[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 [FQ(z)], and Nuclear Enthalpy Rise Hot
Channel Factor (FN 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: 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