Biweekly Notice
Federal RegisterJul 19, 1995
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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 June 23, 1995, through July 7, 1995. The
last biweekly notice was published on July 5, 1995 (60 FR 35058).
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.
[[Page 37085]]
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 August 18, 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 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
[[Page 37086]]
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.
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: September 10, 1993, as
supplemented June 16, 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 September 10, 1993, and June 16, 1995, applications proposed to
upgrade only Section 3/4.8 (Plant Systems) 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 3/4.8 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 Plant 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
3/4.8 of the proposed Technical Specifications, performing their
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 3/4.8 is based on STS guidelines or
later operating BWR plants' NRC accepted changes. The proposed
amendment has been reviewed for acceptability at the Dresden or 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. 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. 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 Plant Systems are not
assumed in any 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
Plant 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
[[Page 37087]]
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 3/4.8
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 or 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 or Quad Cities and maintain necessary
levels of system or component 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 Plant
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
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 requests: September 17, 1993, as
supplemented June 30, 1995
Description of amendment requests: 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 operation
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 September 17, 1993, and June 30, 1995, applications proposed to
upgrade only Section 3/4.6 (Primary System Boundary) 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
amendments for Dresden and Quad Cities Station's Technical
Specification Section 3/4.6 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 Primary System Boundary
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 3/4.6 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
[[Page 37088]]
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 3/4.6 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. 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. 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 Primary System Boundary
are not assumed in any 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 Primary System Boundary 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 3/4.6
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 or component 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 Primary
System Boundary 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 requests involve no significant hazards consideration.
Local Public Document Room location: for Dresden, Morris Public
Library, 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 60690
NRC Project Director: Robert A. Capra
Entergy Operations, Inc., Docket No. 50-368, Arkansas Nuclear One,
Unit No. 2, Pope County, Arkansas
Date of amendment request: March 17, 1995
Description of amendment request: The proposed amendment transfers
requirements for a cycle specific core operating limit from the
Technical Specifications to the Core Operating Limits Report.
Additionally, a reference to a statistical methodology for determining
uncertainties is being changed to reference a methodology that was
recently approved by the NRC.
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:
Criterion 1 - Does Not Involve a Significant Increase in the
Probability of an Accident Previously Evaluated.
The removal of the cycle-dependent value for the departure from
nucleate boiling ratio (DNBR) reduction from technical
specifications and placing it into the Core Operating Limits Report
(COLR) has no impact on plant operation or accident analyses. The
proposed change is considered to be administrative in nature.
Technical specifications will continue to require operation within
the core operational limits for each cycle reload calculated by the
approved reload design methodologies. The appropriate actions
required if limits are violated will remain in the technical
specifications. The reload report presents the results of a cycle-
specific evaluation of accidents and transients addressed in the
ANO-2 Safety Analysis Report (SAR). The cycle-specific evaluation
demonstrates that changes in the fuel cycle design and the
corresponding COLR do not involve a significant increase in the
probability or consequences of an accident previously evaluated.
The Modified Statistical Combination of Uncertainties (MSCU)
methodology statistically combines uncertainties to at least a 95/95
probability/confidence level. The Proposed change to reference the
MSCU is administrative in nature. The currently referenced
methodology is being replaced with a more recently approved
methodology which has been determined to be applicable to ANO-2. The
new methodology has been independently reviewed and approved by the
NRC. This change does not impact either the manner in which the
operating margin to limits on linear heat rate and DNBR is
maintained or the manner in which the CPCs respond to transients and
provide trips. Therefore, this change does not adversely impact
transient analysis assumptions or results. In addition, the physical
design or operation of the plant is not impacted by this change. The
safety analyses will continue to be performed utilizing NRC-approved
methodologies and specific reload changes will be evaluated per
10CFR50.59.
Therefore, these changes do not involve a significant increase
in the probability or consequences of any accident previously
evaluated.
Criterion 2 - Does Not Create the Possibility of a New or
Different Kind of Accident from any Previously Evaluated.
The proposed change to relocate the cycle-specific value for the
DNBR reduction from technical specifications to the COLR is
administrative in nature. No change to the design, configuration, or
method of operation of the plant is made by this change. This
parameter will be determined using NRC-approved methods. Technical
specifications will continue to require operation within the
required core operating limits and appropriate actions will be taken
if the limits are exceeded. The relocation of a cycle-specific
parameter does not create the possibility of a new or different of
accident from any accident previously evaluated.
The proposed change to reference the NRC-approved MSCU
methodology is administrative in nature. The currently
[[Page 37089]]
referenced methodology is being replaced with a more recently approved
methodology which has been determined to be applicable to ANO-2. No
physical alterations of plant configuration, changes to plant
operating procedures, or operating parameters are proposed. The
safety analyses are still performing utilizing NRC-approved
methodologies and specific reload changes will be evaluated per
10CFR50.59. No new equipment is being introduced, and no equipment
is being operated in a manner inconsistent with its design.
Therefore, these changes do not create the possibility of a new
or different kind of accident from any previously evaluated.
Criterion 3 - Does Not Involve a Significant Reduction in the
Margin of Safety.
Existing technical specification operability and surveillance
requirements are not reduced by the proposed change to relocate the
cycle-specific value for DNBR reduction to the COLR. The development
of limits for a particular cycle will continue to conform to methods
described in NRC-approved documentation. Technical specifications
will still require that the core be operated within these limits and
specify appropriate actions to be taken if the limits are violated.
The cycle-specific COLR limits for future reloads will be developed
based on NRC-approved methodologies. Each reload undergoes a
10CFR50.59 safety review to assure that operating of the unit within
the cycle-specific limits will not involve a significant reduction
in a margin of safety.
The proposed change to reference the MSCU methodology is
administrative in nature. The currently referenced methodology is
being replaced with a more recently approved methodology which has
been determined to be applicable to ANO-2. The resultant overall
uncertainty factors using the MSCU methodology are determined and
applied to at least the same 95/95 probability/confidence level as
the overall uncertainty factors using the current methodology. NRC
review and approval of the methodologies used to perform the cycle-
specific reload analyses is not affected by this change. The safety
analyses are still performed utilizing NRC-approved methodologies
and specific reload changes will be evaluated per 10CFR50.59.
Therefore, these 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: Tomlinson Library, Arkansas
Tech University, Russellville, AR 72801
Attorney for licensee: Nicholas S. Reynolds, Esquire, Winston and
Strawn, 1400 L Street, N.W., Washington, DC 20005-3502
NRC Project Director: William D. Beckner
Entergy Operations, Inc., Docket No. 50-368, Arkansas Nuclear One,
Unit No. 2, Pope County, Arkansas
Date of amendment request: March 17, 1995
Description of amendment request: The proposed amendment deletes
requirements associated with surveillances to verify position stops for
High Pressure Safety Injection Emergency Core Cooling System throttle
valves.
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:
Criterion 1 - Does Not Involve a Significant Increase in the
Probability or Consequences of an Accident Previously Evaluated.
The HPSI system is not an initiator of a previously evaluated
accident; therefore, the probability of a previously evaluated
accident will not be increased by the proposed change. Accidents
which require the use of HPSI will not have any increased
consequences since the new injection/isolation valve arrangement is
at least as reliable as the previous valve arrangement. No part of
the proposed change has any adverse effect upon the HPSI system
response or function. The new manual valves will perform the
throttling function previously performed by the HPSI isolation MOVs
without reliance upon any electrical equipment (MOV limit switches).
The proposed change does not affect routing of HPSI piping or affect
total flow characteristics of the system. The proposed change to
remove the requirement to verify the correct settings of position
stops for the HPSI throttle valves is consistent with NUREG-1432,
restructured ``Standard Technical Specifications - Combustion
Engineering Plants,'' since the manual throttle valves fixed into
position serve the function of, and are equivalent to, flow limiting
orifices.
Therefore, this change does not involve a significant increase
in the probability or consequences of any accident previously
evaluated.
Criterion 2 - Does Not Create the Possibility of a New or
Different Kind of Accident from any Previously Evaluated.
The proposed change does not change the function or mode of
operation of the HPSI system. The failure of the new MOVs to
function will have no different effect than failure of the
previously installed MOVs and such failure is enveloped by
assumptions in the existing safety analysis, i.e., redundant trains
will still be able to function. The new manual valves are less
likely to fail in operation since they are fixed into position by
tack-welded locking devices and therefore perform their function
passively. Inadvertent manipulation of the manual valves will be
prevented by the locking arrangement. There are no new functions or
modes being accomplished by the MOVs. The throttling function to be
performed by the manual valves will be more reliably performed by
passive components than by active electrical circuits. The change
eliminates uncertainty in throttle valve position as a result of
limit switch tolerances and repeatability which form the basis for
the current surveillance requirement for periodic verification.
Therefore, this change does not create the possibility of a new
or different kind of accident from any previously evaluated.
Criterion 3 - Does Not Involve a Significant Reduction in the
Margin of Safety.No margin of safety will be reduced or affected by
the proposed deletion of the surveillance requirement. The new
manual valves will be throttled to produce a system flow balance
equivalent to the current one, and the balance will continue to be
confirmed by surveillance testing in accordance with TS
requirements.
Therefore, this change does 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: Tomlinson Library, Arkansas
Tech University, Russellville, AR 72801
Attorney for licensee: Nicholas S. Reynolds, Esquire, Winston and
Strawn, 1400 L Street, N.W., Washington, D.C. 20005-3502
NRC Project Director: William D. Beckner
Entergy Operations, Inc., Docket No. 50-368, Arkansas Nuclear One,
Unit No. 2, Pope County, Arkansas
Date of amendment request: March 17, 1995
Description of amendment request: The proposed amendment revises
requirements associated with the frequency of containment post-entry
visual inspections.
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:
Criterion 1 - Does Not Involve a Significant Increase in the
Probability or Consequences of an Accident Previously Evaluated.
The proposed change to the Arkansas Nuclear One-Unit 2 (ANO-2)
technical specifications (TS) does not involve any system or
component or condition evaluated as an accident initiator; therefore
there is no increase in the probability of an accident previously
evaluated.
The purpose of this change is to reduce the required number of
containment inspections following entries at operational modes above
cold shutdown. This reduction in the number of inspections will
reduce personnel
[[Page 37090]]
exposure to radiation and potential heat stress. These inspections are
to verify that no debris that might be transported to the
containment sump is left behind at the conclusion of the entry.
Typically, containment entries above cold shutdown are for specific
purposes and involve a limited area of containment. The expectation
for job performance at ANO-2 is that a job site is left cleaner than
found. The inspection serves as a verification that any materials
taken into the containment building which might foul the sump
screens have been removed or have been properly anchored. Performing
this inspection on a daily frequency will not result in changing the
work practices at ANO-2, therefore the amount of debris generated or
left in containment should not increase. The daily inspection will
be sufficient verification that conditions in containment are not
degrading; therefore, there will be no significant increase in the
consequences of an accident previously evaluated.
Therefore, this change does not involve a significant increase
in the probability or consequences of any accident previously
evaluated.
Criterion 2 - Does Not Create the Possibility of a New or
Different Kind of Accident from any Previously Evaluated.
Because the proposed amendment will not change the design,
configuration, or method of operation of the plant, this change does
not create the possibility of a new or different kind of accident
from any previously evaluated.
Criterion 3 - Does Not Involve a Significant Reduction in the
Margin of Safety.
There will be no adverse effects on margins of safety since
materials that are considered acceptable to remain in containment
has not changed. By reducing the number of inspections, no mechanism
has been created that will generate more debris in containment nor
have work practices been altered to allow less stringent controls
over what is taken in or left in containment. Therefore, this change
does 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: Tomlinson Library, Arkansas
Tech University, Russellville, AR 72801
Attorney for licensee: Nicholas S. Reynolds, Esquire, Winston and
Strawn, 1400 L Street, N.W., Washington, DC 20005-3502
NRC Project Director: William D. Beckner
Entergy Operations, Inc., Docket No. 50-368, Arkansas Nuclear One,
Unit No. 2, Pope County, Arkansas
Date of amendment request: April 4, 1995
Description of amendment request: The proposed amendment deletes
requirements associated with part length control element assemblies.
During the upcoming refueling outage all part length control assemblies
will be removed from the reactor.
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:
Criterion 1 - Does Not Involve a Significant Increase in the
Probability or Consequences of an Accident Previously Evaluated.
The proposed changes maintain conservative restrictions on the
operation of those control element assemblies (CEAs) formerly
specified as part length CEAs (PLCEAs) and are considered to be
administrative in nature. The Arkansas Nuclear One - Unit 2 (ANO-2)
Safety Analysis Report (SAR) Chapter 15 accident analyses identify
four families of analyses associated with the CEAs. Each of these
analyses is evaluated in the development of the Reload Report for
each fuel cycle, and the appropriate limitations to insure
acceptable analysis results are incorporated in the Core Operating
Limits Report (COLR) for the fuel cycle. The modification replacing
the PLCEAs with full length CEAs will be evaluated under the
Arkansas Nuclear One (ANO) 10CFR50.59 process prior to
implementation. The Reload Report and changes to the COLR are also
evaluated under the ANO 10CFR50.59 process prior to incorporating
the identified changes. Movement of the PLCEAs during power
operation has typically resulted in more dropped CEAs than movement
of the full length CEAs due to the greater weight of the PLCEAs.
Replacement of the PLCEAs with full length CEAs should result in a
reduction in the probability of a dropped CEA.
Therefore, this change does not involve a significant increase
in the probability or consequences of any accident previously
evaluated.
Criterion 2 - Does Not Create the Possibility of a New or
Different Kind of Accident from any Previously Evaluated.
The proposed changes introduce no new mode of plant operation
and are considered to be administrative in nature. Operating
experience has shown that the full length CEAs are capable of
controlling the axial power distribution function intended for the
PLCEAs. The PLCEAs will be replaced with the same type of full
length CEAs used in shutdown and regulating CEA groups.
Therefore, this change does not create the possibility of a new
or different kind of accident from any previously evaluated.
Criterion 3 - Does Not Involve a Significant Reduction in the
Margin Safety.
The proposed changes may improve overall safety margins.
Replacement of the PLCEAs with full length CEAs and including these
Group P CEAs in the CEA drop time testing will allow ANO-2 to credit
these CEAs in the shutdown margin calculations. This should result
in an increase in the available shutdown margin during reactor
operation.
Therefore, this change does 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: Tomlinson Library, Arkansas
Tech University, Russellville, AR 72801
Attorney for licensee: Nicholas S. Reynolds, Esquire, Winston and
Strawn, 1400 L Street, N.W., Washington, DC 20005-3502
NRC Project Director: William D. Beckner
Entergy Operations, Inc., Docket No. 50-368, Arkansas Nuclear One,
Unit No. 2, Pope County, Arkansas
Date of amendment request: April 4, 1995
Description of amendment request: The proposed amendment revises
the containment cooling response time to reduce the likelihood of a
water hammer event in service water piping.
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:
Criterion 1 - Does Not Involve a Significant Increase in the
Probability or consequences of an Accident Previously Evaluated.
The containment cooling system and the service water system are
not considered to be accident initiators for any analyzed accident.
The containment cooling system functions to mitigate the effects of
a Main Steam Line Break (MSLB) or Loss of Coolant Accident (LOCA) on
the containment environment. The proposed change does not affect the
limiting MSLB analysis as the proposed increase in containment
cooling response time is only instituted on a loss of off-site
power. The limiting LOCA analysis has been evaluated with respect to
the proposed containment cooling response time. Although the
analysis shows an increase in the containment peak pressure
(approximately 0.1 psig), this increase in the peak containment
pressure is not considered significant since the MSLB accident with
off-site power available is still the overall limiting accident
condition with respect to containment peak pressure. The containment
peak conditions for the LOCA and MSLB analyses remain below the
original Final Safety Analysis Report (FSAR) conditions of 53.4 psig
and 288 deg.F.
Therefore, this change does not involve a significant increase
int he probability or consequences of any accident previously
evaluated.
[[Page 37091]]
Criterion 2 - Does Not Create the Possibility of a New or
Different Kind of Accident from any Previously Evaluated.
The proposed change in containment cooling response time
introduces no new mode of plant operation. Containment cooling
response time is an analytical input and is not considered to be the
initiator of any accident condition.
Therefore, this change does not create the possibility of a new
or different kind of accident from any previously evaluated.
Criterion 3 - Does Not Involve a Significant Reduction in the
Margin of Safety.
The increase in containment cooling response time has been
evaluated with respect to the accident analyses resulting in peak
containment pressures. This evaluation has shown no significant
increase in the resulting peak containment pressure since the
overall limiting accident with respect to containment pressure is
still the MSLB with off-site power available. The containment peak
conditions for the LOCA and MSLB analyses remain below the original
FSAR conditions of 53.4 psig and 288 deg.F.
Therefore, this change does not involve a significant reduction
in the margin of safety.
Therefore, based upon the reasoning presented above and the
previous discussion of the amendment request, Entergy Operations has
determined that the requested change does not involve a significant
hazards consideration.
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: Tomlinson Library, Arkansas
Tech University, Russellville, AR 72801
Attorney for licensee: Nicholas S. Reynolds, Esquire, Winston and
Strawn, 1400 L Street, N.W., Washington, DC 20005-3502
NRC Project Director: William D. Beckner
Gulf States Utilities Company, Cajun Electric Power Cooperative,
and Entergy Operations, Inc., Docket No. 50-458, River Bend
Station, Unit 1, West Feliciana Parish, Louisiana
Date of amendment request: May 25, 1995
Description of amendment request: The proposed amendment revises
the Physical Security Plan vital island requirements.
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 accident mitigation features of the plant are not affected
by the proposed change. This change provides an equivalent level of
protection to the plant and is adequate for preventing an
unacceptable risk to public health and safety. This is due to
continued compliance with existing regulatory requirements, the
integral defense in depth design of the security program, including
programs in place to minimize the threat of insiders, and
historically high system reliability. The SBO (Station Blackout
diesel) is designed with sufficient capacity to accommodate station
blackout needs as well as those required for security. Ample
protection against a design basis security threat continues to be
provided. Therefore, this change does not increase the probability
or consequences of an accident previously evaluated.
The Station Blackout diesel generator has been approved and
accepted by the Staff pursuant 10CFR50.63. New systems, modes of
equipment operation, failure modes, or other plant perturbations are
not introduced by this change. The change provides an equivalent
level of protection, does not decrease the effectiveness of the
overall security program and is adequate for preventing an
unacceptable risk to public health and safety. Ample protection
against a design basis security threat continues to be provided with
overall physical protection of the plant maintained. Therefore, this
change does not create the possibility of a new or different kind of
accident from any previously evaluated.
This change does not change a safety limit, an LCO (Limiting
Condition of Operation), or a surveillance requirement on equipment
required to operate the plant. It is equivalent in level of
protection, does not decrease the effectiveness of the security
program and is adequate for preventing an unacceptable risk to
public health and safety. The SBO diesel generator will provide an
adequate alternative source of power to security systems. Ample
protection against a design basis security threat continues to be
provided. Therefore, this change does not involve a 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: Government Documents
Department, Louisiana State University, Baton Rouge, Louisiana 70803
Attorney for licensee: Mark Wetterhahn, Esq., Winston & Strawn,
1400 L Street, N.W., Washington, D.C. 20005
NRC Project Director: William D. Beckner
Houston Lighting & Power Company, City Public Service Board of San
Antonio, Central Power and Light Company, City of Austin, Texas,
Docket Nos. 50-498 and 50-499, South Texas Project, Units 1 and 2,
Matagorda County, Texas
Date of amendment request: May 22, 1995
Description of amendment request: The proposed amendment would
revise Technical Specification (TS) 4.8.1.1.2.e.7 to allow the
performance of the 24-hour surveillance test of the diesel generators
during power operation.
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 to permit the 24 hour surveillance test of
the diesels to be performed during power operation does not increase
the chances for a previously analyzed accident to occur. The
function of the diesels is to supply emergency power in the event of
a loss of offsite power. Operation of the diesels is not a precursor
to any accident. Furthermore, the diesel generator being tested will
remain operable and will be available to supply emergency loads
within the required time. In addition, the two remaining diesel
generators will be operable during the test. Consequently, if an
offsite disturbance were to occur that affected the operability of
the diesel being tested, the two remaining diesels would be capable
of feeding the loads necessary for safe shutdown of the plant. This
addresses the concerns raised in Information Notice 84-69 regarding
the operation of emergency diesel generators connected in parallel
with offsite power. In summary, the proposed changes do not
adversely affect the performance or the ability of the diesel
generators to perform their intended function.
Therefore, the proposed change will not involve a significant
increase in 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.
The proposed amendment to the 24 hour surveillance test will not
affect the operation of any safety system or alter its response to
any previously analyzed accident. The diesel will automatically
transfer from the test mode if necessary to supply emergency loads
in the required time. The test mode is used for the monthly
surveillance of the diesel generators as well, therefore, no new
plant operating modes are introduced. In the event the diesel fails
the surveillance test, it will be declared inoperable and the
actions
[[Page 37092]]
required for an inoperable diesel will be performed. The remaining two
diesel generators will be operable and are capable of feeding the
loads necessary for safe shutdown of the plant.
Therefore, the proposed change will 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 the margin of safety.
The proposed amendment will not reduce availability of the
diesel generator being tested to provide emergency power in the
event of a loss of offsite power. If a loss of offsite power occurs
during the surveillance test, the diesel generator output breaker
will be tripped by the directional over-current relay on the ESF
transformer. The diesel generator will transfer to the emergency
mode, and the ESF undervoltage logic will initiate Mode II (Loss of
Offsite Power) operation of the ESF load sequencer to supply
emergency loads from the diesel generator. If a Loss of Coolant
Accident occurs during the surveillance test, the diesel generator
output breaker will be opened by a signal from the Solid State
Protection System and the preferred offsite source will continue to
provide power to the ESF bus. The diesel generator will continue to
run in the emergency mode and would be available to automatically
supply safety-related loads during any loss of offsite power
condition. The test mode to emergency mode transfer is tested once
per cycle in accordance with Surveillance Requirement
4.8.1.1.2.e.10. In addition, the two remaining generators will be
operable during the test. Consequently, if an offsite disturbance
were to occur that affected the operability of the diesel being
tested, the two remaining diesels would be capable of feeding the
loads necessary for safe shutdown of the plant. The time required
for the diesel being tested to pick up emergency loads will not be
affected by performing the 24 hour surveillance test during power
operation.
Therefore, the proposed change will not involve a significant
reduction in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
request for amendments involves no significant hazards consideration.
Local Public Document Room location: Wharton County Junior College,
J. M. Hodges, Learning Center, 911 Boling Highway, Wharton, Texas 77488
Attorney for licensee: Jack R. Newman, Esq., Newman & Holtzinger,
P.C., 1615 L Street, N.W., Washington, D.C. 20036
NRC Project Director: William D. Beckner
Houston Lighting & Power Company, City Public Service Board of San
Antonio, Central Power and Light Company, City of Austin, Texas,
Docket Nos. 50-498 and 50-499, South Texas Project, Units 1 and 2,
Matagorda County, Texas
Date of amendment request: May 25, 1995
Description of amendment request: The proposed amendment would
revise the technical specifications (TSs) on containment leakage, to
make the action statement consistent with the need to perform Type C
testing at power, and to replace the surveillance requirements with a
single requirement to apply the requirements of Appendix J as modified
by approved exemptions. The proposed amendment would also revise the
TSs on containment integrity, containment leakage, and containment air
locks, to eliminate the numerical value of calculated peak containment
internal pressure related to the design basis accident. In addition,
there is an associated proposed exemption, from the requirements of 10
CFR Part 50, Appendix J, to allow the performance of the required
periodic Type C tests during power operation.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a) the licensee has provided
its analysis of the issue of no significant hazards consideration,
which is presented below:
1. Does the change involve a significant increase in the
probability or consequences of an accident previously evaluated?
The change to the action statement of Technical Specification
3.6.1.2 does not significantly increase the probability of an
accident because leakage rate testing is not an accident initiator.
The consequences of an accident previously evaluated are not
increased by changing the ACTION statement of Technical
Specification 3.6.1.2 because the requirements for CONTAINMENT
INTEGRITY are not reduced. The consequences of an accident
previously evaluated are not increased by the change in the
surveillance wording because no technical changes are proposed. The
underlying purpose of the proposed change to the Technical
Specifications and requested exemption to Appendix J, to allow
surveillance credit for at-power Type C testing, will not increase
the consequences of an accident because there are no reductions in
the requirements to maintain containment integrity.
The proposed change to delete the numeric value of Pa is purely
administrative, and has no potential effect on accident initiation
or consequences.
2. Does the change create the possibility of a new or different
kind of accident from any previously evaluated?
Nothing associated with the requested changes will physically
change the configuration of the plant or impose new operating
configurations not previously considered. Leakage rate testing will
remove components and trains from service; however, this is not
operationally different from other testing and maintenance
evolutions that remove components or trains from service, and which
were previously considered. Consequently, the possibility of a new
or different kind of accident from any previously evaluated is not
created.
3. Does this change involve a significant reduction in the
margin of safety?
The margin of safety is not significantly reduced by changing
the ACTION statement of Technical Specification 3.6.1.2 because the
requirements for CONTAINMENT INTEGRITY are not reduced. The margin
of safety is not reduced by the change in the surveillance wording
because no technical changes are proposed. The underlying purpose of
the proposed change to the Technical Specifications and requested
exemption to Appendix J, to allow surveillance credit for at-power
Type C testing, will not reduce the margin of safety because there
are no reductions in the requirements to maintain containment
integrity.
The proposed change to delete the numeric value of Pa is
purely administrative, and has no potential effect on the margin of
safety because the value itself is unchanged.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
request for amendments involves no significant hazards consideration.
Local Public Document Room location: Wharton County Junior College,
J. M. Hodges, Learning Center, 911 Boling Highway, Wharton, Texas 77488
Attorney for licensee: Jack R. Newman, Esq., Newman & Holtzinger,
P.C., 1615 L Street, N.W., Washington, D.C. 20036
NRC Project Director: William D. Beckner
Houston Lighting & Power Company, City Public Service Board of San
Antonio, Central Power and Light Company, City of Austin, Texas,
Docket Nos. 50-498 and 50-499, South Texas Project, Units 1 and 2,
Matagorda County, Texas
Date of amendment request: May 30, 1995
Description of amendment request: The proposed amendment would
increase the spent fuel pool heat load licensing basis to provide
greater flexibility for normal refueling practices.
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:
[[Page 37093]]
1. The proposed changes do not involve a significant increase in
the probability or consequences of an accident previously evaluated
because:
(a) The Spent Fuel Pool conditions are not indicative of
accident initiators.
(b) Design and operability requirements of equipment important
to safety are not affected.
(c) If only one Spent Fuel Pool cooling train is available,
boiling would not occur and the Spent Fuel Pool components would
remain within their design basis.
(d) The complete loss of Spent Fuel Pool cooling event has
previously been analyzed and described in Supplement 6 to the Safety
Evaluation Report, Appendix BB. The dose consequences for this event
have been evaluated and the safety evaluation is described in
Updated Safety Analysis Report Section 9.1.3.3.4. The results of the
evaluation show that the Spent Fuel Pool components would remain
within their design bases. Also, the dose consequences of iodine
release as a result of Spent Fuel Pool boiling are significantly
below the allowable dose limits of 10 CFR 100.
2. The proposed changes do not create the possibility of a new
or different kind of accident from any accident previously because:
(a) The operability of safety-related equipment is not impacted.
(b) The probability of safety-related equipment malfunctioning
is not increased.
(c) The scope of the change does not establish a potential new
accident precursor.
(d) The Spent Fuel Pool design considers design basis heat loads
for the modified refueling procedure which includes a full-core
offload.
(e) For the design basis case, the integrity of the Spent Fuel
Pool Boraflex is not adversely impacted.
3. The proposed changes do not involve a significant reduction
in the margin of safety because:
(a) No fuel damage would occur as a result of the proposed
change.
(b) Technical Specification operability and surveillance
requirements are not reduced.
(c) The Spent Fuel Pool boiling doses would be significantly
below the allowable dose limits of 10 CFR 100.
(d) The modified refueling procedure (full-core offload)
continues to have acceptable margins of safety.
(e) For the design basis case, the integrity of the Spent Fuel
Pool Boraflex is not adversely impacted.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
request for amendments involves no significant hazards consideration.
Local Public Document Room location: Wharton County Junior College,
J. M. Hodges, Learning Center, 911 Boling Highway, Wharton, Texas 77488
Attorney for licensee: Jack R. Newman, Esq., Newman & Holtzinger,
P.C., 1615 L Street, N.W., Washington, D.C. 20036
NRC Project Director: William D. Beckner
Illinois Power Company and Soyland Power Cooperative, Inc., Docket
No. 50-461, Clinton Power Station, Unit No. 1, DeWitt County,
Illinois
Date of amendment request: June 9, 1995
Description of amendment request: The proposed amendment would
modify Technical Specification (TS) 4.1, ``Site Location,'' to
incorporate a description of the exclusion area boundary. The proposed
change is necessary to ensure the content of the TS conforms to Section
182 of the Atomic Energy Act of 1954.
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 change to the plant
design or operation. As a result, the proposed change does not
affect any of the parameters or conditions that could contribute to
the initiation of any accidents previously evaluated. In addition,
the physical location of the [exclusion area boundary] EAB has not
been changed; a description of its location has merely been added to
the TS. Thus, the proposed change cannot increase the probability or
the consequences of any accident previously evaluated.
(2) The proposed change does not involve a change to the plant
design or operation. As a result, the proposed change does not
affect any parameter or condition that could contribute to the
initiation of any accidents. Thus, the proposed change cannot create
the possibility of a new or different kind of accident from any
accident previously evaluated.
(3) The proposed change only affects regulatory controls on the
accepted configuration of the EAB. The proposed change does not
involve an actual change to the location of the EAB. The proposed
change will restore compliance with the Atomic Energy Act of 1954
and require prior NRC approval of any changes to the physical
location of the EAB. As a result, IP has concluded that the proposed
change will not result in a 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: Vespasian Warner Public
Library, 120 West Johnson Street, Clinton, Illinois 61727
Attorney for licensee: Leah Manning Stetzner, Vice President,
General Counsel, and Corporate Secretary, 500 South 27th St., Decatur,
Illinois 62525
NRC Project Director: Gail H. Marcus
Indiana Michigan Power Company, Docket No. 50-315, Donald C. Cook
Nuclear Plant, Unit No. 1, Berrien County, Michigan
Date of amendment request: February 3, 1995, as supplemented April
25, 1995 (AEP:NRC:1166Q and 1166R)
Description of amendment request: The proposed amendment would
allow continued use of a steam generator (SG) tube support plate
interim plugging criteria for fuel cycle 15. The change would allow SG
tubes with bobbin coil eddy current indications less than or equal to
2.0 volts at tube support plate intersections to remain in service if
the projected end-of-cycle distribution of crack indications is shown
to result in primary-to-secondary leakage less than 12.6 gpm during a
postulated steam line break (SLB). The change would also allow
indications greater than 2.0 volts but less than or equal to 5.6 volts
to remain in service if a motorized rotating pancake coil probe
inspection does not detect degradation.
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:
In accordance with the three factor test of 10 CFR 50.92(c),
implementation of the proposed license amendment is analyzed using
the following standards and found not to 1) involve a significant
increase in the probability or consequences of an accident
previously evaluated; 2) create the possibility of a new or
different kind of accident from any accident previously evaluated;
or 3) involve a significant reduction in margin of safety.
Conformance of the proposed amendment to the standards for a
determination of no significant hazards as defined in 10 CFR 50.92
(three factor test) is shown in the following paragraphs.
1) Operation of Cook Nuclear Plant Unit 1 in accordance with the
proposed license amendment does not involve a significant increase
in the probability or consequences of an accident previously
evaluated. Testing of model boiler specimens for free span tubing
(no tube support plate restraint) at room temperature conditions
show burst pressures in excess of 5000 psi for indications of outer
diameter stress corrosion cracking with voltage measurements as high
as 19 volts. Burst testing performed on pulled tubes from Cook
Nuclear Plant Unit 1 with up to a 2.02 volt indication shows
measured burst pressure in excess of 10,000 psi at room temperature.
Burst testing performed on pulled tubes from other plants with up to
7.5 volt indications show burst pressures in excess of 6,300 psi at
room temperatures. Correcting for the effects of temperature on
[[Page 37094]]
material properties and minimum strength levels (as the burst testing
was done at room temperature), tube burst capability significantly
exceeds the safety factor requirements of RG [Regulatory Guide]
1.121 [``Bases for Plugging Degraded PWR Steam Generator Tubes''].
As stated earlier, tube burst criteria are inherently satisfied
during normal operating conditions due to the proximity of the tube
support plate [TSP]. Test data indicates that tube burst cannot
occur within the tube support plate, even for tubes which have 100%
throughwall electric-discharge machined notches 0.75 inch long,
provided the tube support plate is adjacent to the notched area.
Since tube-to-tube support plate proximity precludes tube burst
during normal operating conditions, use of the criteria must,
therefore, retain tube integrity characteristics which maintain the
RG 1.121 margin of safety of 1.43 times the bounding faulted
condition (steam line break) pressure differential.
During a postulated main steam line break, the TSP has the
potential to deflect during blowdown, thereby uncovering the
intersection. Based on the existing data base, the RG 1.121
criterion requiring maintenance of a safety factor of 1.43 times the
steam line break pressure differential on tube burst is satisfied by
7/8 inch diameter tubing with bobbin coil indications with signal
amplitudes less than 9.6 volts, regardless of the indicated depth
measurement. A 2.0 volt plugging criteria compares favorably with
the 9.6 volt structural limit considering the previously calculated
growth rates for ODSCC [outer diameter stress corrosion cracking]
within Cook Nuclear Plant Unit 1 SGs. Considering a voltage growth
component of 0.8 volts (40% voltage growth based on 2.0 volts BOC
[beginning of cycle]), and a nondestructive examination uncertainty
of 0.40 volts (20% voltage uncertainty based on 2.0 volts BOC), when
added to the BOC IPC [interim plugging criteria] of 2.0 volts,
results in a bounding EOC [end of cycle] voltage of approximately
3.2 volts for cycle 15 operation. A 6.4 volt safety margin exists
(9.6 structural limit - 3.2 volt EOC - 6.4 volt margin).
For the voltage/burst correlation, the EOC structural limit is
supported by a voltage of 9.6 volts. Using this structural limit of
9.6 volts, a BOC maximum allowable repair limit can be established
using the guidance of RG 1.121. The BOC maximum allowable repair
limit should not permit a significant number of EOC indications to
exceed the 9.6 volt structural limit and should assure that
acceptable tube burst probabilities are attained. By adding NDE
[nondestructive examination] uncertainty allowances and an allowance
for crack growth to the repair limit, the structural limit can be
validated. The previous IPC submittal established the conservative
NDE uncertainty limit of 20% of the BOC repair limit. For
consistency, a 40% voltage growth is extremely conservative for Cook
Nuclear Plant Unit 1. Therefore, the maximum allowable BOC repair
limit (RL) based on the structural limit of 9.6 volts can be
represented by the expression:
RL + (0.2 x RL) + (0.4 x RL) = 9.6 volts, or,the maximum
allowable BOC repair limit can be expressed as,
RL = 9.6 volt structural limit/1.6 = 6.0 volts
This structural repair limit supports this application for cycle
15 IPC implementation to repair bobbin indications greater than 2.0
volts based on RPC [rotating pancake coil] confirmation of the
indication. Conservatively, an upper limit of 5.6 volts will be used
to repair bobbin indications which are above 2.0 volts but do not
have confirming RPC calls.
The conservatism of this repair limit is shown by the EOC 13
(Spring 1994) eddy current data. The overall average voltage growth
was determined to be on 1.4% (of the BOC voltage). In addition, the
EOC 13 maximum observed voltage increase was 0.40 volts, and
occurred in a tube with a BOC indication of 0.96 volts. The
applicability of cycle 14 growth rates for cycle 15 operation will
be confirmed prior to return to service of Cook Nuclear Plant Unit
1. Similar large structural margins are anticipated.
Relative to the expected leakage during accident condition
loadings, it has been previously established that a postulated main
steam line break outside of containment but upstream of the main
steam isolation valve represent the most limiting radiological
condition relative to the IPC. In support of implementation of the
IPC, it will be determined whether the distribution of crack
indications at the tube support plate intersections at the end of
cycle 15 are projected to be such that primary to secondary leakage
would result in site boundary doses within a small fraction of the
10 CFR 100 guidelines. A separate calculation has determined this
allowable steam line break leakage limit to be 12.6 gpm. Although
not required by the Cook Nuclear Plant design basis, this
calculation uses the recommended Iodine-131 transient spiking values
consistent with NUREG-0800 [Standard Review Plan], and the T/S
[technical specification] reactor coolant system activity limit of
1.0 micro curie per gram dose equivalent Iodine-131. The projected
steam line break leakage rate calculation methodology prescribed in
[Draft] GL 94-XX [``Voltage-Based Repair Criteria for the Repair of
Westinghouse Steam Generator Tubes Affected by Outside Diameter
Stress Corrosion Cracking,'' August 12, 1994] and WCAP 14277 [``SLB
Leak Rate and Tube Burst Probability Analysis Methods for ODSCC at
TSP Intersections''] will be used to calculate EOC 15 leakage, based
on actual EOC 14 distributions and EOC 15 projected distributions.
Due to the relatively low voltage growth rates at Cook Nuclear Plant
Unit 1 and the relatively small number of indications affected by
the IPC, steam line break leakage prediction per GL 94-XX is
expected to be significantly less than the acceptance limit of 12.6
gpm in the faulted loop.
Prior to issue of GL 94-XX, projected EOC 14 leak rates were
calculated, based on draft NUREG-1477 [``Voltage-Based Interim
Plugging Criteria for DG Tubes, Draft for Comments''], for a total
of twelve cases, the combination of six probability-of-leak
correlations and two leak rate calculation methodologies. Results of
the calculations show that the projected EOC 14 leak rates ranged
from 0.001 gpm to 1.360 gpm. These results are well below the 12.6
gpm allowable; therefore, implementation of the 2 volt IPC during
cycle 15 would not adversely affect SG tube integrity and results in
acceptable dose consequences.
Current GL 94-XX methodology requires only the log-logistic
probability of leakage correlation be used. Projected EOC 14 SLB
leakage using this function was calculated to be only 0.001 gpm.
Based on the relatively few numbers of intersections at Cook Nuclear
Plant Unit 1 to which the IPC are applied and extremely small Cook
Nuclear Plant Unit 1 plant-specific growth rate, a similar value
would be expected based on the EOC 14 eddy current data. The
inclusion of all IPC intersections in the leakage model, along with
application of a probability of detection of 0.6, will result in
extremely conservative leakage estimations, especially so since
close examination of the available data shows that indications of
less than 2.8 volts will not be expected to leak during SLB
conditions. All Unit 1 IPC indications are expected to be below 2.8
volts at the EOC 15 conditions.
The proposed amendment does not result in any increase in the
probability or consequences of an accident previously evaluated
within the Cook Nuclear Plant Unit 1 FSAR.
2) The proposed license amendment does not create the
possibility of a new or different kind of accident previously
evaluated.
Implementation of the proposed SG tube IPC does not introduce
any significant changes to the plant design basis. Use of the
criteria does not provide a mechanism which could result in an
accident outside of the region of the tube support plate elevations.
Neither a single or multiple tube rupture event would be expected in
a SG in which the plugging criteria has been applied (during all
plant conditions).
Specifically, we will continue to implement a maximum leakage
rate limit of 150 gpd (0.1 gpm) per SG to help preclude the
potential for excessive leakage during all plant conditions. The
cycle 15 T/S limits imposed on primary to secondary leakage at
operating conditions are: a maximum of 0.4 gpm (600 gpd) for all SGs
with a maximum of 150 gpd allowed for any one SG.
The RG 1.121 criteria for establishing operational leakage rate
limits that require plant shutdown are based upon leak-before-break
considerations to detect a free span crack before potential tube
rupture during faulted plant conditions. The 150 gpd limit should
provide for leakage detection and plant shutdown in the event of the
occurrence of an unexpected single crack resulting in leakage that
is associated with the longest permissible crack length. Regulatory
Guide 1.121 acceptance criteria for establishing operating leakage
limits are based on leak-before-break considerations such that plant
shutdown is initiated if the leakage associated with the longest
permissible crack is exceeded. The longest permissible crack is the
length that provides a factor of safety of 1.43 against bursting at
faulted conditions maximum pressure differential. A voltage
amplitude of 9.6 volts for typical ODSCC corresponds to meeting this
tube burst requirement at a lower 95% prediction limit on the burst
correlation
[[Page 37095]]
coupled with 95/95 lower tolerance limit material properties. Alternate
crack morphologies can correspond to 9.6 volts so that a unique
crack length is not defined by the burst pressure versus voltage
correlation. Consequently, typical burst pressure versus through-
wall crack length correlations are used below to define the
``longest permissible crack'' for evaluating operating leakage
limits.
Consistent with the Cycle 13 and Cycle 14 license amendment
requests for IPC and Section 5 of Enclosure 1 of the GL, operational
leakage limits will remain at 150 gpd per SG. Axial cracks leaking
at this level are expected to provide leak before break (LBB)
protection at both the SLB pressure differential of 2560 psi and,
while not part of any established LBB methodology, LBB protection
will also be provided at a value of 1.43 times the SLB pressure
differential. Thus, the 150 gpd limit provides for plant shutdown
prior to reaching critical crack lengths for steam line break
conditions. Additionally, this leak-before-break evaluation assumes
that the entire crevice area is uncovered during blowdown. Partial
uncovery will provide benefit to the burst capacity of the
intersection.
3) The proposed license amendment does not involve a significant
reduction in margin of safety.
The use of the voltage based bobbin probe interim tube support
plate elevation plugging criteria at Cook Nuclear Plant Unit 1 is
demonstrated to maintain SG tube integrity commensurate with the
criteria of RG 1.121. Regulatory Guide 1.121 describes a method
acceptable to the NRC staff for meeting GDC [General Design
Criteria] 14, 15, 31, and 32 by reducing the probability or the
consequences of SG tube rupture. This is accomplished by determining
the limiting conditions of degradation of SG tubing, as established
by inservice inspection, for which tubes with unacceptable cracking
should be removed from service. Upon implementation of the criteria,
even under the worst case conditions, the occurrence of ODSCC at the
tube support plate elevations is not expected to lead to a SG tube
rupture event during normal or faulted plant conditions. The EOC 15
distribution of crack indications at the tube support plate
elevations will be confirmed by analysis and calculation to result
in acceptable primary to secondary leakage during all plant
conditions and that radiological consequences are not adversely
impacted.
In addressing the combined effects of a LOCA [loss-of-coolant
accident] and SSE [safe-shutdown earthquake] on the SG component (as
required by GDC 2), it has been determined that tube collapse may
occur in the SGs at some plants. This is the case as the tube
support plates may become deformed as a result of lateral loads at
the wedge supports at the periphery of the plate due to the combined
effects of the LOCA rarefaction wave and SSE loadings. Then, the
resulting pressure differential on the deformed tubes may cause some
of the tubes to collapse.
There are two issues associated with SG tube collapse. First,
the collapse of SG tubing reduces the RCS [reactor coolant system]
flow area through the tubes. The reduction in flow area increases
the resistance to flow of steam from the core during a LOCA which,
in turn, may potentially increase peak clad temperature. Second,
there is a potential that partial through-wall cracks in tubes could
progress to through-wall cracks during tube deformation or collapse.
Consequently, since the leak-before-break methodology is
applicable to the Cook Nuclear Plant Unit 1 reactor coolant loop
piping, the probability of breaks in the primary loop piping is
sufficiently low that they need not be considered in the structural
design of the plant. The limiting LOCA event becomes either the
accumulator line break or the pressurizer surge line break. Loss of
coolant accident loads for the primary pipe breaks were used to
bound the Cook Nuclear Plant Unit 1 smaller breaks. The results of
the analysis using the larger break inputs show that the LOCA loads
were found to be of insufficient magnitude to result in SG tube
collapse or significant deformation.
Addressing RG 1.83 [``Inservice Inspection of PWR Steam
Generator Tubes''] considerations, implementation of the bobbin
probe voltage based interim tube plugging criteria of 2.0 volts is
supplemented by enhanced eddy current inspection guidelines to
provide consistency in voltage normalization, a 100% eddy current
inspection sample size at the tube support plant elevations per T/S,
and MRPC [Motorized Rotating Pancake Coil] inspection requirements
for the larger indications left in service to characterize the
principal degradation as ODSCC.
As noted previously, implementation of the tube support plate
elevation plugging criteria will decrease the number of tubes which
must be repaired. The installation of SG tube plugs reduces the RCS
flow margin. Thus, implementation of the IPC will maintain the
margin of flow that would otherwise be reduced in the event of
increased tube plugging.
Based on the above, it is concluded that the proposed license
amendment request does not result in a significant reduction in
margin with respect to plant safety as defined in the Final Safety
Analysis Report or any Bases of the plant T/Ss.
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: Maud Preston Palenske Memorial
Library, 500 Market Street, St. Joseph, Michigan 49085
Attorney for licensee: Gerald Charnoff, Esq., Shaw, Pittman, Potts
and Trowbridge, 2300 N Street, NW, Washington, DC 20037
NRC Project Director: John N. Hannon
Indiana Michigan Power Company, Docket Nos. 50-315 and 50-316,
Donald C. Cook Nuclear Plant, Unit Nos. 1 and 2, Berrien County,
Michigan
Date of amendment requests: May 26, 1995 (AEP:NRC:1207)
Description of amendment requests: The proposed amendments would
change multiple operating limits on both units. The primary change
would allow operation of Cook Unit 1 with steam generator plugging
levels as high as 30% in each steam generator. The second group of
changes would modify the overtemperature delta T and overpower delta T
reactor trip setpoints for both units and increase the allowed
degradation of the Unit 1 auxiliary feedwater pumps consistent with
Unit 2. The third group of changes would reduce the required shutdown
margin in modes 1, 2, 3, and 4, increase the allowable centrifugal
charging pump head degradation, reduce the minimum refueling water
storage tank temperature, and revise the peak pressure of the long-term
containment integrity analysis in the bases. Finally, certain
administrative changes are also proposed.
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:
10 CFR 50.92 specifies that the holder of an operating license
or construction permit of a nuclear power facility participate in
determining whether a change to the T/S's current licensing basis
(CLB) involves a significant hazards consideration. Prior to
implementation of a change to the CLB, the Nuclear Regulatory
Commission must review and make a final determination, pursuant to
the procedures in 10 CFR 50.91, that a proposed amendment to the
operating license involves no significant hazards considerations. In
order to satisfactorily complete the review, the proposed amendment
to the CLB must not:
1. involve a significant increase in the probability or
consequences of an accident previously evaluated,
2. create the possibility of a new or different kind of accident
from any accident previously evaluated, or
3. involve a significant reduction in a margin of safety.
For the purpose of performing a significant hazards
consideration analysis, the four groups of technical specification
changes discussed under Description of Changes can be reduced to
three groups. In evaluating significant hazards, the first three
groups of proposed technical specifications will be considered
together. The miscellaneous change and the administrative change
will each be considered separately.
[[Page 37096]]
Determination Of No Significant Hazards For Changes Based On
Analyses And Evaluations (Groups 1, 2, and 3)
Criterion 1
Do the proposed changes involve a significant increase in the
probability or consequences of an accident previously evaluated?
No. The analyses which were performed to support the first three
groups of proposed changes were performed in accordance with
approved methodologies and acceptance criteria applicable to Cook
Nuclear Plant. The proposed technical specification changes do not
involve postulated initiators for analyzed events. Therefore, the
probability of accidents can not be affected. The analyses and
evaluations performed all met applicable acceptance criteria.
Therefore, the consequences of accidents previously evaluated are
unaffected.
Criterion 2
Do the proposed changes create the possibility of a new or
different kind of accident from any accident previously evaluated?
No. The analyses which were performed to support the second and
third groups of proposed changes address increases in operating
margin for accident mitigators. They do not create the possibility
of new accidents. The first group of proposed changes to reduce
minimum measured primary flow, increase the DNB [departure from
nucleate boiling] temperature limit, and reduce the reactor coolant
system volume have been analyzed or evaluated. The proposed DNB
limit is consistent with the DNB design and does not constitute an
accident initiator. The new volume results from the new value of
allowed tube plugging and is consistent with the analysis. It is not
an accident initiator.
The impact of the reduced primary flow in the primary system was
analyzed or evaluated, as appropriate. All applicable criteria were
satisfied. No new or different kind of accident resulted.
Criterion 3
Do the proposed changes involve a significant reduction in a
margin of safety?
No. The margin of safety is provided for the primary pressure
boundary and other components in part by applicable design codes.
The margin of safety for the various accidents and transients is
maintained by the analysis acceptance criteria. Since the components
remain in compliance with the codes and standards in effect when
Cook Nuclear Plant was licensed and applicable acceptance criteria
are met, the margin of safety is not reduced by the 30% SGTP [steam
generator tube plugging] program.
Determination Of No Significant Hazards For Administrative
Changes (Group 4)
Criterion 1
Does the proposed change involve a significant increase in the
probability or consequences of an accident previously evaluated?
No. The proposed change involves the surveillances for
mitigating equipment. Therefore, it has no impact on probability.
The proposed change also has no impact on the consequences of an
accident because the criteria for operable RHR [residual heat
removal] and SI [safety injection] pumps does not change. The change
is only in the parameter that will be compared with the required
criteria, the differential pressure instead of the discharge
pressure.
Criterion 2
Does the proposed change create the possibility of a new or
different kind of accident from any accident previously evaluated?
No. Nothing is changed with regard to accident initiators. The
surveillance criteria for the RHR and SI pumps, which are mitigating
equipment, is unchanged. The proposed change can have no impact on
accident initiators.
Criterion 3
Does the proposal involve a significant reduction in a margin of
safety?
No. The proposal does not change the requirements for a pump to
be operable. Only the parameter compared to acceptance criteria
changes. The underlying criteria is unchanged. Therefore, there is
no change in the margin of safety.
Conclusion
It is concluded that operation of Cook Nuclear Plant units 1 and
2 with the changes proposed above does not involve any significant
hazards as defined in 10 CFR 50.92
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 requests involve no significant hazards consideration.
Local Public Document Room location: Maud Preston Palenske Memorial
Library, 500 Market Street, St. Joseph, Michigan 49085
Attorney for licensee: Gerald Charnoff, Esq., Shaw, Pittman, Potts
and Trowbridge, 2300 N Street, NW, Washington, DC 20037
NRC Project Director: John N. Hannon
Indiana Michigan Power Company, Docket Nos. 50-315 and 50-316,
Donald C. Cook Nuclear Plant, Unit Nos. 1 and 2, Berrien County,
Michigan
Date of amendment requests: June 15, 1995 (AEP:NRC:0896V)
Description of amendment requests: The proposed amendments would
change the 18 month emergency diesel generator (EDG) surveillance test
from a 24-hour run to an 8-hour run.
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:
Per 10 CFR 50.92, a proposed change does not involve a
significant hazards consideration if the change does not:
1. involve a significant increase in the probability or
consequences of an accident previously evaluated,
2. create the possibility of a new of different kind of accident
from any accident previously evaluated, or
3. involve a significant reduction in a margin of safety.
Criterion 1
The safety function of the EDGs is to supply ac electrical power
to plant safety systems whenever the preferred ac power supply is
unavailable. Through surveillance requirements, the ability of the
EDGs to meet their load and timing requirements is tested and the
quality of the fuel and the availability of the fuel supply are
monitored. Reduction of the 24 hour run to 8 hours will not reduce
the surveillance factors under consideration and will sufficiently
exercise the EDG and its support systems to identify potential
conditions that could lead to performance degradation. Based on
these considerations, it is concluded that the proposed changes do
not involve a significant increase in the probability or
consequences of an accident previously evaluated.
Criterion 2
The proposed changes do not involve physical changes to the
plant or changes in plant operating configuration. The changes only
involve the reduction of 18 month 24 hour EDG surveillance test
duration. Thus, it is concluded that the proposed changes do not
create the possibility of a new or different kind of accident from
any accident previously evaluated.
Criterion 3
Although the duration of the 18 month 24 hour EDG surveillance
test would be reduced, the EDG components will continue to be
sufficiently exercised such that the ability to detect incipient and
degraded conditions will be maintained. The proposed changes have
been determined to be compatible with our plant operating experience
and commensurate with past surveillance test results. Based on these
considerations, it is concluded that 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 10 CFR 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: Maud Preston Palenske Memorial
Library, 500 Market Street, St. Joseph, Michigan 49085
Attorney for licensee: Gerald Charnoff, Esq., Shaw, Pittman, Potts
and Trowbridge, 2300 N Street, NW, Washington, DC 20037
NRC Project Director: John N. Hannon
Nebraska Public Power District, Docket No. 50-298, Cooper Nuclear
Station, Nemaha County, Nebraska
Date of amendment request: June 15, 1995
Description of amendment request: The proposed amendment would
revise
[[Page 37097]]
the definition for logic system functional test and revise the
surveillance interval for emergency core cooling system logic system
functional testing from 6 months to 18 months.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a) the licensee has provided
its analysis of the issue of no significant hazards consideration,
which is presented below:
1. Does the proposed change involve a significant increase in
the probability or consequences of an accident previously evaluated?
The proposed revisions to change the Cooper Nuclear Station
(CNS) Emergency Core Cooling System (ECCS) logic system functional
testing surveillance intervals from once/6 months to once/18 months
do not involve a significant increase in the probability or
consequences of an accident previously evaluated. The change in
surveillance interval to once/18 months is necessary to coincide
with scheduled refueling outages. The expansion of the scope of the
logic system functional tests will ensure that once/18 months all
contacts providing an automatic safety function in the ECCS logic
systems will be tested. Revising the test frequency to once/18
months will prevent CNS from being required to install jumpers and/
or test blocks during power operation, temporarily rendering various
safety functions inoperable, and potentially challenging safety
systems.
This proposed change will not result in any hardware changes to
the facility, nor will it introduce any new mode of operation.
Conversely, not changing the surveillance frequency would contribute
to a slight, but measurable increase in the probability of an
accident. Therefore, this change will not result in a significant
increase in the probability of any accident previously evaluated.
This change will not result in a significant increase in the
consequences of any accident previously evaluated. The District has
evaluated the change in logic system reliability due to the
increased proposed surveillance interval and determined it to be
negligible. This conclusion is supported by a review of the
surveillance history associated with the ECCS logic system
functional tests which demonstrates that the logic systems perform
reliably. Therefore, this change will not result in a significant
reduction in the reliability or performance of the ECCS, and
therefore, will not result in a significant increase in the
consequences of any accident previously evaluated.
The change to the definition for ``Logic System Functional
Test'' will not result in an increase in the probability or
consequences of any accident previously evaluated. This change will
only provide clarification of the definition for performing these
tests.
These changes are also consistent with the NUREG-1433,
``Standard Technical Specifications, General Electric Plants, BWR/
4,'' dated September, 1992. Therefore, these changes have been
previously reviewed and accepted by the NRC, and have been
implemented at other plants.
2. Does the proposed change create the possibility for a new or
different kind of accident from any accident previously evaluated?
The proposed changes revise the ECCS logic system functional
testing surveillance intervals and the definition of that testing to
be consistent with the Standard Technical Specifications, and
therefore reflect current NRC guidance. The proposed changes do not
involve any plant design changes nor any new mode of operation.
Therefore, these proposed changes do not create the possibility for
a new or different kind of accident from any accident previously
evaluated.
3. Does the proposed change create a significant reduction in
the margin of safety?
The proposed changes to the CNS ECCS logic system functional
testing surveillance intervals do not create a significant reduction
in the margin of safety. As discussed above, the District has
revised its logic system functional testing to ensure that all
contacts providing an automatic safety function in the ECCS logic
systems are tested during this surveillance; thus, this change in
testing scope will ensure that all essential functions in these
logic systems are periodically tested.
The proposed changes will extend the ECCS logic system
functional testing intervals to coincide with refueling outages.
This will prevent CNS from being required to install jumpers and/or
test blocks during power operation which would temporarily defeat
safety system capability, and have the potential of challenging
plant safety systems and/or degrading logic system reliability. The
District has also determined that the change in test frequency will
have a negligible impact on logic system reliability. Therefore,
since these changes will continue to ensure the reliability of the
ECCS logic systems, and thereby the capability of those systems to
respond to accidents, these proposed changes do not create 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: Auburn Public Library, 118
15th Street, Auburn, NE 68305
Attorney for licensee: Mr. G.D. Watson, Nebraska Public Power
District, Post Office Box 499, Columbus, Nebraska 68602-0499
NRC Project Director: William D. Beckner
Northeast Nuclear Energy Company (NNECO), Docket No. 50-245,
Millstone Nuclear Power Station, Unit 1, New London County,
Connecticut
Date of amendment request: June 15, 1995
Description of amendment request: The proposed amendment would
change the definition for an alteration of the reactor core to one that
is consistent with the intent of the improved standard technical
specifications. The proposed amendment also makes administrative
changes to several technical specification pages.
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:
NNECO has reviewed the proposed changes in accordance with
10CFR50.92 and concluded that the changes do not involve a
significant hazards consideration (SHC). The basis for this
conclusion is that the three criteria of 10CFR50.92(c) are not
compromised. The proposed changes do not involve an SHC because the
changes would not:
1. Involve a significant increase in the probability or
consequences of an accident previously analyzed.
Revising the definition of core alteration would not affect the
probability or consequences of a fuel handling accident, since the
movement of fuel within the reactor vessel would still be considered
a CORE ALTERATION. Additionally, movement of a fuel assembly
continues to be performed under the supervision of a senior licensed
operator. Therefore, the potential for inadvertent positioning of a
fuel assembly would not be affected by the change to the definition
of a core alteration.
Other activities which were not specifically excluded as core
alterations in the existing technical specifications are now
excluded. These activities do not affect the reactivity of the core.
Based upon the above, the proposed changes do not involve a
significant increase in the probability or consequences of an
accident previously evaluated.
2. Create the possibility of a new or different kind of accident
from any previously analyzed.
All required systems will continue to operate as before.
Therefore, there is no possibility of a new or different kind of
accident. The change in definition of a core alteration cannot
create the possibility of a new type of accident since those
activities which affect reactivity and could affect the initiating
events for accidents will remain classified as core alterations.
3. Involve a significant reduction in the margin of safety.
Refueling operations which have the potential to alter the
reactivity potential of the core will continue to be defined as core
alterations. The margin of safety associated with those evolutions
will not be altered as a result of the revised definition. As a
result of the revised definition, evolutions which take place within
the reactor vessel core region with the vessel head installed, or
with the reactor vessel completely defueled, will not be considered
core alterations. This does not constitute a reduction in the margin
of safety since there is no impact on core reactivity potential
during these conditions.
[[Page 37098]]
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: Learning Resources Center,
Three Rivers Community-Technical College, 574 New London Turnpike,
Norwich, CT 06360
Attorney for licensee: Ms. L. M. Cuoco, Senior Nuclear Counsel,
Northeast Utilities Service Company, Post Office Box 270, Hartford, CT
06141-0270
NRC Project Director: Phillip F. McKee
North Atlantic Energy Service Corporation, Docket No. 50-443,
Seabrook Station, Unit No. 1, Rockingham County, New Hampshire
Date of amendment request: June 7, 1995
Description of amendment request: The proposed amendment would
increase the temperature limit below which reactor coolant sampling and
analysis for dissolved oxygen is not required. Specifically, the
temperature limit stated in the footnotes to Technical Specification
Surveillance Requirement 4.4.7 and to Table 3.4-2 would be increased to
250 deg.F from 180 deg.F.
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. The
NRC staff has reviewed the licensee's analysis against the standards of
10 CFR 50.92(c). The NRC staff's review is presented below.
A. The changes do not involve a significant increase in the
probability or consequences of an accident previously evaluated (10
CFR 50.92(c)(1)) because the proposed changes merely increase the
temperature limit below which sampling of reactor coolant for
dissolved oxygen and maintaining the dissolved oxygen below the
specified limit would not be required. The proposed limit is
consistent with data which shows that there is no significant
oxygen-induced corrosion to reactor coolant system (RCS) components
at or below the limit. The changes do not affect the manner by which
the facility is operated and do not change any structures, systems,
or components. Since there is no change to the facility or to the
way it is operated, there is no effect upon the probability or
consequences of any accident previously analyzed.
B. The changes do not create the possibility of a new or
different kind of accident from any accident previously evaluated
(10 CFR 50.92(c)(2)) because they do not affect the manner by which
the facility is operated or change any structure, system, or
component. The proposed changes merely raise the temperature limit
above which dissolved oxygen must be maintained within the specified
limit. The changes are consistent with data for oxygen-induced
corrosion of RCS components.
C. The changes do not involve a significant reduction in a
margin of safety (10 CFR 50.92(c)(3)) because the proposed changes
are consistent with data for oxygen-induced corrosion of RCS
components.
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: Exeter Public
Library, Founders Park, Exeter, NH 03833
Attorney for licensee: Thomas Dignan, Esquire, Ropes & Gray, One
International Place, Boston MA 02110-2624
NRC Project Director: Phillip F. McKee
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: June 29, 1995 (Reference LAR 95-04)
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 add Mode 1 applicability to
TS 3/4.4.2.2, ``Safety Valves - Operating,'' and to change the low-
temperature overpressure protection (LTOP) system enable temperature
for Mode 4 applicability from 323 degrees F to 270 degrees F in TS 3/
4.4.2.1, ``Safety Valves - Shutdown.''
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 changes have no effect on plant operation. The
proposed changes correct the applicability of TS 3/4.4.2.2,
consistent with the NRC safety evaluation for License Amendments
(LAs) 98 for Unit 1 and 97 for Unit 2, and LAs 100 for Unit 1 and 99
for Unit 2 dated March 9, 1995, and April 13, 1995, respectively.
Therefore, the proposed changes do not involve a significant
increase in 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.
The proposed changes are administrative in nature. Further, the
proposed changes would not result in any physical alteration to any
plant system, and would not be a change in the method by which any
safety-related system performs its 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 change does not involve a significant reduction
in a margin of safety.
The proposed administrative changes correct TS 3/4.4.2.2
applicability, consistent with previous NRC review and approval of
LAs 98 and 97 and LAs 100 and 99, as described in the associated
safety evaluations. Further, these proposed changes have no effect
on current operating methodologies or actions that govern plant
performance.
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 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, P.O. Box 7442, San Francisco, California 94120
NRC Project Director: William H. Bateman
Philadelphia Electric Company, Docket Nos. 50-352 and 50-353,
Limerick Generating Station, Units 1 and 2, Montgomery County,
Pennsylvania
Date of amendment request: June 5, 1995
Description of amendment request: The proposed changes will revise
Technical Specification (TS) Section 3/4.1.5, ``Standby Liquid Control
System,'' (SLCS), to remove the minimum flow rate requirement for the
SLCS pumps from TS Section 3/4.1.5.
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.
[[Page 37099]]
The proposed TS change will remove the minimum flow rate
requirement for the Standby Liquid Control System (SLCS) pumps from
Technical Specifications Section 3/4.1.5. The proposed TS change
does not involve any physical change in the plant configuration or
the SLCS pumps operation. The SLCS is not used during normal plant
operation; therefore, there is no impact on any accident initiators.
The proposed TS change does not change the plant response to
transients in any way that could increase the likelihood of an
accident. The consequences of previously evaluated accidents are not
affected since the SLCS pumps and the balance of the SLCS will
continue to perform as designed, in accordance with the Anticipated
Transient Without Scram (ATWS) Rule specified in 10CFR50.62. The
SLCS pumps will continue to be tested periodically for operability
in accordance with TS 4.0.5 Surveillance Requirements for American
Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel (B
& PV) Code Class 2 pumps, and the testing frequency remains
unchanged.
Therefore, the proposed TS change does 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 TS change will remove the minimum flow rate
requirement for the Standby Liquid Control System (SLCS) pumps from
Technical Specifications Section 3/4.1.5. The SLCS and the SLCS
pumps will continue to function as currently designed. There are no
physical changes being performed to the SLCS or plant configuration.
The proposed TS change does not introduce a new failure mode for the
SLCS pumps. Physical and electrical redundancy and separation
criteria are not impacted by this proposed TS change. There is no
change to the Redundant Reactivity Control System (RRCS) logic which
could create an accident or transient of a different type.
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 following TS Bases were reviewed for potential reduction in
the margin of safety:
3/4.1.5 Standby Liquid Control System
4.0.5 Surveillance Requirements
The margin of safety as defined in the TS Bases will remain the
same. The specific flow rate requirement for the Standby Liquid
Control System (SLCS) pumps is being removed from the TS since the
Anticipated Transient Without Scram (ATWS) equation ensures
acceptable flow rates. The SLCS pumps, which are safety-related, are
not physically modified or impacted by the proposed TS change. The
pumps will continue to be tested for operability, in accordance with
TS 4.0.5 Surveillance Requirements for ASME B & PV Code Class 2
pumps, and the testing frequency remains unchanged. This testing
will ensure that the SLCS pumps operate in accordance with the
existing design basis for the SLCS.
Therefore, the proposed TS change does 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 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: Pottstown Public Library, 500
High Street, Pottstown, Pennsylvania 19464
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
Sacramento Municipal Utility District (SMUD), Docket No. 50-312,
Rancho Seco Nuclear Station, Sacramento County, California
Date of amendment request: June 20, 1995
Description of amendment request: The proposed amendment (PA-190)
would permit SMUD to change the reviewer qualifications of the
Permanently Defueled Technical Specification (PDTS) D6.5.3 from those
required by ANSI N18.1-1971, Section 4.4 to those of Section 4. In
addition, PDTS D6.9.6b, Environmental Reports, would be changed to
permit annual reporting instead of the current semi-annual schedule.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a) the licensee has reviewed
the proposed changes against each of the no significant hazards
consideration criteria in 10 CFR 50.92, and, based on their safety
analysis, concludes:
A significant increase in the probability or consequences of an
accident previously evaluated will not be created, because the
proposed PDTS changes (1) are administrative in nature, (2) have no
effect on any credible accidents previously evaluated in the Rancho
Seco Defueled Safety Analysis Report (DSAR) (i.e., the dropped fuel
assembly accident, the loss of off-site power condition, or a
radwaste tank rupture), (3) will not reduce the effectiveness of the
reviews conducted because the Rancho Seco Qualified Reviewer
training program ensures Qualified Reviewers have adequate skills to
competently perform the required reviews and the Plant Review
Committee will continue to conduct their second level review
function, and (4) will only affect the timing and management of the
required Environmental Reports submittals to the NRC.
PA-190 will not create the possibility of a new or different
type of accident than previously evaluated, because the proposed
PDTS changes (1) do not modify the configuration of the facility or
affect facility operation during the PDM [permanently defueled
mode], (2) are administrative in nature, and (3) do not provide any
new mechanisms by which an accident can occur.
The proposed PDTS amendment will not involve a significant
reduction in the margin of safety, because the proposed changes do
not affect the operation of Rancho Seco or any plant systems. Also,
The PDTS bases do not rely on (1) Qualified Reviewer qualification
requirements or (2) submittal of PDTS D6.9.6b Environmental Reports
to the NRC to provide a margin of safety for plant operation during
the PDM. The Rancho Seco Qualified Reviewer program relies on
training and not the ANSI N18.1 qualification requirements to ensure
the PDTS D6.5.3 required reviews are competently performed.
Therefore, the proposed changes will 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 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: Central Library, Government
Documents, 828 I Street, Sacramento, CA 95814
Attorney for licensee: Dana Appling, Esq., Sacramento Municipal
Utility District, P. O. Box 15830, Sacramento, CA 95852-1830
NRC Project Director: Seymour H. Weiss
South Carolina Electric & Gas Company, South Carolina Public
ServiceAuthority, Docket No. 50-395, Virgil C. Summer Nuclear
Station, Unit No. 1, Fairfield County, South Carolina
Date of amendment request: June 19, 1995
Description of amendment request: The proposed amendment would
revise the Technical Specifications (TS) to delete the scheduler
requirements for Type A testing (Overall Integrated Containment Leakage
Rate) to be performed at 40 plus or minus 10 month intervals and to
delete the schedular requirements for Type B and C tests to be
performed at 24 month intervals.
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 probability or consequences of an accident previously
evaluated is not significantly increased.
There is no increase in the probability of an accident since
there is no work planned that would affect containment integrity.
The testing of containment isolation valves and
[[Page 37100]]
other containment penetration sealing devices is not postulated as an
accident precursor or initiating event.
Type A testing is capable of determining the total leakage from
both local leak paths as well as gross containment leakage paths.
Our Type B and C testing has consistently provided accurate leakage
rates for valves and penetrations.
Administrative controls govern maintenance and testing such that
there is very low probability that unacceptable maintenance or
alignments can occur. After maintenance on containment isolation
valves (CIVs) and penetrations, a local leak rate test (LLRT) is
required to be performed. All work on valves also requires that an
independent valve lineup be performed. As a result, Type A testing
is not required to accurately quantify the leakage through
containment penetrations.
Any specific exemptions to the requirements of Appendix J will
require approval by the NRC before implementation. The proposed
change in itself does not affect reactor operations and does not
change radiological consequences.
Therefore, this proposed change does not involve a significant
increase in the possibility or consequences of an accident
previously evaluated.
2. The possibility of an accident or a malfunction of a
different type than any previously evaluated is not created.
The proposed TS change request (TSCR) does not involve any
physical changes to the plant, affect the operation of the plant, or
change testing methods or acceptance criteria. The history of
containment testing verifies that containment integrity has been
maintained.
The scheduler change that is proposed should not significantly
decrease the level of confidence in the ability of the reactor
building to limit offsite doses to allowable values. No accident or
malfunction can be the result of the change in test schedule or
frequency.
Since the proposed TSCR will not directly impact equipment,
procedures or operations, the changes will not create the
possibility of any new or different kind of accident from any
previously evaluated.
3. The margin of safety has not been significantly reduced.
The reason for performing ILRTs [integrated leakage rate tests]
is to assure that the leakage paths are identified, and any accident
release will be restricted to those paths assumed in
the safety analysis. The purpose for the schedule is to assure
that containment integrity is verified on a periodic basis.
Revising the schedule does not mean that containment integrity
will be compromised. Type B and C testing will still be performed.
The requirements in 10 CFR 50 Appendix J still require the testing
to be performed periodically.
The testing previously performed has shown that acceptable
results were obtained. The ILRT results minus the LLRT results
demonstrate that most of the increases in leakage are the result of
LLRT increases. These changes in Type B and C leakage are tracked
and corrective action is initiated at a specific action level.
Therefore, the margin of safety has not been significantly
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: Fairfield County Library, 300
Washington Street, Winnsboro, SC 29180
Attorney for licensee: Randolph R. Mahan, South Carolina Electric &
Gas Company, Post Office Box 764, Columbia, South Carolina 29218
NRC Project Director: Frederick J. Hebdon
South Carolina Electric & Gas Company, South Carolina Public
Service Authority, Docket No. 50-395, Virgil C. Summer Nuclear
Station, Unit No. 1, Fairfield County, South Carolina
Date of amendment request: June 19, 1995
Description of amendment request: The proposed amendment would
revise the Technical Specifications to change the required test
frequency for the reactor building spray nozzle flow test from once per
five years to once per ten years.
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 probability or consequences of an accident previously
evaluated is not significantly increased.
This change does not effect the probability or consequences of
an accident. The Reactor Building Spray System is normally idle,
with the exception of testing. The possibility of the introduction
of foreign material or corrosion products to restrict flow is
minimized because of the use of 304 stainless steel as construction
material. This change results in an extension of the testing
periodicity only.
2. The possibility of an accident or a malfunction of a
different type than any previously evaluated is not created.
This change results in an extension of the testing periodicity
only and does not result in an accident not previously evaluated.
3. The margin of safety has not been significantly reduced.
The Reactor Building Spray System is normally idle, with the
exception of testing. The possibility of the introduction of foreign
material or corrosion products to restrict flow is minimized because
of the use of 304 stainless steel as construction material. Industry
wide spray system reliability, as demonstrated by the performance of
these tests, justifies this change in the frequency of the nozzle
flow test. This results in an extension of the testing periodicity
only and will not significantly reduce 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: Fairfield County Library, 300
Washington Street, Winnsboro, SC 29180
Attorney for licensee: Randolph R. Mahan, South Carolina Electric &
Gas Company, Post Office Box 764, Columbia, South Carolina 29218
NRC Project Director: Frederick J. Hebdon
Tennessee Valley Authority, Docket Nos. 50-327 and 50-328, Sequoyah
Nuclear Plant, Units 1 and 2, Hamilton County, Tennessee
Date of amendment request: June 29, 1995 (TS 95-14)
Description of amendment request: The proposed change would, under
certain stated administrative controls, allow both sets of containment
personnel airlock doors to be open during core alterations and fuel
movements. The administrative controls that would be added to Limiting
Condition for Operation 3.9.4.b would allow both airlock doors to be
open if one personnel airlock door in each airlock is capable of
closure, and one train of the Auxiliary Building Gas Treatment System
is operable in accordance with Specification 3.9.12. In addition,
proposed changes to Surveillance Requirement 4.9.4 and 4.9.4.a would
replace the requirement to determine that the containment building
penetrations are in the ``closed/isolated'' condition with the need to
determine that they are in the ``required'' condition, and delete the
requirement to verify that the penetrations are in their required
condition and the requirement to test the Containment Ventilation
isolation valves ``within 100 hours prior to the start of'' core
alterations or movement of irradiated fuel in the containment building.
Related changes to the Bases would supply amplifying information.
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:
TVA has evaluated the proposed technical specification (TS)
change and has determined
[[Page 37101]]
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 change to TS 3.9.4, Containment Building
Penetrations, would allow the containment personnel airlocks (PALs)
to be open during fuel movement and core alteration. The PALs are
not an initiator to any accident. The position of the PAL doors
(open or closed) during fuel movement and core alterations has no
affect on the probability of any accident previously evaluated.
All doses from a fuel handling accident (FHA) for the proposed
change remain well below the 10 CFR 100 limits. The proposed change
will reduce the dose to workers inside containment in the event of a
FHA by allowing more rapid egress from containment. The wear on the
PAL doors will significantly be decreased; therefore, increasing the
reliability of the PAL doors in the event of an accident.
Since the probability of a FHA is not affected by the airlock
door positions, and the doses remain within acceptable limits, the
proposed change does not involve a significant increase in the
probability or consequences of an accident previously evaluated.
2. Create the possibility of a new or different kind of accident
from any previously analyzed.
As previously stated, the PAL doors are not accident initiators.
The open PAL doors do not represent a significant change in the
configuration of the plant; therefore, does not create a new or
different type of accident from any previously analyzed.
3. Involve a significant reduction in a margin of safety.
The margin of safety provided for an FHA inside containment
remains well below the 10 CFR 100 limits. Therefore, this proposed
change to allow the PAL doors to remain open during fuel movement or
core alterations does not involve a significant reduction in the
margin of safety.
The NRC has reviewed the licensee's analysis and, based on
thisreview, 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
Washington Public Power Supply System, Docket No. 50-397, Nuclear
Project No. 2, Benton County, Washington
Date of amendment request: September 2, 1992
Description of amendment request: The proposed amendment revises
the surveillance criteria for the source range monitors (SRMs) to
incorporate a more conservative signal-to-noise (S/N) ratio, as
recommended by General Electric for this system.
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. The
NRC staff has reviewed the licensee's analysis against the standards of
10 CFR 50.92(c). The NRC staff's review is presented below:
1.
The proposed change does not involve a significant increase in the
probability or consequences of an accident previously evaluated.
The SRM instrumentation is not assumed to be an initiator of any
analyzed event. The SRM instrumentation provides monitoring of neutron
flux levels to give the control room operator early indication of
unexpected subcritical multiplication that could be indicative of an
approach to criticality. As such, action could be taken on the
indication to avert or minimize the consequences of the event. However,
the SRM function is not relied upon in any design bases or transient
analysis. Rod motion interlocks and other instrumentation are relied on
in the accident analysis to avert an accident. The change in acceptable
count rate and signal-to-noise ratio preserves the confidence level of
the General Electric design. As a result, the consequences of any
analyzed events are unaffected because the change does not alter any
system or component design assumptions or operation. Therefore, no
significant increase in the probability or consequences of an accident
previously evaluated will be involved.
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 in SRM count rate and S/N ratio values does not
change modes of plant operation or require physical modifications. The
WNP-2 design basis accident and transient analyses do not rely on the
SRMs to assume plant safety. Therefore, the proposed change does not
create the possibility of a new or different kind of accident.
3.
The proposed change does not involve a significant reduction in a
margin of safety.
The proposed change does not involve a significant reduction in a
margin of safety. The design basis to assure SRM operability is based
on an instrument count rate that will assure the SRMs will provide
early indication of subcritical multiplication with a 95-percent
confidence level. Requiring the count rate to be greater than or equal
to 0.7 counts per second (cps) with a S/N ratio greater than or equal
to 20, or greater than or equal to 3 cps with a S/N ratio greater than
or equal to 2 (vs. a count rate of greater than or equal to 0.5 cps
with a S/N ratio greater than or equal to 2 in current TS) ensures the
design 95-percent confidence level is maintained when verifying SRM
operability. Therefore, the margin of safety is not affected by this
change.
Based on this review, it appears that the three standards of
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: Richland Public Library, 955
Northgate Street, Richland, Washington 99352
Attorney for licensee: M. H. Philips, Jr., Esq., Winston & Strawn,
1400 L Street, N.W., Washington, D.C. 20005-3502
NRC Project Director: William H. Bateman
Washington Public Power Supply System, Docket No. 50-397, Nuclear
Project No. 2, Benton County, Washington
Date of amendment request: June 6, 1995
Description of amendment request: The proposed amendment would
change Technical Specification 6.9.3.2. The change would add references
to three topical reports describing analytical methods that may be used
in determining reactor core operating limits for reload licensing
applications.
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 amendment does not remove or modify existing
Technical Specification
[[Page 37102]]
requirements or safety limits. The Technical Specifications will
continue to require operations within analyzed core operating limits
and appropriate actions be taken when, or if, limits are exceeded.
There will be no changes to the physical design of the plant as a
result of adding the proposed references to Section 6.9.3.2. The
results of analytical determination of core operating limitations is
not assumed as the initiator of any analyzed event, and the approved
safety analysis is still applicable. Therefore, the proposed
amendment to Technical Specification 6.9.3.2 does not involve a
significant increase in 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.
The proposed amendment does not remove or modify existing
Technical Specification requirements or safety limits. The Technical
Specifications will continue to require operation within analyzed
core operating limits and appropriate actions be taken when, or if,
limits are exceeded. The technical methodology outlined in the three
new reports is in accordance with the accepted principals, and the
specific reports proposed for inclusion in the Technical
Specifications by this request have been previously approved by NRC
for use at WNP-2 as a basis for core reload analyses. Therefore, the
proposed amendment to Technical Specification 6.9.3.2 does not
create the possibility of a new or different type of accident from
any accident previously evaluated.
3. The proposed change does not involve a significant reduction
in a margin of safety.
Plant safety limits are established through LCOs, limiting
safety systems settings, and safety limits specified in the
Technical Specifications. There will be no changes to either the
physical design of the plant or to any of these settings and limits
as a result of adding the proposed references to Section 6.9.3.2.
The ability to mitigate the consequences of all accidents previously
evaluated will be maintained and nuclear safety is not adversely
affected because the technical methodology outlined in the three new
reports is in accordance with accepted principals, and the specific
reports proposed for inclusion in the Technical Specifications by
this request have been previously approved by NRC for use at WNP-2
as a basis for core reload analyses. Therefore, the proposed
amendment to Technical Specification 6.9.3.2 does not significantly
reduce any margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 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: Richland Public Library, 955
Northgate Street, Richland, Washington 99352
Attorney for licensee: M. H. Philips, Jr., Esq., Winston & Strawn,
1400 L Street, N.W., Washington, D.C. 20005-3502
NRC Project Director: William H. Bateman
Washington Public Power Supply System, Docket No. 50-397, Nuclear
Project No. 2, Benton County, Washington
Date of amendment request: June 6, 1995
Description of amendment request: The proposed amendment would
change the Index of the WNP-2 technical specifications by deleting
reference to the Bases pages. Consistent with the requirements of 10
CFR 50.36(a), which states that the Bases shall not become part of the
technical specifications, the Bases information will be consolidated
into a controlled plant document.
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 changes are administrative and do not remove or
modify existing Technical Specification requirements or safety
limits. There will be no changes to the physical design of the plant
as a result of the proposed change. The Bases information, per 10
CFR 50.36(a), is not part of the Technical Specifications and will
be consolidated into a controlled plant document. Future changes to
the Bases will be evaluated per 10 CFR 50.59. Therefore, the
proposed changes to the Technical Specification Index do not involve
a significant increase in 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.
The proposed changes are administrative and do not remove or
modify existing Technical Specification requirements or safety
limits. There will be no changes to the physical design of the plant
or alteration of any operational practice as a result of the
proposed change. The Bases information, per 10 CFR 50.36(a), is not
part of the Technical Specifications and will be consolidated into a
controlled document. Future changes to the Bases will be evaluated
under 10 CFR 50.59. Therefore, the proposed changes to the Technical
Specifications Index do not create the possibility of a new or
different type of accident.
3. The proposed change does not involve a significant reduction
in a margin of safety.
Plant safety limits are established through LCOs, limiting
safety system settings, and safety limits specified in the Technical
Specifications. There will be no changes to either the physical
design of the plant or to any of these settings and limits as a
result of modifying the Technical Specification Index. The ability
to mitigate the consequences of all accidents previously evaluated
will be maintained and nuclear safety is not impacted. The Bases
information, per 10 CFR 50.36(a), is not part of the Technical
Specifications and will be consolidated into a controlled document.
Future changes to the Bases will be evaluated under 10 CFR 50.59.
Therefore, the proposed amendment does not significantly reduce any
margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 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: Richland Public Library, 955
Northgate Street, Richland, Washington 99352
Attorney for licensee: M. H. Philips, Jr., Esq., Winston & Strawn,
1400 L Street, N.W., Washington, D.C. 20005-3502
NRC Project Director: William H. Bateman
Washington Public Power Supply System, Docket No. 50-397, Nuclear
Project No. 2, Benton County, Washington
Date of amendment request: June 6, 1995
Description of amendment request: The proposed amendment would
change Technical Specification (TS) 6.0, ``Administrative Controls''
for WNP-2. Specifically, the changes would (a) reflect Supply System
titles for senior management throughout TS 6.0, (b) modify the Plant
Operations Committee (POC) composition to specify members according to
functional areas rather than by organizational titles (c) replace the
Plant Manager as the POC Chairman with an individual appointed by the
Plant General Manager, and (d) make an editorial correction.
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 senior management title changes are title changes only and
will not impact the plant safety responsibilities associated with
these positions. The removal of the Plant Operations Committee (POC)
organizational
[[Page 37103]]
titles and replacement with functional areas, and the elimination of
the Plant Manager as the POC Chairman, will not impact the POC
function because membership qualifications will continue to be
consistent with the unit staff qualifications in TS 6.3.1 for those
POC members and alternates considered part of the unit staff. Those
designated POC members and alternates not considered part of the
unit staff will possess skills and knowledge commensurate with their
organizational positions. The proposed change ensures that POC will
continue to be comprised of personnel who are experienced, have
varied expertise, and are involved in daily plant activities. In
maintaining the qualification requirements for members of POC, the
POC will continue to fulfill its review and advisory
responsibilities specified in TS 6.5.1.6 and TS 6.5.1.7. The
proposed changes do not involve any physical changes to plant
systems, structures, or components (SSC) or the manner in which the
SSC are operated, maintained, modified, tested, or inspected. The
changes therefore do not involve a significant increase in 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.
Because the proposed changes are of an organizational nature and
their implementation does not involve physical changes to the plant
SSC or the manner in which the SSC are operated and maintained, the
proposed changes do not create the possibility of a new or different
kind of accident. The proposed changes do not introduce any new
modes of operation or alter system setpoints which could create a
new or different kind of accident. Therefore, the possibility of a
new or different kind of accident from any accident previously
evaluated is not created.
3. The proposed change does not involve a significant reduction
in a margin of safety.
The senior management title changes do not impact the management
responsibilities or functions associated with ensuring plant safety.
Changes proposed in the POC composition will allow the scope of
available expertise to be expanded without changing the POC function
or responsibilities. Maintaining the current level of personnel
qualifications and experience ensures the POC will continue to meet
its TS review and advisory requirements. The proposed changes will
not impact the basis for any Technical Specification related to the
establishment of, or maintenance of, nuclear safety margins.
Therefore, operation of the facility in accordance with the proposed
amendment does not involve a significant reduction in a margin of
safety.
The NRC staff has revieywed the licensee's analysis and, based on
this review, it appears that the three standards of 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: Richland Public Library, 955
Northgate Street, Richland, Washington 99352
Attorney for licensee: M. H. Philips, Jr., Esq., Winston & Strawn,
1400 L Street, N.W., Washington, D.C. 20005-3502
NRC Project Director: William H. Bateman
Notice Of Issuance Of Amendments To Facility Operating Licenses
During the period since publication of the last biweekly notice,
the Commission has issued the following amendments. The Commission has
determined for each of these amendments that the application complies
with the standards and requirements of the Atomic Energy Act of 1954,
as amended (the Act), and the Commission's rules and regulations. The
Commission has made appropriate findings as required by the Act and the
Commission's rules and regulations in 10 CFR Chapter I, which are set
forth in the license amendment.
Notice of Consideration of Issuance of Amendment to Facility
Operating License, Proposed No Significant 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.
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: March 28, 1994
Brief description of amendments: The amendments change the minimum
condensate storage tank indicated level from 25 feet to 29.5 feet to
ensure that the condensate storage tank contains a sufficient volume of
water. In addition, an editorial change was made to Technical
Specification 3.7.1.3 for Unit 3 to be consistent with Units' 1 and 2
technical specifications.
Date of issuance: July 6, 1995
Effective date: July 6, 1995
Amendment Nos.: Unit 1 - Amendment No. 94; Unit 2 - Amendment No.
82; Unit 3 - Amendment No. 65
Facility Operating License Nos. NPF-41, NPF-51, and NPF-74: The
amendments revised the Technical Specifications.
Date of initial notice in Federal Register: June 8, 1994 (59 FR
29625) The Commission's related evaluation of the amendments is
contained in a Safety Evaluation dated July 6, 1995.No significant
hazards consideration comments received: No.
Local Public Document Room location: Phoenix Public Library, 12
East McDowell Road, Phoenix, Arizona 85004.
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, as supplemented
by letter dated June 7, 1995.
Brief description of amendments: These amendments involve
improvements delineated in Generic Letter 93-07, ``Modification of the
Technical Specification Administrative Control Requirements for
Emergency and Security Plans,'' changes in plant review board, and
miscellaneous minor changes.
Date of issuance: July 7, 1995
Effective date: July 7, 1995
Amendment Nos.: Unit 1 - Amendment No. 95; Unit 2 - Amendment No.
83; Unit 3 - Amendment No. 66
Facility Operating License Nos. NPF-41, NPF-51, and NPF-74: The
amendments revised the Technical Specifications.
Date of initial notice in Federal Register: May 23, 1995 (60 FR
27335) The June 7, 1995, letter provided clarifying information and did
not change the initial no sigificant hazards consideration
determination. The Commission's related evaluation of the amendments is
contained in a Safety Evaluation dated July 7, 1995.No
[[Page 37104]]
significant hazards consideration comments received: No.
Local Public Document Room location: Phoenix Public Library, 12
East McDowell Road, Phoenix, Arizona 85004.
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 amendments: February 16, 1993, as
supplemented by letter dated May 2, 1995.
Brief description of amendments: This application upgrades the
current custom Technical Specifications (TS) for Dresden and Quad
Cities to the Standard Technical Specification contained in NUREG-0123,
``Standard Technical Specification General Electric Plants BWR/4.''
This application upgrades only Section 3/4.10 (Refueling Operations).
Date of issuance: June 23, 1995
Effective date: Immediately, to be implemented no later than
December 31, 1995, for Dresden Station and June 30, 1996, for Quad
Cities Station.
Amendment Nos.: 136, 130, 157, and 153
Facility Operating License Nos. DPR-19, DPR-25, DPR-29 and DPR-30.
The amendments revised the Technical Specifications.
Date of initial notice in Federal Register: May 23, 1995 (60 FR
27337) The Commission's related evaluation of the amendments is
contained in a Safety Evaluation dated June 23, 1995. No significant
hazards consideration comments received: No 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.
Commonwealth Edison Company, Docket Nos. 50-295 and 50-304, Zion
Nuclear Power Station Units 1 and 2, Lake County, Illinois
Date of application for amendments: December 23, 1994
Brief description of amendments: The amendments revise the
Technical Specifications by increasing the allowable U-235 enrichment
of fuel to be stored in the new fuel storage vault.
Date of issuance: June 22, 1995
Effective date: June 22, 1995
Amendment Nos.: 164 and 152
Facility Operating License Nos. DPR-39 and DPR-48: The amendments
revised the Technical Specifications.
Date of initial notice in Federal Register: February 15, 1995 (60
FR 8742) The Commission's related evaluation of the amendments is
contained in an Environmental Assessment dated June 8, 1995, and a
Safety Evaluation dated June 22, 1995. No significant hazards
consideration comments received: No
Local Public Document Room location: Waukegan Public Library, 128
N. County Street, Waukegan, Illinois 60085.
Commonwealth Edison Company, Docket Nos. 50-295 and 50-304, Zion
Nuclear Power Station Units 1 and 2, Lake County, Illinois
Date of application for amendments: March 24, 1995
Brief description of amendments: The amendments recognize
performing containment leakage rate tests in accordance with 10 CFR
Part 50, Appendix J, and approved exemptions.
Date of issuance: June 30, 1995
Effective date: June 30, 1995
Amendment Nos.: 165 and 153
Facility Operating License Nos. DPR-39 and DPR-48: The amendments
revised the Technical Specifications.
Date of initial notice in Federal Register: April 26, 1995 (60 FR
20516) The Commission's related evaluation of the amendments i
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