[Federal Register Volume 59, Number 131 (Monday, July 11, 1994)]
[Unknown Section]
[Page 0]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: 94-16695]
[[Page Unknown]]
[Federal Register: July 11, 1994]
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NUCLEAR REGULATORY COMMISSION
[Docket No. STN 50-456]
Commonwealth Edison Co.; Consideration of Issuance of Amendment
to Facility Operating License, Proposed No Significant Hazards
Consideration Determination, and Opportunity for a Hearing
The U.S. Nuclear Regulatory Commission (the Commission) is
considering issuance of an amendment to Facility Operating License No.
NPF-72 issued to Commonwealth Edison Company (CECo, the licensee) for
operation of the Braidwood Station, Unit 1, located in Will County,
Illinois.
The proposed amendment would revise the Braidwood, Unit 1,
Technical Specifications (TSs) to remove the condition limiting
operation of the facility to 100 days during the present fuel cycle
when Thot is greater than 500 deg.F and to restore the reactor
coolant dose equivalent Iodine-131 limit to 1 microcurie per gram of
coolant from the present value of 0.35. Both the limit on permissible
operational time and the reduction in the permissible level of Iodine-
131 were incorporated into the TSs by Amendment No. 50 issued to
Facility Operating License No. NPF-72 for Braidwood Station, Unit 1, on
May 7, 1994.
Before issuance of the proposed license amendment, the Commission
will have made findings 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 amendment
request involves no significant hazards consideration. Under the
Commission's regulations in 10 CFR 50.92, this means that operation of
the facility in accordance with the proposed amendment would not (1)
involve a significant increase in the probability or consequences of an
accident previously evaluated; or (2) create the possibility of a new
or different kind of accident from any accident previously evaluated;
or (3) involve a significant reduction in a margin of safety. As
required by 10 CFR 50.91(a), the licensee has provided its analysis of
the issue of no significant hazards consideration, which is presented
below:
1. The proposed change does not involve a significant increase in
the probability or consequences of an accident previously evaluated.
Braidwood Unit 1 TS Amendment 50 imposed a 100 calendar days with
Thot greater than 500 deg.F operating limit on Unit 1. This
limitation was a consequence of the amount of main steam line break
(MSLB) leakage predicted in Braidwood Station's April 30, 1994,
submittal. These predictions were made using the Log-Logistic method of
draft NUREG 1477, ``Voltage Based Interim Plugging for Steam Generator
Tubes--Task Group Report,'' with the Dose Equivalent Iodine-131 limit
of Specification 3.4.8 reduced from 1.0 microcurie per gram
(Ci/gm) to 0.35 Ci/gm. However, WCAP 14046;
``Braidwood Unit 1 Technical Support for Cycle 5 Steam Generator
Interim Plugging Criteria'' (WCAP-14046), docketed June 10, 1994, as
required by Braidwood Station's April 25, 1994, submittal, has shown
using the Electric Power Research Institute (EPRI) Leakrate Correlation
that projected End Of Cycle (EOC)-5 MSLB leakage is 3.1 gallons per
minute (gpm) which is less than the allowable limit of 9.1 gpm for
Braidwood Unit 1. This analysis is discussed in detail in WCAP-14046.
Thus the Unit 1 100-day operating limit and reactor coolant dose
equivalent iodine restriction imposed by Amendment 50 on the basis of
MSLB leakage is no longer required.
In addition to the 100-day, leakage based limit, the Nuclear
Regulatory Commissions (NRC) Safety Evaluation Report (SER), issued May
7, 1994, in support of Braidwood Station's Unit 1 TS Amendment 50
discusses a 4.6 month (138 day) limit derived from a deterministic
assessment of SG tube burst probability. To address the issue of tube
burst for full cycle operation Braidwood Station's April 25, 1994,
submittal provided a probabilistic risk assessment which is restated
below.
As part of ComEd's evaluation of the operability of Braidwood Unit
1 Cycle 5, a risk evaluation was completed. The objective of this
evaluation was to compare core damage frequency, with containment
bypass, with and without the interim plugging criteria applied at
Braidwood 1.
ComEd has evaluated the impact of operation using the proposed
interim plugging criteria against the results of insights from the
draft Braidwood Individual Plant Examination (IPE). Braidwood Station
is scheduled to docket its IPE June 30, 1994. Byron Station's IPE was
docketed April 20, 1994. The SG sections of these documents are
identical.
While the Braidwood IPE is not in its final form, it is believed
that the quantification in hand is sufficiently robust to allow a
validation assessment of the impact of such operation. The ComEd
evaluation parallels that described in the NRC Staff's SER for Palo
Verde Unit 2 dated August 19, 1993.
The values calculated in WCAP-14046, for Beginning of Cycle (BOC) 5
and EOC 5 using 0.6 Probability Of Detection (POD) were used to develop
a cycle average burst probability. Another BOC 5 burst probability
assuming a POD of 0.6 for indications less than 3 volts and 1.0 for
indications greater than 3 volts was used to evaluate the impact of POD
on core damage frequency.
The total Braidwood core damage frequency is estimated to be 2.74E-
5 per reactor year with a total contribution from containment bypass
sequences of 2.9E-8 per reactor year in the current IPE. Operation with
the alternate repair criteria with a variable POD is expected to
increase the MSLB with containment bypass sequence frequency
contribution by a factor of only 10%. An upper bound increase of a
factor of two is derived when the fixed POD of 0.6 is employed in the
calculation. Neither increase is significant from a risk perspective.
The reason for a reduced core damage frequency with a higher POD is
that large voltage indications have a high assurance of being
identified and removed from service during inspection. Therefore, the
calculation of burst probability during MSLB changes because of
differences in the assumed distribution of indications left in service
at BOC. The EOC burst probability also changes because the growth
distribution is added to the new BOC distribution of indications. The
result of this change is a significant reduction in burst probability
during MSLB.
Therefore, the operation of Braidwood Unit 1 Cycle 5 for a complete
18 month fuel cycle with the application of the one volt IPC does not
significantly increase the core damage frequency even with the
conservative assumption of a POD of 0.6 and application of the full
growth rate distribution observed during Cycle 4.
To further address SG tube burst probability, the following
qualitative discussion of limited tube support plate (TSP) displacement
is provided. As part of ComEd's technical support for the
implementation of IPC at Braidwood Unit 1, numerous quantitative
analyses were completed to assure the structural integrity of the SG
tubing. These quantitative determinations were provided as part of
WCAP-14046. These analyses focused on the quantifiable elements of the
IPC to evaluate the impact of crack length on steam generator tube
leakage and burst, and were completed consistent with the guidance
provided in draft RG 1.121, ``Bases for Plugging Degraded SG Tubes.''
The bases for these calculations are the analyses completed by the
utility industry and reported to the NRC in the EPRI Draft Report TR-
100407, ``PWR Steam Generator Tube Repair Limits--Technical Support
Document for Outside Diameter Stress Corrosion Cracking at Tube Support
Plates'', Revision 1, August 1993. As explained in this document, the
analyses have been completed to assure that the general design criteria
and the requirements of RG 1.121 are met during plant operation.
In the preparation of these industry documents and the Braidwood
Unit 1 specific WCAP-14046, all analyses for leakage and burst
potential were completed using the extremely conservative assumption
that all Outside Diameter Stress Corrosion Cracking (ODSCC) indications
occur on the tubing freespan. In fact, as indicated in both WCAP-14046
and EPRI Draft Report TR-100407, ODSCC degradation is confined to the
region of the tube/TSP intersection. The burst capability of a section
of tube containing ODSCC indications and located within the tube/TSP
intersection substantially exceeds the burst capability of a freespan
tube section without ODSCC indications. Therefore, tubing left in
service by Braidwood's Unit 1 IPC amendment will not burst when
confined by the tube support plates.
In fact, it is highly unlikely that a section of tubing within the
tube support plate will leak, even with through wall cracks.
To assure structural integrity of the tubing, even during a MSLB
accident, ComEd undertook extensive analyses, presented as part of
WCAP-14046, to show analytically that the TSP's do not move far enough
during a MSLB to allow degraded tubes to uncover, and subsequently,
result in increased leakage.
A Generic Model D-4 SG Limited Support Plate Motion Analysis is
also being conducted and should be submitted to the NRC by the end of
August, 1994.
This analysis is being performed using the following assumptions:
1. The TSP crevices are clean,
2. The TSPs are free to move, depending on applied loads, along the
length of the SG tube, and
3. Movement of the TSPs along the length of the tube is not
restricted by bending or distortion of the SG tube hole.
Each of these base assumptions is extremely conservative in its own
right:
1. For assumption 1, visual inspections of the secondary side of
the tube bundles of Braidwood Unit 1 SGs show some quality of deposits
in the tube to TSP crevice, and along the length of the tube. Since
these deposits are considered to be a possible factor in causing ODSCC,
it is likely that any tube having ODSCC indications has deposits in the
tube/TSP intersection. These deposits would tend to close the tube to
TSP crevice, restricting by friction the ability of the TSP to move
along the tubes as loads are applied to the TSP during a MSLB.
2. With regards to assumptions 2 and 3, the TSPs tend to flex and
in some locations, are constrained by tie-rods and wedges attached to
the tube bundle shroud. These constraints tend to cause the TSPs to
ripple under the applied loads as indicated in WCAP-14046. This effect
tends to distort the shape of the tube holes, which are fitted to a
tight tolerance around the tubes. Therefore, any distortion of these
tube holes caused by motion of the TSP will tend to cause the TSP to
bind against the outside diameter of the tube, further constraining its
movement away from the degraded area of the tubing.
The impact of these facts will lessen the ability of the TSP to
move, thereby significantly reducing the possibility that a degraded
section of tubing would become uncovered during a MSLB.
Thus, this proposed license amendment request does not result in
any increase in the probability or consequences of an accident
previously evaluated within the Braidwood Updated Final Safety Analysis
Report (UFSAR).
2. The proposed change does not create the possibility of a new or
different kind of accident from any accident previously evaluated.
Approval of this proposed change does not introduce any significant
changes to the plant design basis. Removal of the Amendment 50 and SER
operating limits for Unit 1 does not provide a mechanism which could
result in a new or different kind of accident. Neither a single or
multiple tube rupture event would be expected in a SG in which the IPC
has been applied.
ComEd has implemented a maximum leakage rate limit of 150 gallons
per day (gpd) through any one SG to help preclude the potential for
excessive leakage during all plant conditions. The RG 1.121 criterion
for establishing operational leakage rate limits that require plant
shutdown are based upon leak-before-break considerations to detect a
free span crack before potential tube rupture during faulted plant
conditions. The 150 gpd limit will provide for leakage detection and
plant shutdown in the event of the occurrence of an unexpected single
crack resulting in leakage that is associated with the longest
permissible free span crack length. Since tube burst is precluded
during normal operation due to the proximity of the TSP to the tube and
the potential exists for the crevice to become uncovered during MSLB
conditions, the leakage from the maximum permissible crack must
preclude tube burst at MSLB conditions. Thus, the 105 gpd limit
provides for plant shutdown prior to reaching critical crack lengths
for MSLB conditions.
As SG tube integrity will continue to be maintained upon approval
of this amendment request through inservice inspection and primary-to-
secondary leakage monitoring, the possibility of a new or different
kind of accident from any previously evaluated is not created.
3. The proposed change does not involve a significant reduction in
a margin of safety.
Braidwood Unit 1 TS Amendment 50 imposed a 100 calendar days with
Thot greater than 500 deg.F operating limit on Unit 1. This
limitation was a consequence of the amount of MSLB leakage predicted in
Braidwood Station's April 30, 1994, submittal. These predictions were
made using the Log-Logistic method of draft NUREG 1477, with the Dose
Equivalent Iodine-131 limit of Specification 3.4.8 reduced from 1.0
Ci/gm to 0.35 Ci/gm. However, WCAP 14046, docketed
June 10, 1994, as required by Braidwood Station's April 25, 1994,
submittal, has shown using the EPRI Leakrate Correlation that projected
EOC-5 MSLB leakage is 3.1 gpm which is less than the allowable limit of
9.1 gpm for Braidwood Unit 1. This analysis is discussed in detail in
WCAP-14046.
Thus the Unit 1 operating limit imposed by Amendment 50 on the
basis of MSLB leakage is no longer required.
In addition to the 100 day, leakage based limit, the Nuclear
Regulatory Commissions (NRC) Safety Evaluation Report (SER), issued May
7, 1994, in support of Braidwood Station's Unit 1 TS Amendment 50
discusses a 4.6 month (138 day) limit derived from a deterministic
assessment of SG tube burst probability. To address the issue of tube
burst for full cycle operation Braidwood Station's April 25, 1994,
submittal provided a probabilistic risk assessment which is restated
below.
As part of ComEd's evaluation of the operability of Braidwood Unit
1 Cycle 5, a risk evaluation was completed. The objective of this
evaluation was to compare core damage frequency, with containment
bypass, with and without the interim plugging criteria applied at
Braidwood 1.
ComEd has evaluated the impact of operation using the proposed
interim plugging criteria against the results of insights from the
draft Braidwood IPE. Braidwood Station is scheduled to docket its IPE
June 30, 1994. Byron Station's IPE was docketed April 20, 1994. The SG
sections of these documents are identical. While the Braidwood IPE is
not in its final form, it is believed that the quantification in hand
is sufficiently robust to allow a validation assessment of the impact
of such operation. The ComEd evaluation parallels that described in the
NRC Staff's SER for Palo Verde Unit 2 dated August 19, 1993.
The values calculated in WCAP-14046, for BOC 5 and EOC 5 using 0.6
POD were used to develop a cycle average burst probability. Another BOC
5 burst probability assuming a POD of 0.6 for indications less than 3
volts and 1.0 for indications greater than 3 volts was used to evaluate
the impact of POD on core damage frequency.
The total Braidwood core damage frequency is estimated to be 2.74E-
5 per reactor year with a total contribution from containment bypass
sequences of 2.9E-8 per reactor year in the current IPE. Operation with
the alternate repair criteria with a variable POD is expected to
increase the MSLB with containment bypass sequence frequency
contribution by a factor of only 10%. An upper bound increase of a
factor of two is derived when the fixed POD of 0.6 is employed in the
calculation. Neither increase is significant from a risk perspective.
The reason for a reduced core damage frequency with a higher POD is
that large voltage indications have a high assurance of being
identified and removed from service during inspection. Therefore, the
calculation of burst probability during MSLB changes because of
differences in the assumed distribution of indications left in service
at BOC. The EOC burst probability also changes because the growth
distribution is added to the new BOC distribution of indications. The
result of this change is a significant reduction in burst probability
during MSLB.
Therefore, the operation of Braidwood Unit 1 Cycle 5 for a complete
18 month fuel cycle with the application of the one volt IPC does not
significantly increase the core damage frequency even with the
conservative assumption of a POD of 0.6 and application of the full
growth rate distribution observed during Cycle 4.
To further address SG tube burst probability, the following
qualitative discussion of limited TSP displacement is provided.
As part of ComEd's technical support for the implementation of IPC
at Braidwood Unit 1, numerous quantitative analyses were completed to
assure the structural integrity of the SG tubing. These quantitative
determinations were provided as part of WCAP-14046. These analyses
focused on the quantifiable elements of the IPC to evaluate the impact
of crack length on steam generator tube leakage and burst, and were
completed consistent with the guidance provided in draft RG 1.121.
The bases for these calculations are the analyses completed by the
utility industry and reported to the NRC in the EPRI draft report TR-
100407. As explained in this document, the analyses have been completed
to assure that the general design criteria and the requirements of RG
1.121 are met during plant operation.
In the preparation of these industry documents and the Braidwood
Unit 1 specific WCAP-14046, all analyses for leakage and burst
potential were completed using the extremely conservative assumption
that all ODSCC indications occur on the tubing freespan. In fact, as
indicated in both WCAP-14046 and EPRI Draft Report TR-100407, ODSCC
degradation is confined to the region of the tube/TSP intersection. The
burst capability of a section of tube containing ODSCC indications and
located within the tube/TSP intersection substantially exceeds the
burst capability of a freespan tube section without ODSCC indications.
Therefore, tubing left in service by Braidwood's Unit 1 IPC amendment
will not burst when confined by the tube support plates.
In fact, it is highly unlikely that a section of tubing within the
tube support plate will leak, even with through wall cracks.
To assure structural integrity of the tubing, even during a MSLB
accident, ComEd undertook extensive analyses, presented as part of
WCAP-14046, to show analytically that the TSPs do not move far enough
during a MSLB to allow degraded tubes to uncover, and subsequently,
result in increase leakage.
A Generic Model D-4 SG Limited Support Plate Motion Analysis is
also being conducted and should be submitted to the NRC by the end of
August, 1994.
This analysis is being performed using the following assumptions:
1. The TSP crevices are clean,
2. The TSPs are free to move, depending on applied loads, along the
length of the SG tube, and
3. Movement of the TSPs along the length of the tube is not
restricted by bending or distortion of the SG tube hole.
Each of these base assumptions is extremely conservative in its own
right:
1. For assumption 1, visual inspections of the secondary side of
the tube bundles of Braidwood Unit 1 SGs show some quantity of deposits
in the tube to TSP crevice, and along the length of the tube. Since
these deposits are considered to be a possible factor in causing ODSCC,
it is likely that any tube having ODSCC indications has deposits in the
tube/TSP intersection. These deposits would tend to close the tube to
TSP crevice, restricting by friction the ability of the TSP to move
along the tubes as loads are applied to the TSP during a MSLB.
2. With regards to assumptions 2 and 3, the TSPs tend to flex and
in some locations, are constrained by tie-rods and wedges attached to
the tube bundle shroud. These constraints tend to cause the TSPs to
ripple under the applied loads as indicated in WCAP-14046. This effect
tends to distort the shape of the tube holes, which are fitted to a
tight tolerance around the tubes. Therefore, any distortion of these
tube holes caused by motion of the TSP will tend to cause the TSP to
bind against the outside diameter of the tube, further constraining its
movement away from the degraded area of the tubing.
The impact of these facts will lessen the ability of the TSP to
move, thereby significantly reducing the possibility that a degraded
section of tubing would become uncovered during a MSLB.
This evidence, in conjunction with the probability of occurrence of
a MSLB, and the probabilistic assessment of the consequences of a MSLB,
results in the substantially increased assurance that the consequences
of a MSLB will be significantly less severe than those assessed in
WCAP-14046 and the generic Model D-4 SG Limited Support Plate Motion
Analyses.
Thus, this proposed change does not involve a significant reduction
in a margin of safety.
The NRC staff has reviewed the pertinent portions of the licensee's
analysis and, based on this review, it appears that the three standards
of 10 CFR 50.92(c) are satisfied. This staff finding is partially based
on the licensee's usage of a constant value for the Probability of
Detection (POD) of 0.6 as recommended in draft NUREG-1447. This is
consistent with the staff's position in the Safety Evaluation (SE) it
issued in support of Amendment No. 50 to the Braidwood, Unit 1,
operating license. While the licensee also discussed in its analysis
the usage of a higher value for the POD, the staff did not rely on
this. Therefore, the NRC staff proposes to determine that the amendment
request involves 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 amendment 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 amendment before the expiration of
the 30-day notice period, provided that its final determination is that
the amendment involves 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 Rules Review and
Directives Branch, Division of Freedom of Information and Publications
Services, Office of Administration, U.S. Nuclear Regulatory Commission,
Washington, DC 20555, and should cite the publication date and page
number of this Federal Register notice. Written comments may also be
delivered to Room 6D22, Two White Flint North, 11545 Rockville Pike,
Rockville, Maryland, from 7:30 a.m. to 4:15 p.m. Federal workdays.
Copies of written comments received may be examined at the NRC Public
Document Room, the Gelman Building, 2120 L Street, NW., Washington, DC
20555.
The filing of requests for hearing and petitions for leave to
intervene is discussed below.
By August 10, 1994, the licensee may file a request for a hearing
with respect to issuance of the amendment to the subject facility
operating license and any person whose interest may be affected by this
proceeding and who wishes to participate as a party in the proceeding
must file a written request for a hearing and a petition for leave to
intervene. Requests for a hearing and a petition for leave to intervene
shall be filed in accordance with the Commission's ``Rules of Practice
for Domestic Licensing Proceedings'' in 10 CFR Part 2. Interested
persons should consult a current copy of 10 CFR 2.714 which is
available at the Commission's Public Document Room, the Gelman
Building, 2120 L Street, NW., Washington, DC 20555 and at the local
public document room located at Wilmington Township Public Library, 201
S. Kankakee Street, Wilmington, Illinois 60481. 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 amendment 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 amendment request involves
no significant hazards consideration, the Commission may issue the
amendment and make it immediately effective, notwithstanding the
request for a hearing. Any hearing held would take place after issuance
of the amendment.
If the final determination is that the amendment request involves a
significant hazards consideration, any hearing held would take place
before the issuance of any amendment.
A request for a hearing or a petition for leave to intervene must
be filed with the Secretary of the Commission, U.S. Nuclear Regulatory
Commission, Washington, DC 20555, Attention: Docketing and Services
Branch, or may be delivered to the Commission's Public Document Room,
the Gelman Building, 2120 L Street, NW., Washington, DC 20555, by the
above date. Where petitions are filed during the last 10 days of the
notice period, it is requested that the petitioner promptly so inform
the Commission by a toll-free telephone call to Western Union at 1-
(800) 248-5100 (in Missouri 1-(800) 342-6700). The Western Union
operator should be given Datagram Identification Number N1023 and the
following message addressed to Robert A. Capra: petitioner's name and
telephone number, date petition was mailed, plant name, and publication
date and page number of this Federal Register notice. A copy of the
petition should also be sent to the Office of the General Counsel, U.S.
Nuclear Regulatory Commission, Washington, DC 20555, and to Michael I.
Miller, Esquire; Sidley and Austin, One First National Plaza, Chicago,
Illinois 60690, 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).
For further details with respect to this action, see the
application for amendment dated June 20, 1994, which is available for
public inspection at the Commission's Public Document Room, the Gelman
Building, 2120 L Street, NW., Washington, DC 20555 and at the local
public document room located at Wilmington Township Public Library, 201
S. Kankakee Street, Wilmington, Illinois 60481.
Dated at Rockville, Maryland, this 1st day of July 1994.
For the Nuclear Regulatory Commission.
Ramin R. Assa,
Acting Project Manager, Project Directorate III-2, Division of Reactor
Projects--IV/V, Office of Nuclear Reactor Regulation.
[FR Doc. 94-16695 Filed 7-8-94; 8:45 am]
BILLING CODE 7590-01-M
