[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