[Federal Register Volume 62, Number 88 (Wednesday, May 7, 1997)]
[Notices]
[Pages 24984-24997]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: X97-10507]


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NUCLEAR REGULATORY COMMISSION

Biweekly Notice


Applications and Amendments to Facility Operating Licenses 
Involving No Significant Hazards Considerations

I. Background

    Pursuant to Public Law 97-415, the U.S. Nuclear Regulatory 
Commission (the Commission or NRC staff) is publishing this regular 
biweekly notice. Public Law 97-415 revised section 189 of the Atomic 
Energy Act of 1954, as amended (the Act), to require the Commission to 
publish notice of any amendments issued, or proposed to be issued, 
under a new provision of section 189 of the Act. This provision grants 
the Commission the authority to issue and make immediately effective 
any amendment to an operating license upon a determination by the 
Commission that such amendment involves no significant hazards 
consideration, notwithstanding the pendency before the Commission of a 
request for a hearing from any person.
    This biweekly notice includes all notices of amendments issued, or 
proposed to be issued from April 12, 1997, through April 25, 1997. The 
last biweekly notice was published on April 23, 1997 (62 FR 19825).

Notice Of Consideration Of Issuance Of Amendments To Facility 
Operating Licenses, Proposed No Significant Hazards Consideration 
Determination, And Opportunity For A Hearing

    The Commission has made a proposed determination that the following 
amendment requests involve no significant hazards consideration. Under 
the Commission's regulations in 10 CFR 50.92, this means that operation 
of the facility in accordance with the proposed 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. The basis 
for this proposed determination for each amendment request is shown 
below.
    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 
before action is taken. Should the Commission take this action, it will 
publish in the Federal Register a notice of issuance and provide for 
opportunity for a hearing after issuance. The Commission expects that 
the need to take this action will occur very infrequently.
    Written comments may be submitted by mail to the Chief, Rules 
Review and Directives Branch, Division of Freedom of Information and 
Publications Services, Office of Administration, U.S.

[[Page 24985]]

Nuclear Regulatory Commission, Washington, DC 20555-0001, and should 
cite the publication date and page number of this Federal Register 
notice. Written comments may also be delivered to Room 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 a hearing and 
petitions for leave to intervene is discussed below.
    By June 6, 1997, 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 and at the local public 
document room for the particular facility involved. 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 a 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 a 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 20555-0001, 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 (Project Director): 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 20555-0001, 
and to the attorney for the licensee.
    Nontimely filings of petitions for leave to intervene, amended 
petitions, supplemental petitions and/or requests for a hearing will 
not be entertained absent a determination by the Commission, the 
presiding officer or the Atomic Safety and Licensing Board that the 
petition and/or request should be granted based upon a balancing of 
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 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 for 
the particular facility involved.

Duquesne Light Company, et al., Docket No. 50-334, Beaver Valley 
Power Station, Unit No. 1, Shippingport, Pennsylvania

    Date of amendment request: March 10, 1997
    Description of amendment request: The proposed amendment would 
modify the Technical Specifications (TSs) by reducing the reactor 
coolant system (RCS) specific activity limits in accordance with 
Generic Letter 95-05. The definition of DOSE EQUIVALENT I-131 would be 
replaced with the Improved Standard TS definition wording in the first 
sentence and an

[[Page 24986]]

equation added based on dose conversion factors derived from 
International Commission on Radiation Protection (ICRP) ICRP-30. TS 
3.4.8, Specific Activity, would be revised by reducing the DOSE 
EQUIVALENT I-131 limit from 1.0 [micro]Ci[curies]/gram to 0.35 
[micro]Ci[curies]/gram. Item 4.a in TS Table 4.4-12, Primary Coolant 
Specific Activity Sample and Analysis Program, TS Figure 3.4-1, and the 
Bases for TS 3/4.4.8 would be modified to reflect the reduced DOSE 
EQUIVALENT I-131 limit.
    Basis for proposed no significant hazards consideration 
determination: 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. Does the change involve a significant increase in the 
probability or consequences of an accident previously evaluated?
    The proposed change reduces the reactor coolant system (RCS) 
specific activity limits of Specification 3.4.8 from 1.0 [micro]Ci/
gram to 0.35 [micro]Ci/gram and lowers the graph in Figure 3.4-1 by 
39 [micro]/Ci gram following the guidance provided in Generic Letter 
(GL) 95-05. This reduces the RCS activity allowed to leak to the 
secondary side when the plant is operating so that additional margin 
is available to support a higher allowable accident-induced leakage 
value as justified by analysis.
    The proposed changes to Specification 3.4.8 and the definition 
of DOSE EQUIVALENT I-131 ensure these requirements are consistent 
the latest analyses.
    These changes implement the more restrictive RCS activity limits 
in accordance with applicable analyses and GL 95-05 to ensure the 
regulations are satisfied. Therefore, these changes do not involve a 
significant increase in the probability or consequences of an 
accident previously evaluated.
    2. Does the change create the possibility of a new or different 
kind of accident from any accident previously evaluated?
    The proposed change does not alter the configuration of the 
plant or affect the operation with the reduced specific activity 
limit. By reducing the specific activity limit, the limit would be 
reached sooner to initiate evaluation of the out of limit condition. 
The proposed changes will not result in any additional challenges to 
the main steam system or the reactor coolant system pressure 
boundary. Consequently, no new failure modes are introduced as a 
result of the proposed changes. As a result, the main steam line 
break, steam generator tube rupture and loss of coolant accident 
analyses remain bounding. Therefore, the proposed change will not 
create the possibility of a new or different kind of accident from 
any accident previously evaluated.
    3. Does the change involve a significant reduction in a margin 
of safety?
    The proposed change reduces the RCS specific activity limit to 
0.35 [micro]Ci/gram along with lowering the Figure 3.4-1 limits by 
39 [micro]Ci/gram. Reduction of the RCS specific activity limits 
allows an increase in the limit for the projected SG [steam 
generator] leakage following SG tube inspection and repair in 
accordance with the voltage-based SG tube alternate repair criteria 
(ARC) incorporated by Amendment No. 198. This follows the guidance 
provided in GL 95-05 and effectively takes margin available in the 
specific activity limits and applies it to the projected SG leakage 
for the ARC. This has been determined to be an acceptable means for 
accepting higher projected leakage rates while still meeting the 
applicable limits of 10 CFR [Part] 100 and GDC [General Design 
Criterion] 19 with respect to offsite and control room doses.
    The capability for monitoring the specific activity and 
complying with the required actions remains unchanged. In addition, 
there is no resultant change in dose consequences. Therefore, the 
proposed change does not involve a significant reduction in a margin 
of safety.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: B. F. Jones Memorial Library, 
663 Franklin Avenue, Aliquippa, PA 15001
    Attorney for licensee: Jay E. Silberg, Esquire, Shaw, Pittman, 
Potts & Trowbridge, 2300 N Street, NW., Washington, DC 20037
    NRC Project Director: John F. Stolz

Duquesne Light Company, et al., Docket Nos. 50-334 and 50-412, 
Beaver Valley Power Station, Unit Nos. 1 and 2, Shippingport, 
Pennsylvania

    Date of amendment request: March 14, 1997
    Description of amendment request: The proposed amendment would 
relocate the following administrative control technical specifications 
(TSs) from the Beaver Valley Power Station, Unit Nos. 1 and 2 (BVPS-1 
and BVPS-2) TSs to the quality assurance program description, which is 
presented in Section 17.2 of the BVPS-2 Updated Final Safety Analysis 
Report (UFSAR). Section 17.2 of the BVPS-2 UFSAR contains the quality 
assurance program description for both BVPS-1 and BVPS-2. The licensee 
stated that the proposed changes are based on NRC Administrative Letter 
95-06, ``Relocation of Technical Specification Administrative Controls 
Related to Quality Assurance.''
    BVPS-2 TS 6.2.3 (Independent Safety Evaluation Group)
    BVPS-1 and BVPS-2 TS 6.5.1 (Onsite Safety Committee)
    BVPS-1 and BVPS-2 TS 6.5.2 (Offsite Review Committee)
    BVPS-1 and BVPS-2 TS 6.8.2 (Procedures, Review and Approval)
    BVPS-1 and BVPS-2 TS 6.8.3 (Temporary Procedure Changes, Review and 
Approval)
    BVPS-1 and BVPS-2 TS 6.10.1 (Records Retention, At least 5 years)
    BVPS-1 and BVPS-2 TS 6.10.2 (Records Retention, Duration of 
Operating License)
    Basis for proposed no significant hazards consideration 
determination: 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. Does the change involve a significant increase in the 
probability or consequences of an accident previously evaluated?
    This proposed change would relocate technical specification 
administrative controls to the quality assurance program 
description. Adequate controls are provided by the established 
quality assurance program change process in 10 CFR 50.54(a).
    The provisions of Technical Specification 6.2.3.2 which states 
that: ``The ISEG [Independent Safety Evaluation Group] shall be 
composed of at least five, dedicated, full-time engineers located on 
site,'' would be omitted from the provisions relocated to the 
quality assurance program description. Since no system, component or 
operational procedure changes are involved, and the ISEG function 
will continue to be implemented, the change can have no effect on 
safe operation of the plant.
    The likelihood that an accident will occur is not increased by 
this proposed technical specification change which involves 
administrative controls. No systems, equipment, or components are 
affected by the proposed change. Thus, the consequences of a 
malfunction of equipment important to safety previously evaluated in 
the Updated Final Safety Analysis Report (UFSAR) are not increased 
by this change.
    Relocation of technical specification provisions and related 
changes do not affect possible initiating events for accidents 
previously evaluated or any system functional requirement. The 
proposed changes have no impact on accident initiators or plant 
equipment, and do not affect the probabilities or consequences of an 
accident.
    Therefore, the proposed changes will not involve a significant 
increase in the probability or consequences of a previously 
evaluated accident.
    2. Does the change create the possibility of a new or different 
kind of accident from any accident previously evaluated?
    The proposed relocation of technical specification provisions to 
the quality assurance program description and related changes do not 
involve changes to the physical plant or operations. Since the 
proposed changes to administrative controls do not affect equipment 
or its operation, they cannot contribute to accident initiation and

[[Page 24987]]

cannot produce a new accident scenario or a new type of equipment 
malfunction.
    Therefore, the proposed change does not create the possibility 
of a new or different kind of accident from any accident previously 
evaluated.
    3. Does the change involve a significant reduction in a margin 
of safety?
    The proposed changes are administrative in nature and do not 
directly affect plant equipment or operation. Safety limits and 
limiting safety system settings are not affected by this proposed 
change. The proposed changes do not affect the UFSAR design bases, 
accident assumptions, or technical specification bases. In addition, 
the proposed changes do not affect release limits, monitoring 
equipment or practices.
    Therefore, the proposed changes would not involve a significant 
reduction in the margin of safety.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: B. F. Jones Memorial Library, 
663 Franklin Avenue, Aliquippa, PA 15001
    Attorney for licensee: Jay E. Silberg, Esquire, Shaw, Pittman, 
Potts & Trowbridge, 2300 N Street, NW., Washington, DC 20037.
    NRC Project Director: John F. Stolz

Entergy Operations Inc., Docket No. 50-382, Waterford Steam 
Electric Station, Unit 3, St. Charles Parish, Louisiana

    Date of amendment request: April 11, 1997
    Description of amendment request: The proposed amendment modifies 
Technical Specification (TS) 3.3.3.7.3 and Surveillance Requirement 
4.3.3.7.3 for the broad range gas detection system at Waterford Steam 
Electric Station, Unit 3. The proposed change also includes changes in 
TS Basis 3/4.3.3.7.3 to support the changes.
    Basis for proposed no significant hazards consideration 
determination: 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. Will operation of the facility in accordance with this 
proposed change involve a significant increase in the probability or 
consequences of an accident previously evaluated?
    Response: No.
    The broad range gas detection system has no effect on the 
accidents analyzed in chapter 15 of the Final Safety Analysis 
Report. It's only effect is on habitability of the control room, 
which will be enhanced by installation of the new monitoring system 
and this change to the Technical Specifications. Analysis has shown 
that the impact on operator incapacitation and subsequent core 
damage risk of this background check is negligible.
    Therefore, the proposed change will not involve a significant 
increase in the probability or consequences of any accident 
previously evaluated.
    2. Will operation of the facility in accordance with this 
proposed change create the possibility of a new or different type of 
accident from any accident previously evaluated?
    Response: No.
    The proposed Technical Specification change in itself does not 
change the design or configuration of the plant. The new system for 
broad range toxic gas monitoring performs the same function as the 
old system, but it accomplishes this with a more sophisticated 
system that increases reliability.
    Therefore, the proposed change will not create the possibility 
of a new or different kind of accident from any accident previously 
evaluated.
    3. Will operation of the facility in accordance with this 
proposed change involve a significant reduction in a margin of 
safety?
    Response: No.
    The broad range gas detection system has no effect on a margin 
of safety as defined by Section 2 of the Technical Specifications. 
It's only effect is on habitability of the control room, which will 
be enhanced by installation of the new monitoring system and this 
change to the Technical Specifications.
    Therefore, the proposed change will not involve a significant 
reduction in a margin of safety.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: University of New Orleans 
Library, Louisiana Collection, Lakefront, New Orleans, LA 70122
    Attorney for licensee: N.S. Reynolds, Esq., Winston & Strawn 1400 L 
Street N.W., Washington, D.C. 20005-3502
    NRC Project Director: William D. Beckner

Northeast Nuclear Energy Company, et al., Docket No. 50-336, 
Millstone Nuclear Power Station, Unit No. 2, New London County, 
Connecticut

    Date of amendment request: April 10, 1997
    Description of amendment request: The proposed changes would modify 
the Technical Specifications (TSs) for the Enclosure Building. The 
Enclosure Building is a limited-leakage, steel-framed structure that 
completely surrounds the containment. It is designed and constructed to 
ensure that any leakage of radioactive materials to the environment 
would not exceed an acceptable upper limit in the event of a design 
basis loss-of-coolant accident or movement of loads over the spent fuel 
pool. A slight negative pressure is maintained by the Enclosure 
Building Filtration System and the system exhausts the filtered air 
through charcoal and high-efficiency particulate air (HEPA) filters.
    Specifically, the proposed changes would relocate the surveillance 
requirement for attaining a negative pressure in the Enclosure Building 
from TS 3.6.5.1 ``Enclosure Building Filtration System,'' to TS 
3.6.5.2, ``Enclosure Building Integrity.'' TS 3.6.5.2 would also be 
changed to address operability, which includes integrity requirements, 
and the Definition 1.25, ``Enclosure Building Integrity,'' would be 
deleted. TS 4.6.5.2, ``Surveillance Requirements,'' would be modified 
to require each access opening in the Enclosure Building to be closed 
instead of the current requirement to close each door (some access 
openings have two doors in series) in each access opening. This TS 
would also be renumbered as 4.6.5.2.1.
    In addition, editorial changes are proposed for consistency and the 
index pages would be updated to reflect the proposed changes. The TS 
Bases would also be updated to reflect the proposed changes including 
the need to maintain the integrity of the Enclosure Building and to 
support previously approved laboratory testing requirements for 
charcoal filter sample testing.
    Basis for proposed no significant hazards consideration 
determination: 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. Involve a significant increase in the probability or 
consequences of an accident previously evaluated.
    The proposed changes to Technical Specifications 3.6.5.1 and 
3.6.5.2, relocation of Surveillance Requirement 4.6.5.1.d.3 to 
Specification 3.6.5.2, changes to Bases Sections 3.6.5.1 and 
3.6.5.2, and deletion of Definition 1.25 will resolve the conflict 
that currently exists between Specifications 3.6.5.1 and 3.6.5.2. 
Specifically, the requirement to establish and maintain a negative 
pressure in the Enclosure Building boundary included in 
Specification 3.6.5.1 belongs in Specification 3.6.5.2. In the event 
Enclosure Building operability is not maintained in Modes 1-4, the 
Action Statement for LCO [limiting condition for operation] 3.6.5.2 
requires that Enclosure Building operability must be restored within 
24 hours. Twenty-four hours is a reasonable completion time 
considering the limited leakage design of containment and the low

[[Page 24988]]

probability of a DBA [design-basis accident] occurring during this 
time period. Therefore, it is considered that there exists no loss 
of safety function. The
    proposed changes do no modify the LCO or surveillance acceptance 
criterion, nor do they change the frequency of the surveillances. 
The proposed changes do not involve any physical changes to the 
plant, do not alter the way any structure, system, or component 
functions. Therefore, the structures, systems, or components will 
perform their intended function when called upon. (The redundancy of 
the double doors has not been credited in the radiological dose 
calculations for any Design Basis Accident.) Additionally, the 
proposed changes are consistent with the new, improved Standard 
Technical Specifications for Combustion Engineering plants (NUREG-
1432).
    The editorial changes to Technical Specifications 3.6.5.1, 
3.6.5.2, and 3.9.15 do not change any technical aspect of these 
specifications. Therefore the proposed changes do not affect the 
probability of any previously evaluated accident.
    Based on the above, the proposed changes do not involve a 
significant increase in the probability or consequences of an 
accident previously evaluated.
    2. Create the possibility of a new or different kind of accident 
from any accident previously evaluated.
    The proposed changes do not make any physical or operational 
changes to existing plant structures, systems, or components. The 
proposed changes do not introduce any new failure modes. The 
proposed changes simply resolve a conflict which currently exits 
between Specifications 3.6.5.1 and 3.6.5.2. Thus, the proposed 
changes do not create the possibility of a new or different kind of 
accident from any accident previously evaluated.
    3. Involve a significant reduction in a margin of safety.
    The proposed changes do not have any adverse impact on the 
accident analyses. Also, the proposed changes resolve a conflict 
which currently exists between Specifications 3.6.5.1 and 3.6.5.2. 
The structures, systems, or components covered under Specifications 
3.6.5.1 and 3.6.5.2 will perform their intended safety function when 
called upon.
    Based on the above, there is no significant reduction in the 
margin of safety.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: Learning Resources Center, 
Three Rivers Community-Technical College, 574 New London Turnpike, 
Norwich, CT 06360, and the Waterford Library, ATTN: Vince Juliano, 49 
Rope Ferry Road, Waterford, CT 06385
    Attorney for licensee: Lillian M. Cuoco, Esq., Senior Nuclear 
Counsel, Northeast Utilities Service Company, P.O. Box 270, Hartford, 
CT 06141-0270
    NRC Deputy Director: Phillip F. McKee

PECO Energy Company, Public Service Electric and Gas Company, 
Delmarva Power and Light Company, and Atlantic City Electric 
Company, Dockets Nos. 50-277 and 50-278, Peach Bottom Atomic Power 
Station, Units Nos. 2 and 3, York County, Pennsylvania

    Date of application for amendments: March 31, 1997
    Description of amendment request: The proposed change revises the 
Peach Bottom Atomic Power Station, Units 2 and 3 technical 
specifications to extend the surveillance interval for calibration of 
Average Power Range Monitor (APRM) flow bias instrumentation from 18 
months to 24 months.
    Basis for proposed no significant hazards consideration 
determination: 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 changes do not involve a significant increase in 
the probability or consequences of an accident previously evaluated 
because the accidents previously evaluated take credit only for the 
clamped 120% high neutron flux scram setpoint. Credit is not taken 
for the flow biased APRM scram setpoint. Failure or inaccuracy of 
the flow biased feature of the APRM scram setpoint will in no way 
affect the clamped high flux scram setpoint. The 120% high flux 
scram setpoint is derived internal to the APRM circuitry and 
calibrated separately as part of the APRM trip circuitry. The APRM 
clamped high flux scram setpoint is not being impacted by the 
proposed changes and will be automatically enforced regardless of 
the status or accuracy of the APRM flow bias circuitry.
    Because there is no impact on the clamped 120% high neutron flux 
scram setpoint which is the only APRM scram setpoint with any 
analytical safety basis, the proposed changes will not involve a 
significant increase in the probability or consequences of an 
accident previously evaluated.
    2. The proposed changes do not create the possibility of a new 
or different kind of accident from any accident previously evaluated 
because the proposed changes do not allow plant operation in any 
mode that is not already evaluated. The APRM system provides 
monitoring and accident mitigation functions to limit peak flux in 
the core during Modes 1 and 2. No pressure boundary interfaces or 
process control parameters will be challenged in any way as to 
create the possibility of a new or different type of accident than 
any previously evaluated. Also, failure of the sensing line 
associated with flow transmitters to measure recirculation drive 
flow has already been accounted for in the initial plant design by 
including excess flow check valves for sensing line break isolation. 
Therefore, these changes will not create the possibility of a new or 
different kind of accident than any accident previously evaluated.
    3. The proposed changes do not involve a significant reduction 
in a margin of safety because the APRM flow biased high flux scram 
is not credited in the PBAPS safety analysis. Because the proposed 
changes do not impact safety analysis assumptions, these proposed 
changes will not involve a significant reduction in a margin of 
safety.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: Government Publications 
Section, State Library of Pennsylvania, (REGIONAL DEPOSITORY) Education 
Building, Walnut Street and Commonwealth Avenue, Box 1601, Harrisburg, 
PA 17105.
    Attorney for licensee: J. W. Durham, Sr., Esquire, Sr. V. P. and 
General Counsel, PECO Energy Company, 2301 Market Street, Philadelphia, 
PA 19101
    NRC Project Director: John F. Stolz

Wisconsin Public Service Corporation, Docket No. 50-305, Kewaunee 
Nuclear Power Plant, Kewaunee County, Wisconsin

    Date of amendment request: April 22, 1997
    Description of amendment request: The proposed amendment would 
revise Kewaunee Nuclear Power Plant (KNPP) Technical Specification (TS) 
Section 4.2.b, ``Steam Generator Tubes,'' to allow a laser-welded 
repair of Westinghouse hybrid expansion joint (HEJ) sleeved steam 
generator (SG) tubes. The proposed repair process would fuse the tube 
to the sleeve in the upper joint of the existing HEJ sleeved tubes. The 
repair weld would be made in either the hardroll (HR) expansion or the 
upper hydraulic expansion (HE) region of the HEJ. By fusing the tube to 
the sleeve, parent tube degradation below the weld would be isolated 
and a new pressure boundary would be formed. The new pressure boundary 
would satisfy both the structural and leakage integrity requirements of 
the sleeved tube assembly with no change in the flow or heat transfer 
characteristics of the sleeved tube. The proposed amendment supersedes 
in its entirety a previously submitted proposed amendment dated 
September 6, 1996, which was noticed in the Federal Register on October 
15, 1996 (61 FR 53769).

[[Page 24989]]

    Basis for proposed no significant hazards consideration 
determination: 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. Operation of the KNPP in accordance with the proposed license 
amendment does not involve a significant increase in the probability 
or consequences of an accident previously evaluated.
    The laser-weld repair of HEJ sleeved tubes in either the HR or 
HE location will not affect the tube, sleeve, or weld stress 
conditions or fatigue usage factors such that the limits of the ASME 
Boiler and Pressure Vessel Code are exceeded. Accelerated corrosion 
testing performed on prototypic HR welds, and a corrosion assessment 
performed for the HE welds concluded that the repair welds will not 
result in aggravated stress corrosion cracking at the weld-repair 
location. Any postulated sleeve joint degradation would occur at a 
relatively slow rate and would be detectable by routine non-
destructive examination (NDE) inspection prior to reaching any 
applicable safety margins. Therefore, use of the laser-weld repair 
process will not result in an increased probability of an accident 
previously evaluated.
    A post-weld stress relief ultrasonic test inspection is required 
to verify minimum acceptable weld thickness to ensure that the weld 
stresses do not exceed ASME Code limits for both stress intensity 
and fatigue usage. Leakage testing of laser-welded sleeve joints, 
and in-situ leakage testing of the laser-welded repairs (LWR) at 
KNPP, demonstrate a leak-tight joint at pressures up to main steam 
line break. Mechanical testing of 7/8 inch laser-welded tubesheet 
sleeves installed in roll-expanded tubes has shown that the 
individual joint structural strength of Alloy 690 laser-welded 
sleeves under normal, upset, and faulted conditions provides margin 
to acceptable limits. These acceptable limits bound the most 
limiting (3 times normal operating pressure differential) 
recommended by Regulatory Guide (RG) 1.121.
    The HEJ sleeve plugging limit currently defined in the TS is 
reduced from 31% to 24% throughwall due to the use of ASME code 
minimum material properties values for the sleeve material. Minimum 
wall thickness requirements (used for developing the depth-based 
plugging limit for the sleeve) are determined using the guidance of 
RG 1.121 and the pressure stress equation of Section 3 of the ASME 
Code.
    The hypothetical consequences of failure of the laser-welded 
repaired HEJ would be bounded by the current SG tube rupture (SGTR) 
analysis covered in the KNPP Updated Safety Analysis Report. Due to 
the slight reduction in diameter caused by the sleeve wall 
thickness, primary coolant release rates would be slightly less than 
assumed for the SGTR, and, therefore, would result in lower primary 
fluid mass release to the secondary system. The laser-weld repair 
process does not change the existing reactor coolant system flow 
conditions; therefore, existing loss of coolant accident (LOCA) and 
non-LOCA analysis results will be unaffected. Plant response to 
design basis accidents for the current tube plugging and flow 
conditions are not affected by the repair process; no new tube 
diameter restrictions are introduced. Therefore, the application of 
the repair weld will not increase the consequences of a previously 
evaluated accident.
    2. The proposed license amendment request does not create the 
possibility of a new or different kind of accident from any accident 
previously evaluated.
    Application of laser-welded repair for the HEJ sleeved tubes 
will not introduce significant or adverse changes to the plant 
design basis. The general configuration of the HEJ sleeve is 
unaffected by the repair process. The repair process also does not 
represent a potential to affect any other plant component. Stress 
and fatigue analysis of the repair has shown that the ASME Code and 
RG 1.121 criteria are not exceeded. Application of the laser-weld 
repair to the HEJ sleeved tubes maintains overall tube bundle 
structural and leakage integrity. Extensive testing and evaluation 
including examination of actual pulled tube samples verified 
adequate structural and leakage integrity of repair HEJs, which had 
acceptable NDE.
    Any hypothetical accident as a result of potential tube or 
sleeve degradation in the repaired portion of the joint is bounded 
by the existing tube rupture accident analysis. Therefore, use of 
the laser-welded repair process will not create the possibility of a 
new or different kind of accident from any accident previously 
evaluated.
    3. The proposed license amendment does not involve a significant 
reduction in the margin of safety.
    The laser-weld repair of the HEJ sleeved tubes has been shown to 
restore integrity of the tube bundle consistent with its original 
design basis conditions; i.e., tube/sleeve operational and faulted 
load stresses and cumulative fatigue usage factors are bounded by 
ASME Code requirements and the tubes are leak tight under all plant 
conditions. Based on the results of the structural and leakage 
testing performed on LWR joints pulled from the KNPP SGs and 
supporting analytical evaluations, application of laser-welded 
repair will not result in a significant reduction in the margin of 
safety.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: University of Wisconsin, 
Cofrin Library, 2420 Nicolet Drive, Green Bay, Wisconsin 54311-7001.
    Attorney for licensee: Bradley D. Jackson, Esq., Foley and Lardner, 
P.O. Box 1497, Madison, Wisconsin 53701-1497.
    NRC Project Director: Gail H. Marcus

Wisconsin Public Service Corporation, Docket No. 50-305, Kewaunee 
Nuclear Power Plant, Kewaunee County, Wisconsin

    Date of amendment request: April 24, 1997
    Description of amendment request: The proposed amendment would 
revise Kewaunee Nuclear Power Plant (KNPP) Technical Specification (TS) 
Section 4.2.b, ``Steam Generator Tubes,'' to allow repair of steam 
generator (SG) tubes with Combustion Engineering (CE) leak-tight 
sleeves in accordance with CE generic topical report CEN-629-P, 
Revision 2, ``Repair of Westinghouse Series 44 and 51 Steam Generator 
Tubes Using Leak-Tight Sleeves.'' The TS would also be revised to allow 
re-sleeving of tubes with existing sleeve joints in accordance with 
KNPP specific topical report CEN-632-P, ``Repair of Kewaunee Steam 
Generator Tubes Using a Re-Sleeving Technique.''
    Basis for proposed no significant hazards consideration 
determination: 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. Operation of the KNPP in accordance with the proposed license 
amendment does not involve a significant increase in the probability 
or consequences of an accident previously evaluated.
    The supporting technical evaluation and safety evaluation for 
the CE leak-tight sleeves demonstrates that the sleeve configuration 
will provide SG tube structural and leakage integrity under normal 
operating and accident conditions. The sleeve configurations have 
been designed and analyzed in accordance with the requirements of 
the ASME Code. Mechanical testing has shown that the sleeve and 
sleeve joints provide margin above acceptance limits. Ultrasonic 
testing is used to verify the leak tightness of the weld above the 
tubesheet. Testing has demonstrated the leak tightness of the 
hardroll joint as well as the structural integrity of the hardroll 
joint. Tube rupture cannot occur at the hardroll joint due to the 
reinforcing effect of the tubesheet. Tests have demonstrated that 
tube collapse will not occur due to postulated loss of coolant 
accident loadings.
    The existing TS leak-rate requirements and accident analysis 
assumptions remain unchanged in the event that significant leakage 
does occur from the sleeve joint or the sleeve assembly ruptures. 
Any leakage through the sleeve assembly is fully bounded by the 
existing SG tube rupture analysis included in the KNPP Updated Final 
Safety Analysis Report. The proposed sleeving and re-sleeve repair 
processes do not adversely impact any other previously evaluated 
design basis accidents.
    2. The proposed license amendment request does not create the 
possibility of a new or different kind of accident from any accident 
previously evaluated.

[[Page 24990]]

    Installation of the sleeves or re-sleeves does not introduce any 
significant changes to the plant design basis. The use of a sleeve 
to span the area of degradation of the SG tube restores the 
structural and leakage integrity of the tubing to meet the original 
design basis. Stress and fatigue analysis of the sleeve assembly 
shows that the requirements of the ASME Code are met. Mechanical 
testing has demonstrated that margin exists above the design 
criteria. Any hypothetical accident as a result of any degradation 
in the sleeved tube would be bounded by the existing tube rupture 
accident analysis.
    3. The proposed license amendment does not involve a significant 
reduction in the margin of safety.
    The use of sleeves to repair degraded SG tubing has been 
demonstrated to maintain the integrity of the tube bundle 
commensurate with the requirements of the ASME Code and draft 
Regulatory Guide 1.121, and to maintain the primary to secondary 
pressure boundary under normal and postulated accident conditions. 
The safety factors used in the verification of the strength of the 
sleeve assembly are consistent with the safety factors in the ASME 
Boiler and Pressure Vessel Code used in SG design. The operational 
and faulted condition stresses and cumulative usage factors are 
bounded by the ASME Code requirements. The sleeve assembly has been 
verified by testing to prevent both tube pullout and significant 
leakage during normal and postulated accident conditions. A test 
program was conducted to ensure the lower hardrolled joint design 
was leak tight and capable of withstanding the design loads. The 
primary coolant pressure boundary of the sleeve assembly will be 
periodically inspected by non-destructive examination to identify 
sleeve degradation due to operation.
    Installation of the sleeves and re-sleeves will decrease the 
number of tubes that must be taken out-of-service due to plugging. 
There is a small amount of primary coolant flow reduction due to the 
sleeve for which an equivalent plugging sleeve to plug ratio is 
assigned based on sleeve length. The ratio is used to assess the 
final equivalent plugging percentage as an input to other safety 
analyses. Because the sleeve maintains the design basis requirements 
for the SG tubing, it is concluded that the proposed change does not 
result in a significant reduction in the margin of safety.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: University of Wisconsin, 
Cofrin Library, 2420 Nicolet Drive, Green Bay, Wisconsin 54311-7001.
    Attorney for licensee: Bradley D. Jackson, Esq., Foley and Lardner, 
P.O. Box 1497, Madison, Wisconsin 53701-1497.
    NRC Project Director: Gail H. Marcus

Previously Published Notices Of Consideration Of Issuance Of 
Amendments To Facility Operating Licenses, Proposed No Significant 
Hazards Consideration Determination, And Opportunity For A Hearing

    The following notices were previously published as separate 
individual notices. The notice content was the same as above. They were 
published as individual notices either because time did not allow the 
Commission to wait for this biweekly notice or because the action 
involved exigent circumstances. They are repeated here because the 
biweekly notice lists all amendments issued or proposed to be issued 
involving no significant hazards consideration.
    For details, see the individual notice in the Federal Register on 
the day and page cited. This notice does not extend the notice period 
of the original notice.

Carolina Power & Light Company, et al., Docket Nos. 50-325 and 50-
324, Brunswick Steam Elecric Plant, Units 1 and 2, Brunswick 
County, North Carolina

    Date of amendments request: March 27, 1997
    Brief description of amendments: The proposed amendments would 
revise the Technical Specifications for the Brunswick Steam Electric 
Plant Units 1 and 2 to eliminate certain instrumentation response time 
testing requirements in accordance with NRC-approved BWR Owners Group 
Topical Report NEDO-32291-A, ``System Analysis for the Elimination of 
Selected Response Time Testing Requirements.''
    Date of publication of individual notice in Federal Register: April 
1, 1997(62 FR 15542)
    Expiration date of individual notice: May 1, 1997
    Local Public Document Room location: University of North Carolina 
at Wilmington, William Madison Randall Library, 601 S. College Road, 
Wilmington, North Carolina 28403-3297.

Northern States Power Company, Docket Nos. 50-282 and 50-306, 
Prairie Island Nuclear Generating Plant, Unit Nos. 1 and 2, Goodhue 
County, Minnesota, and Docket No. 50-263, Monticello Nuclear 
Generating Plant, Wright County, Minnesota

    Date of amendment requests: December 6, 1996
    Description of amendment requests: The licensee requests amendments 
to the Prairie Island and Monticello operating licenses to reflect the 
Commission's approval of the transfer of control over the subject NRC 
licenses held by Northern States Power Company (NSP). On October 20, 
1995, as supplemented August 28, 1996, NSP requested NRC approval for 
the transfer of control of licenses. The Commission is considering the 
issuance of amendments to the licenses to reflect the above transfer 
approved by the Commission on April 1, 1997 (62 FR 17882, dated April 
11, 1997).
    Date of individual notice in the Federal Register: April 11, 1997 
(62 FR 17882)
    Expiration date of individual notice: May 12, 1997
    Local Public Document Room location: Minneapolis Public Library, 
Technology and Science Department, 300 Nicollet Mall, Minneapolis, 
Minnesota 55401

Pennsylvania Power and Light Company, Docket Nos. 50-387 and 50-388 
Susquehanna Steam Electric Station, Units 1 and 2, Luzerne County, 
Pennsylvania

    Date of amendment request: April 4, 1997
    Brief description of amendment request: The proposed amendment 
would clarify the scope of the surveillance requirements for response 
time testing of instrumentation in the reactor protection system, 
isolation actuation system, and emergency core cooling system in the 
Technical Specifications for each unit (Sections 4.3.1.3, 4.3.2.3, and 
4.3.3.3).
    Date of publication of individual notice in Federal Register: April 
17, 1997 (62 FR 17885)
    Expiration date of individual notice: May 19, 1997
    Local Public Document Room location: Osterhout Free Library, 
Reference Department, 71 South Franklin Street, Wilkes-Barre, PA 18701

Notice Of Issuance Of Amendments To Facility Operating Licenses

    During the period since publication of the last biweekly notice, 
the Commission has issued the following amendments. The Commission has 
determined for each of these amendments that the application complies 
with the standards and requirements of the Atomic Energy Act of 1954, 
as amended (the Act), and the Commission's rules and regulations. The 
Commission has made appropriate findings as required by the Act and the 
Commission's rules and regulations in 10 CFR Chapter I, which are set 
forth in the license amendment.
    Notice of Consideration of Issuance of Amendment to Facility 
Operating License, Proposed No Significant

[[Page 24991]]

Hazards Consideration Determination, and Opportunity for A Hearing in 
connection with these actions was published in the Federal Register as 
indicated.
    Unless otherwise indicated, the Commission has determined that 
these amendments satisfy the criteria for categorical exclusion in 
accordance with 10 CFR 51.22. Therefore, pursuant to 10 CFR 51.22(b), 
no environmental impact statement or environmental assessment need be 
prepared for these amendments. If the Commission has prepared an 
environmental assessment under the special circumstances provision in 
10 CFR 51.12(b) and has made a determination based on that assessment, 
it is so indicated.
    For further details with respect to the action see (1) the 
applications for amendment, (2) the amendment, and (3) the Commission's 
related letter, Safety Evaluation and/or Environmental Assessment as 
indicated. All of these items are 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 rooms for 
the particular facilities involved. Boston Edison Company, Docket No. 
50-293, Pilgrim Nuclear Power Station, Plymouth County, Massachusetts
    Date of application for amendment: January 24, 1997, as 
supplemented March 27, 1997
    Brief description of amendment: The proposed amendment will update 
the Safety Limit Minimum Critical Power Ratio (SLMCPR) in Technical 
Specification 2.1.2 and the associated Bases section to reflect the 
results of the latest cycle-specific calculation performed for the 
Pilgrim Nuclear Power Station Operating Cycle 12. In addition, the 
values provided in Note 5 of Table 3.2.C.1, which are based on the 
SLMCPR values, have been revised as a result of the changes to the 
SLMCPR value.
    Date of issuance: April 7, 1997
    Effective date: April 7, 1997
    Amendment No.: 171
    Facility Operating License No. DPR-35: Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: February 12, 1997 (62 
FR 6568) The March 27, 1997, supplemental letter provided clarifying 
information that did not change the initial proposed no significant 
hazards consideration. The Commission's related evaluation of the 
amendment is contained in a Safety Evaluation dated April 7, 1997 No 
significant hazards consideration comments received: No
    Local Public Document Room location: Plymouth Public Library, 11 
North Street, Plymouth, Massachusetts 02360.

Carolina Power & Light Company, et al., Docket Nos. 50-325 and 50-
324, Brunswick Steam Electric Plant, Units 1 and 2, Brunswick 
County, North Carolina

    Date of application for amendment: March 27, 1997, as supplemented 
April 11, 1997.
    Brief description of amendment: The amendments revise the Technical 
Specifications relating to response time testing requirements 
associated with the reactor protection system, isolation system, and 
emergency core cooling system.
    Date of issuance: April 18, 1997
    Effective date: April 18, 1997
    Amendment Nos.: 184 and 215
    Facility Operating License Nos. DPR-71 and DPR-62. Amendments 
revised the Technical Specifications. Public comments requested as to 
proposed no significant hazards consideration (NSHC): Yes (62 FR 15542 
dated April 1, 1997). The notice provided an opportunity to submit 
comments on the Commission's proposed NSHC determination. No comments 
have been received. The notice also provided for an opportunity to 
request a hearing by May 1, 1997, but indicated that if the Commission 
makes a final NSHC determination, any such hearing would take place 
after issuance of the amendments. The Commission's related evaluation 
of the amendment, finding of exigent circumstances, and final 
determination of NSHC are contained in a Safety Evaluation dated April 
18, 1997.
    Attorney for licensee: William D. Johnson, Vice President and 
Senior Counsel, Carolina Power & Light Company, Post Office Box 1551, 
Raleigh, North Carolina 27602
    Local Public Document Room location: University of North Carolina 
at Wilmington, William Madison Randall Library, 601 S. College Road, 
Wilmington, North Carolina 28403-3297.

Commonwealth Edison Company, Docket Nos. STN 50-454 and STN 50-455, 
Byron Station, Unit Nos. 1 and 2, Ogle County, Illinois Docket Nos. 
STN 50-456 and STN 50-457, Braidwood Station, Unit Nos. 1 and 2, 
Will County, Illinois

    Date of application for amendments: December 21, 1995, as 
supplemented on October 24, 1996, and March 24, 1997.
    Brief description of amendments: The amendments relocate certain 
cycle-specific parameter limits from the Technical Specifications (TS) 
to the Operating Limits Report. The cycle-specific parameter limits to 
be relocated are for Shutdown Rod Insertion Limit, Control Rod 
Insertion Limits, Axial Flux Difference Target Band, Heat Flux Hot 
Channel Factor [F<INF>Q</INF>(z)], and Nuclear Enthalpy Rise Hot 
Channel Factor (F<INF>N</INF> delta H). In addition, your March 24, 
1997, submittal contained supplementary revisions to the Bases section 
associated with the above TS change. The supplementary Bases pages will 
be reviewed and transmitted to you under separate cover. Finally, 
Braidwood's TS 6.9.1.7 title was corrected.
    Date of issuance: April 16, 1997
    Effective date: Immediately, to be implemented within 30 days.
    Amendment Nos.: 88, 88, 80, 80
    Facility Operating License Nos. NPF-37, NPF-66, NPF-72 and NPF-77: 
The amendments revised the Technical Specifications.
    Date of initial notice in Federal Register: February 20, 1997 (62 
FR 7804). The March 24, 1997, submittal provided clarifying information 
that did not change the initial proposed no significant hazards 
consideration determination. The Commission's related evaluation of the 
amendments is contained in a Safety Evaluation dated April 16, 1997. No 
significant hazards consideration comments received: No
    Local Public Document Room location: For Byron, the Byron Public 
Library District, 109 N. Franklin, P.O. Box 434, Byron, Illinois 61010; 
for Braidwood, the Wilmington Public Library, 201 S. Kankakee Street, 
Wilmington, Illinois 60481.

Commonwealth Edison Company, Docket Nos. STN 50-454 and STN 50-455, 
Byron Station, Unit Nos. 1 and 2, Ogle County, Illinois Docket Nos. 
STN 50-456 and STN 50-457, Braidwood Station, Unit Nos. 1 and 2, 
Will County, Illinois

    Date of application for amendments: April 29, 1996, as supplemented 
on January 21 and March 25, 1997.
    Brief description of amendments: The amendments would: (1) revise 
Technical Specification (TS) 3.7.1.1, Action a., to require the unit to 
be in hot shutdown, rather than cold shutdown, for consistency with 
NUREG-1431, ``Standard Technical Specifications for Westinghouse 
Plants,'' and add a new Action b. to clarify the shutdown requirements 
when there are more than three inoperable main steam line American 
Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code 
(Code) safety valves on any

[[Page 24992]]

one steam generator; (2) revise TS Surveillance Requirement 4.7.1.1 to 
clarify that Specification 4.0.4 does not apply for entry into Mode 3 
for Byron and Braidwood and for Braidwood only, delete the one-time 
requirements for Unit 1, Cycle 5 and Unit 2 after outage A2F27; (3) 
revise the maximum allowable power range neutron flux high trip 
setpoints in Table 3.7-1; (4) revise Table 3.7-2 to increase the as-
found main steam safety valve (MSSV) lift setpoint tolerance to plus or 
minus 3 percent, provide an as-left setpoint tolerance of plus or minus 
1 percent, and change a table notation; (5) delete the orifice size 
column from Table 3.7-2; and (6) revise the Bases for TS 3.7.1.1 to be 
consistent with the proposed changes to TS 3.7.1.1.
    Date of issuance: April 15, 1997
    Effective date: Immediately, to be implemented within 30 days.
    Amendment Nos.: 87, 87, 79, and 79
    Facility Operating License Nos. NPF-37, NPF-66, NPF-72 and NPF-77: 
The amendments revised the Technical Specifications.
    Date of initial notice in Federal Register: March 12, 1997 (62 FR 
11486). The March 25, 1997, submittal provided additional information 
that did not change the initial proposed no significant hazards 
consideration determination. The Commission's related evaluation of the 
amendments is contained in a Safety Evaluation dated April 15, 1997. No 
significant hazards consideration comments received: No
    Local Public Document Room location: For Byron, the Byron Public 
Library District, 109 N. Franklin, P.O. Box 434, Byron, Illinois 61010; 
for Braidwood, the Wilmington Public Library, 201 S. Kankakee Street, 
Wilmington, Illinois 60481.

Consolidated Edison Company of New York, Docket No. 50-247, Indian 
Point Nuclear Generating Unit No. 2, Westchester County, New York

    Date of application for amendment: August 7, 1996, as supplemented 
March 12, 1997.
    Brief description of amendment: The amendment revises Technical 
Specifications to allow the use of 10 CFR Part 50, Appendix J, Option 
B, ``Performance-Based Containment Leak Rate Testing.''
    Date of issuance: April 10, 1997
    Effective date: As of the date of issuance to be implemented within 
30 days.
    Amendment No.: 190
    Facility Operating License No. DPR-26: Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: September 11, 1996 (61 
FR 47976) The Commission's related evaluation of the amendment is 
contained in a Safety Evaluation dated April 10, 1997. No significant 
hazards consideration comments received: No
    Local Public Document Room location: White Plains Public Library, 
100 Martine Avenue, White Plains, New York 10610.

Detroit Edison Company, Docket No. 50-341, Fermi-2, Monroe County, 
Michigan

    Date of application for amendment: March 27, 1997, as supplemented 
on April 4, 1997
    Brief description of amendment: The amendment revises technical 
specification surveillance requirement (SR) 4.3.1.3 for the Reactor 
Protection System Instrumentation to indicate that certain sensors are 
exempt from response time testing. A similar revision is made to SR 
4.3.2.3 for the Isolation Actuation Instrumentation. Finally, SR 
4.3.3.3 for the Emergency Core Cooling System Actuation Instrumentation 
is revised to indicate that the emergency core cooling system actuation 
instrumentation is exempt from response time testing.
    Date of issuance: April 18, 1997
    Effective date: April 18, 1997, with full implementation prior to 
entry into Operation Condition 2 or 3
    Amendment No.: 111
    Facility Operating License No. NPF-43. Amendment revises the 
Technical Specifications.
    Public comments requested as to proposed no significant hazards 
considerations (NSHC): Yes (62 FR 15731 dated April 2, 1997). The 
notice provided an opportunity to submit comments on the Commission's 
proposed NSHC determination. No comments have been received. The notice 
also provided for an opportunity to request a hearing by May 2, 1997, 
but indicated that if the Commission makes a final NSHC determination, 
any such hearing would take place after issuance of the amendment. The 
Commission's related evaluation of the amendment, finding of exigent 
circumstances, and final determination of NSHC are contained in a 
Safety Evaluation dated April 18, 1997.
    Attorney for licensee: John Flynn, Esq., Detroit Edison Company, 
2000 Second Avenue, Detroit, Michigan 48226
    Local Public Document Room location: Monroe County Library System, 
3700 South Custer Road, Monroe, Michigan 48161

Duke Power Company, Docket Nos. 50-369 and 50-370, McGuire Nuclear 
Station, Units 1 and 2, Mecklenburg County, North Carolina Date of 
application for amendments: January 6, 1997, as supplemented by 
letters dated April 10 and 15, 1997

    Brief description of amendments: The amendments revise portions of 
the Technical Specifications to permit a one-time operation of the 
Containment Purge Ventilation System during Modes 3 and 4 after the 
current and forthcoming steam generator replacement outages.
    Date of issuance: April 24, 1997
    Effective date: As of the date of issuance to be implemented within 
30 days
    Amendment Nos.: 174 and 156
    Facility Operating License Nos. NPF-9 and NPF-17: Amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: February 12, 1997 (62 
FR 6574) The Commission's related evaluation of the amendments is 
contained in a Safety Evaluation dated April 24, 1997. No significant 
hazards consideration comments received: No
    Local Public Document Room location: J. Murrey Atkins Library, 
University of North Carolina at Charlotte, 9201 University City 
Boulevard, North Carolina 28223-0001

Duquesne Light Company, et al., Docket Nos. 50-334 and 50-412, 
Beaver Valley Power Station, Unit Nos. 1 and 2, Shippingport, 
Pennsylvania

    Date of application for amendments: September 9, 1996
    Brief description of amendments: These amendments modify the design 
features section (Section 5.0) of the Technical Specifications (TSs) to 
make the design features section consistent with the intent of 10 CFR 
50.36 and with the guidance provided in the NRC's Standard Technical 
Specifications, Westinghouse Plants (NUREG-1431, Revision 1).
    Date of issuance: April 14, 1997
    Effective date: Both units, as of date of issuance, to be 
implemented within 60 days.
    Amendment Nos.: 202 and 83
    Facility Operating License Nos. DPR-66 and NPF-73: Amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: December 4, 1996 (61 FR 
64384) The Commission's related evaluation of the amendments is 
contained in a Safety Evaluation dated April 14, 1997. No significant 
hazards consideration comments received: No.
    Local Public Document Room location: B. F. Jones Memorial Library,

[[Page 24993]]

663 Franklin Avenue, Aliquippa, PA 15001

Entergy Operations, Inc., Docket No. 50-368, Arkansas Nuclear One, 
Unit No. 2, Pope County, Arkansas

    Date of application for amendment: December 19, 1996
    Brief description of amendment: The amendment deletes the specific 
value for the total reactor coolant system volume from the Design 
Features section of the Technical Specifications.
    Date of issuance: April 16, 1997
    Effective date: April 16, 1997
    Amendment No.: 181
    Facility Operating License No. NPF-6: Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: January 29, 1997 (62 FR 
4348) The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated April 16, 1997. No significant hazards 
consideration comments received: No.
    Local Public Document Room location: Tomlinson Library, Arkansas 
Tech University, Russellville, AR 72801

Entergy Operations, Inc., Docket No. 50-368, Arkansas Nuclear One, 
Unit No. 2, Pope County, Arkansas

    Date of application for amendment: December 19, 1996
    Brief description of amendment: Request to add CENTS code as a 
Reference to the Technical Manual used for determining Core Operating 
Limits Report in the Technical Specifications.
    Date of issuance: April 24, 1997
    Effective date: April 24, 1997
    Amendment No.: 182
    Facility Operating License No. NPF-6: Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: January 29, 1997 (62 FR 
4347) The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated April 24, 1997. No significant hazards 
consideration comments received: No.
    Local Public Document Room location: Tomlinson Library, Arkansas 
Tech University, Russellville, AR 72801

Entergy Gulf States, Inc., Cajun Electric Power Cooperative, and 
Entergy Operations, Inc., Docket No. 50-458, River Bend Station, 
Unit 1, West Feliciana Parish, Louisiana

    Date of amendment request: January 10, 1997
    Brief description of amendment: The amendment revises the technical 
specifications for reactor pressure vessel pressure and temperature 
limits by providing new limits that are valid to 12 effective full 
power years.
    Date of issuance: April 14, 1997
    Effective date: April 14, 1997
    Amendment No.: 93
    Facility Operating License No. NPF-47: The amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: February 26, 1997 (62 
FR 8798) The Commission's related evaluation of the amendment is 
contained in a Safety Evaluation dated April 14, 1997. No significant 
hazards consideration comments received. No.
    Local Public Document Room location: Government Documents 
Department, Louisiana State University, Baton Rouge, LA 70803

Entergy Operations, Inc., Docket No. 50-382, Waterford Steam 
Electric Station, Unit 3, St. Charles Parish, Louisiana

    Date of amendment request: November 7, 1995, as supplemented by 
letters dated July 17, and December 26, 1996, and February 27, March 
14, April 7, and April 17, 1997.
    Brief description of amendment: The amendment changes the Appendix 
A Technical Specifications by revising TS 3/4.8.1, ``Electrical Power 
Systems - A.C. Sources,'' to incorporate recommendations and 
suggestions from (1) Generic Letter (GL) 93-05, ``Line-Item Technical 
Specifications Improvements to Reduce Surveillance Requirements for 
Testing During Power Operations;'' (2) GL 94-01, ``Removal of 
Accelerated Testing and Special Reporting Requirements for Emergency 
Diesel Generators from Plant Technical Specifications;'' and (3) NUREG-
1432, ``Standard Technical Specifications Combustion Engineering 
Plants.''
    Date of issuance: April 21, 1997
    Effective date: April 21, 1997, to be implemented within 60-days.
    Amendment No.: 126
    Facility Operating License No. NPF-38: Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: January 3, 1996 (61 FR 
180) The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated April 21, 1997. No significant hazards 
consideration comments received: No.
    Local Public Document Room location: University of New Orleans 
Library, Louisiana Collection, Lakefront, New Orleans, LA 70122

GPU Nuclear Corporation, et al., Docket No. 50-219, Oyster Creek 
Nuclear Generating Station, Ocean County, New Jersey

    Date of application for amendment: October 10, 1996 (TSCR 243)
    Brief description of amendment: The amendment modifies the 
Technical Specifications (TS) by replacing the description of the 
existing permissive interlock from AC Voltage to Core Spray Booster 
Pump d/p Permissive: <ls-thn-eq> 21.2 psid for initiation of the 
automatic depressurization system, adds corresponding surveillance 
requirements, and adds notes clarifying functional requirements.
    Date of Issuance: April 14, 1997
    Effective date: April 14, 1997, with full implementation within 60 
days
    Amendment No.: 190
    Facility Operating License No. DPR-16.
    Date of initial notice in Federal Register: November 6, 1996 (61 FR 
57485). The Commission's related evaluation of this amendment is 
contained in a Safety Evaluation dated April 14, 1997 No significant 
hazards consideration comments received: No.
    Local Public Document Room location: Ocean County Library, 
Reference Department, 101 Washington Street, Toms River, NJ 08753.

Northeast Nuclear Energy Company, Docket No. 50-245, Millstone 
Nuclear Power Station, Unit 1, New London County, Connecticut

    Date of application for amendment: September 5, 1996
    Brief description of amendment: The amendment deletes License 
Condition 2.C.(5), ``Integrated Implementation Schedule'' from the 
Millstone Unit 1 Operating License.
    Date of issuance: April 15, 1997
    Effective date: As of the date of issuance, to be implemented 
within 60 days.
    Amendment No.: 100
    Facility Operating License No. DPR-21: Amendment revised the 
Operating License.
    Date of initial notice in Federal Register: October 23, 1996 (61 FR 
55036) The Commission's related evaluation of the amendment is 
contained in a Safety Evaluation dated April 15, 1997. No significant 
hazards consideration comments received: No.
    Local Public Document Room location: Learning Resources Center, 
Three Rivers Community-Technical College, 574 New London Turnpike, 
Norwich, Connecticut 06360 and at the Waterford Library, ATTN: Vince 
Juliano, 49 Rope Ferry Road, Waterford, Connecticut 06385

Northeast Nuclear Energy Company, et al., Docket No. 50-423, 
Millstone Nuclear Power Station, Unit No. 3, New London County, 
Connecticut

    Date of application for amendment: February 5, 1996

[[Page 24994]]

    Brief description of amendment: The amendment deletes a clause from 
Technical Specification 4.0.5.a. Specifically, this change deletes the 
clause ``(g), except where specific written relief has been granted by 
the Commission pursuant to 10 CFR Part 50, Section 50.55a(g)(6)(i).'' 
The amendment also makes the appropriate changes to the Bases section.
    Date of issuance: April 21, 1997
    Effective date: As of the date of issuance, to be implemented 
within 60 days.
    Amendment No.: 138
    Facility Operating License No. NPF-49: Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: February 26, 1997 (62 
FR 8800) The Commission's related evaluation of the amendment is 
contained in a Safety Evaluation dated April 21, 1997. No significant 
hazards consideration comments received: No.
    Local Public Document Room location: Learning Resources Center, 
Three Rivers Community-Technical College, 574 New London Turnpike, 
Norwich, Connecticut 06360, and the Waterford Library, ATTN: Vince 
Juliano, 49 Rope Ferry Road, Waterford, Connecticut 06385

Northeast Nuclear Energy Company, et al., Docket No. 50-423, 
Millstone Nuclear Power Station, Unit No. 3, New London County, 
Connecticut

    Date of application for amendment: March 4, 1996
    Brief description of amendment: The amendment modifies Surveillance 
Requirements 4.8.1.1.2.a.6, 4.8.1.1.2.b, and 4.8.1.1.2.g.7 by 
specifying load bands in loading the diesel generator (DG) in lieu of 
the present requirement to load the DG greater than or equal to a given 
value. A footnote is being added to the three surveillance 
rerquirements to indicate that a momentary transient outside the load 
range shall not invalidate the test. The aassociated Bases sections 
have been revised to reflect the above changes.
    Date of issuance: April 15, 1997
    Effective date: As of the date of issuance, to be implemented 
within 60 days.
    Amendment No.: 137
    Facility Operating License No. NPF-49: Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: March 12, 1997 (62 FR 
11496) The Commission's related evaluation of the amendment is 
contained in a Safety Evaluation dated April 15, 1997. No significant 
hazards consideration comments received: No
    Local Public Document Room location: Learning Resources Center, 
Three Rivers Community-Technical College, 574 New London Turnpike, 
Norwich, Connecticut 06360, and the Waterford Library, ATTN: Vince 
Juliano, 49 Rope Ferry Road, Waterford, Connecticut 06385

Pacific Gas and Electric Company, Docket Nos. 50-275 and 50-323, 
Diablo Canyon Nuclear Power Plant, Unit Nos. 1 and 2, San Luis 
Obispo County, California

    Date of application for amendments: February 14, 1996, as 
supplemented by letter dated February 24, 1997.
    Brief description of amendments: The amendments revised the 
combined Technical Specifications (TS) for the Diablo Canyon Power 
Plant (DCPP) Unit Nos. 1 and 2 to revise 30 TS and add two new TS 
surveillance requirements to support implementation of extended fuel 
cycles at DCPP Unit Nos. 1 and 2. The specific TS changes include those 
for 9 trip actuating device tests, 12 fluid system actuation tests, and 
11 miscellaneous tests. Two of the fluid system actuation tests are new 
TS surveillance requirements. The TS changes also involve adding a new 
frequency notation, ``R24, REFUELING INTERVAL,'' to Table 1.1 of the 
TS. Also, a revision that applies to all subsequent TS changes involves 
revising the Bases Section of TS 4.0.2 to change the surveillance 
frequency from an 18-month surveillance interval to at least once each 
refueling interval.
    Date of issuance: April 14, 1997
    Effective date: April 14, 1997, to be implemented within 90 days 
from the date of issuance.
    Amendment Nos.: Unit 1 - 118; Unit 2 - 116
    Facility Operating License Nos. DPR-80 and DPR-82: The amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: June 19, 1996 (61 FR 
31183) The February 24, 1997, supplemental letter provided additional 
clarifying information and did not change the staff's initial no 
significant hazards consideration determination. The Commission's 
related evaluation of the amendments is contained in a Safety 
Evaluation dated April 14, 1997. No significant hazards consideration 
comments received: No.
    Local Public Document Room location: California Polytechnic State 
University, Robert E. Kennedy Library, Government Documents and Maps 
Department, San Luis Obispo, California 93407

Pacific Gas and Electric Company, Docket Nos. 50-275 and 50-323, 
Diablo Canyon Nuclear Power Plant, Unit Nos. 1 and 2, San Luis 
Obispo County, California

    Date of application for amendments: May 31, 1996, as supplemented 
by letter dated December 16, 1996.
    Brief description of amendments: The amendments revised the 
combined Technical Specifications (TS) for the Diablo Canyon Power 
Plant (DCPP) Unit Nos. 1 and 2 to revise 23 TS surveillance frequencies 
from at least once every 18 months to at least once per refueling 
outage (nominally 24 months) and to make administrative changes for 6 
other TS to maintain consistency for TS that are not proposed for 
surveillance extension. The specific TS changes proposed include those 
for 2 response time tests, 3 containment spray system tests, and 24 
ventilation system tests.
    Date of issuance: April 14, 1997
    Effective date: April 14, 1997, to be implemented within 90 days of 
issuance.
    Amendment Nos.: Unit 1 - 119; Unit 2 - Amendment No. 117
    Facility Operating License Nos. DPR-80 and DPR-82: The amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: October 9, 1996 (61 FR 
52966) The December 16, 1996, supplemental letter provided additional 
clarifying information and did not change the staffs initial no 
significant hazards consideration determination. The Commission's 
related evaluation of the amendments is contained in a Safety 
Evaluation dated April 14, 1997. No significant hazards consideration 
comments received: No.
    Local Public Document Room location: California Polytechnic State 
University, Robert E. Kennedy Library, Government Documents and Maps 
Department, San Luis Obispo, California 93407

Power Authority of The State of New York, Docket No. 50-286, Indian 
Point Nuclear Generating Unit No. 3, Westchester County, New York

    Date of application for amendment: November 22, 1996
    Brief description of amendment: The amendment allows an increase in 
the U-235 enrichment of fuel stored in the fresh fuel storage racks or 
the spent fuel storage racks from 4.5 weight percent (w/o) U-235 to 5.0 
w/o U-235.
    Date of issuance: April 15, 1997
    Effective date: As of the date of issuance to be implemented within 
30 days.

[[Page 24995]]

    Amendment No.: 173
    Facility Operating License No. DPR-64: Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: January 15, 1997 (62 FR 
2182) The Commission's related evaluation of the amendment is contained 
in the Safety Evaluation dated April 15, 1997, and an Environmental 
Assessment dated March 25, 1997. No significant hazards consideration 
comments received: Yes
    Local Public Document Room location: White Plains Public Library, 
100 Martine Avenue, White Plains, New York 10610

Southern Nuclear Operating Company, Inc., Georgia Power Company, 
Oglethorpe Power Corporation, Municipal Electric Authority of 
Georgia, City of Dalton, Georgia, Docket No. 50-366, Edwin I. Hatch 
Nuclear Plant, Unit 2, Appling County, Georgia

    Date of application for amendments: December 3, 1996, as 
supplemented by letters dated January 27 and April 4, 1997
    Brief description of amendments: The amendments revise Technical 
Specification 2.1.1.2 to change the Safety Limit Minimum Critical Power 
Ratio based on the cycle-specific analyses of Cycle 13 of a non-
equilibrium core of all General Electric (GE) 9 fuel with varying 
enrichments and Cycle 14 of a non-equilibrium mixed core of GE13 and 
GE9 fuel.
    Date of issuance: April 17, 1997
    Effective date: For Cycle 13, as of the date of issuance; For Cycle 
14, effective upon startup.
    Amendment Nos.: 148 for Cycle 13; 149 for Cycle 14
    Facility Operating License Nos. DPR-57 and NPF-5. Amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: January 29, 1997 (62 FR 
4349) The January 27 and April 4, 1997, letters provided additional 
information that did not change the scope of the December 3, 1996, 
application and the initial proposed no significant hazards 
consideration determination. The Commission's related evaluation of the 
amendments is contained in a Safety Evaluation dated April 17, 1997. No 
significant hazards consideration comments received: No
    Local Public Document Room location: Appling County Public Library, 
301 City Hall Drive, Baxley, Georgia 31513

Tennessee Valley Authority, Docket Nos. 50-327 and 50-328, Sequoyah 
Nuclear Plant, Units 1 and 2, Hamilton County, Tennessee

    Date of application for amendments: April 4, 1996, as supplemented 
by letters dated January 10, February 7, February 13, March 17, March 
19, March 20, March 25, April 1, April 6, April 10, April 11, and April 
18, 1997.
    Brief description of amendments: The amendments revise the Sequoyah 
Technical Specifications (TSs) and associated Bases to allow for the 
conversion from Westinghouse fuel to Framatome Cogema Fuel, designated 
Mark-BW. The planned fuel conversion begin with fuel cycle 9 for each 
unit. The amendments would revise the TSs to reflect the fuel design 
and vendor change. The licensee's evaluation was contained in Topical 
Report BAW-10220P, ``Mark-BW Fuel Assembly Application for Sequoyah 
Nuclear Units 1 and 2.''
    Date of issuance: April 21, 1997
    Effective date: As of the date of issuance to be implemented no 
later than 45 days of its issuance for Unit 1, and implemented upon 
installation of Framatome Cogema Fuel in the Unit 2 reactor vessel for 
Unit 2.
    Amendment Nos.: 223 and 214
    Facility Operating License Nos. DPR-77 and DPR-79: Amendments 
revise the Technical Specifications and License Conditions.
    Date of initial notice in Federal Register: May 8, 1996 (61 FR 
20856) The Commission's related evaluation of the amendment is 
contained in a Safety Evaluation dated April 21, 1997 No significant 
hazards consideration comments received: No
    Local Public Document Room location: Chattanooga-Hamilton County 
Library, 1101 Broad Street, Chattanooga, Tennessee 37402

Notice Of Issuance Of Amendments To Facility Operating Licenses And 
Final Determination Of No Significant Hazards Consideration And 
Opportunity For A Hearing (Exigent Public Announcement Or Emergency 
Circumstances)

    During the period since publication of the last biweekly notice, 
the Commission has issued the following amendments. The Commission has 
determined for each of these amendments that the application for the 
amendment complies with the standards and requirements of the Atomic 
Energy Act of 1954, as amended (the Act), and the Commission's rules 
and regulations. The Commission has made appropriate findings as 
required by the Act and the Commission's rules and regulations in 10 
CFR Chapter I, which are set forth in the license amendment.
    Because of exigent or emergency circumstances associated with the 
date the amendment was needed, there was not time for the Commission to 
publish, for public comment before issuance, its usual 30-day Notice of 
Consideration of Issuance of Amendment, Proposed No Significant Hazards 
Consideration Determination, and Opportunity for a Hearing.
    For exigent circumstances, the Commission has either issued a 
Federal Register notice providing opportunity for public comment or has 
used local media to provide notice to the public in the area 
surrounding a licensee's facility of the licensee's application and of 
the Commission's proposed determination of no significant hazards 
consideration. The Commission has provided a reasonable opportunity for 
the public to comment, using its best efforts to make available to the 
public means of communication for the public to respond quickly, and in 
the case of telephone comments, the comments have been recorded or 
transcribed as appropriate and the licensee has been informed of the 
public comments.
    In circumstances where failure to act in a timely way would have 
resulted, for example, in derating or shutdown of a nuclear power plant 
or in prevention of either resumption of operation or of increase in 
power output up to the plant's licensed power level, the Commission may 
not have had an opportunity to provide for public comment on its no 
significant hazards consideration determination. In such case, the 
license amendment has been issued without opportunity for comment. If 
there has been some time for public comment but less than 30 days, the 
Commission may provide an opportunity for public comment. If comments 
have been requested, it is so stated. In either event, the State has 
been consulted by telephone whenever possible.
    Under its regulations, the Commission may issue and make an 
amendment immediately effective, notwithstanding the pendency before it 
of a request for a hearing from any person, in advance of the holding 
and completion of any required hearing, where it has determined that no 
significant hazards consideration is involved.
    The Commission has applied the standards of 10 CFR 50.92 and has 
made a final determination that the amendment involves no significant 
hazards consideration. The basis for this determination is contained in 
the

[[Page 24996]]

documents related to this action. Accordingly, the amendments have been 
issued and made effective as indicated.
    Unless otherwise indicated, the Commission has determined that 
these amendments satisfy the criteria for categorical exclusion in 
accordance with 10 CFR 51.22. Therefore, pursuant to 10 CFR 51.22(b), 
no environmental impact statement or environmental assessment need be 
prepared for these amendments. If the Commission has prepared an 
environmental assessment under the special circumstances provision in 
10 CFR 51.12(b) and has made a determination based on that assessment, 
it is so indicated.
    For further details with respect to the action see (1) the 
application for amendment, (2) the amendment to Facility Operating 
License, and (3) the Commission's related letter, Safety Evaluation 
and/or Environmental Assessment, as indicated. All of these items are 
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 for the particular facility involved.
    The Commission is also offering an opportunity for a hearing with 
respect to the issuance of the amendment. By June 6, 1997, 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 and at the local public document room for the 
particular facility involved. 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 a 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 a 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. Since the Commission has made a final determination 
that the amendment involves no significant hazards consideration, if a 
hearing is requested, it will not stay the effectiveness of the 
amendment. Any hearing held would take place while the amendment is in 
effect.
    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 20555-001, 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 (Project Director): 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 20555-001, and to the 
attorney for the licensee.
    Nontimely filings of petitions for leave to intervene, amended 
petitions, supplemental petitions and/or requests for a hearing will 
not be entertained absent a determination by the Commission, the 
presiding officer or the 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).

Commonwealth Edison Company, Docket Nos. 50-237 and 50-249, Dresden 
Nuclear Power Station, Unit Nos. 2 and 3, Grundy County, Illinois

    Date of application for amendments: April 14, 1997, as supplemented 
on April 17, April 22, and April 24, 1997.
    Brief description of amendments: The proposed amendments requested 
(1) review and approval of an Unreviewed Safety Question (USQ) 
involving the control room operator dose resulting from an error in the 
secondary containment volume, (2) a change in Technical Specification 
(TS) Surveillance Requirements (SR) 4.7. P.2.b and 4.7. P.3 values for 
the allowed methyl iodide penetration for the standby gas treatment 
charcoal adsorbers, and (3) change of TS 5.2.C to

[[Page 24997]]

reflect the new calculated free volume of the secondary containment. 
The April 17, April 22 and April 24, 1997, submittals provided 
additional clarifying information that did not change the initial 
proposed no significant hazards consideration determination.
    Date of Issuance: April 25, 1997
    Effective date: Immediately, to be implemented within 30 days.
    Amendment Nos.: 158 and 153
    Facility Operating License Nos. DPR-19 and DPR-25: The amendments 
revised the Technical Specifications. Press release issued requesting 
comments as to proposed no significant hazards consideration: Yes. 
April 22, 1997. Joliet Herald News. Comments received: No. The 
Commission's related evaluation of the amendments, finding of exigent 
circumstances, consultation with the State of Illinois and final 
determination of no significant hazards consideration are contained in 
a Safety Evaluation dated April 25, 1997.
    Attorney for licensee: Michael I. Miller, Esquire; Sidley and 
Austin, One First National Plaza, Chicago, Illinois 60690
    Local Public Document Room location: Morris Area Public Library 
District, 604 Liberty Street, Morris, Illinois 60450
    NRC Project Director: Robert A. Capra

Pennsylvania Power and Light Company, Docket No. 50-388, 
Susquehanna Steam Electric Station, Unit 2, Luzerne County, 
Pennsylvania

    Date of application for amendment: April 16, 1997, and as 
supplemented by a letter dated April 18, 1997
    Brief description of amendment: This amendment changes the footnote 
in the Design Features Section 5.3.1 of the Technical Specifications to 
allow the use of ATRIUM-10 fuel in Operational Conditions 3 and 4.
    Date of issuance: April 25, 1997
    Effective date: As of the date of issuance to be implemented upon 
receipt.
    Amendment No.: 138
    Facility Operating License No. NPF-22: This amendment revised the 
Technical Specifications. Public comments requested as to proposed no 
significant hazards consideration: Yes. The NRC published a public 
notice of the proposed amendment, issued a proposed finding of no 
significant hazards consideration and reqeusted that any comments on 
the proposed no significant hazards consideration be provided to the 
staff by the close of business on April 24, 1997. The notice was 
published in the Wilkes-Barre Times Leader and the Berwick Press 
Enterprise on April 22-24, 1997. Public comments were received and have 
been addressed in the staff's safety evaluation.
    The Commission's related evaluation of the amendment, finding of 
exigent circumstances, consultation with the State of Pennsylvania and 
final no significant hazards consideration determination are contained 
in a Safety Evaluation dated April 25, 1997.
    Local Public Document Room location: Osterhout Free Library, 
Reference Department, 71 South Franklin Street, Wilkes-Barre, PA 18701
    Attorney for licensee: Jay Silberg, Esquire, Shaw, Pittman, Potts 
and Trowbridge, 2300 N Street NW., Washington, DC 20037
    NRC Project Director: John F. Stolz
    Dated at Rockville, Maryland, this 30th day of April 1997.
    For the Nuclear Regulatory Commission
Elinor G. Adensam,
Deputy Director, Division of Reactor Projects III/IV, Office of Nuclear 
Reactor Regulation.
[Doc. 97-11725 Filed 5-6-97; 8:45 am]
BILLING CODE 7590-01-F