[Federal Register Volume 60, Number 193 (Thursday, October 5, 1995)]
[Notices]
[Pages 52222-52226]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: 95-24766]



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NUCLEAR REGULATORY COMMISSION
[Docket Nos. STN 50-456 And STN 50-457]


Commonwealth Edison Company; Notice of Consideration of Issuance 
of Amendments to Facility Operating Licenses, Proposed No Significant 
Hazards Consideration Determination, and Opportunity for a Hearing

    The U.S. Nuclear Regulatory Commission (the Commission) is 
considering issuance of amendments to Facility Operating License Nos. 
NPF-72 and NPF-77, issued to Commonwealth Edison Company for operation 
of the Braidwood Station, Units 1 and 2, located in Will County, 
Illinois.
    The proposed amendments would effectively renew the present 
voltage-based repair criteria in the Braidwood, Unit 1, Technical 
Specifications (TS) which were added to the existing steam generator 
(SG) tube repair criteria by License Amendment No. 54, issued on August 
18, 1994. The differences between the present repair criteria in the 
Braidwood, Unit 1, TSs and those in the pending request to continue 
their use, are discussed below. The need to take action on this matter 
arises partly from the limit placed on the use of the present voltage-
based criteria for only one operating cycle when the license amendment 
cited above was issued.
    The voltage-based repair criteria in the subject TSs are applicable 
only to a specific type of SG tube degradation which is predominantly 
axially-oriented outer diameter stress corrosion cracking (ODSCC). This 
particular form of SG tube degradation occurs entirely within the 
intersections of the SG tubes with the tube support plates (TSP).
    The need to effectively renew the present voltage-based SG tube 
repair criteria is also predicated on the possibility that the NRC 
staff may not find acceptable, a pending request for license amendments 
dated September 1, 1995, for the Byron and Braidwood Stations in 
sufficient time to be applicable for the forthcoming refueling outage 
for Braidwood, Unit 1, presently scheduled to start on September 30, 
1995.
    This request for a 3.0 volt lower voltage limit was first submitted 
on February 13, 1995, and was subsequently superseded by requests for 
license amendments submitted on July 7, 1995, and September 1, 1995. 
All three of these requests for license amendments propose to raise the 
present value of the lower voltage repair limit from 1.0 volt to 3.0 
volts. The license amendment request dated September 1, 1995, 
supersedes the prior two requests on this matter in their entirety.
    The license amendment request dated September 1, 1995, is under 
active review by the staff; however, a number of technical issues 
associated with this pending revision to the present TSs may require 
considerable time to resolve. In the event that the staff is not able 
to resolve these outstanding technical issues prior to the repair of 
the Braidwood, Unit 1, SG tubes presently scheduled to start on or 
about October 15, 1995, the licensee proposes in its request dated 
August 15, 1995, to adopt the SG tube repair criteria contained in 
Generic Letter (GL) 95-05, ``Voltage-Based Repair Criteria for 
Westinghouse Steam Generator Tubes Affected by Outside Diameter Stress 
Corrosion Cracking,'' dated August 3, 1995. 

[[Page 52223]]

    The SG tube voltage-based repair criteria presently in the 
Braidwood, Unit 1, TSs differ slightly from those proposed in the 
licensee's submittal dated August 15, 1995, in that the present repair 
criteria in the TSs were similar to those in the draft generic letter 
on the issue of ODSCC published by the staff on August 12, 1994, while 
the pending proposal is consistent with GL 95-05. This generic letter 
contains repair criteria slightly different from those contained in the 
earlier draft version. These differences reflect the staff's further 
review of this matter, including a review of comments by industry and 
the public.
    In summary, the request for license amendments dated August 15, 
1995, to adopt the voltage-based repair criteria in GL 95-05 will be 
considered by the staff only in the event that the pending request to 
raise the lower voltage limit from 1.0 volt to 3.0 volts can not be 
addressed in a timely manner.
    While the voltage-based repair criteria for ODSCC flaws are 
applicable only to Braidwood, Unit 1, the pending request for license 
amendments involves both units in that the Braidwood Station has a set 
of TSs applicable to both units. Before issuance of the proposed 
license amendments, 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 amendments would not (1) 
involve a significant increase in the probability or consequences of an 
accident previously evaluated; or (2) create the possibility of a new 
or different kind of accident from any accident previously evaluated; 
or (3) involve a significant reduction in a margin of safety.
    As required by 10 CFR 50.91(a), the licensee has provided its 
analysis of the issue of no significant hazards consideration, which is 
presented below:

    1. The proposed change does not involve a significant increase 
in the probability or consequences of an accident previously 
evaluated.
    Consistent with Regulatory Guide (RG) 1.121, ``Basis for 
Plugging Degraded PWR Steam Generator Tubes,'' Revision 0, August 
1976, the traditional depth-based criteria for SG tube repair 
implicitly ensures that tubes accepted for continued service will 
retain adequate structural and leakage integrity during normal 
operating, transient, and postulated accident conditions. It is 
recognized that defects in tubes permitted to remain in service, 
especially cracks, occasionally grow entirely through-wall and 
develop small leaks. Limits on allowable primary-to-secondary 
leakage established in Technical Specifications ensure timely plant 
shutdown before the structural and leakage integrity of the affected 
tube is challenged.
    The proposed license amendment request to implement voltage 
amplitude SG tube support plate APC for Braidwood Unit 1 meets the 
requirements of RG 1.121. The APC methodology demonstrates that tube 
leakage is acceptably low and tube burst is a highly improbable 
event during either normal operation or the most limiting accident 
condition, a postulated main steam line break (MSLB) event.
    During transients, the tube support plate (TSP) is 
conservatively assumed to displace due to the thermal-hydraulic 
loads associated with the transient. This may partially expose a 
crack which is within the boundary of the TSP during normal 
operations to free span conditions. Burst is therefore 
conservatively evaluated assuming the crack is fully exposed to free 
span conditions. The structural eddy current bobbin coil voltage 
limit for free-span burst is 4.75 volts. This limit takes into 
consideration a 1.43 safety factor applied to the steam line break 
differential pressure that is consistent with RG 1.121 requirements. 
With additional considerations for growth rate assumptions and an 
upper 95% confidence estimate on voltage variability, the maximum 
voltage indication that could remain in service is given by the 
upper voltage repair limit equation in Generic Letter 95-05. For 
added conservatism, the allowable indication voltage is further 
reduced in the proposed amendment to a 1.0 volt confirmed ODSCC 
indication limit. All indications greater than 1.0 volt will be 
subject to an RPC examination. Tubes with RPC confirmed outside 
diameter stress corrosion cracking (ODSCC) indications will be 
plugged or sleeved. Any ODSCC indications between 1.0 volt and the 
upper voltage repair limit which are not confirmed as ODSCC will be 
allowed to remain in service since these indications are not as 
likely to affect tube structural integrity or leakage integrity over 
the next operating cycle as the indications that are detectable by 
both bobbin and rotating pancake coil (RPC) inspections.
    The eddy current inspection process has been enhanced to address 
RG 1.83, ``Inservice Inspection of PWR Steam Generator Tubes,'' 
Revision 1, July 1975, considerations as well as the EPRI SG 
Inspection Guidelines. Enhancements in accordance with Generic 
Letter 95-05 are in place to increase detection of ODSCC indications 
and to ensure reliable, consistent acquisition and analysis of data. 
Based on the conservative selection of the voltage criteria and the 
increased ability to identify ODSCC, the probability of tube failure 
during an accident is also not significantly increased due to 
application of requested APC.
    Modification of the Braidwood Specifications for conformance 
with Generic Letter 95-05 requirements does not impact any accidents 
previously evaluated. The decrease in the allowed burst probability 
from 2.5 x 10-2 to 1.0 x 10-2 is conservative.
    Calculations conducted for Braidwood have shown that the 
resulting 2-hour doses at the site boundaries will not currently 
exceed an appropriately small fraction of 10 CFR 100 dose guideline 
values in conjunction with the predicted MSLB leakage calculated in 
accordance with this submittal and a DE I-131 level of 1.0 
Ci/gm. The site allowable leakage calculated using a DE I-
131 level of 1.0 Ci/gm is 9.4 gallons per minute (gpm). 
This leakage includes accident leakage and the allowed 0.1 gpm 
primary-to-secondary leakage of the 3 unfaulted SGs per TS 
3.4.6.2.c. However, in order to provide a defense in depth approach 
to application of this requested APC and to envelope any future 
increases in MSLB leakage due to tube degradation, Braidwood is 
lowering the RCS DE I-131 levels to 0.35 Ci/gm for all 
future cycles until SG replacement. The site allowable leak rate 
calculated using 0.35 Ci/gm DE I-131 is 26.8 gpm. This 
leakage also includes accident leakage and the allowed 0.1 gpm 
primary-to-secondary leakage of the 3 unfaulted SGs per TS 
3.4.6.2.c. Lowering the limit to 0.35 Ci/gm DE I-131 is 
conservative and will not increase the probability or consequences 
of any accidents previously evaluated.
    Renewal of the 1.0 volt IPC for Braidwood Unit 1 does not 
adversely affect steam generator tube integrity and results in 
acceptable dose consequences. Therefore, the proposed license 
amendment request does not result in any significant increase in the 
probability or consequences of an accident previously evaluated 
within the Braidwood Updated Final Safety Analysis Report.
    2. The proposed change does not create the possibility of a new 
or different kind of accident from any accident previously 
evaluated.
    Renewal of the proposed SG tube APC for Braidwood Unit 1 does 
not introduce any significant changes to the plant design basis. Use 
of the criteria does not provide a mechanism which could result in 
an accident outside the tube support plate elevations since industry 
experience indicates that ODSCC originating within the tube support 
plate does not extend significantly beyond the thickness of the 
support plate. This criteria only applies to ODSCC contained within 
the region of the tube bounded by the tube support plate. Therefore, 
neither a single or multiple tube rupture event would be expected in 
a steam generator in which APC has been applied.
    In addressing the combined effects of Loss of Coolant Accident 
(LOCA) coincident with a Safe Shutdown Earthquake (SSE) on the SG 
(as required by General Design Criteria 2), it has been determined 
that tube collapse of select tubes may occur in the SGs at some 
plants, including Braidwood Unit 1. There are two issues associated 
with SG tube collapse. First, the collapse of SG tubing reduces the 
RCS flow area through the tubes. The reduction in flow area 
increases the resistance to flow of steam from the core during a 
LOCA which, in turn, may potentially increase Peak Clad Temperature 
(PCT). Second, there is a potential that partial through-wall cracks 
in tubes could progress to through-wall cracks during tube 
deformation or collapse. A number of tubes 

[[Page 52224]]
have been identified, in the ``wedge'' locations of the SG TSPs, that 
demonstrate the potential for tube collapse during a LOCA + SSE 
event. Because of this potential, these tubes have been excluded 
from application of the voltage-based SG TSP APC.
    ComEd has implemented a maximum primary to secondary leakage 
limit of 150 gallons per day (gpd) through any one SG at Braidwood 
to help preclude the potential for excessive leakage during all 
plant conditions. The 150 gpd limit provides for leakage detection 
and plant shutdown in the event of an unexpected single crack leak 
associated with the longest permissible free span crack length. The 
150 gpd limit provides adequate leakage detection and plant shutdown 
criteria in the event an unexpected single crack results 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 150 gpd limit provides a conservative limit to 
prompt plant shutdown prior to reaching critical crack lengths under 
MSLB conditions.
    Calculations conducted for Braidwood have shown that the 
resulting 2-hour doses at the site boundaries will not currently 
exceed an appropriately small fraction of 10 CFR 100 dose guideline 
values in conjunction with the predicted MSLB leakage calculated in 
accordance with this submittal and a DE I-131 level of 1.0 
Ci/gm. The site allowable leakage calculated using a DE I-
131 level of 1.0 Ci/gm is 9.4 gpm. This leakage includes 
accident leakage and the allowed 0.1 gpm primary-to-secondary 
leakage of the 3 unfaulted SGs per TS 3.4.6.2.c. However, in order 
to provide a defense in depth approach to application of this 
requested APC and to envelope any future increases in MSLB leakage 
due to tube degradation, Braidwood is lowering the RCS DE I-131 
levels to 0.35 Ci/gm for all future cycles until SG 
replacement. The site allowable leak rate calculated using 0.35 
Ci/gm DE I-131 is 26.8 gpm. This leakage also includes 
accident leakage and the allowed 0.1 gpm primary-to-secondary 
leakage of the 3 unfaulted SGs per TS 3.4.6.2.c. Lowering the 
Braidwood Unit 1 RCS DE I-131 concentration limit to the 0.35 
Ci/gm is conservative and will not introduce any changes to 
the design basis for Braidwood Station.
    Modification of the Braidwood Specifications for conformance 
with Generic Letter 95-05 requirements will not alter the plant 
design basis. The decrease in the allowed burst probability from 
2.5 x 10-2 to 1.0 x 10-2 is conservative.
    Upon renewal of the 1.0 volt APC for Braidwood Unit 1, steam 
generator tube integrity continues to be maintained through 
inservice inspection and primary-to-secondary leakage monitoring. 
Therefore, 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.
    The use of the voltage based bobbin coil probe SG TSP APC for 
Braidwood Unit 1 will maintain steam generator tube integrity 
commensurate with the criteria of RG 1.121 as discussed above. Upon 
implementation of the criteria, even under the worst case 
conditions, the occurrence of ODSCC at the TSP elevations is not 
expected to lead to a steam generator tube rupture event during 
normal or faulted plant conditions. The distribution of crack 
indications at the TSP elevations results in acceptable primary-to-
secondary leakage during all plant conditions and radiological 
consequences are not adversely impacted by the application of APC.
    The installation of SG tube plugs and sleeves reduces the RCS 
flow margin. As noted previously, renewal of the SG TSP APC will 
decrease the number of tubes which must be repaired by plugging or 
sleeving. Thus, renewal of APC will retain additional flow margin 
that would otherwise be reduced due to increased tube plugging. 
Therefore, no significant reduction in the margin of safety will 
occur as a result of this proposed license amendment request.
    Although not relied upon to prove adequacy of the proposed 
amendment request, the following analyses demonstrate that 
significant conservatisms exist in the methods and justifications 
described above:

Limited Tube Support Plate Displacement

    An analysis was performed to verify the extent of limited TSP 
displacement during accident conditions (MSLB). Application of 
minimum TSP displacement assumptions provides conservatism and 
reduces the likelihood of a tube burst to negligible levels. 
Consideration of limited TSP displacement would also reduce 
potential MSLB leakage when compared to the leakage calculated 
assuming free span indications.

Probability of Detection

    The Electric Power Research Institute (EPRI) Performance 
Demonstration Program analyzed the performance of approximately 20 
eddy current data analysts evaluating data from a unit with \3/4\'' 
inside diameter and 0.043'' wall thickness tubes. The results of 
this analysis clearly show that the detectability of larger voltage 
indications is increased which lends creditability for application 
of a POD of > 0.6 for ODSCC indications larger than 1.0 volt.

Risk Evaluation of Core Damage

    As part of ComEd's evaluation of the operability of Braidwood 
Unit 1, a risk evaluation was completed. The objective of this 
evaluation was to compare core damage frequency under containment 
bypass conditions, with and without the APC applied at Braidwood 
Unit 1. The total Braidwood core damage frequency is estimated to be 
3.09E-5 per reactor year with a total contribution from containment 
bypass sequences of 3.72E-8 per reactor year according to the 
results of the current individual plant evaluation (IPE). Operation 
with the requested APC resulted in an insignificant increase in core 
damage frequency resulting from MSLB with containment bypass 
conditions.
    Calculations conducted for Braidwood have shown that the 
resulting 2-hour doses at the site boundaries will not currently 
exceed an appropriately small fraction of 10 CFR 100 dose guideline 
values in conjunction with the predicted MSLB leakage calculated in 
accordance with this submittal and a DE I-131 level of 1.0 
Ci/gm. The site allowable leakage calculated using a DE I-
131 level of 1.0 Ci/gm is 9.4 gpm. This leakage includes 
accident leakage and the allowed 0.1 gpm primary-to-secondary 
leakage of the 3 unfaulted SGs per TS 3.4.6.2.c. However, in order 
to provide a defense in depth approach to application of this 
requested APC and to envelope any future increases in MSLB leakage 
due to tube degradation, Braidwood is lowering the RCS DE I-131 
levels to 0.35 Ci/gm for all future cycles until SG 
replacement. The site allowable leak rate calculated using 0.35 
Ci/gm DE I-131 is 26.8 gpm. This leakage also includes 
accident leakage and the allowed 0.1 gpm primary-to-secondary 
leakage of the 3 unfaulted SGs per TS 3.4.6.2.c. Lowering the 
Braidwood Unit 1 RCS DE I-131 concentration limit to the 0.35 
Ci/gm is conservative and will not introduce any changes to 
the design basis for Braidwood Station. Thus this change is in 
conformance with Braidwood's current TS and does not involve a 
reduction in a margin of safety.
    Modification of the Braidwood Specifications for conformance 
with Generic Letter 95-05 requirements will not reduce any safety 
margins. The decrease in the allowed burst probability from 
2.5 x 10-2 to 1.0 x 10-2 is conservative.

    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 
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 amendments until the 
expiration of the 30-day notice period. However, should circumstances 
change during the notice period such that failure to act in a timely 
way would result, for example, in derating or shutdown of the facility, 
the Commission may issue the license amendments before the expiration 
of the 30-day notice period, provided that its final determination is 
that the amendments involve no significant hazards consideration. The 
final determination will consider all public and State comments 
received. Should the Commission take this action, it will publish in 
the Federal Register a notice of issuance and provide for opportunity 
for a hearing after issuance. The Commission expects that the need to 

[[Page 52225]]
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.
    The filing of requests for hearing and petitions for leave to 
intervene is discussed below.
    By November 6, 1995, the licensee may file a request for a hearing 
with respect to issuance of the amendments to the subject facility 
operating licenses 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, and at the local public 
document room located at the Wilmington 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 amendments under consideration. The contention must be 
one which, if proven, would entitle the petitioner to relief. A 
petitioner who fails to file such a supplement which satisfies these 
requirements with respect to at least one contention will not be 
permitted to participate as a party.
    Those permitted to intervene become parties to the proceeding, 
subject to any limitations in the order granting leave to intervene, 
and have the opportunity to participate fully in the conduct of the 
hearing, including the opportunity to present evidence and cross-
examine witnesses.
    If a hearing is requested, the Commission will make a final 
determination on the issue of no significant hazards consideration. The 
final determination will serve to decide when the hearing is held.
    If the final determination is that the 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 amendments.
    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 amendments.
    A request for a hearing or a petition for leave to intervene must 
be filed with the Secretary of the Commission, U.S. Nuclear Regulatory 
Commission, Washington, DC 20555, 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, 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 Mr. 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 60603, 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 amendments dated August 15, 1995, 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 Wilmington Public Library, 201 S. 
Kankakee Street, Wilmington, Illinois 60481.

    Dated at Rockville, Maryland, this 29th day of September 1995.


[[Page 52226]]

    For the Nuclear Regulatory Commission.
George F. Dick,
Senior Project Manager, Project Directorate III-2, Division of Reactor 
Projects--III/IV, Office of Nuclear Reactor Regulation.
[FR Doc. 95-24766 Filed 10-4-95; 8:45 am]
BILLING CODE 7590-01-P