament 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
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, l which the petitioner intends to rely in proving the contention at the hearing. The petitioner must I
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 j
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 preserit 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 ac
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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 R. 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 i134 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
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l Civilian Nuclear Power Reactors"(published at 50 FR 41662 dated October 15,1985). Under
-l 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, l
the riquest 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 lequest 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 entimely request. If the presiding officer grants a request for oral argument, any hearing held on the application must be conducted in accorcance with the hybrid hearing procedures, in essence, those procedures limit the time available for discovery i
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 recuests oral j
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 fne Wilmington Public Library,201 S. Kankakee Street, Wilmington, Illinois 60481 for Braidwood Station.
r t
. Dated at Rockville, Maryland, this 10th day of June 1999.
FOR THE NUCLEAR REGULATORY COMMISSION 7
Stewart N. Bailey, Project Manager, Section 2 Project Directorate 3 Division of Licensing Project Management Office of Nuclear Reactor Regulation
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.p June'10. 1999 I
I MEMORANDUM TO:
81ules and Directives Branch Division of Administrative Services i
Office'of Administration FROM:
Office of Nuclear Reactor Regulation
SUBJECT:
NOTICE OF CONSIDERATI0h 0F ISSUANCE - BRAIDWOOD STATION UNITS 1 AND 2. AND BYRON STATION, UNITS 1 AND 2 j
One signed original of the erbegi.
Oce entbfe I w is ac e r your transmittal to the Office of the Federal Register for publication. Additional conformed copies ( 5
) of the Notice are enclosed for your use.
Notice of Receipt of Application for Construction Permit (s) and Operating License (s).
- Notice of Receipt of Partial Application for Construction Permit (s) and Facility License (s):
Time for submission of Views on Antitrust matters.
Notice of Consideration of issuance of Amendment to Facility Operating License. (Call with 1
30-day insert date).
Notice of Receipt of Application for Facility License (s); Notice of Availability of Applicant's
' Environmental Report; and Notice of Consideration of lasuance of Facility License (s) and Notice of Opportesity for Hearing.
I Notice of Availability of NRC Draft / Final Environmental Statement.
Notice of Limited Work Authorization.
j Notice of Availability of Safety Evaluation Report.
.' Notice of lasuance of Construction Permit (s).
'I i
Notice of.lssuance of Facility Operating License (s) or Amendment (s).
C Order.
Exemption.
Notice of Granting Exemption.
Environmental Assessment.
Notice of Preparation of Environmental Assessment.
Receipt of Petition for Directori Decision Under 10 CFR 2.206.
Issuance of Final Director's D.ecision Under 10 CFR,2.206.
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' Othar:
DOCKET NOS STN 50-454. STN 50-455 STN 50=456 and ~ STN 50-457 Attachment (s): As stated i
Contact:
C. Meere Telephone: 415-3100 DOCUMENT NAME:
Ts receive a copy of this document Indicate in the box:
"C' = Copy withs,ut ettechment/ enclosure
- E' = Copy with attachment / enclosure
- N' = No copy l
On:lCE LPM /PDIIS2 /
l l-l l
,1 l
NAME DWills 220) 1 DATE 06/10/19 OFFICIAL RECORD COPY