[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 Keff 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 Keff 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 Keff (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 Keff [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, FQ, 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
(MWt) to 3467 MWt. 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