[Federal Register Volume 60, Number 99 (Tuesday, May 23, 1995)]
[Notices]
[Pages 27334-27353]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: 95-12538]



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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 May 1, 1995, through May 12, 1995. The last 
biweekly notice was published on May 10, 1995 (60 FR 24904).

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 Rules Review and 
Directives Branch, Division of Freedom of Information and Publications 
Services, Office of Administration, U.S. Nuclear Regulatory Commission, 
Washington, DC 20555, and should cite the publication date and page 
number of this Federal Register notice. Written comments may also be 
delivered to Room 6D22, Two White Flint North, 11545 Rockville Pike, 
Rockville, Maryland from 7:30 a.m. to 4:15 p.m. Federal workdays. 
Copies of written comments received may be examined at the NRC Public 
Document Room, the Gelman Building, 2120 L Street, NW., Washington, DC. 
The filing of requests for a hearing and petitions for leave to 
intervene is discussed below.
    By June 23, 1995, 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 [[Page 27335]] bases of the contention and a 
concise statement of the alleged facts or expert opinion which support 
the contention and on which the petitioner intends to rely in proving 
the contention at the hearing. The petitioner must also provide 
references to those specific sources and documents of which the 
petitioner is aware and on which the petitioner intends to rely to 
establish those facts or expert opinion. Petitioner must provide 
sufficient information to show that a genuine dispute exists with the 
applicant on a material issue of law or fact. Contentions shall be 
limited to matters within the scope of the amendment under 
consideration. The contention must be one which, if proven, would 
entitle the petitioner to relief. A petitioner who fails to file such a 
supplement which satisfies these requirements with respect to at least 
one contention will not be permitted to participate as a party.
    Those permitted to intervene become parties to the proceeding, 
subject to any limitations in the order granting leave to intervene, 
and have the opportunity to participate fully in the conduct of the 
hearing, including the opportunity to present evidence and cross-
examine witnesses.
    If a hearing is requested, the Commission will make a final 
determination on the issue of no significant hazards consideration. The 
final determination will serve to decide when the hearing is held.
    If the final determination is that the amendment request involves 
no significant hazards consideration, the Commission may issue the 
amendment and make it immediately effective, notwithstanding the 
request for a hearing. Any hearing held would take place after issuance 
of the amendment.
    If the final determination is that the amendment request involves a 
significant hazards consideration, any hearing held would take place 
before the issuance of any amendment.
    A request for a hearing or a petition for leave to intervene must 
be filed with the Secretary of the Commission, U.S. Nuclear Regulatory 
Commission, Washington, DC 20555, Attention: Docketing and Services 
Branch, or may be delivered to the Commission's Public Document Room, 
the Gelman Building, 2120 L Street, NW., Washington DC, 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, 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.

Arizona Public Service Company, et al., Docket Nos. STN 50-528, STN 50-
529, and STN 50-530, Palo Verde Nuclear Generating Station, Units 1, 2, 
and 3, Maricopa County, Arizona

    Date of application for amendments: April 6, 1995.
    Brief description of amendments: The proposed amendment involves 
changes in personnel titles, implementation of line item improvements 
delineated in Generic Letter 93-07, ``Modification of the Technical 
Specification Administrative Control Requirements for Emergency and 
Security Plans,'' changes in the Plant Review Board, and miscellaneous 
minor 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) The proposed changes do not involve a significant increase 
in the probability or consequences of an accident previously 
evaluated.
    These changes involve (1) minor changes in the organization of 
PVNGS, (2) line item improvements recommended by the NRC, or (3) 
clarification or corrections to existing specifications. It is 
expected that the organizational changes will have a positive effect 
on the conduct of plant operations and safety-related work. 
Functions which are necessary to operate the facility safely and in 
accordance with the operating licenses, remain in the new 
organization. The line item improvements to the Technical 
Specifications will not affect the safe operation of the plant and 
continue to ensure proper control of administrative activities. The 
proposed changes will not affect the operation of structures, 
systems and components, and will not reduce programmatic controls 
such that plant safety would be affected. Therefore, the proposed 
changes do 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 analyzed.
    The proposed changes will not affect the operation of 
structures, systems and components, and will not reduce programmatic 
controls such that plant safety would be affected. The changes in 
the organization and as a result of line item improvements will 
continue to provide necessary oversight and control of 
administrative processes. Therefore, the proposed changes do not 
create the possibility of a new or different kind of accident from 
any previously evaluated.
    (3) The proposed changes do not involve a significant reduction 
in a margin of safety.

    These changes are administrative and will not diminish any 
organizational or administrative controls currently in place. The 
proposed changes will not affect the operation of structures, 
systems and components, and will not reduce programmatic controls 
such that plant safety would be affected. Therefore, the proposed 
changes do not involve a significant reduction in a margin of 
safety.

    Local Public Document Room location: Phoenix Public Library, 12 
East McDowell Road, Phoenix, Arizona 85004.
    Attorney for licensee: Nancy C. Loftin, Esq., Corporate Secretary 
and Counsel, Arizona Public Service Company, P.O. Box 53999, Mail 
Station 9068, Phoenix, Arizona 85072-3999.
    NRC Project Director: William H. Bateman.

Arizona Public Service Company, et al., Docket Nos. STN 50-528, STN 50-
529, and STN 50-530, Palo Verde Nuclear Generating Station, Unit Nos. 
1, 2, and 3, Maricopa County, Arizona

    Date of amendment requests: April 18, 1995.
    Description of amendment requests: The proposed Technical 
Specification amendments would revise the surveillance requirements for 
Technical Specification 3/4.4.4, ``Steam Generators,'' and the 
associated Bases. These amendments would allow the installation of tube 
sleeves as an alternative to plugging defective steam generator tubes.
    Basis for proposed no significant hazards consideration 
determination: [[Page 27336]] As required by 10 CFR 50.91(a), the 
licensee has provided its analysis of the issue of no significant 
hazards consideration, which is presented below:

    1. The proposed change does not involve a significant increase 
in the probability or consequences of an accident previously 
evaluated.
    The proposed amendment to permit the use of steam generator tube 
sleeves as an alternative to tube plugging is a safe and effective 
repair procedure that does not require removing a tube from service. 
Mechanical strength, corrosion resistance, installation methods, and 
inservice inspection techniques of sleeves have been shown to meet 
NRC acceptance criteria.
    Analytical verifications were performed using design and 
operating transient parameters selected to envelope loads imposed 
during normal operating and accident conditions. Fatigue and stress 
analysis of sleeved tube assemblies were completed in accordance 
with the requirements of Section III of the ASME Code. The results 
of qualification testing, analysis and plant operating experience at 
other facilities demonstrates that the sleeving process is an 
acceptable means of maintaining steam generator tube integrity. The 
sleeve configuration has been designed and analyzed in accordance 
with the structural margins specified in Regulatory Guide (RG) 
1.121. Furthermore, the installed sleeve will be monitored through 
periodic inspections on a sample basis with eddy current techniques. 
A sleeve-specific plugging margin, per the recommendations of RG 
1.121, has been specified with appropriate allowances for NDE 
(nondestructive examination) uncertainty and defect growth rate.
    The consequences of accidents previously analyzed are not 
increased as a result of sleeving activities. The hypothetical 
failure of the sleeve would be bounded by the current steam 
generator tube rupture analysis contained in the PVNGS (Palo Verde 
Nuclear Generating Station) UFSAR (updated final safety analysis 
report). Due to the slight reduction in diameter caused by the 
sleeve wall thickness, it is expected that the primary release rates 
would be less than assumed for the steam generator tube rupture 
analysis, and therefore would result in lower total primary fluid 
mass release to the secondary system. Additionally, further 
conservatism is introduced if the break were postulated to occur at 
a location on the tube higher than the location where a sleeve is 
installed. The overall effect would be reduced steam generator tube 
rupture release rates. The minimal reduction in flow area associated 
with a tube sleeve has no significant affect on steam generator 
performance with respect to heat transfer or system flow resistance 
and pressure drop. The installation of sleeves rather than plugging 
also maintains a greater heat transfer surface in the steam 
generator. In any case, the impacts are bounded by evaluations which 
demonstrate the acceptability of tube plugging which totally removes 
the tube from service. Therefore, in comparison to plugging, tube 
sleeving is considered a significant improvement with respect to 
steam generator performance. The cumulative impact of multiple 
sleeved tubes was evaluated to ensure the effects remain within the 
analytical design bases.
    Recent industry experience with forced shutdown events 
associated with tube failures at sleeve junctions was assessed by 
ASP and ABB-CE. The root cause of these events has been attributed 
to the lack of proper post-installation stress relief and/or the 
imposition of high stresses due the tube growth restrictions at 
locked tube support. The material and design of the PVNGS steam 
generator supports minimizes the potential for locked supports. The 
tube supports are of eggcrate design and are constructed of ferritic 
stainless steel. The large flow area in the eggcrate design provides 
better irrigation and reduces the potential for steam blanketing, 
therefore, the tube-to-tube support crevices are less likely to be 
blocked by crud, boiler water deposits and corrosion products. Since 
the support material is type 409 ferritic stainless steel, it is not 
susceptible to magnetite corrosion which has resulted in denting and 
lockup at plants with carbon steel supports. These conclusions have 
been substantiated via tube pull activities conducted in PVNGS Unit 
2. Although ABB-CE does not require post-weld heat treatment in all 
applications, APS will require that a post-weld stress relief be 
conducted for all sleeve installations.
    APS has incorporated an integrated leakage monitoring program, 
utilizing equipment, procedure upgrades and administrative shutdown 
limits significantly lower than Technical Specification 
requirements. The program is designed to provide plant operators 
with the ability to detect and respond to changes in primary-to-
secondary leakage and shutdown the unit prior to a significant leak 
or steam generator tube rupture, should sleeve or tube degradation 
exceed expected values. The program is designed to reduce the 
probability of steam generator tube rupture events.
    Therefore, based on the above, the proposed amendment does not 
significantly increase the probability or consequences of an 
accident previously evaluated.
    2. The proposed change does not create the possibility of a new 
or different kind of accident from any accident previously analyzed.
    A sleeved steam generator tube performs the same function in the 
same passive manner as an unsleeved steam generator tube. Tube 
sleeves are designed, qualified, and maintained under the stress and 
pressure limits of Section III of the ASME Code and Regulatory Guide 
1.121.
    The installation of the sleeve, including weld and welder 
qualification and nondestructive examination (NDE), meets or exceeds 
the requirements of ASME Section XI. Three types of NDE are 
conducted. Ultrasonic Testing (UT) is performed to verify adequacy 
of the tube to sleeve weld assuring proper fusion. Eddy current 
testing (ET) is performed following each installation to establish 
baseline data for each sleeve in order to monitor future degradation 
of the primary to secondary pressure boundary. Visual inspections 
may be performed to verify or ascertain the mechanical and 
structural condition of a weld. Critical conditions which are 
checked include weld width and completeness, and the absence of 
visibly noticeable indications such as cracks, pits, and burn 
through.
    ABB-Combustion Engineering Inc., Report CEN-613-P, ``Arizona 
Public Service Co., Palo Verde Units 1, 2, and 3, Steam Generator 
Tube Repair Using Leak Tight Sleeves,'' Revision 01, January 1995, 
demonstrates that the repair of degraded steam generator tubes using 
tube sleeves will result in tube bundle integrity consistent with 
the original design basis. An extensive analysis and corrosion and 
mechanical test programs were undertaken to prove the adequacy of 
tube sleeve repair. The proposed amendments have no significant 
effect on the configuration of the plant, and the change does not 
effect the way in which the plant is operated. Based upon the 
results of the analytical and test programs described in the ABB 
Combustion Engineering Inc. report, the tube sleeve fulfills its 
intended function and meets or exceeds established design criteria. 
Therefore, the proposed change does not create the possibility of a 
new or different kind of accident from any accident previously 
evaluated.
    3. The proposed change does not involve a significant reduction 
in a margin of safety.
    Evaluation of the sleeved tubes indicates no detrimental effects 
on the sleeve-tube assembly resulting from reactor system flow, 
coolant chemistries, or thermal and pressure conditions. Structural 
analyses of the sleeve-tube assembly, using demonstrated margins of 
safety, have established sleeve-tube integrity under normal and 
accident conditions. Structural analyses have been performed for 
sleeves which span the tube at the top of the tubesheet and which 
span the flow distribution plate or eggcrate support. Mechanical 
testing has been performed to support the analyses. Corrosion 
testing of typical sleeve-tube assemblies has been completed and 
reveals no evidence of sleeve or tube corrosion considered 
detrimental under anticipated service conditions.
    Based upon the testing and analyses performed, the installation 
of tube sleeves will not result in a significant reduction in a 
margin of safety.
    Steam generator tube integrity is maintained under the same 
limits for sleeved tubes as for unsleeved tubes, i.e., Section III 
of the ASME Code and Regulatory Guide 1.121. The portions of the 
installed sleeve assembly which represents the reactor coolant 
pressure boundary can be monitored for the initiation and 
progression of sleeve/tube wall degradation, thus satisfying the 
requirements of Regulatory Guide 1.83. The degradation limit at 
which a sleeve/tube boundary is considered inoperable has been 
analyzed in accordance with Regulatory Guide 1.121 and is specified. 
Eddy current detectability of flaws has been verified by ABB 
Combustion Engineering. The Technical Specifications continue to 
require monitoring and restriction of primary to secondary system 
leakage through the steam generators. A conservative integrated 
leakage program employed by APS provides reasonable assurance than 
an orderly unit shutdown will [[Page 27337]] occur prior to a 
significant increase in leakage due to failure of a sleeved or 
unsleeved tube. The minimal reduction in reactor coolant system 
flow, due to sleeving, is considered to have an insignificant impact 
on steam generator operation during normal operation or accident 
conditions and is bounded by tube plugging evaluations. Therefore, 
this change does not involve a significant reduction in a margin of 
safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
that review, it appears that the three standards of Sec. 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment requests involve no significant hazards consideration.
    Local Public Document Room location: Phoenix Public Library, 12 
East McDowell Road, Phoenix, Arizona 85004.
    Attorney for licensees: Nancy C. Loftin, Esq., Corporate Secretary 
and Counsel, Arizona Public Service Company, P.O. Box 53999, Mail 
Station 9068, Phoenix, Arizona 85072-3999.
    NRC Project Director: William H. Bateman.

Commonwealth Edison Company, Docket Nos. 50-237 and 50-249, Dresden 
Nuclear Power Station, Units 2 and 3, Grundy County, Illinois, Docket 
Nos. 50-254 and 50-265, Quad Cities Nuclear Power Station, Units 1 and 
2, Rock Island County, Illinois

    Date of application for amendment request: February 16, 1993, as 
supplemented by letter dated May 2, 1995.
    Description of amendment request: As a result of findings by a 
Diagnostic Evaluation Team inspection performed by the NRC staff at the 
Dresden Nuclear Power Station in 1987, Commonwealth Edison Company 
(ComEd, the licensee) made a decision that both the Dresden Nuclear 
Power Station and sister site Quad Cities Nuclear Power Station, needed 
attention focused on the existing custom Technical Specifications (TS) 
used.
    The licensee made the decision to initiate a Technical 
Specification Upgrade Program (TSUP) for both Dresden and Quad Cities. 
The licensee evaluated the current TS for both Dresden and Quad Cities 
against the Standard Technical Specifications (STS) contained in NUREG-
0123, ``Standard Technical Specifications General Electric Plants BWR/
4.'' The licensee's evaluation identified numerous potential 
improvements such as clarifying requirements, changing TS to make them 
more understandable and to eliminate interpretation, and deleting 
requirements that are no longer considered current with industry 
practice. As a result of the evaluation, ComEd has elected to upgrade 
both the Dresden and Quad Cities TS to the STS contained in NUREG-0123.
    The TSUP for Dresden and Quad Cities is not a complete adaption of 
the STS. The TSUP focuses on (1) integrating additional information 
such as equipment operability requirements during shutdown conditions, 
(2) clarifying requirements such as limiting conditions for operations 
and action statements utilizing STS terminology, (3) deleting 
superseded requirements and modifications to the TS based on the 
licensee's responses to Generic Letters (GL), and (4) relocating 
specific items to more appropriate TS locations.
    The February 16, 1993, and May 2, 1995, applications proposed to 
upgrade only Section 3/4.10 (Refueling Operations) of the Dresden and 
Quad Cities TS.
    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:

    The proposed changes do not involve a significant increase in 
the probability or consequences of an accident previously evaluated 
because:
    In general, the proposed amendment represents the conversion of 
current requirements to a more generic format, or the addition of 
requirements which are based on the current safety analysis. 
Implementation of these changes will provide increased reliability 
of equipment assumed to operate in the current safety analysis, or 
provide continued assurance that specified parameters remain within 
their acceptance limits, and as such, will not significantly 
increase the probability or consequences of a previously evaluated 
accident.
    Some of the proposed changes represent minor curtailments of the 
current requirements which are based on generic guidance or 
previously approved provisions for other stations. The proposed 
amendment for Dresden and Quad Cities Station's Technical 
Specification Section 3/4.10 are based on STS guidelines or later 
operating BWR plant's NRC accepted changes. Any deviations from STS 
requirements do not significantly increase the probability or 
consequences of any previously evaluated accidents for Dresden or 
Quad Cities Stations. The proposed amendment is consistent with the 
current safety analyses and has been previously determined to 
represent sufficient requirements for the assurance and reliability 
of equipment assumed to operate in the safety analysis, or provide 
continued assurance that specified parameters remain within their 
acceptance limits. As such, these changes will not significantly 
increase the probability or consequences of a previously evaluated 
accident.
    The associated systems that make up the Refueling Systems are 
not assumed in any safety analysis to initiate any accident sequence 
for Dresden or Quad Cities Stations; therefore, the probability of 
any accident previously evaluated is not increased by the proposed 
amendment. In addition, the proposed surveillance requirements for 
the proposed amendments to these systems are generally more 
prescriptive than the current requirements specified within the 
Technical Specifications. The additional surveillance requirements 
improve the reliability and availability of all affected systems and 
therefore, reduce the consequences of any accident previously 
evaluated as the probability of the systems outlined within Section 
3/4.10 of the proposed Technical Specifications, performing its 
intended function is increased by the additional surveillances.
    Create the possibility of a new or different kind of accident 
from any previously evaluated because:
    In general, the proposed amendment represents the conversion of 
current requirements to a more generic format, or the addition of 
requirements which are based on the current safety analysis. Others 
represent minor curtailments of the current requirements which are 
based on generic guidance or previously approved provisions for 
other stations. These changes do not involve revisions to the design 
of the station. Some of the changes may involve revision in the 
operation of the station; however, these provide additional 
restrictions which are in accordance with the current safety 
analysis, or are to provide for additional testing or surveillances 
which will not introduce new failure mechanisms beyond those already 
considered in the current safety analyses.
    The proposed amendment for Dresden and Quad Cities Station's 
Technical Specification Section 3/4.10 is based on STS guidelines or 
later operating BWR plants' NRC accepted changes. The proposed 
amendment has been reviewed for acceptability at the Dresden and 
Quad Cities Nuclear Power Stations considering similarity of system 
or component design versus the STS or later operating BWRs. Any 
deviations from STS requirements do not create the possibility of a 
new or different kind of accident previously evaluated for Dresden 
or Quad Cities Stations. No new modes of operation are introduced by 
the proposed changes, considering the acceptable operational modes 
in present specifications, the STS, or later operating BWRs. 
Surveillance requirements are changed to reflect improvements in 
technique, frequency of performance or operating experience at later 
plants. Proposed changes to action statements in many places add 
requirements that are not in the present technical specifications or 
adopt requirements that have been used successfully at other 
operating BWRs with designs similar to Dresden and Quad Cities. The 
proposed changes maintain at least the present level of operability. 
Therefore, the proposed changes do not create the possibility of a 
new or different kind of accident from any previously evaluated.
    The associated systems that make up the Refueling Systems are 
not assumed in any [[Page 27338]] safety analysis to initiate any 
accident sequence for Dresden or Quad Cities Stations. In addition, 
the proposed surveillance requirements for affected systems 
associated with the Refueling Systems are generally more 
prescriptive than the current requirements specified within the 
Technical Specifications; therefore, the proposed changes do not 
create the possibility of a new or different kind of accident from 
any previously evaluated.
    Involve a significant reduction in the margin of safety because:
    In general, the proposed amendment represents the conversion of 
current requirements to a more generic format, or the addition of 
requirements which are based on the current safety analysis. Others 
represent minor curtailments of the current requirements which are 
based on generic guidance or previously approved provisions for 
other stations. Some of the later individual items may introduce 
minor reductions in the margin of safety when compared to the 
current requirements. However, other individual changes are the 
adoption of new requirements which will provide significant 
enhancement of the reliability of the equipment assumed to operate 
in the safety analysis, or provide enhanced assurance that specified 
parameters remain with their acceptance limits. These enhancements 
compensate for the individual minor reductions, such that taken 
together, the proposed changes will not significantly reduce the 
margin of safety.
    The proposed amendment to Technical Specification Section 3/4.10 
implements present requirements, or the intent of present 
requirements in accordance with the guidelines set forth in the STS. 
Any deviations from STS requirements do not significantly reduce the 
margin of safety for Dresden or Quad Cities Stations. The proposed 
changes are intended to improve readability, usability, and the 
understanding of technical specification requirements while 
maintaining acceptable levels of safe operation. The proposed 
changes have been evaluated and found to be acceptable for use at 
Dresden and Quad Cities based on system design, safety analysis 
requirements and operational performance. Since the proposed changes 
are based on NRC accepted provisions at other operating plants that 
are applicable at Dresden and Quad Cities and maintain necessary 
levels of system, component or parameter (reliability), the proposed 
changes do not involve a significant reduction in the margin of 
safety.
    The proposed amendment for Dresden and Quad Cities Stations will 
not reduce the availability of systems associated with the Refueling 
Systems when required to mitigate accident conditions; therefore, 
the proposed changes do 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: For Dresden, Morris Area 
Public Library District, 604 Liberty Street, Morris, Illinois 60450; 
for Quad Cities, Dixon Public Library, 221 Hennepin Avenue, Dixon, 
Illinois 61021.
    Attorney for licensee: Michael I. Miller, Esquire; Sidley and 
Austin, One First National Plaza, Chicago, Illinois 60603.
    NRC Project Director: Robert A. Capra.
Duke Power Company, et al., Docket Nos. 50-413 and 50-414, Catawba 
Nuclear Station, Units 1 and 2, York County, South Carolina

    Date of amendment request: September 19, 1994, as supplemented by 
letter dated April 26, 1995.
    Description of amendment request: The amendments would change the 
Technical Specifications (TS) to increase the enrichment limits for 
fuel stored in the fuel pools and establish restricted loading patterns 
and associated burnup criteria for qualifying fuel in the spent fuel 
pools. In addition, several administrative changes have been included 
in order to provide clarity to the TS and bring them more in line with 
the Standard Technical Specifications format. These changes are as 
follows:
    (1) The TS index is changed to add TS 3/4.9.12 and 3/4.9.13, Tables 
3.9-1 and 3.9-2 and Figure 3.9-1.
    (2) TS 3/4.9.12, Spent Fuel Pool (SFP) Boron Concentration, is 
added to establish a boron concentration limit and to establish a 
Limiting Condition for Operation (LCO) for all modes of operation and 
to allow the numerical value of the limit to be specified in the Core 
Operating Limits Report (COLR).
    (3) TS 3/4.9.13, Tables 3.9-1 and 3.9-2 and Figure 3.9-1 are being 
added to establish restricted loading patterns for spent fuel storage 
and associated burnup criteria.
    (4) Corresponding BASES for TSs 3/4.9.12 and 3/4.9.13 are added to 
explain the basis for each LCO, Action Statement, and Surveillance 
Requirement covered by the subject TSs.
    (5) TS 5.6, Fuel Storage, is changed to reflect limits for 
criticality analysis for fuel storage.
    (6) TS 6.9, Reporting Requirements, is changed to reflect the 
inclusion of the SFP boron concentration limit values in the COLR as 
established by TS 3/4.9.12.
    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.
    There is no increase in the probability or consequences of an 
accident in the new fuel vault since the only credible accidents for 
this area are criticality accidents and it has been shown that 
calculated, worst case K<INF>eff for this area is (less than or 
equal to) 0.95 under all conditions.
    There is no increase in the probability of a fuel drop accident 
in the Spent Fuel Storage Pool since the mass of an assembly will 
not be affected by the increase in fuel enrichment. The likelihood 
of other accidents, previously evaluated and described in Section 
9.1.2 of the FSAR (Final Safety Analysis Report), is also not 
affected by the proposed changes. In fact, it could be postulated 
that since the increase in fuel enrichment will allow for extended 
fuel cycles, there will be a decrease in fuel movement and the 
probability of an accident may likewise be decreased. There is also 
no increase in the consequences of a fuel drop accident in the Spent 
Fuel Pool since the fission product inventory of individual fuel 
assemblies will not change significantly as a result of increased 
initial enrichment. In addition, no change to safety related systems 
is being made.
    Therefore, the consequences of a fuel rupture accident remain 
unchanged. In addition, it has been shown that K<INF>eff is (less 
than or equal to) 0.95, under all conditions. Therefore, the 
consequences of a criticality accident in the Spent Fuel Pool remain 
unchanged as well.
    2. The proposed changes do not create the possibility of a new 
or different kind of accident from any accident previously 
evaluated.
    The proposed changes do not create the possibility of a new or 
different kind of accident since fuel handling accidents (fuel drop 
and misplacement) are not new or different kinds of accidents. Fuel 
handling accidents are already discussed in the FSAR for fuel with 
enrichments up to 4.0 weight %. As described in Section IV.9 of 
Attachment IV, additional analyses have been performed for fuel with 
enrichment up to 5.00 weight %. Worst case misloading accidents 
associated with the new loading patterns were evaluated. It was 
shown that the negative reactivity provided by soluble boron 
maintains K<INF>eff (less than or equal to) 0.95.
    3. The proposed changes do not involve a significant reduction 
in the margin of safety.
    The proposed change does not involve a significant reduction in 
the margin of safety since, in all cases, a K<INF>eff [less than or 
equal to] 0.95 is being maintained. Criticality analyses have been 
performed which show that the new fuel storage vault will remain 
subcritical under a variety of moderation conditions, from fully 
flooded to optimum moderation. As discussed above, the Spent Fuel 
Pool will remain sufficiently subcritical during any fuel 
misplacement accident.

    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 
[[Page 27339]] satisfied. Therefore, the NRC staff proposes to 
determine that the amendment request involves no significant hazards 
consideration.
    Local Public Document Room location: York County Library, 138 East 
Black Street, Rock Hill, South Carolina 29730.
    Attorney for licensee: Mr. Albert Carr, Duke Power Company, 422 
South Church Street, Charlotte, North Carolina 28242.
    NRC Project Director: Herbert N. Berkow.

Florida Power and Light Company, Docket Nos. 50-250 and 50-251, Turkey 
Point Plant Units 3 and 4, Dade County, Florida

    Date of amendment request: March 30, 1995, and supplemented May 5, 
1995.
    Description of amendment request: The licensee proposes to change 
Turkey Point Units 3 and 4 Technical Specifications (TS) by separation 
of the 24-hour emergency diesel generator (EDG) run from the hot 
restart EDG test.
    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 facility in accordance with the proposed 
amendments would not involve a significant increase in the 
probability or consequences of an accident previously evaluated.
    The proposed TS changes would revise the EDG surveillance 
criteria to allow the EDG hot-start test with full ESF load 
acceptance to be performed separately and independently from the 24-
hour EDG run. The proposed SRs (surveillance requirements) would 
continue to demonstrate that the objectives of these two tests are 
met. Specifically, the EDGs are shown to be: (1) Capable of starting 
and running continuously at full load capability for an interval not 
less than 24 hours, and (2) capable of restarting from a full load 
temperature condition. The proposed changes would not affect the 
EDGs' ability to support mitigation of the consequences of any 
previously evaluated accident. Additionally, the proposed changes to 
the SRs do not affect the initiating assumptions or progression of 
any accident sequence.
    Therefore, operation of the facility would not involve a 
significant increase in the probability or consequences of an 
accident previously analyzed.
    (2) Operation of the facility in accordance with the proposed 
amendments would not create the possibility of a new or different 
kind of accident from any accident previously evaluated.
    The proposed TS SR changes do not require any physical changes 
to the plant or equipment, and do not impact any design or 
functional requirements of the EDGs. The proposed changes do not 
create any plant configurations which are prohibited by the TS. The 
proposed changes continue to meet the EDG test objectives associated 
with demonstrating EDG operability.
    Therefore, operation of the facility in accordance with the 
proposed amendments would not create the possibility of a new or 
different kind of accident from any accident previously evaluated.
    (3) Operation of the facility in accordance with the proposed 
amendments would not involve a significant reduction in a margin of 
safety.
    The proposed TS SR changes do not require any physical changes 
to the plant or equipment and do not impact any design or functional 
requirements of the EDGs. Surveillance testing in accordance with 
the proposed TS will continue to demonstrate the ability of the EDGs 
to perform their intended function of providing electrical power to 
mitigate design basis transients, consistent with the plant safety 
analyses.
    Therefore, operation of the facility in accordance with the 
proposed amendments would not involve a 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 Sec. 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: Florida International 
University, University Park, Miami, Florida 33199.
    Attorney for licensee: J. R. Newman, Esquire, Morgan, Lewis & 
Bockius, 1800 M Street, NW., Washington, DC 20036.
    NRC Project Director: David B. Matthews.

Omaha Public Power District, Docket No. 50-285, Fort Calhoun Station, 
Unit No. 1, Washington County, Nebraska

    Date of amendment request: April 7, 1995.
    Description of amendment request: The proposed amendment would 
revise the technical specifications (TS) to relocate the axial power 
distribution limits to the Core Operating Limits Report (COLR).
    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 change does not involve a significant increase 
in the probability or consequences of an accident previously 
evaluated.
    The proposed change relocates the cycle-specific Axial Power 
Distribution (APD) limits contained in Figure 1-2 of the Technical 
Specifications (TS), to the Core Operating Limits Report (COLR). 
This change is consistent with the NRC recommendations of Generic 
Letter 88-16, and will not modify the methodology used in generating 
the limits nor the manner in which they are implemented. The 
methodology used to determine the APD limits is reviewed and 
approved by the NRC in accordance with TS 5.9.5. The APD limits will 
continue to be determined by analyzing the same postulated events as 
previously analyzed. The plant will continue to operate within the 
limits specified in the COLR and will take the same remedial actions 
if the APD limit is exceeded as required by the current TS. 
Therefore, the proposed change would not increase the probability or 
consequences of an accident previously evaluated.
    (2) The proposed change does not create the possibility of a new 
or different kind of accident from any accident previously 
evaluated.
    There will be no physical alterations to the plant 
configuration, changes to setpoint values, or changes to the 
implementation of setpoints or limits as a result of this proposed 
change. The proposed change only relocates the APD figure from the 
TS to the COLR consistent with NRC Generic Letter 88-16. Therefore, 
the proposed change does not create the possibility of a new or 
different kind of accident from any previously evaluated.
    (3) The proposed change does not involve a significant reduction 
in a margin of safety.
    As indicated above, the implementation of the APD into the COLR, 
consistent with the guidance of NRC Generic Letter 88-16, makes use 
of the existing safety analysis methodologies and the resulting 
limits and setpoints for plant operation. Additionally, the safety 
analysis acceptance criteria for operations with the proposed change 
have not changed from that use in the current reload analysis. 
Therefore, the margin of safety is not reduced due to the relocation 
of the APD from the TS and implementation in the COLR.

    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: W. Dale Clark Library, 215 
South 15th Street, Omaha, Nebraska 68102.
    Attorney for licensee: LeBoeuf, Lamb, Leiby, and MacRae, 1875 
Connecticut Avenue, NW., Washington, DC 20009-5728.
    NRC Project Director: William Bateman. [[Page 27340]] 

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 amendment requests: April 19, 1995 (Reference LAR 95-03).
    Description of amendment requests: The proposed amendments would 
revise the combined Technical Specifications (TS) for the Diablo Canyon 
Nuclear Power Plant, Unit Nos. 1 and 2 to revise TS 3/4.8.1.1, ``A.C. 
Sources, Operating.'' The specific TS changes proposed are as follows:
    (1) TS 4.8.1.1.2b.8), emergency diesel generator (EDG) 24-hour load 
run and hot restart surveillance, would be revised to delete the 
requirement to perform TS 4.8.1.1.2b.5)b), loss of offsite power (LOOP) 
load sequencing surveillance within 5 minutes following the 24-hour 
test.
    (2) New TS 4.8.1.1.2e. would be added to perform an EDG hot restart 
test within 5 minutes of shutting down the EDG after the EDG has 
operated for at least 2 hours at a load of greater than or equal to 
2484 kW.
    (3) TS 4.8.1.1.2b.8), TS 4.8.1.1.2e., and footnote ``*'' on page 3/
4 8-5 would be changed to be cycle-specific with the new TS 
requirements effective for Units 1 and 2, Cycle 8 and after.
    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.
    Demonstrating emergency diesel generator (EDG) hot restart 
capability without sequencing loss of offsite power (LOOP) loads 
does not invalidate or reduce the effectiveness of the hot restart 
test, since normal operating temperatures are achieved prior to the 
hot restart test. Sequencing the LOOP loads does not contribute to 
verifying that the EDG will start from normal operating 
temperatures. The proposed TS 4.8.1.1.[2]e may be performed in any 
plant condition since performance of this new surveillance will have 
no adverse effect on plant operations. The reliability of the EDGs 
is not affected by the proposed changes.
    Therefore, the proposed changes do 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.
    The proposed changes do not involve any physical alterations to 
the plant. The proposed changes will not have any adverse effect on 
the ability of the EDGs to perform their required safety function.
    Therefore, the proposed changes do not create the possibility of 
a new or different kind of accident from any accident previously 
evaluated.
    3. The proposed changes do not involve a significant reduction 
in a margin of safety.
    The proposed changes will not alter any accident analysis 
assumptions, initial conditions, or results. Consequently, the 
proposed changes do not have any effect on the margin of safety. The 
proposed changes to the surveillance requirements would continue to 
demonstrate the ability of the EDGs to perform their intended safety 
function.
    Therefore, the proposed changes do 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 Sec. 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment requests involve no significant hazards consideration.
    Local Public Document Room location: California Polytechnic State 
University, Robert E. Kennedy Library, Government Documents and Maps 
Department, San Luis Obispo, California 93407.
    Attorney for licensee: Christopher J. Warner, Esq., Pacific Gas and 
Electric Company, PO Box 7442, San Francisco, California 94120.
    NRC Project Director: William H. Bateman.

Philadelphia Electric Company, Public Service Electric and Gas Company, 
Delmarva Power and Light Company, and Atlantic City Electric Company, 
Docket No. 50-278, Peach Bottom Atomic Power Station, Unit No. 3, York 
County, Pennsylvania

    Date of application for amendment: November 21, 1994.
    Description of amendment request: The proposed change would extend 
the Type A test (i.e., Containment Integrated Leak Rate Test (CILRT)) 
interval on a one-time basis.
    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 Technical Specifications (TS) change does not 
involve a significant increase in the probability or consequences of 
an accident previously evaluated.
    The accidents which are potentially adversely impacted by the 
proposed change are any Loss of Coolant Accident (LOCA) inside 
primary containment as described in the PBAPS, Units 2 and 3 UFSAR.
    The proposed change increases the surveillance interval of the 
10 CFR part 50, appendix J Type A test (i.e., Containment Integrated 
Leakage Rate Test (CILRT)) from 46 months to 70 months. This test is 
performed to determine that the total leakage from containment does 
not exceed the maximum allowable primary containment leakage rate 
(i.e., designated La) at a calculated peak containment internal 
pressure (Pa), as defined in 10 CFR part 50, appendix J. The primary 
containment limits the leakage of radioactive material during and 
following design bases accidents in order to comply with the offsite 
does limits specified in 10 CFR part 100. Accordingly, the primary 
containment is not an accident initiator. It is an accident 
mitigator. No physical or operational changes to the containment 
structure, plant systems, or components would be made as a result of 
the proposed change. Therefore, the probability of occurrence of an 
accident previously evaluated is not increased.
    The failure effects that are potentially created by the proposed 
one-time TS change have been considered. The relevant components 
important to safety which are potentially affected are the 
containment structure, plant systems, and containment penetrations. 
There are no physical or operational changes to any plant equipment 
associated with the proposed TS change. Therefore, the probability 
or consequences of a malfunction of equipment important to safety is 
not increased.
    The proposed change introduces the possibility that primary 
containment leakage in excess of the allowable value (i.e., La) 
would remain undetected during the proposed 24 month extension of 
the interval between the Type A tests. The types of mechanisms which 
would cause degradation of the primary containment can be 
categorized into two types. These are: (1) Degradation due to work 
which is performed as part of a modification or maintenance activity 
on a component or system (i.e., activity-based), or; (2) degradation 
resulting from a time-based failure mechanism.
    A review of the history of the PBAPS, Unit 3 CILRT results was 
performed to evaluate the risk of activity-based and time-based 
degradation. This review has determined that the potential for a 
time-based and activity-based failure is minimal. The PBAPS LLRT 
program would identify most types of penetration leakage. The LLRT 
program involves measurement of leakage from Type B and Type C 
primary containment penetrations as defined in 10 CFR part 50, 
appendix J.
    The 10 CFR part 50, appendix J, Type B tests are intended to 
detect local leaks and to measure leakage across pressure containing 
or leakage-limiting boundaries other than values, such as 
containment penetrations incorporating resilient seals, gaskets, 
expansion bellows, flexible seal assemblies, door operating 
mechanism penetrations that are part of the containment system, 
doors, and hatches. 10 CFR part 50, appendix J, Type C testing is 
intended to measure reactor system primary containment isolation 
valve leakage rates. The frequency of the Type B and Type C testing 
is not being altered by the [[Page 27341]] proposed TS change. The 
acceptance criterion for Type B and Type C leakage is 0.6 La (i.e., 
0.3% wt/day) which, when compared to the Type A test acceptance 
criterion of 0.75 La (i.e., 0.375% wt/day), is a significant portion 
of the Type A test allowable leakage.
    The proposed TS change only extends the interval between two 
consecutive Type A tests. The Type B and Type C tests will be 
performed as required. The Type B and Type C tests will continue to 
be used to confirm that the containment isolation valves and 
penetrations have not degraded. Containment system components that 
would not be tested are the containment structure itself and small-
diameter instrumentation lines. Time-based degradation of any of the 
instrumentation lines would not likely be identified by faulty 
instrument indication or during instrument calibrations that will be 
performed during the PBAPS, Unit 3 refueling outage 10. In examining 
the potential for a time-based failure mechanism that could cause 
significant degradation of the containment structure, we concluded 
that the risk, if any, of such a mechanism is small since the design 
requirements and fabrication specifications established for the 
containment structure are in themselves adequate to ensure 
containment leak tight integrity.
    Based on the above evaluation, we have concluded that the 
proposed TS change will have a negligible impact on the consequences 
of any accident previously evaluated.
    Although this review concluded that the risk of undetected 
primary containment degradation is not increased, the Individual 
Plan Examination (IPE) for PBAPS, Units 2 and 3, was also reviewed 
in order to access the impact of exceeding the primary containment 
allowable leakage rate, if a non-mechanistic activity type (i.e., 
time-based) failure were to occur. The IPE included an evaluation of 
the effect of various containment leakage sizes under different 
scenarios. The IPE results showed that a containment leakage rate of 
35% wt/day would represent less than a 5% increase in risk to the 
public of being exposed to radiation. This evaluation was based on a 
study performed by Oak Ridge National Laboratory for light water 
reactors that evaluated the impact of leakage rates on public risk. 
As stated earlier, the current value of La for PBAPS, Unit 3, is 
0.5% wt/day, which is significantly less than the 35% wt/day 
discussed in the IPE evaluation.
    Therefore, the proposed TS change involving a one-time extension 
of the Type A test interval and performing the Type A test after the 
second appendix J 10-year service period will not involve an 
increase in the probability or consequences of an accident 
previously evaluated.
    2. The proposed TS change does not create the possibility of a 
new or different kind of accident from any accident previously 
evaluated.
    The proposed change is an increase of a surveillance test 
interval and does not make any physical or operational changes to 
existing plant systems or components. Primary containment acts as an 
accident mitigator not initiator. Therefore, the possibility of a 
different type of accident than any previously evaluated or the 
possibility of a different type of equipment malfunction is not 
introduced.
    Therefore, the proposed TS change does not create the 
possibility of a new or different kind of accident from any accident 
previously evaluated.
    3. The proposed TS change does not involve a significant 
reduction in a margin of safety.
    The total primary containment leakage rate ensures that the 
total containment leakage volume will not exceed the value assumed 
in the safety analyses at the peak accident pressure. As an added 
conservatism, the measured overall leakage rate is further limited 
to less than or equal to 0.75 La during performance of periodic 
tests to account for possible degradation of the containment leakage 
barriers between leakage tests. There is the potential that 
containment degradation could remain undetected during the proposed 
24 month surveillance interval extension and result in the 
containment leakage exceeding this allowable value assumed in safety 
analysis. A review of the history of the PBAPS, Unit 3 CILRT results 
was performed to evaluate the risk of activity-based and time-based 
degradation. This review has determined that the potential for a 
time-based and activity-based failure is minimal. The PBAPS LLRT 
program would identify most types of penetration leakage. The LLRT 
program involves measurement of leakage from Type B and Type C 
primary containment penetrations as defined in 10 CFR part 50, 
appendix J.
    The 10 CFR part 50, appendix J, Type B tests are intended to 
detect local leaks and to measure leakage across pressure containing 
or leakage-limiting boundaries other than valves, such as 
containment penetrations incorporating resilient seals, gaskets, 
expansion bellows, flexible seal assemblies, door operating 
mechanism penetrations that are part of the containment system, 
doors, and hatches. 10 CFR part 50, appendix J, Type C testing is 
intended to measure reactor system primary containment isolation 
valve leakage rates. The frequency of the Type B and Type C testing 
is not being altered by the proposed TS change.
    Therefore, we have concluded that the proposed extended test 
interval would not result in a non-detectable PBAPS, Unit 3 primary 
containment leakage rate in excess of the allowable value (i.e., 
0.5% wt/day) established by the TS and 10 CFR part 50, appendix J.
    Therefore, the proposed TS 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: Government Publications 
Section, State Library of Pennsylvania, (REGIONAL DEPOSITORY) Education 
Building, Walnut Street and Commonwealth Avenue, Box 1601, Harrisburg, 
Pennsylvania 17105.
    Attorney for Licensee: J.W. Durham, Sr., Esquire, Sr. V.P. and 
General Counsel, Philadelphia Electric Company, 2301 Market Street, 
Philadelphia, Pennsylvania 19101.
    NRC Project Director: John F. Stolz.

Public Service Electric & Gas Company, Docket No. 50-272, Salem Nuclear 
Generating Station, Unit No. 1, Salem County, New Jersey

    Date of amendment request: April 4, 1995.
    Description of amendment request: The amendment would provide a 
one-time interval extension for the Type A test required by 10 CFR part 
50, appendix J. The extension would allow the test to be conducted 
during the thirteenth refueling outage, rather than the twelfth 
refueling outage.
    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 not involve a significant increase in the probability or 
consequences of an accident previously evaluated.
    The proposed change will provide a one-time exemption from 10 
CFR part 50, appendix J Section III.D.1(a) leak rate test schedule 
requirement. This change will allow for a one-time test interval for 
Type A Integrated Leak Rate Tests (ILRTs) of 65+/-10 months.
    Leak rate testing is not an initiating event in any accident, 
therefore, this proposed change does not involve a significant 
increase in the probability of a previously evaluated accident.
    Type A tests are capable of detecting both local leak paths and 
gross containment failure paths. The history at Salem Generating 
Station Unit 1 (SGS1) demonstrates that Type B and C Local Leak Rate 
Tests (LLRTs) have consistently detected any excessive local 
leakages. SGS1 has passed all of its ILRTs with significant margin.
    Administrtive controls govern the maintenance and testing of 
containment penetrations such that the probability of excessive 
penetration leakage due to improper maintenance or valve 
misalignment is very low. Following any maintenance that could 
affect the leakage characteristics of any containment penetration, 
an LLRT is performed to ensure acceptable leakage levels. Following 
any LLRT on a containment isolation valve, an independent valve 
alignment check is performed before declaring the penetration 
OPERABLE. Therefore, Type A testing is not necessary to ensure 
acceptable leakage rates through containment penetrations.
    While Type A testing is not necessary to ensure acceptable 
leakage rates through [[Page 27342]] containment penetrations, Type 
A testing is necessary to demonstrate that there are no gross 
containment failures. Structural failure of the containment is 
considered to be a very unlikely event, and in fact, since SGS1 has 
been in operation, it has never failed a Type A ILRT. Therefore, a 
one-time exemption increasing the interval for performing an ILRT 
does not result in a significant decrease in the confidence in the 
leak tightness of the containment structure.
    Therefore, this proposed change does not result in a significant 
increase of the probability or consequences of any previously 
evaluated accident.
    2. Will not create the possibility of a new or different kind of 
accident from any accident previously evaluated.
    This proposed change allows a one-time interval of 65+/-10 
months for the next ILRT. The method of performing the test is not 
changed. No new accident modes are created by extending the testing 
intervals. No safety-related equipment or safety functions are 
altered as a result of this change. A one-time extension of the ILRT 
test interval has no influence on, nor does it contribute in any way 
to, the possibility of a new or different kind of accident or 
malfunction from those previously analyzed.
    3. Will not involve a significant reduction in a margin of 
safety.
    The purpose of the existing schedule of ILRTs is to ensure that 
the release of radioactive materials will be restricted to those 
leak paths and leak rates assumed in accident analyses. The relaxed 
schedule for ILRTs does not allow for relaxation of Type B and C 
LLRTs. Therefore, methods for detecting local containment leak paths 
and leak rates are unaffected by this proposed change. Given that 
the test history for ILRTs shows no failure during plant life, a 
one-time increase of the test interval does not lead to a 
significant probability of creating a new leakage path or increased 
leakage rates, and the margin of safety inherent in existing 
accident analyses is maintained. Therefore, this change does not 
involve a significant reduction kin 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: Salem Free Public Library, 112 
West Broadway, Salem, New Jersey 08079.
    Attorney for licensee: Mark J. Wetterhahn, Esquire, Winston and 
Strawn, 1400 L Street, NW, Washington, DC 20005-3502.
    NRC Project Director: John F. Stolz.

Public Service Electric & Gas Company, Docket No. 50-272, Salem Nuclear 
Generating Station, Unit No. 1, Salem County, New Jersey

    Date of amendment request: May 4, 1995.
    Description of amendment request: The amendment would authorize a 
one-time extension for the Type A test (overall integrated containment 
leakage rate) that is required by 10 CFR part 50, appendix J. The 
current Technical Specification would require that this test be 
conducted by July 7, 1995. The amendment would allow this test to be 
conducted by November 30, 1995.
    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 not involve a significant increase in the probability or 
consequences of an accident previously evaluated.
    The proposed change involves no hardware changes, no changes to 
the operation of any systems or components, and no changes to 
existing structures. This change is temporary, allowing a one-time 
extension of a specific surveillance requirement for cycle 12 to 
allow surveillance testing to coincide with the twelfth refueling 
outage. The proposed surveillance interval extension is short and 
will not result in any significant reduction in structural 
reliability nor will the extension affect the ability of the 
structure in performing its intended functions. to preclude the 
possibility of an undetected containment failure/leakage at a valve 
or penetration seal, Type ``B'' and ``C'' tests will continue to be 
performed as required by the Technical Specifications. Therefore, 
this change will not involve a significant increase in the 
probability or consequences of any accidents previously evaluated.
    2. Will not create the possibility of a new or different kind of 
accident from any previously evaluated.
    Extending the surveillance interval for the performance of 
specific testing will not create the possibility of any new or 
different kinds of accident. No changes are required to any system 
configurations, plant equipment, or analyses. Therefore, this change 
will not create the possibility of a new or different kind of 
accident from any accident previously evaluated.
    3. Will not involve a significant reduction in a margin of 
safety.
    The proposed change will not alter any assumptions, initial 
conditions, or results of any accident analyses. The safety limits 
assumed in the accident analyses and the design function of the 
structure required to mitigate the consequences of any postulated 
accidents will not be changed since only the surveillance interval 
is being extended. Historical performance indicates a high degree of 
reliability, and surveillance testing performed during continued 
plant operation will verify that Salem 1 will remain within analyzed 
limits. Consequently, the 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: Salem Free Public library, 112 
West Broadway, Salem, New Jersey 08079.
    Attorney for licensee: Mark J. Wetterhahn, Esquire, Winston and 
Strawn, 1400 L Street, NW., Washington, DC 20005-3502.
    NRC Project Director: John F. Stolz.

Public Service Electric & Gas Company, Docket Nos. 50-272 and 50-311, 
Salem Nuclear Generating Station, Unit Nos. 1 and 2, Salem County, New 
Jersey

    Date of amendment request: April 18, 1995.
    Description of amendment request: The amendments would delete the 
quarterly leak rate test for the containment pressure-vacuum relief 
valves which is presently required because of the valves' resilient 
seat material. The resilient valve seat material will be replaced with 
a hard seat (metal to metal) design. The valves would still remain in 
the 10 CFR part 50 appendix J, Type C leak rate test program.
    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. Do not involve a significant increase in the probability or 
consequences of an accident previously evaluated.
    The containment pressure/vacuum relief valves are normally 
closed, and are used under administrative control to maintain 
containment internal pressure within -1.5 psig and +0.3 psig, as 
required by SGS Technical Specifications. The pressure/vacuum relief 
valves are relied upon for containment isolation and automatically 
close on high containment pressure or high containment atmosphere 
radioactivity. The pressure/vacuum relief system does not affect the 
probability of any previously evaluated accident.
    The containment isolation function of the pressure/vacuum relief 
valves limits the consequences of a radiological release inside 
containment (i.e., Loss of Coolant Accident). The proposed changes 
to eliminate quarterly pressure drop (leak rate) testing would not 
increase the consequences of any previously evaluated accident. The 
valve flow characteristics and closure time requirements are not 
affected. The valves will continue to be subject to the Type C leak 
rate test criteria of 10 CFR part 50, appendix J. The deletion of 
the augmented quarterly test requirement is justified by replacement 
of the resilient [[Page 27343]] valve seat material (which has a 
history of degradation and loss of leaktightness) with a metal to 
metal seating design.
    2. Do not create the possibility of a new or different kind of 
accident from any accident previously evaluated.
    Eliminating quarterly leak rate testing based on improved valve 
design would not result in any new or different kind of accident. 
The valves would continue to perform the containment isolation 
function consistent with the plant safety analyses, and would not 
adversely affect the initiation or progression of any accident 
sequence.
    (3) Do not involve a significant reduction in a margin of 
safety.
    This proposal involves replacement of the existing pressure/
vacuum relief valves, which have resilient seating material, with 
valves using a hard seat (metal to metal design). Based on the 
improved design and operating experience of the replacement valves, 
augmented quarterly leak rate testing is no longer necessary or 
appropriate to verify leaktightness of the valves. Periodic leak 
rate testing will continue to be performed in accordance with 10 CFR 
part 450, appendix J. The pressure/vacuum relief valves will 
continue to maintain their containment isolation capability such 
that no margin of safety is affected by the proposed changes.

    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: Salem Free Public library, 112 
West Broadway, Salem, New Jersey 08079.
    Attorney for licensee: Mark J. Wetterhahn, Esquire, Winston and 
Strawn, 1400 L Street, NW., Washington, DC 20005-3502.
    NRC Project Director: John F. Stolz.

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

    Date of amendment request: May 3, 1995 (TS 93-09).
    Descripton of amendment request: The proposed change would revise 
the implementation schedule for Amendment Nos. 182 and 174 from that 
stated in the amendments when they were approved by the Commission by 
letter dated May 24, 1994. As issued, the amendments reflected the 
licensee's plans to implement the changes for both units during the 
Unit 2 Cycle 6 refueling outage. However, by letter dated August 19, 
1994, the licensee requested that implementation be delayed to 1995. 
This request was granted by Amendment Nos. 188 and 180 for Units 1 and 
2 respectively by letter dated October 17, 1994. By letter dated May 3, 
1995, the licensee informed the staff that evaluation of the design 
changes have concluded that significant safety risks would be involved 
with modification activities associated with installation. Therefore, 
the licensee has requested that implementation of the amendment be 
changed to specify that the amendment will be implemented along with 
the related plant modifications, without specifying the date when the 
modifications would be performed. No changes to the technical 
specification pages other than those approved when the amendments were 
issued are needed.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
determined that the no significant hazards consideration exists. This 
analysis was provided in the original submittal for the amendment from 
the licensee dated October 1, 1993, and was used in the preparation of 
the amendments. The licensee has determined that this analysis remains 
valid for the proposed revision and that the changes do not constitute 
a significant hazard. The staff previously issued the proposed finding 
in the Federal Register (59 FR 4947 and 59 FR 47182) and there were no 
public comments on the finding. This analysis is reproduced as follows:

    TVA has evaluated the proposed technical specification (TS) 
change and has determined that it does not represent a significant 
hazards consideration based on criteria established in 10 CFR 
50.92(c). Operation of Sequoyah Nuclear Plant (SQN) in accordance 
with the proposed amendment will not:
    1. Involve a significant increase in the probability or 
consequences of an accident previously evaluated.
    The proposed revision supports the implementation of design 
logic and setpoint changes to the loss-of-power relaying. This 
relaying is designed to ensure adequate voltage is available to 
safety-related loads in order to enhance their operability and 
support accident mitigation functions and to provide for auxiliary 
feedwater (AFW) pump starts. The design changes alter relay logic 
and delete unnecessary relaying, but do not change the diesel 
generator (D/G) start and load-shedding actuations that result from 
loss-of-power conditions. Therefore, no new actuations or functions 
have been created; and because the existing and proposed functions 
provide for accident mitigation considerations that are not the 
source of an accident, the probability of an accident is not 
increased. The deletion of the 6.9-kilovolt shutdown board normal-
feeder undervoltage relays actually reduces the potential for 
inadvertent shutdown board blackouts as a result of short-duration 
voltage transients or instrument failures.
    The setpoints and time delays for loss-of-power functions have 
been modified based on the guidelines developed by the Electrical 
Distribution System Clearinghouse as evaluated and determined 
through detailed analysis by TVA. This design is documented in TVA 
Calculations SQN-EEB-MS-TI06-0008, 27DAT, and DS-1-2 and is 
available for NRC review at the SQN site. The assigned values are 
conservative settings that will ensure adequate voltage is supplied 
to safety-related loads for accident mitigation and safety functions 
under normal, degraded, and loss-of-offsite-power voltage conditions 
with appropriate time delays to prevent damage to electrical loads 
and minimize premature or unnecessary actuations. The identification 
of loss-of-voltage conditions is enhanced by the design changes to 
ensure the timely sequencing of loads onto the D/G and the 
initiation of AFW pump starts for accident mitigation. Because there 
are no reductions in safety functions resulting from the design 
logic, setpoint, and time-delay changes to the loss-of-power 
instrumentation and offsite dose levels for postulated accidents 
will not be increased, the consequences of an accident are not 
increased.
    The applicable mode addition, TS 3.0.4 exclusion deletion, and 
response time measurement clarification incorporated in the proposed 
change do not affect plant functions. These changes reflect the 
requirements that SQN has been maintaining and serve to clarify the 
requirements to provide consistency of application and easier 
understanding. The AFW footnote addition and bases revision only 
clarify operability conditions that are consistent with the plant 
design for the AFW pump and loss-of-power instrumentation. Because 
there are no changes to plant functions or operations, these 
revisions have no impact on accident probabilities or consequences.
    2. Create the possibility of a new or different kind of accident 
from any previously analyzed.
    As described above, the loss-of-power instrumentation ensures 
adequate voltage to safety-related loads by initiating D/G starts 
and load shedding and provides for AFW pump starting, but is not 
considered to be the source of an accident. Although the design 
logic, setpoint, and time-delay actuation criteria have changed, the 
output functions to various plant systems that actuate for load 
shedding and D/G starts remain the same. Therefore, actuation 
criteria have been affected, but not safety functions, and the TVA 
evaluation has confirmed that the new design enhances the ability to 
maintain adequate voltage to support safety functions. Since safety 
functions have not changed and the new loss-of-power instrumentation 
design continues to support operability of safety-related equipment, 
no new or different accident is created.
    The applicable mode addition, TS 3.0.4 exclusion deletion, and 
response time measurement clarification, as well as the AFW 
operability clarifications, do not affect plant functions and will 
not create a new accident.
    3. Involve a significant reduction in a margin of safety.
    The proposed loss-of-power TS changes support design logic, 
setpoint, and time-delay requirements that have been verified by 
[[Page 27344]] TVA analysis to provide acceptable voltage levels for 
safety-related components. In determining the acceptability of these 
voltage levels, the minimum voltage for operation as well as 
detrimental component heating resulting from sustained degraded-
voltage conditions were considered. This design ensures that safety-
related loads will be available and operable for normal and accident 
plant conditions. The applicable mode addition, TS 3.0.4 exclusion 
deletion, response time measurement clarification, and AFW 
operability clarifications provide enhancements to TS requirements 
and do not affect plant functions. Therefore, no safety functions 
are reduced by these changes and there is no reduction in the margin 
of safety.

    The NRC 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: Chattanooga-Hamilton County 
Library, 1101 Broad Street, Chattanooga, Tennessee 37402.
    Attorney for licensee: General Counsel, Tennessee Valley Authority, 
400 West Summit Hill Drive, ET 11H, Knoxville, Tennessee 37902.
    NRC Project Director: Frederick J. Hebdon.

The Cleveland Electric Illuminating Company, Centerior Service Company, 
Duquesne Light Company, Ohio Edison Company, Pennsylvania Power 
Company, Toledo Edison Company, Docket No. 50-440, Perry Nuclear Power 
Plant, Unit No. 1, Lake County, Ohio

    Date of amendment request: April 28, 1995.
    Description of amendment request: The proposed amendment would 
extend for one more operating cycle an exception to Limiting Condition 
for Operation (LCO) 3.0.4 as it applies to the Technical Specification 
for the main steam isolation valve leakage control system. The existing 
LCO 3.0.4 exception was issued by Amendment 63 to the Operating 
License, and will expire upon completion of the fifty cycle of plant 
operation. The extension is proposed for the duration of the sixth 
cycle of operation to permit completion of activities necessary to 
implement the most appropriate permanent resolution for the issue of 
secondary containment bypass leakage through the main steam line 
drains.
    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 change does not involve a significant increase 
in the probability or consequences of an accident previously 
evaluated.
    This License Amendment application proposes an extension for one 
operating cycle of the exception to Limiting Condition for Operation 
for Operation (LCO) 3.0.4 as it applies to the Technical 
Specification for the MSIV [main steam isolation valve] Leakage 
Control system. This extension is proposed for the duration of the 
sixth cycle of PNPP (Perry Nuclear Power Plant) operation, to permit 
completion of activities necessary to implement the most appropriate 
permanent resolution for the issue of secondary containment bypass 
leakage through the Main Steam Line drains. During the sixth cycle, 
the drains will remain in their current configuration, which seals 
off the bypass leakage path. The sealed drain path results in a 
temporary inoperability of the Inboard MSIV Leakage control system 
(MSIV-LCS) subsystem when the plant is operated below 50% power, due 
to condensate build-up in the bottom of the steam lines between the 
MSIVs. The requested 3.0.4 exception is necessary to permit plant 
startups with this temporary inoperability, for the duration of the 
sixth operating cycle.
    The probability of occurrence of a previously evaluated accident 
is not affected by the proposed extension of the LCO 3.0.4 exception 
since no change to the plant or to the manner in which the plant is 
operated is involved. The existing plant configuration will be 
maintained for another operating cycle, and possible concerns 
resulting from that configuration have been analyzed. The extra 
weight of the water pooled between the MSIVs was analyzed with 
respect to piping supports and seismic considerations and was found 
to be acceptable, and any condensate that is carried past the 
outboard MSIVs will be drained to the condenser by drain connections 
downstream of the outboard MSIVs before it can reach the turbine. 
The temporary inoperability of the Inboard MSIV-LCS when below 50% 
power has no impact on accident initiation probability, since LCS 
does not serve to prevent accidents, but is only used in mitigating 
the consequences of Loss of Coolant Accidents that have already 
occurred.
    The consequences of an accident are not significantly increased 
in that the Outboard MSIV-LCS will be available to perform the MSIV-
LCS function by mitigating the consequences of a Loss of Coolant 
Accident (LOCA) during the temporary period in which the Inboard 
MSIV-LCS is unavailable. Any condensate that is carried past the 
outboard MSIVs will be drained to the condenser by drain connections 
downstream of the outboard MSIVs; therefore no impairment of the 
Outboard MSIV-LCS will result from condensed water.
    The Action statement for one inoperable LCS subsystem remains 
the same, and the limits plant operation to the previously 
established 30-day Allowable Outage Time. The Action required if 
both the subsystems of MSIV-LCS were to become inoperable also 
remains the same. The MSIV function of isolating the Main Stream 
Lines is also unaffected by the existing plant configuration, since 
MSIV performance will not be affected by the existence of 
accumulated water in the bottom of the steam lines between the MSIVs 
during the plant operation below 50% power. Therefore, if necessary, 
the Main Steam Lines will be isloated, and leakage past the MSIVs 
will be routed for filtration as in the design-basis radiological 
analyses, and the consequences of previously evaluated accidents 
will remain unaffected.
    2. The proposed change does not create the possibility of a new 
or different kind of accident from any accident previously 
evaluated.
    The proposed change to permit inoperability of the Inboard MSIV-
LCS during periods of startup and power ascension to 50% RTP (rated 
thermal power) and during shutdown below 50% RTP does not create the 
possibility of a new or different kind of accident from any 
previously evaluated. The Inboard MSIV-LCS is only credited during a 
Recirculation Line Break LOCA wherein Reactor Coolant System 
depressurization occurs. The temporary unavailability of the Inboard 
MSIV-LCS. the amendment to the Technical Specifications is an 
administrative change that does not involve any change to the 
current plant design or methods of operation. No new plant equipment 
failure modes or accident initiators are introduced by the extension 
of the LCO 3.0.4 exception.
    3. The proposed change does not involve a significant reduction 
in a margin of safety.
    The response to the Recirculation Line Break LOCA will not be 
significantly affected since the Outboard MSIV-LCS can be assumed to 
be available. Allowing entry into Operational Conditions 1, 2 and 3 
while utilizing the existing Action statement does not significantly 
reduce the margin of safety since the duration of time allowed for 
remaining in that Action statement is not increased. The proposed 
change will have no adverse impact on the reactor coolant system 
pressure boundary nor will any other system protective boundary or 
safety limit be affected.

    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: Perry Public Library, 3753 
Main Street, Perry, Ohio 44081.
    Attorney for licensee: Jay Silberg, Esq., Shaw, Pittman, Potts & 
Trowbridge, 2300 N Street, NW., Washington, DC 20037.
    NRC Project Director: Gail H. Marcus. [[Page 27345]] 

The Cleveland Electric Illuminating Company, Centerior Service Company, 
Duquesne Light Company, Ohio Edison Company, Pennsylvania Power 
Company, Toledo Edison Company, Docket No. 50-440, Perry Nuclear Power 
Plant, Unit No. 1, Lake County, Ohio

    Date of amendment request: May 1, 1995.
    Description of amendment request: The proposed amendment would 
eliminate selected response time testing requirements, and incorporate 
guidance provided by Generic Letter 93-08, ``Relocation of Technical 
Specification Tables of Instrument Response Time Limits.''
    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 changes do not involve a significant increase in the 
probability or consequences of an accident previously evaluated.
    For those proposed changes dealing with the elimination of 
selected response time test requirements, the purpose of the 
proposed Technical Specification change is to eliminate response 
time testing requirements for selected components in the Reactor 
Protection System, Isolation system, and Emergency Core Cooling 
System. The BWR Owners' Group has completed an evaluation which 
demonstrates that the response time testing is redundant to other 
Technical Specification required testing. These other tests, in 
conjunction with actions taken in response to NRC Bulletin 90-01, 
``Loss of Fill-Oil in Transmitters Manufactured by Rosemount,'' and 
Supplement 1, are sufficient to identify failure modes or 
degradations in instrument response time and ensure operation of the 
associated systems within acceptable limits. There are no known 
failure modes that can be detected by response time testing that 
cannot also be detected by the other required Technical 
Specification testing. This evaluation was documented in NEDO-32291, 
``System Analyses for Elimination of Selected Response Time Testing 
Requirements,'' January 1994, and the letter from T. Green to P. 
Loeser dated April 15, 1994 which were approved by an NRC Safety 
Evaluation dated December 28, 1994. The applicability of this 
evaluation to the Perry Nuclear Power Plant (PNPP) has been 
confirmed. In addition, PNPP will complete the additional actions 
identified in the NRC staff's Safety Evaluation of NEDO-32291.
    Because of the continued application of other existing Technical 
Specification required tests such as channel calibrations, channel 
checks, channel functional tests, and logic system functional tests, 
the response times of these systems will be maintained within the 
acceptance limits assumed in plant safety analysis and required for 
successful mitigation of an initiating event. The proposed Technical 
Specification changes do not affect the capability of the associated 
systems to perform their intended function within their required 
response time, nor do the proposed changes themselves affect the 
operation of any equipment. As a result the proposed changes dealing 
with elimination of selected response time tests do not involve a 
significant increase in the probability or the consequences of an 
accident previously evaluated.
    For those changes dealing with moving the surveillance 
requirement for ECCS RESPONSE TIME testing from the instrumentation 
section to the system section of the Technical Specifications, no 
change in testing requirements (other than the elimination of the 
instrument loops implemented as part of the NEDO-32291 changes) has 
been introduced. The relaxation in Applicability does not increase 
the probability or the consequences of an accident previously 
evaluated, since there are no design basis events during OPERATIONAL 
CONDITION 4 and 5 where ECCS systems are relied upon.
    For those changes dealing with relocation of the response time 
limits from Technical Specification Tables and into the Updated 
Safety Analysis Report (USAR), the proposed changes are 
administrative in nature in that the test requirements and time 
limits are still requirements, but the placement of the limits have 
been relocated from the Technical Specifications and into the USAR. 
Therefore these changes do not involve a significant increase in the 
probability or the consequences of an accident previously evaluated.
    2. The changes do not create the possibility of a new or 
different kind of accident from any previously evaluated.
    None of the proposed Technical Specification changes affect the 
capability of the associated systems to perform their intended 
function within the acceptance limits assumed in plant safety 
analyses and required for successful mitigation of an initiating 
event. The proposed changes also do not change the manner in which 
any plant equipment is operated. Therefore, the proposed changes do 
not create the possibility of a new or different kind of accident 
from any previously evaluated.
    3. The changes do not involve a significant reduction in the 
margin of safety.
    The current Technical Specification response times are based on 
the maximum allowable value assumed in the plant safety analyses. 
These analyses conservatively establish the margin of safety. As 
described above, the proposed Technical Specification changes do not 
affect the capability of the associated systems to perform their 
intended function within the allowed response time used as the basis 
for the plant safety analyses. Plant and system response to an 
initiating event will remain in compliance within the assumptions of 
the safety analyses, and therefore the margin of safety is not 
affected.
    Although not explicitly evaluated, the proposed Technical 
Specification changes dealing with response time testing elimination 
will provide an improvement to plant safety and operation by 
reducing the time safety systems are unavailable, reducing safety 
system actuation, reducing plant shutdown risk, limiting radiation 
exposure to plant personnel, and eliminating the diversion of key 
personnel to conduct unnecessary testing. Therefore, the proposed 
changes do not result in a significant reduction in a margin of 
safety, and may result in an overall increase in the margin of 
safety.
    The changes dealing with relocation of the time response limits 
from the Technical Specifications to the USAR is an administrative 
change that does not affect either the requirements to perform 
response time testing or the limits associated with the response 
time tests. Future changes to the limits will be controlled by 10 
CFR 50.59. Therefore, this portion of the change does not result in 
a significant decrease 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: Perry Public Library, 3753 
Main Street, Perry, Ohio 44081.
    Attorney for licensee: Jay Silberg, Esq., Shaw, Pittman, Potts & 
Trowbridge, 2300 N Street, NW., Washington, DC 20037.
    NRC Project Director: Gail H. Marcus.

Union Electric Company, Docket No. 50-483, Callaway Plant, Unit 1, 
Callaway County, Missouri

    Date of amendment request: April 26, 1995.
    Description of amendment request: the proposed amendment would 
revise Technical Specification (TS) Surveillance Requirements 3/4.7.6 
and associated Bases to reduce the upper limit on the control room 
filtration subsystem flow rate. It would also adopt ASTM D-3803-1989 as 
the laboratory testing standard for control room filtration and control 
building pressurization charcoal absorber.
    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:

    The proposed revision does not involve a significant hazards 
consideration because operation of Callaway Plant with this change 
would not:
    (1) Involve a significant increase in the probability or 
consequences of an accident previously evaluated.
    Overall protection system performance will remain within the 
bounds of the accident analysis documented in FSAR Chapter 15 * * * 
since no hardware changes are proposed. [[Page 27346]] 
    The Control Room Emergency Ventilation System (CREVS) will 
continue to function in a manner consistent with the above analysis 
assumptions and the plant design basis. There will be no degradation 
in the performance of or an increase in the number of challenges to 
equipment assumed to function during an accident situation.
    These Technical Specification revisions do not involve any 
hardware changes nor do they affect the probability of any event 
initiators. The change to the control room filtration flow rate is 
consistent with the original licensing basis and will ensure an 
average atmosphere residence time of greater than or equal to 0.25 
sec. There will be no change to ESF (engineered safety feature) 
actuation setpoints or accident mitigation capabilities. The 
laboratory testing will demonstrate the required absorber 
performance after a design basis LOCA (loss-of-coolant accident).
    The control room dose analyses assume a total flow rate through 
the control room filtration units that is less than the proposed 
upper limit. As such, there will be no changes required to the 
control room dose analyses.
    Based on the above, these Technical Specification changes will 
not increase the probability or consequences of an accident or 
malfunction.
    (2) Create the possibility of a new or different kind of 
accident from any accident previously evaluated.
    As discussed above, there are no hardware changes associated 
with these Technical Specification revisions nor are there any 
changes in the method by which any safety-related plant system 
performs its safety function.
    Revisions to the Surveillance Requirements for the CREVS will 
ensure that the control room does analysis assumptions made in 
support of OL (operating license) Amendment No. 96 are valid. 
Changes to the control room filtration unit flow rate are more 
limiting than that currently specified and have already been 
implemented by resetting the open limit switches on the respective 
units' outlet dampers. This flow rate is consistent with the design 
basis for the filtration units as originally licensed.
    No new accident scenarios, transient precursors, failure 
mechanisms, or limiting single failures are introduced as a result 
of these changes. There will be no adverse effect or challenges 
imposed on any safety-related system as a result of these changes. 
Therefore, the possibility of a new or different kind of accident is 
not created.
    (3) Involve a significant reduction in a margin of safety.
    There will be no margin reduction since these changes are in the 
conservative direction and have already been approved by NRC via the 
approval of OL Amendment No. 96. The reduced upper bound flow rate 
for the control room filtration units is consistent with their 
design basis and will maintain an average atmosphere residence time 
greater than or equal to 0.25 sec under both clean and dirty filter 
conditions. The new charcoal absorber sample laboratory testing 
protocol is more stringent than the current testing practice and 
more accurately demonstrates the required performance after a design 
basis LOCA.
    There will be no effect on the manner in which safety limits or 
limiting safety system settings are determined nor will there be any 
effect on those plant systems, necessary to assure the 
accomplishment of protection functions. There will no impact on the 
overpower limit, DNBR (departure from nucleate boiling ratio) 
limits, F<INF>Q, F[delta]H, LOCA PCT (peak cladding temperature), 
peak local power density, or any other margin of safety. These 
changes will ensure that the criteria of GDC 19 are met.

    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: Callaway County Public 
Library, 710 Court Street, Fulton, Missouri 65251.
    Attorney for licensee: Gerald Charnoff, Esq., Shaw, Pittman, Potts 
& Trowbridge, 2300 N Street, NW, Washington, DC 20037.
    NRC Project Director: Gail H. Marcus.

Wisconsin Electric Power Company, Docket Nos. 50-266 and 50-301, Point 
Beach Nuclear Power Plant, Unit Nos. 1 and 2, Town of Two Creeks. 
Manitowoc County, Wisconsin

    Date of amendment request: April 17, 1995.
    Description of amendment request: The proposed amendment would 
modify Technical Specification (TS) Section 15.6.2, ``Organization,'' 
and TS Section 15.6.3, ``Facility Staff Qualifications.'' The training 
requirements for the Operations Manager and other staff would be 
changed to provide staffing flexibility.
    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:

    The proposed change does not involve a significant increase in 
the probability or consequences of an accident previously evaluated; 
create the possibility of a new or different kind of accident from 
any previously evaluated; or create the possibility of a new or 
different kind of accident from any previously evaluated.
    1. The proposed change affects only an administrative control, 
which was based on industry guidance in ANSI N18.1-1971, that 
recommended the Operations Manager hold an SRO (senior reactor 
operator) license. This administrative control is being updated to 
meet the current guidance in ANSI/ANS 3.1-1987.
    2. The proposed qualification requirements for the Operations 
Manager ensures the individual filling the position meets knowledge 
levels equivalent to the present requirements. It also ensures that 
individuals responsible for directing the activities of licensed 
operators continue to hold SRO licenses as required by 10 CFR 
50.54(l).
    3. Since the proposed specifications ensure regulatory 
requirements are met and ensures knowledge levels equivalent to 
existing license requirements for operations management, the 
proposed changes are considered administrative. The design of plant 
systems and equipment is not being altered. Plant operations will 
continue to be directed and performed by qualified personnel. 
Therefore, the probability or consequences of accidents previously 
evaluated are not affected, a new or different type of accident is 
not created, nor is a margin of safety reduced.

    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: Joseph P. Mann Library, 1516 
Sixteenth Street, Two Rivers, Wisconsin 54241.
    Attorney for licensee: Gerald Charnoff, Esq., Shaw, Pittman, Potts, 
and Trowbridge, 2300 N Street, NW., Washington, DC 20037.
    NRC Project Director: Gail H. Marcus.

Wisconsin Electric Power Company, Docket Nos. 50-266 and 50-301, Point 
Beach Nuclear Power Plant, Unit Nos. 1 and 2, Town of Two Creeks, 
Manitowoc County, Wisconsin

    Date of amendment request: April 27, 1995.
    Description of amendment request: The proposed amendment would 
modify Technical Specification (TS) Table 15.3.5-1, ``Engineered Safety 
Features Initiation Instrument Setting Limits,'' and TS Table 15.35-3, 
``Engineered Safety Features.'' Setting limits would be modified and 
references would be changed. The bases for TS Section 15.3.5, 
``Instrumentation System,'' would also be changed to be consistent with 
the TS 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. Operation of this facility under the proposed Technical 
Specifications will not create a significant increase in the 
probability or consequences of an accident previously evaluated.
    The probabilities of accidents previously evaluated are based on 
the probability of initiating events for these accidents. 
[[Page 27347]] Initiating events for accidents previously evaluated 
for Point Beach include: control rod withdrawal and drops, CVCS 
(chemical and volume control system) malfunction (Boron Dilution), 
startup of an inactive reactor coolant loop, reduction in feedwater 
enthalpy, excessive load increase, losses of reactor coolant flow, 
loss of external electrical load, loss of normal feedwater, loss of 
all AC power to the auxiliaries, turbine overspeed, fuel handling 
accidents, accidental releases of water liquid or gas, steam 
generator tube rupture, steam pipe rupture, control rod ejection, 
and primary coolant system ruptures.
    This license amendment request proposes to correct some minor 
errors, include appropriate operability requirements for the 
modification to include the safety injection signal in the time 
delay for the 4.16KV degraded voltage protection logic, slightly 
lower the degraded voltage setting limit, change the format of the 
4.16 KV degraded voltage and loss of voltage setting limits, and 
change the time delays associated with the 4.16 KV degraded voltage, 
4.16 KV loss of voltage and 480 V loss of voltage protection 
functions.
    These proposed changes do not cause an increase in the 
probabilities of any accidents previously evaluated because these 
changes will not cause an increase in the probability of any 
initiating events for accidents previously evaluated. In particular, 
these proposed changes affect time delay and format of the setting 
limits associated with the 4.16 KV degraded voltage, 4.16 KV loss of 
voltage, and 480 V loss of voltage protection functions. These are 
protection functions and do not cause accidents.
    The consequences of the accidents previously evaluated in the 
PBNP FSAR (Final Safety Analysis Report) are determined by the 
results of analyses that are based on initial conditions of the 
plant, the type of accident, transient response of the plant, and 
the operation and failure of equipment and systems. The changes 
proposed in this license amendment request provide appropriate 
limiting conditions for operation, action settlements, allowable 
outage times, setting limits, and time delays for the Point Beach 
Nuclear Plant Technical Specifications for the 4.16 KV degraded 
voltage, 4.16 KV loss of voltage, and 480 V loss voltage protection 
functions.
    The proposed changes affect functions that are required to 
ensure the proper operation of engineered features equipment. The 
proposed changes do not increase the probability of failure of this 
equipment or its ability to operate as required for the accidents 
previously evaluated in the PBNP FSAR.
    The modifications to reduce the time delay limit associated with 
the 4.16 KV degraded voltage protection function when the degraded 
voltage condition is coincident with a safety injection signal, have 
been designed and installed in accordance with the requirements for 
PBNP. The probability of occurrence of degraded voltage conditions 
at PBNP has not been increased. The modifications and proposed 
Technical Specifications will ensure the proper operation of ESF 
(engineered safety feature) equipment. These changes do not increase 
the possibility of failure of this equipment.
    Therefore, this proposed license amendment does not affect the 
consequences of any accident previously evaluated in the Point Beach 
Nuclear Plant FSAR, because the factors that are used to determine 
the consequences of accidents are not being changed.
    2. Operation of this facility under the proposed Technical 
Specifications change will not create the possibility of a new or 
different kind of accident from any accident previously evaluated.
    New or different kinds of accidents can only be created by new 
or different accident initiators or sequences. New and different 
types of accidents (different from those that were originally 
analyzed for Point Beach) have been evaluated and incorporated into 
the licensing basis for Point Beach Nuclear Plant. Examples of 
different accidents that have been incorporated into the Point Beach 
Licensing basis include anticipated transients without scram and 
station blackout.
    The changes proposed by this license amendment request do not 
create any new or different accident initiators or sequences because 
these changes to the 4.16 KV degraded voltage, 4.16 KV loss of 
voltage, and 480 V loss of voltage protection functions will not 
cause failures of equipment or accident sequences different than the 
accidents previously evaluated. Therefore, these modifications and 
proposed Technical Specification changes do not create the 
possibility of an accident of a different type than any previously 
evaluated in the Point Beach FSAR.
    3. Operation of this facility under the proposed Technical 
Specifications change will not create a significant reduction in a 
margin of safety.
    The margins of safety for Point Beach are based on the design 
and operation of the reactor and containment and the safety systems 
that provide their protection.
    The changes proposed by this license amendment request provide 
the appropriate setting limits and time delays for the 4.16 KV 
degraded voltage, 4.16 KV loss of voltage, and 480 V loss of voltage 
protection functions. This ensures that the safety systems that 
protect the reactor and containment will operate as required. The 
design and operation of the reactor and containment are not affected 
by these proposed changes. Therefore, the margins of safety for 
Point Beach are not being reduced because the design and operation 
of the reactor and containment are not being changed and the safety 
systems that provide their protection that are being changed are 
being modified in accordance with the applicable design and 
installation requirements for Point Beach Nuclear Plant.

    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: Joseph P. Mann Library, 1516 
Sixteenth Street, Two Rivers, Wisconsin 54241.
    Attorney for licensee: Gerald Charnoff, Esq., Shaw, Pittman, Potts, 
and Trowbridge, 2300 N Street, NW., Washington, DC 20037.
    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 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.

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 21, 1995.
    Description of amendment request: The proposed amendment would 
revise the Technical Specifications (TS) 3.1.2.4, ``Charging Pumps-
Operating,'' by adding a note that indicates that the provisions of TS 
3.0.4 and 4.0.4 are not applicable for entry into MODE 4 from MODE 5.
    Date of publication individual notice in Federal Register: May 2, 
1995 (60 FR 21558).
    Expiration date of individual notice: June 1, 1995.
    Local Public Document Room location: Learning Resource Center, 
Three Rivers Community-Technical College, Thames Valley Campus, 574 New 
London Turnpike, Norwich, Connecticut 06360.

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 
[[Page 27348]] 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 Ch. 1, which are set forth in the license amendment.
    Notice of Consideration of Issuance of Amendment to Facility 
Operating License, Proposed No Significant 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.

Carolina Power & Light Company, et al., Docket No. 50-400, Shearon 
Harris Nuclear Power Plant, Unit 1, Wake and Chatham Counties, North 
Carolina

    Date of application for amendment: August 19, 1994, as supplemented 
November 3, 1994.
    Brief description of amendment: The amendment requests a line-item 
improvement to the Radiological Effluent Technical Specifications 
pursuant to the guidance of Generic Letter 89-01 and incorporates the 
requirements of revised 10 CFR part 20 and 10 CFR 50.36a.
    Date of issuance: May 1, 1995.
    Effective date: May 1, 1994.
    Amendment No.: 58.
    Facility Operating License No. NPF-63: Amendment revises the 
Technical Specifications.
    Date of initial notice in Federal Register: October 12, 1994 (60 FR 
51617) The Commission's related evaluation of the amendment, and NRC's 
response to the public comments received, are contained in a Safety 
Evaluation dated May 1, 1995.
    No significant hazards consideration comments received: Yes.
    Local Public Document Room location: Cameron Village Regional 
Library, 1930 Clark Avenue, Raleigh, North Carolina 27605.

Duke Power Company, Docket Nos. 50-269, 50-270, and 50-287, Oconee 
Nuclear Station, Units 1, 2, and 3, Oconee County, South Carolina

    Date of application of amendments: November 22, 1994, as 
supplemented by letters dated January 30, March 2, March 13, and May 2, 
1995.
    Brief description of amendments: The amendments revise Technical 
Specification 3.8 to establish restricted loading patterns and 
associated burnup criteria for placing fuel in the Oconee spent fule 
pools. In addition, the Design Features sections associated with the 
reactor and fuel storage are also revised.
    Date of issuance: May 3, 1995.
    Effective date: As of the date of issuance and shall be implemented 
within 30 days from the date of issuance.
    Amendment Nos.: 209, 209, and 206.
    Facility Operating License Nos. DPR-38, DPR-47, and DPR-55: The 
amendments revised the Technical Specifications.
    Date of initial notice in Federal Register: February 15, 1995 (60 
FR 8746); Re-Noticed March 29, 1995 (60 FR 16185).
    The May 2, 1995, letter did not change the scope of the November 
22, 1994, 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 May 3, 1995.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: Oconee County Library, 501 
West South Broad Street, Walhalla, South Carolina 29691.

Duquesne Light Company, et al., Docket No. 50-412, Beaver Valley Power 
Station, Unit 2, Shippingport, Pennsylvania

    Date of application for amendment: April 10, 1995, as supplemented 
April 12, 1995, and April 20, 1995.
    Brief description of amendment: This amendment revises Technical 
Specification 4.6.2.2.d to delete the reference to the specific test 
acceptance criteria for the Containment Recirculation Spray Pumps and 
replace the specific test acceptance criteria with reference to the 
developed head required by the plant's safety analysis. In addition, 
the 18-month test frequency would be replaced with the test frequency 
requirements specified in the IST Program. The current footnote (1) 
pertaining to the performance of recirculation spray pump 2RSS*P21A 
would be deleted.
    Date of issuance: May 3, 1995.
    Effective date: May 3, 1995.
    Amendment No.: 68.
    Facility Operating License No. NPF-73: Amendment revised the 
Technical Specifications.
    Public comments requested as to proposed no significant hazards 
consideration: Yes (60 FR 19417, April 18, 1995) That notice provided 
an opportunity to submit comments on the Commission's proposed no 
significant hazards consideration determination. No comments have been 
received. The notice also provided for an opportunity to request a 
hearing by May 18, 1995, but indicated that if the Commission makes a 
final no significant hazards consideration 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 no significant 
hazards consideration are contained in a Safety Evaluation dated May 3, 
1995.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: B. F. Jones Memorial Library, 
663 Franklin Avenue, Aliquippa, Pennsylvania 15001.

Entergy Operations, Inc., Docket Nos. 50-313 and 50-368, Arkansas 
Nuclear One, Unit Nos. 1 and 2, Pope County, Arkansas

    Date of amendment request: August 30, 1994 as supplemented January 
19, 1995.
    Brief description of amendments: The amendments changed 
requirements related to the site perimeter security system.
    Date of issuance: April 28, 1995.
    Effective date: April 28, 1995.
    Amendment Nos.: Unit 1--Amendment No. 180; Unit 2--Amendment No. 
161
    Facility Operating License Nos. DPR-51 and NPF-6: Amendments 
revised the licenses.
    Date of initial notice in Federal Register: April 12, 1995 (60 FR 
18625).
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated April 28, 1995.
    No significant hazards consideration comments received: No. 
[[Page 27349]] 
    Local Public Document Room location: Tomlinson Library, Arkansas 
Tech University, Russellville, AR 72801.

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

    Date of amendment request: December 14, 1993, as supplemented by 
letter dated March 3, 1995.
    Brief description of amendment: The amendment changed the Appendix 
A Technical Specifications by removing the reactor vessel material 
specimen withdrawal schedule and by updating the reactor coolant system 
pressure-temperature (P-T) curves.
    Date of issuance: May 8, 1995.
    Effective date: May 8, 1995.
    Amendment No.: 106.
    Facility Operating License No. NPF-38.: Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: January 19, 1994 (59 FR 
2867).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated May 8, 1995.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: University of New Orleans 
Library, Louisiana Collection, Lakefront, New Orleans, Louisiana 70122.

Florida Power and Light Company, Docket Nos. 50-250 and 50-251, Turkey 
Point Plant Units 3 and 4, Dade County, Florida

    Date of application for amendments: October 20, 1994.
    Brief description of amendments: These amendments change the 
definition of ``core alteration'' to exclude the movement of items not 
associated with reactivity. The second change involves allowing the 
personnel airlock (PAL) doors to remain open during fuel movement and 
core alterations under certain conditions.
    Date of issuance: May 11, 1995.
    Effective date: May 11, 1995.
    Amendment Nos.: 173 and 167.
    Facility Operating License No. DPR-31 and DPR-41: Amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: November 9, 1994 (59 FR 
55869).
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated May 11, 1995.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: Florida International 
University, University Park, Miami, Florida 33199.

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

    Date of application for amendment: February 28, 1995.
    Brief description of amendment: The amendment revises Technical 
Specification (TS) Section 6.5.1.12 to delete the requirement to render 
determinations in writing with regard to whether or not activities 
listed in TS Sections 6.5.1.2 and 6.5.1.5 constitute an unreviewed 
safety question. These activities are changes to Appendix A Technical 
Specifications (6.5.1.2) and investigations of all violations of the 
TSs (6.5.1.5). This change is consistent with NUREG-1433 Standard 
Technical Specifications General Electric Plants, BWR/4 Revision 0, 
dated September 28, 1992.
    Date of issuance: May 1, 1995.
    Effective date: May 1, 1995.
    Amendment No.: 180.
    Facility Operating License No. DPR-16.: Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: March 29, 1995 (60 FR 
16188).
    The Commission's related evaluation of this amendment is contained 
in a Safety Evaluation dated May 1, 1995.
    No significant hazards consideration comments received: Yes.
    By letter dated April 5, 1995, Mr. Kent W. Tosch, of the State of 
New Jersey Department of Environmental Protection commented that they 
concur with GPU Nuclear's rationale that these unreviewed safety 
question reviews serve no value since these activities specifically 
require NRC review and approval. The State official had no other 
comments.
    Local Public Document Room location: Ocean County Library, 
Reference Department, 101 Washington Street, Toms River, NJ 08753.

Houston Lighting & Power Company, City Public Service Board of San 
Antonio, Central Power and Light Company, City of Austin, Texas, Docket 
Dos. 50-498 and 50-499, South Texas Projects, Units 1 and 2, Matagorda 
County, Texas

    Date of amendment request: February 15, 1995.
    Brief description of amendment: The amendment modified Technical 
Specification 4.6.2.3.a.2 (and associated Bases) to reflect the reactor 
containment fan cooler flow rate assumed in the accident analysis and 
to specify that this flow is provided by the component cooling water 
system.
    Date of issuance: May 2, 1995.
    Effective date: May 2, 1995, to be implemented within 30 days.
    Amendment Nos.: Unit 1--Amendment No. 74; Unit 2--Amendment No. 63.
    Facility Operating License Nos. NPF-76 AND NPF-80. The amendment 
revised the Technical Specifications.
    Date of initial notice in Federal Register: March 29, 1995 (60 FR 
16189) The Commission's related evaluation of the amendments is 
contained in a Safety Evaluation dated May 2, 1995.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: Wharton County Junior College, 
J.M. Hodge Learning Center, 911 Boling Highway, Wharton, TX 77488.

Illinois Power Company and Soyland Power Cooperative, Inc., Docket No. 
50-461, Clinton Power Station, Unit No. 1. DeWitt County, Illinois

    Date of application for amendment: February 10, 1995.
    Brief description of amendment: The amendment changes Technical 
Specification 3.3.2.1, ``Control Rod Block Instrumentation,'' to revise 
two surveillance requirements and their associated notes for the Rod 
Withdrawal Limiter mode of the Rod Pattern Control System. The changes 
are consistent with the Clinton Power Station Technical Specifications 
prior to implementation of the improved Technical Specifications 
(Amendment No. 95) and eliminates the potential for unnecessary power 
reductions.
    Date of issuance: May 2, 1995.
    Effective date: May 2, 1995.
    Amendment No.: 100.
    Facility Operating License No. NPF-62. The amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: March 29, 1995. (60 FR 
16190)
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated May 2, 1995.
    No significant hazard consideration comments received: No.
    Local Public Document Room location: The Vespasian Warner Public 
Library, 120 West Johnson Street, Clinton, Illinois 61727.

Niagara Mohawk Power Corporation, Docket No. 50-410, Nine Mile Point 
Nuclear Station, Unit 2, Oswego County, New York

    Date of application for amendment: July 22, 1993, as supplemented 
February 4, August 23, September 16, October 6, and December 2, 1994, 
and January 3, January 9, March 8, and April 10, 1995. [[Page 27350]] 
    Brief description of amendment: The amendment modified Facility 
Operating License No. NPF-69 and the NMP-2 TSs to authorize an increase 
in the maximum power level of NMP-2 from 3323 megawatts thermal 
(MW<INF>t) to 3467 MW<INF>t. The amendment also approves changes to the 
TSs to implement uprated power operation.
    Date of issuance: April 28, 1995.
    Effective date: As of the date of issuance to be implemented prior 
to restart from refueling outage number 4.
    Amendment No.: 66.
    Facility Operating License No. NPF-69: Amendment revises the 
Technical Specifications and modifies Facility Operating License No. 
NPF-69.
    Date of initial notice in Federal Register: March 16, 1994 (59 FR 
12360). The letters dated February 4, August 23, September 16, October 
6, and December 2, 1994, and January 3, January 9, March 8, and April 
10, 1995, provided clarifying information that did not change the 
initial proposed no significant hazards consideration determination.
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated April 28, 1995.
    No significant hazards consideration comments received: No
    Local Public Document Room location: Reference and Documents 
Department, Penfield Library, State University of New York, Oswego, New 
York 13126.
Northeast Nuclear Energy Company, et al., Docket No. 50-336, Millstone 
Nuclear Power Station, Unit No. 2, New London County, Connecticut

    Date of application for amendment: October 18, 1994, a supplemented 
February 21, 1995.
    Brief description of amendment: The amendment changes Surveillance 
Requirement 4.6.1.2.a (Overall Integrated Containment Leakage Rate 
Tests) by revising the surveillance interval for Type A tests from 40 
plus or minus 10 months to approximately equal intervals during each 
10-year inservice period. The amendment also removes a note that 
expired upon completion of Cycle II refueling outage.
    Date of issuance: May 3, 1995.
    Effective date: As of the date of issuance to be implemented within 
30 days.
    Amendment No.: 187.
    Facility Operating License No. DPR-65. Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: March 29, 1995 (60 FR 
16191).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated may 3, 1995.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: Learning Resource Center, 
Three Rivers Community-Technical College, Thames Valley Campus, 574 New 
London turnpike, Norwich, CT 06360.

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: December 23, 1994.
    Brief description of amendment: The amendment changes the 
acceptance criteria for the peak transient generator voltage from 4784 
volts to 5000 volts during full load rejection tests of the diesel 
generator (DG), and also deletes the 10-year surveillance requirement 
to perform a 110% pressure test of the DG fuel oil system.
    Date of issuance: May 1, 1995.
    Effective date: As of the date of issuance to be implemented within 
30 days.
    Amendment No.: 110.
    Facility Operating License No. NPF-49. Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: February 15, 1995 (60 
FR 8751).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated May 1, 1995.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: Learning Resources Center, 
Three Rivers Community-Technical College, Thames Valley Campus, 574 New 
London Turnpike, Norwich, CT 06360.

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: September 28, 1994.
    Brief description of amendment: The amendment revises Surveillance 
Requirement 4.6.1.2.a of the Technical Specification to eliminate the 
requirement to perform Type A tests on an interval of 40 plus or minus 
10 months while reiterating the Appendix J requirement that the Type A 
tests be performed three times, at approximately equal intervals, 
during each 10 year service period. In addition, a footnote is added 
which states that the third Type A test will be performed during the 
sixth refueling outage. This reflects an exemption to Appendix J which 
separates the third Type A test from the 10 year inservice inspection.
    Date of issuance: May 8, 1995.
    Effective date: As of the date of issuance to be implemented within 
30 days.
    Amendment No.: 111.
    Facility Operating License No. NPF-49. Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: November 23, 1994 (59 
FR 60384)
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated May 8, 1995.
    No significant hazards consideration comments received: NO.
    Local Public Document Room location: Learning Resources Center, 
Three Rivers Community-Technical College, Thames Valley Campus, 574 New 
London Turnpike, Norwich, CT 06360.

Public Service Electric & Gas Company, Docket Nos. 50-272 and 50-311, 
Salem Nuclear Generating Station, Unit Nos. 1 and 2, Salem County, New 
Jersey

    Date of application for amendment: August 19, 1994, as supplemented 
March 15, 1995.
    Brief description of amendment: The amendments add a new action 
statement to Technical Specification 3.1.3.2.1., ``Position Indication 
Systems--Operating''.
    Date of issuance: May 3, 1995.
    Effective date: May 3, 1995.
    Amendment No.: 166 and 148.
    Facility Operating License Nos. DPR-70 and DPR-75. The amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: October 12, 1994 (59 FR 
51626) The March 15, 1995 supplement 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 may 3, 1995.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: Salem Free Public Library, 112 
West Broadway, Salem, New Jersey 08079.

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: March 19, 1993; superseded May 
16, 1994; superseded February 10, 1995; supplemented February 17, 1995 
(TS 93-04).
    Brief description of amendment: The amendments clarify the Limiting 
[[Page 27351]] Conditions for Operation applicable to the dual function 
of the containment vacuum relief isolation lines by specifying the 
actions that would be required should one or more of the vacuum relief 
isolation lines by specifying the actions that would be required should 
one or more of the vacuum relief lines be incapable of performing the 
containment isolation function or incapable of performing the vacuum 
relief function.
    Date of issuance: April 28, 1995.
    Effective date: April 28, 1995.
    Amendment No.: 197 and 188.
    Facility Operating License Nos. DPR-77 and DPR-79: Amendments 
revise the technical specifications.
    Date of initial notice in Federal Register: May 12, 1994 (58 FR 
28060); renoticed June 22, 1994 (59 FR 32237), and March 29, 1995 (60 
FR 16202).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated April 28, 1995.
    No significant hazards consideration comments received: None.
    Local Public Document Room location: Chattanooga-Hamilton County 
Library, 1101 Broad Street, Chattanooga, Tennessee 37402.

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

    Date of application for amendment: November 15, 1994; superseded 
March 7, 1995 (TS 94-12).
    Brief description of amendments: The amendments remove the 
frequencies specified in the Technical Specifications for performing 
audits and delete the requirement to perform the Radiological Emergency 
Plan, Physical Security Plan, and Safeguard Contingency Plan reviews.
    Date of issuance: May 10, 1995.
    Effective date: May 10, 1995.
    Amendment No.: 198 and 189.
    Facility Operating License Nos. DPR-77 and DPR-79: Amendments 
revise the technical specifications.
    Date of initial notice in Federal Register: December 21, 1994 (59 
FR 65823); renoticed March 29, 1995 (60 FR 16203)
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated May 10, 1995.
    No significant hazards consideration comments received: None.
    Local Public Document Room location: Chattanooga-Hamilton County 
Library, 1101 Broad Street, Chattanooga, Tennessee 37402.

Toledo Edison Company, Centerior Service Company, and The Cleveland 
Electric Illuminating Company, Docket No. 50-346, Davis-Besse Nuclear 
Power Station, Unit No. 1, Ottawa County, Ohio

    Date of application for amendment: January 30, 1995.
    Brief description of amendment: This amendment revises Technical 
Specification (TS) 4.6.1.2.a, ``Containment Systems, Containment 
Leakage, Surveillance Requirements (SR)'' and Bases 3/4.6, 
``Containment Systems,'' to state that Type A tests for overall 
integrated containment leakage rate testing shall be conducted in 
accordance with the requirements specified in appendix J of 10 CFR part 
50, as modified by NRC-approved exemptions. Additionally, TS SR 
4.6.1.2.a.
    Date of issuance: May 3, 1995.
    Effective date: May 3, 1995.
    Amendment No.: 198.
    Facility Operating License No. NPF-3. Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: March 15, 1995 (60 FR 
14028).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated May 3, 1995.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: University of Toledo Library, 
Documents Department, 2801 Bancroft Avenue, Toledo, Ohio 43606.

Virginia Electric and Power Company, et al., Docket Nos. 50-338 and 50-
339, North Anna Power Station, Units No. 1 and No. 2, Louisa County, 
Virginia

    Date of application for amendments: July 8, 1993, as supplemented 
by letters dated July 12, 1994, and March 7, 1995.
    Brief description of amendments: The amendments revise the NA-1&2 
Technical Specifications by deleting the requirements to periodically 
review certain administrative and technical procedures.
    Date of issuance: May 1, 1995.
    Effective date: May 1, 1995.
    Amendment Nos.: 190 and 171.
    Facility Operating License Nos. NPF-4 and NPF-7: Amendments revised 
the Technical Specifications.
    Date of initial notice in Federal Register: August 4, 1993 (58 FR 
41518).
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated May 1, 1995.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: The Alderman Library, Special 
Collections Department, University of Virginia, Charlottesville, 
Virginia 22903-2498.

Virginia Electric and Power Company, et al., Docket Nos. 50-338 and 50-
339; North Anna Power Station, Units No. 1 and No. 2, Louisa County, 
Virginia

    Date of application for amendments: December 27, 1993, as 
supplemented September 6, 1994, and March 7, 1995.
    Brief description of amendments: The amendments revise the NA-1&2 
Technical Specifications regarding the review responsibilities of the 
Station Nuclear Safety and Operating Committee and the Management 
Safety Review Committee.
    Date of issuance: May 2, 1995.
    Effective date: May 2, 1995.
    Amendment Nos.: 191 and 172.
    Facility Operating License Nos. NPF-4 and NPF-7: Amendments revised 
the Technical Specifications.
    Date of initial notice in Federal Register: February 16, 1994 (59 
FR 7700).
    The September 6, 1994, and March 7, 1995 submittals provided 
additional information only, and did not change the staff's initial 
proposed determination of no significant hazards consideration.
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated May 2, 1995.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: The Alderman Library, Special 
Collections Department, University of Virginia, Charlottesville, 
Virginia 22903-2498.
Virginia Electric and Power Company, Docket Nos. 50-280 and 50-281, 
Surry Power Station, Unit Nos. 1 and 2, Surry County, Virginia

    Date of application for amendments: June 28, 1991.
    Brief description of amendments: These amendments incorporate 
operability and surveillance requirements for power-operated relief 
valves to conform with Generic Letter 90-06.
    Date of issuance: May 2, 1995.
    Effective date: May 2, 1995.
    Amendment Nos.: 198 and 198.
    Facility Operating License Nos. DPR-32 and DPR-37: Amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: October 2, 1991 (56 FR 
49929).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated May 2, 1995.
    No significant hazards consideration comments received: 
No. [[Page 27352]] 
    Local Public Document Room location: Swem Library, College of 
William and Mary, Williamsburg, Virginia 23185.

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 Ch. 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 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 23, 1995, 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 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 
[[Page 27353]] 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 lest 
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, 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 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, 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. 50-295 and 50-304, Zion 
Nuclear Power Station, Units 1 and 2

    Date of application for amendments: April 24, 1995.
    Brief description of amendments: the amendments change the 
Technical Specifications by modifying the surveillance testing 
periodicity requirements of the automatic actuation logic of engineered 
safeguards equipment.
    Date of issuance: May 5, 1995.
    Effective date: May 5, 1995.
    Amendment Nos.: 162 and 150.
    Facility Operating Licenses Nos. DPR-39 and DPR-48. The amendments 
revised the Technical Specifications.
    Public comments requested as to proposed no significant hazards 
consideration: No.
    The Commission's related evaluation of the amendments, finding of 
emergency circumstances, and final determination of no significant 
hazards consideration are contained in a Safety Evaluation dated May 5, 
1995.
    Attorney for licensee: Michael I. Miller, Esquire; Sidley and 
Austin, One First National Plaza, Chicago, Illinois 60690.
    Local Public Document Room location: Waukegan Public Library, 128 
N. County Street, Waukegan, Illinois 60085.
    NRC Project Director: Robert A. Capra.

Baltimore Gas and Electric Company, Docket No. 50-317, Calvert Cliffs 
Nuclear Power Pant, Unit No. 1, Calvert County, Maryland

    Date of application for amendment: April 28, 1995.
    Brief description of amendment: The amendment revises the control 
room emergency ventilation system TS 3.7.6.1, Limiting Condition For 
Operation. The revision extends the one-time increase in the allowed 
outage time for loss of emergency power only, from the 30 days 
previously approved, to 45 days. This extension is necessary to allow 
time to repair the Number 21 emergency diesel generator which failed 
its operability tests subsequent to modifications which have been 
recently completed.
    Date of issuance: May 2, 1995.
    Effective date: As of the date of issuance to be implemented upon 
receipt.
    Amendment No.: 205.
    Facility Operating License No. DPR-53: Amendment revised the 
Technical Specifications.
    Public comments requested as to proposed no significant hazards 
consideration: No.
    The Commission's related evaluation of the amendment, consultation 
with the State, and final determination of no significant hazards 
consideration are continued in a Safety Evaluation dated May 2, 1995.
    Local Public Document Room location: Calvert County Library, Prince 
Frederick, Maryland 20678.
    Attorney for licensee: Jay E. Silbert, Esquire, Shaw, Pittman, 
Potts and Trowbridge, 2300 N. Street, NW., Washington, DC 20037.
    NRC Project Director: Ledyard B. Marsh.

    Dated at Rockville, MD, this 17th day of May, 1995.

    For the Nuclear Regulatory Commission,
Elinor G. Adensam,
Acting Director, Division of Reactor Projects--III/IV, Office of 
Nuclear Reactor Regulation.
[FR Doc. 95-12538 Filed 5-22-95; 8:45 am]
BILLING CODE 7590-01-M