Biweekly Notice
Federal RegisterMay 10, 1995
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UNITED STATES NUCLEAR REGULATORY COMMISSION
Biweekly Notice
Applications and Amendments to Facility Operating LicensesInvolving
No Significant Hazards Considerations
I. Background
Pursuant to Public Law 97-415, the U.S. Nuclear Regulatory
Commission (the Commission or NRC staff) is publishing this regular
biweekly notice. Public Law 97-415 revised section 189 of the Atomic
Energy Act of 1954, as amended (the Act), to require the Commission to
publish notice of any amendments issued, or proposed to be issued,
under a new provision of section 189 of the Act. This provision grants
the Commission the authority to issue and make immediately effective
any amendment to an operating license upon a determination by the
Commission that such amendment involves no significant hazards
consideration, notwithstanding the pendency before the Commission of a
request for a hearing from any person.
This biweekly notice includes all notices of amendments issued, or
proposed to be issued from April 17, 1995, through April 28, 1995. The
last biweekly notice was published on April 26, 1995. [[Page 24905]]
Notice of Consideration of Issuance of Amendments to Facility
Operating Licenses, Proposed No Significant Hazards Consideration
Determination, and Opportunity for a Hearing
The Commission has made a proposed determination that the following
amendment requests involve no significant hazards consideration. Under
the Commission's regulations in 10 CFR 50.92, this means that operation
of the facility in accordance with the proposed amendment would not (1)
involve a significant increase in the probability or consequences of an
accident previously evaluated; or (2) create the possibility of a new
or different kind of accident from any accident previously evaluated;
or (3) involve a significant reduction in a margin of safety. The basis
for this proposed determination for each amendment request is shown
below.
The Commission is seeking public comments on this proposed
determination. Any comments received within 30 days after the date of
publication of this notice will be considered in making any final
determination.
Normally, the Commission will not issue the amendment until the
expiration of the 30-day notice period. However, should circumstances
change during the notice period such that failure to act in a timely
way would result, for example, in derating or shutdown of the facility,
the Commission may issue the license amendment before the expiration of
the 30-day notice period, provided that its final determination is that
the amendment involves no significant hazards consideration. The final
determination will consider all public and State comments received
before action is taken. Should the Commission take this action, it will
publish in the Federal Register a notice of issuance and provide for
opportunity for a hearing after issuance. The Commission expects that
the need to take this action will occur very infrequently.
Written comments may be submitted by mail to the Rules Review and
Directives Branch, Division of Freedom of Information and Publications
Services, Office of Administration, U.S. Nuclear Regulatory Commission,
Washington, DC 20555, and should cite the publication date and page
number of this Federal Register notice. Written comments may also be
delivered to Room 6D22, Two White Flint North, 11545 Rockville Pike,
Rockville, Maryland from 7:30 a.m. to 4:15 p.m. Federal workdays.
Copies of written comments received may be examined at the NRC Public
Document Room, the Gelman Building, 2120 L Street, NW., Washington, DC.
The filing of requests for a hearing and petitions for leave to
intervene is discussed below.
By June 9, 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
[[Page 24906]] Document Room, the Gelman Building, 2120 L Street, NW.,
Washington DC, by the above date. Where petitions are filed during the
last 10 days of the notice period, it is requested that the petitioner
promptly so inform the Commission by a toll-free telephone call to
Western Union at 1-(800) 248-5100 (in Missouri 1-(800) 342-6700). The
Western Union operator should be given Datagram Identification Number
N1023 and the following message addressed to (Project Director):
petitioner's name and telephone number, date petition was mailed, plant
name, and publication date and page number of this Federal Register
notice. A copy of the petition should also be sent to the Office of the
General Counsel, U.S. Nuclear Regulatory Commission, Washington, DC
20555, and to the attorney for the licensee.
Nontimely filings of petitions for leave to intervene, amended
petitions, supplemental petitions and/or requests for a hearing will
not be entertained absent a determination by the Commission, the
presiding officer or the Atomic Safety and Licensing Board that the
petition and/or request should be granted based upon a balancing of
factors specified in 10 CFR 2.714(a)(1)(i)-(v) and 2.714(d).
For further details with respect to this action, see the
application for amendment which is available for public inspection at
the Commission's Public Document Room, the Gelman Building, 2120 L
Street, NW., Washington, DC, and at the local public document room for
the particular facility involved.
Carolina Power & Light Company, et al., Docket No. 50-400, Shearon
Harris Nuclear Power Plant, Unit 1, Wake and Chatham Counties,
North Carolina
Date of amendment request: April 5, 1995
Description of amendment request: The licensee proposes to revise
Technical Specification (TS) 3/4.9, Refueling Operations, to be
consistent with NUREG-1431, Standard Technical Specifications,
Westinghouse Plants, and to relocate the applicable sections from the
TS that do not meet the Commission's screening criteria for retention.
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:
This change does not involve a significant hazards consideration
for the following reasons:
The proposed amendment does not involve a significant increase
in the probability or consequences of an accident previously
evaluated.
The proposed changes will have no significant impact on the
safety, reliability, or operation of fuel handling equipment or
activities. These changes will simplify the Technical Specifications
and implement the recommendations of the Commission's Final Policy
Statement on Technical Specification Improvements based upon the
assumptions and analyses contained in the bases of NUREG-1431. Those
elements that involve relocations to plant procedures are
administrative in nature and do not involve any modifications to
plant equipment or operation. Therefore, there would be no increase
in the probability or consequences of an accident previously
evaluated.
2. The proposed amendment does not create the possibility of a
new or different kind of accident from any accident previously
evaluated.
The proposed changes do not introduce any new equipment or
require existing equipment to operate to perform a function
different from that previously evaluated in the Final Safety
Analysis Report or Technical Specifications. The changes are
consistent with the new Standard Techical Specification and
assumptions contained in NUREG-1431 and in the Commission's Final
Policy Statement on Technical Specification Improvements. Therefore,
the proposed changes would not increase the possibility of a new or
different type of accident from any accident previously evaluated.
3. The proposed amendment does not involve a significant
reduction in the margin of safety.
The proposed changes do not affect any of the parameters which
relate to the margin of safety as described in the [Bases] of the
Technical Specifications or the Final Safety Analysis Report.
Accordingly, NRC Acceptance Limits are not affected by these
changes. For those specifications being relocated to other plant
documents, these changes are purely administrative. 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: Cameron Village Regional
Library, 1930 Clark Avenue, Raleigh, North Carolina 27605
Attorney for licensee: R. E. Jones, General Counsel, Carolina Power
& Light Company, Post Office Box 1551, Raleigh, North Carolina 27602
NRC Project Director: David B. Matthews
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 15, 1992, as
supplemented April 21, 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
(TSs).
The licensee made the decision to initiate a Technical
Specification Upgrade Program (TSUP) for both Dresden and Quad Cities.
The licensee evaluated the current TSs for both Dresden and Quad Cities
against the Standard Technical Specifications (STSs) 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 TSs 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 TSs to the STSs contained in NUREG-
0123.
The TSUP for Dresden and Quad Cities is not a complete adaption of
the STSs. 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 TSs based on the
licensee's responses to Generic Letters (GLs), and (4) relocating
specific items to more appropriate TS locations.
The application dated September 15, 1992, as supplemented April 21,
1995, proposed to upgrade only Sections 2.0 (Safety Limits and Limiting
Safety System Settings), 3/4.11 (Power Distribution Limits), and 3/4.12
(Special Test Exceptions) 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 [[Page 24907]] issue of no significant
hazards consideration, which is presented below:
Section 2.0
The proposed changes do not involve a significant increase in
the probability or consequences of an accident previously evaluated
because:
The proposed changes to Specifications 1/2.1 and 1/2.2 to delete
the present Applicability and Objective sections represent
administrative changes to format and presentation of material. The
proposed changes provide the user with a format that will allow
better access to needed information and provides concise Safety
Limit, Limiting Safety System Settings, Applicability and Action
requirements. The additions of Applicability and Action requirements
represent clarification of intended requirements that do not
presently state all required conditions of operability or provide
clearly stated Action statements if the requirements are not met.
The combining of the two sections and added requirements follow STS
guidelines that are in use at many operating BWRs with similar
design and operating configurations as Dresden and Quad Cities
Stations. Operability requirements for Safety Limits have been
chosen to reflect only those Operational Modes where the Safety
Limits apply. Operability requirements for Limiting Safety System
Settings are already stated in other sections of the Technical
Specifications, thus reference to the appropriate operability
requirement is made rather than repeating the requirement in the
Limiting Safety System Setting Specification.
Deletion of the Power Transient Safety Limit does not impact any
safety analyses. The safety analyses assume the Reactor Protection
System (RPS) operates as designed and the reactor scrams when the
neutron flux exceeds the limiting safety system setting. The
proposed Technical Specifications will continue to provide a highly
reliable system to operate as assumed in the safety analyses.
Therefore, this change does not involve a significant increase in
the probability or consequences of an accident previously evaluated.
The reactor water level low scram setpoint is changed (for Quad
Cities) to be consistent with other reactor water level setpoints in
the Technical Specifications and the STS. The setpoint is equivalent
to the current requirement but is expressed as the reactor water
level above the top of active fuel.
The scram discharge volume scram level is converted for Dresden
Unit 2 and Unit 3 to gallons to be consistent with the Quad Cities
Units. The proposed setpoints are consistent with the current
specifications. The change in the units does not represent a change
in the physical setpoint.
The proposed change to delete the APRM Downnscale Scram trip
function for Quad Cities has been evaluated by Commonwealth Edison
and General Electric and previously approved for Dresden Station.
The events of concern with respect to the APRM/IRM companion trip
are the Control Rod Drop Accident and the low power Rod Withdrawal
Error. The FSAR and reload safety analyses do not credit this scram
function in the termination of either of these events. Since this
scram function is not credited in the termination of these events,
the elimination of this scram function has no adverse effect on
previously evaluated accidents.
The change to the low condenser vacuum scram setpoint from 23
inches Hg to 21 inches of Hg is consistent with an identical change
made to Quad Cities Units 1 and 2. The low condenser vacuum scram is
an anticipatory scram and is not credited in any transient analysis.
Thus the reduction in the setpoint will not affect any transient
analysis.
The proposed changes do not alter the intent of existing
setpoints or accident assumptions and follow existing requirements
at other operating BWRs for operability and Action statements.
Therefore, the proposed changes do not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
The proposed changes do not create the possibility of a new or
different kind of accident from any previously evaluated because:
The proposed administrative changes to the format and
arrangement of material do not affect technical requirements or
assumptions of any potential accident and; therefore, cannot create
the possibility of a new or different kind of accident from any
previously evaluated.
The proposed addition of Applicability and Action requirements
enhance the understanding and usability of the Technical
Specifications and thus represent an improvement over present
specifications. New requirements are modeled after those in use at
operating BWRs and do not represent requirements that will adversely
affect potential accident analyses or assumptions. Therefore, the
proposed changes do not create the possibility of a new or different
kind of accident from any previously evaluated.
Deletion of the Power Transient Safety Limit does not involve a
change in the design or operation of any systems assumed to operate
in the safety analyses. Therefore, this change does not involve a
significant increase in the probability or consequences of an
accident previously evaluated.
The change in the units for the Reactor Water Level scram
function do not change any physical plant setpoints. The setpoint
will remain the same but will be expressed as the level above the
top of active fuel. The change does not create the possibility of a
new or different kind of accident.
The conversion of the Scram Discharge Volume scram setpoint from
inches to gallons does not alter any physical plant setpoints. The
setpoint will remain the same but will be expressed in gallons
rather than inches. The change will provide consistency between
Dresden and Quad Cities.
The deletion of the APRM Downscale Scram Trip Function does not
introduce any new accident. The limiting accidents, Control Rod
Drop, Rod Withdrawal Error, in the operating region of transition
between the Startup and Run Operational Modes are well understood
and are evaluated in FSAR and reload analyses. Other control rod
initiated events which are less limiting in this region are subsets
of the low power Rod Withdrawal Error event and are bounded by it
and the design basis Control Rod Drop Accident. General Electric has
indicated that, for reactivity insertion mechanisms at very low
power, the only effect of the deletion of the APRM downscale scram
would be that the initial power level could be a few percent lower
which would not have a significant effect on the severity of the
event. In addition, proper overlap between the IRMs and APRMs is not
affected since the calibration requirements are not being changed.
The change in the low condenser vacuum scram function will not
create the possibility of a new or different kind of accident
because the function is not recognized in any of the transient
analysis. The low condenser vacuum scram function is an anticipatory
scram.
The proposed changes do not involve a significant reduction in
the margin of safety because:
The proposed administrative changes to format, arrangement of
material, clarification of requirements and other non-technical
changes do not affect any safety aspects of the plant and as such
can not involve a significant reduction in the margin of safety.
The proposed Applicability statements require availability of
Safety Limits and Limiting Safety System Settings when required to
perform their respective functions. Proposed Actions for Safety
Limits allow only 2 hours to be in Hot Shutdown and then reference
Specification 6.4 to ensure that proper reports are made and restart
is prohibited until approved by the NRC. These provisions help
ensure that present margins are not significantly reduced.
Deletion of the Power Transient Safety Limit does not impact the
margin assumed in the safety analyses. The safety analyses assume
the RPS operates as designed and the reactor scrams when the neutron
flux exceeds the limiting safety system setting. The margins assumed
in the design of the RPS and in the safety and transient analyses
calculations have not been revised. Therefore, this change does not
involve a significant reduction in the margin of safety.
The change in units to the Reactor Water Level scram setpoint
and the Scram Discharge Volume scram setpoint do not involve a
significant reduction in the margin of safety because the changes do
not represent a change in the physical setpoints.
The reduction in the Low Condenser Vacuum scram setpoint does
not represent a reduction in the margin of safety because the scram
is not credited in any transient analysis.
The APRM Downscale Scram Trip Function is not credited in the
termination of any FSAR or reload safety analysis event. As such,
the elimination of this scram function has no effect on any margin
of safety.
Section 3/4.11
The proposed changes do not involve a significant increase in
the probability or consequences of an accident previously evaluated
because:
In general, the proposed changes represent the conversion of
current requirements to a [[Page 24908]] 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. These proposed
changes are consistent with the current safety analyses and have
been previously determined to represent sufficient requirements for
the assurance of 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 Generic Changes to the technical specifications involve
administrative changes to format and arrangement of the material. As
such, these changes cannot involve a significant increase in the
probability or consequences of an accident previously evaluated.
The current specifications require the reactor to be placed in
cold shutdown when a thermal limit was exceeded and not restored
within the allotted 2 hours, but the proposed specifications require
the reactor to be less than 25% of rated thermal power if this
condition occurred. The change eliminates a shutdown and requires
the power level to be reduced to the point that the limits are no
longer applicable.
Therefore, the change will not increase the probability or
consequences of an accident.
Create the possibility of a new or different kind of accident
from any previously evaluated because:
In general, the proposed changes represent 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 stations; however, these changes provide additional
restrictions which are in accordance with the current safety
analyses, or are to provide for additional testing or surveillance
which will not introduce new failure mechanisms beyond those already
considered in the current safety analyses. Therefore, these changes
will not create the possibility of a new or different kind of
accident from any accident previously evaluated.
Since the Generic Changes proposed to the technical
specifications are administrative in nature, they cannot create the
possibility of a new or different kind of accident from any
previously evaluated.
The requirement to reduce thermal power to less than 25% of
rated thermal power rather than place the reactor in cold shutdown
will not create a new or different kind of accident because the
thermal limits are not required in operational mode 1 when thermal
power is less than 25% of rated power.
Involve a significant reduction in the margin of safety because:
In general, the proposed changes represent 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 latter 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 within 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 Generic Changes proposed in this amendment request are
administrative in nature and, as such, do not involve a reduction in
the margin of safety.
Section 3/4.12
The proposed changes do not involve a significant increase in
the probability or consequences of an accident previously evaluated
because:
The proposed Specification 3/4.12 is a new section which will
provide the user with a format that will allow better access to
needed information and provide concise Applicability and Action
requirements. The additions of Applicability and Action requirements
represent classification of intended requirements that do not
presently state all required conditions of operability or provide
clearly stated Action statements if the requirements are not met.
The combining of the two sections and the added requirements follow
Standard Technical Specifications (STS) guidelines that are in use
at many operating BWRs with similar design and operating
configurations as Dresden and Quad Cities Stations.
The proposed Section 3/4.12 involves the relocation of present
requirements into one section identical to STS provisions. The
changes also implement the Applicability and Action provisions of
the STS and later operating BWR plants that have been evaluated and
found acceptable for use at Dresden and Quad Cities. Present
Surveillance Requirements are replaced, where applicable, with
proven STS guidelines that are being used at plants with a system
similar to that at Dresden and Quad Cities. The changes in the
present Surveillance Requirements add testing requirements that are
not presently in the Dresden and Quad Cities technical
specifications. The proposed changes do not affect accident
assumptions other than a minor increase in the initial power level
(approximately 0.2% to 1%) and as such, do not involve a significant
increase in the probability of an accident previously evaluated. The
proposed specifications add additional requirements to
specifications currently contained in the Technical Specifications.
Since the proposed changes to the Technical Specifications implement
requirements that have been demonstrated to provide acceptable
operability provisions at other facilities with a design similar to
that at Dresden and Quad Cities, the proposed changes do not
significantly increase the consequences of an accident previously
evaluated.
The proposed changes do not create the possibility of a new or
different kind of accident from any previously evaluated because:
The proposed administrative changes to the format and
arrangement of material do not affect technical requirements or
assumptions of any potential accident and; therefore, cannot create
the possibility of a new or different kind of accident from any
previously evaluated.
The proposed addition of Applicability and Action requirements
enhance the understanding and usability of the Technical
Specifications and thus represent an improvement over present
specifications. New requirements are modeled after those in use at
operating BWRs and do not represent requirements that will adversely
affect potential accident analyses or assumptions. Therefore, the
proposed changes do not create the possibility of a new or different
kind of accident from any previously evaluated.
The proposed changes do not involve a significant reduction in
the margin of safety because:
The proposed administrative changes to format, arrangement of
material, clarification of requirements and other non technical
changes do not affect any safety aspects of the plant and as such
can not involve a significant reduction in the margin of safety.
In addition, the commission has provided guidance concerning the
application of standards for determining whether significant hazards
consideration exists by providing certain examples (51 FR 7751) of
amendments that are considered not likely to involve significant
hazards considerations. Commonwealth Edison has reviewed the
proposed changes against these examples and believes that the
proposed changes fall within the scope of example (ii) ``a change
that constitutes an additional limitation, restriction, or control
not presently included in the technical specifications''.
The proposed amendment does not involve a significant relaxation
of the criteria used to establish safety limits, a significant
relaxation of the bases for the limiting safety system settings or a
significant relaxation of the bases for the limiting conditions for
operations. Therefore, based on the guidance provided in the Federal
Register and the criteria established in 10 CFR 50.92(c), the
proposed change does not constitute a significant hazards
consideration.
The NRC staff has reviewed the licensee's analysis and, based on
this [[Page 24909]] 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, IllinoisDocket
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: December 15, 1993, as
supplemented by letter dated April 21, 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 (TSs)
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 TSs for both Dresden and Quad Cities
against the Standard Technical Specifications (STSs) 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 TSs 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 TSs to the STSs contained in NUREG-
0123.
The TSUP for Dresden and Quad Cities is not a complete adaption of
the STSs. 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 TSs based on the
licensee's responses to Generic Letters (GLs), and (4) relocating
specific items to more appropriate TS locations.
The December 15, 1993, and April 21, 1995, applications proposed to
upgrade only Section 5.0 (Design Features) of the Dresden and Quad
Cities TSs.
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 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 to the current Technical Specifications (CTS)
represent minor curtailments of the current requirements which are
based on generic guidance or previously approved provisions for
other stations. The proposed amendment for current Dresden and Quad
Cities Station's Technical Specifications Section 5.0 represent a
minor relaxation of the current requirements, and is based on BWR-
STS (NUREG-0123) guidelines or later operating BWR plant's NRC
accepted changes. The proposed changes are consistent with the
current safety analyses and have been previously determined to
represent sufficient requirements for the assurance and reliability
of equipment assumed to operate in the safety analysis. Any
deviations from CTS or STS requirements do not significantly
increase the probability or consequences of any previously evaluated
accidents for Dresden or Quad Cities Stations.
Details describing the plant's design are presented in TSUP
Section 5.0. There are no Limiting Conditions for Operation (LCO) or
Surveillance Requirements (SR) encompassed within TSUP Section 5.0.
This information is administrative in nature and consistent to the
UFSAR; therefore, the probability of any accident previously
evaluated is not increased by the proposed amendment.
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 relaxations 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. The proposed changes are administrative in nature
and do not involve a revision in the operation of the station. As
such, there are no changes to the current safety analysis.
Therefore, the proposed changes 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 Specifications Section 5.0 is based on BWR-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 BWR-STS or later operating BWRs. Any
deviations from CTS or BWR-STS requirements do not create the
possibility of a new or different kind of accident previously
evaluated for Dresden and Quad Cities Stations. No new modes of
operation are introduced by the proposed changes. The proposed
changes maintain at least the present level of operability, and in
some cases are more conservative. 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. The proposed amendment to Technical Specification
Section 5.0 implements present requirements, or the intent of
present requirements in accordance with the guidelines set forth in
the STS. Any deviations from CTS or BWR-STS requirements do not
significantly reduce the margin of safety for Dresden or Quad Cities
Stations. These changes do not involve revisions to the design of
the station. The proposed changes are administrative in nature and
do not involve a revision in the operation of the station. As such,
there are no changes to the current safety analysis. Therefore, the
proposed changes will not introduce new failure mechanisms beyond
those already considered in the current safety analyses. Therefore,
because the proposed changes are administrative in nature, do not
involve a revision in the operation of the station and maintains the
current design requirements specified in the UFSAR, 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. [[Page 24910]]
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
Consumers Power Company, Docket No. 50-255, Palisades Plant, Van
Buren County, Michigan
Date of amendment request: December 13, 1994
Description of amendment request: The proposed amendment would
revise the Palisades' technical specifications (TSs) to add a high
thermal performance (HTP) departure from nucleate boiling correlation
to Safety Limit 2.1. The HTP correlation is used for the high thermal
performance fuel loaded during recent fuel cycles.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. Involve a significant increase in the probability or
consequences of an accident previously evaluated.
The proposed change to the TS adds the HTP critical heat flux
correlation to the Safety Limit - Reactor Core Section 2.1. The HTP
correlation is an NRC approved methodology for a Departure from
Nucleate Boiling (DNB) Correlation for high thermal performance
(HTP) fuel as is used at Palisades. The HTP correlation is an
extension of the currently approved ANFP correlation. There are no
associated changes in plant operation. Palisades fuel loaded in
cycle 9 and later meet the requirements of the HTP correlation.
Therefore, operation of the facility in accordance with the proposed
TS would not result in 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 evaluated.
The HTP correlation will allow for more accurate DNB predictions
within the applicable operating conditions for fuels with the HTP
design used at Palisades. There are no changes in plant operation.
Therefore operation of the facility in accordance with the proposed
TS would not create the possibility of a new or different kind of
accident from any previously evaluated.
3. Involve a significant reduction in a margin of safety.
As stated previously, the HTP correlation will allow for more
accurate DNB predictions within the applicable operating conditions
for fuel with the HTP design. There are no associated changes in
plant operation. Therefore, operation of the facility in accordance
with the proposed TS would not involve a significant reduction in a
margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: Van Wylen Library, Hope
College, Holland, Michigan 49423.
Attorney for licensee: Judd L. Bacon, Esquire, Consumers Power
Company, 212 West Michigan Avenue, Jackson, Michigan 49201
NRC Project Director: Cynthia A. Carpenter, Acting
Duke Power Company, et al., Docket Nos. 50-413 and 50-414, Catawba
Nuclear Station, Units 1 and 2, York County, South Carolina
Date of amendment request: January 18, 1995
Description of amendment request: The proposed amendments would
relocate the requirements for the seismic instrumentation,
meteorological instrumentation, and loose-part detection system from
the Technical Specifications to the Selected Licensee Commitment (SCL)
Manual. This will allow future changes to these controls to be
performed under the provisions of 10 CFR 50.59. No changes are being
made to the technical content of the affected 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:
Criterion 1
The requested amendments will not involve a significant increase
in the probability or consequences of an accident previously
evaluated. Relocation of the affected TS sections to the SLC Manual
will have no effect on the probability of any accident occurring. In
addition, the consequences of an accident will not be impacted since
the above instrumentation will continue to be utilized in the same
manner as before. No impact on the plant response to accidents will
be created.
Criterion 2
The requested amendments will not create the possibility of a
new or different kind of accident from any accident previously
evaluated. No new accident causal mechanisms will be created as a
result of relocating the affected TS requirements to the SLC Manual.
Plant operation will not be affected by the proposed amendments and
no new failure modes will be created.
Criterion 3
The requested amendments will not involve a significant
reduction in a margin of safety. No impact upon any plant safety
margins will be created. Relocation of the affected TS requirements
to the SLC Manual is consistent with the content of the Westinghouse
RSTS [Revised Standard Technical Specifications], as the NRC did not
require technical specification controls for the affected
instrumentation in the RSTS. The proposed amendments are consistent
with the NRC philosophy of encouraging utilities to propose
amendments that are consistent with the content of the RSTS.
Based upon the preceding analyses, Duke Power Company concludes
that the requested amendments do 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: York County Library, 138 East
Black Street, Rock Hill, South Carolina 29730
Attorney for licensee: Mr. Albert Carr, Duke Power Company, 422
South Church Street, Charlotte, North Carolina 28242
NRC Project Director: Herbert N. Berkow
Florida Power and Light Company, et al., Docket Nos. 50-335 and 50-
389, St. Lucie Plant, Unit Nos. 1 and 2, St. Lucie County, Florida
Date of amendment request: April 3, 1995
Description of amendment request: The amendments will incorporate
line-item TS improvements to Specifications 3/4.8.1 ``Electrical Power
Systems-A.C. Sources,'' and 4.8.1.2.2 ``Electrical Power Systems-
Shutdown.'' The proposed changes are consistent with recommendations
for Emergency Diesel Generator (EDG) Surveillance Requirements in
NUREG-1366, and regulatory guidance provided in Generic Letter (GL) 93-
05 and GL 94-01. This proposal also contains FPL's commitment to
implement a maintenance program for monitoring and maintaining EDG
performance for both St. Lucie Units consistent with 10 CFR 50.65 and
the guidance of Regulatory Guide 1.160.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
(1) Operation of the facility in accordance with the proposed
amendment would not [[Page 24911]] involve a significant increase in
the probability or consequences of an accident previously evaluated.
The license amendments proposed for St. Lucie Units 1 and 2 will
incorporate line-item Technical Specification (TS) improvements for
Emergency Diesel Generators (EDG) pursuant to guidance provided in
Generic Letters (GL) 93-05 and 94-01. The EDGs are not accident
initiators, the proposed TS changes do not involve any assumptions
relative to accident initiators in the plant safety analyses, and
therefore the proposed amendments will not impact the probability of
occurrence for accidents previously analyzed.
The EDG line-item TS improvements associated with GL 93-05 are
based on recommendations designed to remove unwarranted requirements
for testing during power operation and other factors that are
counter-productive to safety in terms of equipment degradation and
availability. These recommendations resulted from a comprehensive
study of industry-wide EDG surveillance requirements and subsequent
findings reported by the NRC in NUREG-1366. The proposed amendments
are consistent with the GL 93-05 guidance for implementing such
recommendations.
Similarly, GL 94-01 provides guidance for a line-item TS
improvement that will remove accelerated testing requirements from
the TS provided that the licensee commits to a maintenance program
for monitoring and maintaining EDG performance that includes the
applicable provisions of the maintenance rule (10 CFR 50.65). Such a
program will further assure EDG availability. Since the availability
of EDGs is assumed in certain success paths for mitigating analyzed
accidents, an improvement in EDG availability will enhance accident
mitigation capabilities.
Therefore, operation of the facility in accordance with the
proposed amendments would not involve a significant increase in the
probability or consequences of an accident previously evaluated.
(2) Operation of the facility in accordance with the proposed
amendment would not create the possibility of a new or different
kind of accident from any accident previously evaluated.
The proposed amendments incorporate line-item TS improvements to
EDG surveillance testing requirements, and will not change the
physical plant or the modes of plant operation defined in the
Facility License. The changes do not involve the addition or
modification of equipment, nor do they alter the design or methods
of operation of plant systems. Plant configurations that are
prohibited by TS will not be created by the amendments. Therefore,
operation of the facility in accordance with the proposed amendment
would not create the possibility of a new or different kind of
accident from any accident previously evaluated.
(3) Operation of the facility in accordance with the proposed
amendment would not involve a significant reduction in a margin of
safety.
The proposed amendments are designed to improve EDG availability
by eliminating unwarranted surveillance testing. The presently
specified surveillance intervals are not changed. The proposed
changes do not otherwise alter the basis for any technical
specification that is related to the establishment of, or the
maintenance of a nuclear safety margin. Therefore, operation of the
facility in accordance with the proposed amendment would not involve
a significant reduction in a margin of safety.
Based on the above discussion and the supporting Evaluation of
Technical Specification changes, FPL has determined that the
proposed license amendment involves no significant hazards
consideration.
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: Indian River Junior College
Library, 3209 Virginia Avenue, Fort Pierce, Florida 34954-9003
Attorney for licensee: J. R. Newman, Esquire, Morgan, Lewis &
Bockius, 1800 M Street, NW., Washington, DC 20036
NRC Project Director: David B. Matthews, Director
Maine Yankee Atomic Power Company, Docket No. 50-309, Maine Yankee
Atomic Power Station, Lincoln County, Maine
Date of amendment request: March 7, 1995
Description of amendment request: The proposed amendment would add
an Exception to Technical Specifications (TS) 3.6.A and 3.6.C. The
Exception would permit reduced component cooling water flow for short
periods of time, while component cooling water heat exchangers are
shifted.
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 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.
Plant experience shows that the component cooling water heat
exchangers can be shifted in a few minutes; well within the time limit
for Remedial Action under this TS 3.6.A or C, or TS 3.0.A. Thus, the
proposed change 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 change does not affect equipment reliability when such
equipment is required to be operable. Existing TS 3.6 and its Remedial
Action statement govern the plant circumstances under which cooling
water subsystems are required, and specify the maximum time such
subsystems may be unavailable. The proposed change does affects neither
operating requirements nor the time limit on restoring system
operability.
3. The proposed change does not involve a significant reduction in
a margin of safety.
The proposed change does not significantly alter the availability
or condition of the cooling water subsystems and, therefore, does not
alter the accident analysis or its associated conclusions. The proposed
change would permit flow in one component cooling water train to be
reduced below that required for operation of the emergency core cooling
systems in the recirculation mode, for a short period of time. The
amount of time that flow is reduced is small, and full flow operation
can be easily restored within the time required for design heat load
removal. Thus, there is no significant reduction in a margin of safety.
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 this amendment request involves no significant hazards
consideration.
Local Public Document Room location: Wiscasset Public Library,
High Street, P.O. Box 367, Wiscasset, ME 04578
Attorney for licensee: Mary Ann Lynch, Esquire, Maine Yankee Atomic
Power Company, 329 Bath Road, Brunswick, ME 04011
NRC Project Director: Phillip F. McKee
Northeast Nuclear Energy Company (NNECO), Docket No. 50-245,
Millstone Nuclear Power Station, Unit 1, New London County,
Connecticut
Date of amendment request: April 18, 1995
Description of amendment request: The proposed amendment would
allow the use of the ANSI/ANS 5.1-1979 decay heat model for post-loss
of coolant accident containment cooling analysis.
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
[[Page 24912]] consideration, which is presented below:
NNECO has reviewed the proposed change in accordance with
10CFR50.92 and concluded that the change does 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 change does not involve an SHC because the
change would not:
1. Involve a significant increase in the probability or
consequences of an accident previously analyzed.
The change to the decay heat model used to determine post-
accident conditions cannot affect the probability of any accident.
No changes to plant operation or design would occur due to the new
analysis.
The new model cannot directly affect the consequences of an
accident, since it is the tool used to predict the temperature
effects of the postulated accident. However, using the ANSI/ANS 5.1-
1979 model could change the anticipated actions necessary to respond
to an event. Changing the response action could possibly affect the
consequences of an accident. This model change will not have such an
effect. Operator actions to throttle LPCI [low pressure coolant
injection], CS [core spray], or ESW [emergency service water] pump
flow are taken based upon observed conditions, not predetermined
data points from the analysis.
Operability of the emergency core cooling systems (ECCS) can be
shown for temperatures that are higher than those predicted by the
containment cooling analysis.
Therefore, the utilization of the ANSI/ANS 5.1-1979 decay heat
model does not involve a significant increase in the probability or
consequences of a previously evaluated accident.
2. Create the possibility of a new or different kind of accident
from any previously analyzed.
The proposed license amendment only revises the predicted
temperature that result from a postulated accident. There is no
change to the design or operation of any system or component. Since
this change only deals with the post-accident effects of currently
analyzed accidents, there is no possibility of creating a new or
different kind of accident.
3. Involve a significant reduction in the margin of safety.
The early design documentation stated that the ECCS components
were designed for post-accident torus temperatures of 203 deg.F. As
this issue evolved, NNECO performed operability determinations which
showed that peak temperatures of 209 deg.F were acceptable.
Utilizing a more accurate decay heat model which results in lower
predicted peak temperatures demonstrates the acceptability of the
plant design. Therefore, replacing the May-Witt decay heat model
with the ANSI/ANS 5.1-1979 model does not result in a decrease in
the margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: Learning Resource Center,
Three Rivers Community-Technical College, Thames Valley Campus, 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
Northeast Nuclear Energy Company, et al., Docket No. 50-423,
Millstone Nuclear Power Station, Unit No. 3, New London County,
Connecticut
Date of amendment request: March 29, 1995
Description of amendment request: The proposed amendment changes
Technical Specifications to revise peaking factor penalties based on
NRC approved methods.
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 an SHC because the changes
would not:
1. Involve a significant increase in the probability or
consequences of an accident previously analyzed.
The proposed changes to the action statements of Sections
3.2.2.1 and 3.2.2.2 are purely administrative and therefore they do
not adversely affect the probability or consequences of an accident
previously analyzed. The proposed changes to Surveillance
Requirements 4.2.2.1.2.e, 4.2.2.1.4.e, 4.2.2.2.2.e and 4.2.2.2.4.e
and Section 6.9.1.6.b are based on the NRC approved methodology for
calculating the penalty to be applied to FQM(Z). The
margin for the FQRTP limit is still maintained by the
proposed changes. In addition, the penalty is included in the COLR
[Core Operating Limits Report] which will be maintained and
controlled per the requirements of 10CFR50.59. Therefore, the
proposed changes do not increase the probability or consequences of
an accident previously analyzed.
2. Create the possibility of a new or different kind of accident
from any previously analyzed.
The proposed changes to the Action Statement of Sections 3.2.2.1
and 3.2.2.2 are purely administrative and therefore, they do not
create the possibility of a new or different kind of accident from
any previously analyzed. The proposed changes to Surveillance
Requirements 4.2.2.1.2.e, 4.2.2.1.4.e, 4.2.2.2.2.e, and 4.2.2.2.4.e
and Section 6.9.1.6.b do not create a malfunction that is different
from those previously evaluated. The changes do not involve
positioning reactivity systems or plant components into any new
configuration or sequence not previously analyzed. Therefore, the
changes will not create the possibility of a new or different kind
of accident from any other previously analyzed.
3. Involve a significant reduction in the margin of safety.
The proposed changes to the action statements of Sections
3.2.2.1 and 3.2.2.2 are purely administrative and therefore they
will not reduce the margin of safety. The proposed changes to
Surveillance Requirements 4.2.2.1.2.e, 4.2.2.1.4.e, 4.2.2.2.2.e and
4.2.2.2.4.e and Section 6.9.1.6.b do not reduce the margin to the
FQRTP limit. The approved methods more distinctly evaluate
the expected changes to FQM than previously existed.
Therefore, there is no impact on the margin of safety as specified
in the Technical Specifications.
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 Resource Center,
Three Rivers Community-Technical College, Thames Valley Campus, 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
PECO Energy Company, Public Service Electric and Gas Company,
Delmarva Power and Light Company, and Atlantic City Electric
Company, Docket No. 50-277, Peach Bottom Atomic Power Station, Unit
No. 2, York County, Pennsylvania
Date of application for amendment: March 30, 1995
Description of amendment request: The proposed change would revise
Technical Specifications Section 4.7.D.1.b.(1) by adding a footnote to
exempt the High Pressure Coolant Injection [HPCI] motor-operated valve
MO-2-23-015 from quarterly stoke testing requirements until refueling
outage 2RO11.
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 [[Page 24913]] consequences of an accident
previously evaluated.
The proposed change does not serve as an initiator or
contributor to any accidents previously evaluated. It does not
decrease the effectiveness of equipment relied upon to mitigate
previously evaluated accidents. A calculation was performed and it
has been determined the leakage through the valve's packing will be
within the allowable limits of containment leakage (La). While
positioning the valve in the backseated position does increase its
stroke time, it has been calculated and demonstrated that the valve
will close within the TS time limit of 20 seconds.
Therefore, the proposed change does not involve a significant
increase in the probability or consequence of an accident previously
evaluated.
2. The proposed change does not create the possibility of a new
or different kind of accident from any previously evaluated.
The proposed change does not serve as an initiator or
contributor to any of the accidents previously evaluated. The
proposed change does not introduce any new modes of plant operation.
Implementation of the proposed changes will not affect the
design function or configuration of any component or introduce any
new operating scenarios or failure modes or accident initiation. It
does not impair or prevent safety systems from performing their
safety function.
Therefore, the proposed change does not create the possibility
of a new or different kind of accident from any previously
evaluated.
3. The proposed change does not involve a significant reduction
in a margin of safety.
The proposed change does not serve as an initiator or
contributor to any accidents evaluated in the [Safety Analysis
Report] SAR. It has no impact on any safety analysis assumptions.
Exempting the HPCI valve MO-2-23-015 from quarterly stroke testing
until 2RO11 does not impact its reliability or affect its ability to
perform its intended safety function. The change does not adversely
affect the assumptions or sequence of events used in any accident
analysis.
Therefore, the proposed change does not involve a significant
reduction in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: Government Publications
Section, State Library of Pennsylvania, (REGIONAL DEPOSITORY) Education
Building, Walnut Street and Commonwealth Avenue, Box 1601, Harrisburg,
Pennsylvania 17105.
Attorney for licensee: J. W. Durham, Sr., Esquire, Sr. V. P. and
General Counsel, PECO Energy Company, 2301 Market Street, Philadelphia,
Pennsylvania 19101
NRC Project Director: John F. Stolz
PECO Energy Company, Public Service Electric and Gas Company,
Delmarva Power and Light Company, and Atlantic City Electric
Company, Dockets Nos. 50-277 and 50-278, Peach Bottom Atomic Power
Station, Units Nos. 2 and 3, York County, Pennsylvania
Date of application for amendments: March 16, 1995
Description of amendment request: This amendment would change the
existing requirements for the Source Range Monitors (SRM) while the
plant is in the refueling condition to requirements based on the
Improved Technical Specifications in NUREG-1433, ``Standard Technical
Specification General Electric Plants, BWR/4.
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 any accident previously
evaluated.
The proposed changes to the SRM requirements will not increase
the probability or consequences of an accident previously evaluated.
The SRMs are not assumed to function during any UFSAR [Updated Final
Safety Analysis Report] design basis accident or transient analysis.
This TS change will not alter any safety limits which ensure the
integrity of fuel barriers, and will not result in any increase to
onsite or offsite dose. Additionally, continued availability of the
SRMs in the refuel mode is ensured through additional testing
requirements being added by this TS change. The changes to the SRM
requirements will not alter the operation of equipment assumed to be
available for the mitigation of accidents or transients.
The proposed changes are based on NUREG-1433, ``Standard
Technical Specifications General Electric Plants, BWR/4,'' and are
consistent with the PECO Energy submittal of September 29, 1994,
requesting an overall conversion, based on NUREG-1433. The overall
conversion to the ITS [Improved Technical Specifications] included
both technically justified deviations from the NUREG, and
technically justified changes from the PBAPS current TS.
2. The proposed change does not create the possibility of a new
or different kind of accident from any previously evaluated.
The proposed changes to the SRM requirements will not create the
possibility of a new or different type of accident from any
previously evaluated. The SRMs are not assumed to function during
any analyzed UFSAR design basis accident or transient analysis.
Additionally, the changes will not involve any changes to plant
systems, structures or components (SCCs) which could act as new
accident initiators. Implementation of the proposed changes will
effect the manner in which these SCCs are tested; however, TS
requirements that govern routine testing and verification of plant
components and variables are not assumed to be initiators of any
analyzed event.
3. The proposed change does not result in a significant
reduction in the margin of safety.
No margins of safety are reduced as a result of the proposed TS
changes. No safety limits will be changed as a result of this TS
change. The proposed change does not involve a reduction in the
margin of safety because SRMs are not credited in any safety
analysis. At least one SRM will remain operable during rod
withdrawal during core alterations and rod withdrawal will not occur
if no SRMs are operable. Excessive reactivity additions will be
quickly identified and mitigated by the Intermediate Range Monitors
and associated rod blocks. The Average Power Range Monitor Flux
scram, and not any SRM function, is credited for mitigating a rod
withdrawal or reactivity addition accident.
Use of a spiral offload or reload pattern will provide assurance
that the SRM will be in the optimum position for monitoring changes
in neutron flux levels during core alternations.
The changes proposed in this TS change do not introduce any
hardware changes, and will not alter the intended operation of plant
structures, systems or components utilized in the mitigation of
accidents or transients. Additionally, these changes will not
introduce any new failure modes of plant equipment not previously
evaluated.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: Government Publications
Section, State Library of Pennsylvania, (REGIONAL DEPOSITORY) Education
Building, Walnut Street and Commonwealth Avenue, Box 1601, Harrisburg,
Pennsylvania 17105.
Attorney for licensee: J. W. Durham, Sr., Esquire, Sr. V. P. and
General Counsel, PECO Energy Company, 2301 Market Street, Philadelphia,
Pennsylvania 19101
NRC Project Director: John F. Stolz
PECO Energy Company, Public Service Electric and Gas Company,
Delmarva Power and Light Company, and Atlantic City Electric
Company, Dockets Nos. 50-277 and 50-278, Peach Bottom Atomic Power
Station, Units Nos. 2 and 3, York County, Pennsylvania
Date of application for amendments: March 22, 1995 [[Page 24914]]
Description of amendment request: The amendment would revise Note
(1) for Technical Specifications Tables 3.7.2 through 3.7.4 by reducing
the Local Leak Rate Test (LLRT) hold time duration from one hour to 20
minutes.
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 does not serve as an initiator or
contributor to any accidents previously evaluated. It does not
decrease the effectiveness of equipment relied upon to mitigate
previously evaluated accidents. The change does not involve any
physical changes to any plant systems, structures, or components.
Therefore, the proposed change does not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
2. The proposed changed does not create the possibility of a new
or different kind of accident from any previously evaluated.
The proposed change does not serve as an initiator or
contributor to any of the accidents previously evaluated. The
proposed change does not introduce any new modes of plant operation.
Implementation of the proposed changes will not affect the
design function or configuration of any component or introduce any
new operating scenarios or failure modes or accident initiation. It
does not impair or prevent safety systems from performing their
safety function.
Therefore, the proposed change does not create the possibility
of a new or different kind of accident from any previously
evaluated.
3. The proposed changes do not involve a significant reduction
in a margin of safety.
The proposed change does not serve as an initiator or
contributor to any accidents evaluated in the SAR [Safety Analysis
Report]. It has no impact on any safety analysis assumptions.
Changing the LLRT duration hold time from one hour to 20 minutes
does not impact equipment reliability. The change does not adversely
affect the assumptions or sequence of events used in any accident
analysis. Therefore, the propose change does not involve a
significant reduction in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: Government Publications
Section, State Library of Pennsylvania, (REGIONAL DEPOSITORY) Education
Building, Walnut Street and Commonwealth Avenue, Box 1601, Harrisburg,
Pennsylvania 17105.
Attorney for licensee: J. W. Durham, Sr., Esquire, Sr. V. P. and
General Counsel, PECO Energy Company, 2301 Market Street, Philadelphia,
Pennsylvania 19101
NRC Project Director: John F. Stolz
Philadelphia Electric Company, Docket Nos. 50-352 and 50-353,
Limerick Generating Station, Units 1 and 2, Montgomery County,
Pennsylvania
Date of amendment request: November 21, 1994, as supplemented by
letter dated April 6, 1995
Description of amendment request: The proposed amendment would make
changes affecting the Administrative Controls Section of the Technical
Specifications (TSs). The areas proposed to be changed are: 1) NEEDS
[Nuclear Effectiveness and efficiency Design Study] Organization Title
Changes, 2) Minimum Shift Crew Composition, 3) Delete Independent
Techincal Review Section from TS, 4) Delete NRB [Nuclear Review Board]
Review Section from TS, and 5) Delete NRB Audit Section from 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:
1. The proposed Technical Specifications changes do not involve
a significant increase in the probability or consequences of an
accident previously evaluated.
The proposed TS changes to revise the organization position
titles, PORC [Plant Operations Review Committee] composition
description, and eliminate the Assistant Superintendent - Operations
position do not involve any physical modifications to plant
structures, systems, or components (SSC), or the manner in which
these SSC are operated, maintained, modified, tested, or inspected.
The proposed changes to position titles will not change the
requirements for the qualifications and training of personnel in any
management or supervisory position. Personnel will continue to meet
the guidance specified in ANSI/ANS 3.1-1978 as required by Technical
Specification 6.3.1. The probability of occurrence of an accident is
based in part on: the training and qualifications of the personnel
filling key plant management and supervisory positions; clear lines
of authority, responsibility and communication; and, adequate
management and corporate oversight of plant performance and
activities. The proposed TS changes do not change any of these
management and organizational elements.
Allowing the Plant Manager to designate appropriately qualified,
trained and experienced members of the LGS [Limerick Generating
Station] staff as members of the PORC, as proposed, will not degrade
the effectiveness of the PORC. The qualifications, training and
experience level of the PORC will meet the requirements listed in
ANSI/ANS 3.1-1978, and the required PORC quorum (including the use
of alternates) will not be affected.
Elimination of the position of Assistant Superintendent -
Operations eliminates a level of supervision between the Plant
Manager and the Shift Managers. The Shift Managers, who hold SRO
licenses, will report directly to the Senior Manager - Operations.
Other organizational changes within the Operations group (i.e.,
establishment of the positions of Manager - Operations Services and
Manager - Operations Support) will ensure that the Senior Manager -
Operations has sufficient time to properly supervise and monitor on-
shift performance. The Senior Manager -Operations and/or an
Operations Manager will be required to hold a Senior Reactor
Operator (SRO) license. Individuals filling these positions will
satisfy the applicable training, qualifications, and experience
requirements of ANSI/ANS 3.1-1978.
The consequences of an accident could be affected by the
qualifications and training of plant management and supervisory
personnel. However, the proposed changes do not change the
qualifications and training of personnel in any management or
supervisory position. Personnel will continue to meet the criteria
specified in ANSI/ANS 3.1-1978 as required by TS 6.3.1.
Therefore, the proposed changes do not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
The proposed TS changes to increase the minimum shift crew
composition do not involve any physical changes to plant SSC.
The probability of the occurrence of an accident is based in
part on the operating crew and their ability to safely operate the
plant. The increase in the minimum on-shift crew composition and the
associated changes improves the capability of the on-shift crew to
safely operate the plant and SSC, thereby reducing the probability
of a situation that could result in an accident. The increase in the
minimum on-shift crew composition will improve the manner in which
the SSC are operated, maintained, tested, and inspected.
The consequences of an accident could be affected by an
operating error. However, the proposed TS changes increase the
number of licensed operators required to be on-shift, and therefore,
increase the capability of the on-shift crew to properly operate the
facility and to implement the appropriate emergency procedures to
reduce the consequences of an accident.
The proposed changes will also delete redundant and/or relocate
existing independent technical review and, Nuclear Review Board
review and audit requirements from TS that are and/or will be
contained in the LGS UFSAR [Updated Final Safety Analysis Report].
Removal of redundant/relocation of existing requirements does not
affect any equipment important to safety, or involve any physical
modifications to plant SSC, therefore, is not associated with an
accident initiator or accident mitigator and [[Page 24915]] can not
affect the probability of occurrence of an accident or increase the
consequences of an accident. The licensee controlled UFSAR
containing the requirements will be maintained using the provisions
of 10 CFR 50.59, or 10 CFR 50.54(a), as appropriate, and are subject
to the change control process in the Administrative Controls Section
(6.0) of the Technical Specifications. Since future changes to
related licensee-controlled documents will be evaluated per 10 CFR
50.59 or 10 CFR 50.54(a), no increase (significant or insignificant)
in the probability or consequences of an accident previously
evaluated will be allowed.
Therefore, these proposed changes do not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
2. The proposed TS changes do not create the possibility of a
new or different kind of accident from any accident previously
evaluated.
The proposed TS changes to revise the organization position
titles, PORC composition description, and eliminate the Assistant
Superintendent - Operations position do not involve any physical
modifications to plant structures, systems, or components (SSC), or
the manner in which these SSC are operated, maintained, modified,
tested, or inspected. The proposed changes to position titles will
not change the requirements for the qualifications and training of
personnel in any management or supervisory position. Personnel will
continue to meet the guidance specified in ANSI/ANS 3.1-1978 as
required by Technical Specification 6.3.1. Therefore, these proposed
TS changes do not create the possibility of a new or different kind
of accident from any accident previously evaluated.
The proposed changes to the on-shift crew composition can not
create the possibility of a new or different type of accident than
previously evaluated in the SAR since implementation of the changes
will not involve any physical changes to the plant SSC. The increase
in the minimum on-shift crew composition increases the ability of
the operating crew to ensure that the SSC are properly operated,
maintained, tested and inspected. An increase in the required number
of licensed operators on each shift improves the ability of the crew
to adequately operate the facility, to respond to accident
conditions, and to implement applicable plant procedures. Therefore,
these proposed TS changes do not create the possibility of a new or
different kind of accident from any accident previously evaluated.
The proposed changes will also delete redundant and/or relocate
existing independent technical review and, Nuclear Review Board
review and audit requirements from TS that are and/or will be
contained in the UFSAR. The changes will not alter the plant
configuration (no new or different type of equipment will be
installed) or create changes in methods governing normal plant
operation that will introduce new failure modes. These changes will
not impose different requirements and proper control of information
will be maintained. These changes will not alter assumptions made in
the safety analysis and licensing basis. Therefore, these changes
will not create the possibility of a new or different kind of
accident from any accident previously evaluated.
3. The proposed TS changes do not involve a significant
reduction in a margin of safety.
The proposed TS changes to revise the organization position
titles, PORC composition description, and eliminate the Assistant
Superintendent - Operations position, do not reduce the margin of
safety because positions with equivalent authority and
responsibility are established and the new positions have equivalent
requirements for education, experience and training. Allowing the
Plant Manager to designate appropriately qualified, trained and
experienced members of the LGS staff as members of the PORC will not
degrade the effectiveness of the PORC because the qualifications,
training and experience level of the PORC will meet the requirements
listed in ANSI/ANS 3.1-1978 and the required PORC quorum (including
the use of alternates) will not be affected. Elimination of the
position of Assistant Superintendent - Operations eliminates a level
of supervision between the Plant Manager and the Shift Managers. If
the Senior Manager - Operations does not hold an SRO license, then
an Operations Manager must hold an SRO license. This individual will
1) be qualified to fill the Senior Manager - Operations position, 2)
have the same management authority over the licensed operators as
the Senior Manager - Operations, and 3) by being designated by
Administrative procedures assures that there is always an individual
holding a current SRO license in one of the Operations management
positions. Other organizational changes (i.e., establishment of the
positions of Manager - Operations Services and Manager - Operations
Support), will ensure that the Senior Manager -Operations has
sufficient time to properly supervise and monitor on-shift
performance. Therefore, these changes do not involve a significant
reduction in a margin of safety.
The proposed changes to the on-shift crew composition increases
the number of licensed SROs per shift to be one (1) above the
minimum number required by the regulations. Additionally, the title
changes are consistent with the organization and reporting
relationships discussed in the regulation and the LGS Updated Final
Safety Analysis Report (UFSAR). The Shift Manager holds a SRO
license for both units and is assigned responsibility for overall
plant operation at all times when there is fuel in any unit. The
other SROs on the shift report to the Shift Manager and at least one
(1) of the SRO licensed individuals is in the Main Control Room when
either unit is in an operating mode other than cold shutdown or
refuel. The increase in the minimum on-shift crew composition and
the associated changes improves the capability of the on-shift crew
to safely operate the plant and SSC. Therefore, these changes do not
involve a significant reduction in a margin of safety.
The proposed changes will also delete redundant and/or relocate
existing independent technical review and, Nuclear Review Board
review and audit requirements from TS that are and/or will be
contained in the LGS UFSAR. The changes will not reduce the margin
of safety since they have no impact on any safety analysis
assumptions. In addition, any future changes to the UFSAR will be
evaluated per the requirements of 10 CFR 50.59 or 10 CFR 50.54(a),
as appropriate. Therefore, these changes will not involve a
significant reduction in a margin of safety.
The existing requirement for NRC review and approval of
revisions, in accordance with 10 CFR 50.90, to these TS details and
requirements proposed for relocation, does not have a specific
margin of safety upon which to evaluate. However, since the proposed
changes to delete redundant and/or relocate requirements are
consistent with the BWR Standard Technical Specifications (NUREG-
1433) and the four criteria set forth in the NRC ``Final Policy
Statement on Technical Specifications Improvements for Nuclear Power
Reactors,'' and since the change controls for proposed relocated
details and requirements provide an equivalent level of regulatory
authority, revising the TS to reflect the approved level of detail
and requirements ensures no 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: 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
Power Authority of the State of New York, Docket No. 50-333, James
A. FitzPatrick Nuclear Power Plant, Oswego County, New York
Date of amendment request: February 22, 1995
Description of amendment request: The proposed changes to the James
A. Fitzpatrick Technical Specifications establish operability and
surveillance requirements for the Reactor Vessel Overfill Protection
Instrumentation that initiates feedwater pump turbine trips, and a main
turbine trip, on high reactor vessel water level.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. Involve a significant increase in the probability or
consequences of an accident previously evaluated
because: [[Page 24916]]
The proposed changes involve the addition of new operability and
surveillance requirements to the Technical Specification regarding
the current high reactor water level trip feature for the feedwater
pump turbines and main turbine. The changes do not introduce any new
modes of plant operation, make any physical changes, or alter any
operational setpoints associated with the plants instrumentation and
controls. Further, the Fitzpatrick UFSAR [Updated Final Safety
Analysis Report], Section 14.5.9, for the Feedwater Controller
Failure operational transient does not take credit for the automatic
high reactor vessel water level trip of the feedwater pump turbines.
The Fitzpatrick UFSAR analysis (Section 14.5.9), for the Feedwater
Controller Failure operational transient assumes an automatic high
reactor vessel water level trip of the main turbine. Incorporating
these requirements into the Technical Specifications provides
additional assurance that a trip feature described in the UFSAR
remains functional. For these reasons the changes do not increase
the probability or consequences of an accident previously evaluated.
2. Create the possibility of a new or different kind of accident
from those previously evaluated because:
The proposed changes do not introduce any new accident
initiators or failure mechanisms since the changes do not introduce
any new modes of plant operation, make any physical changes, or
alter any operational setpoints. Accordingly, the changes do not
create the possibility of a new or different kind of accident from
those previously evaluated.
3. Involve a significant reduction in the margin of safety
because:
The proposed changes establish operability and surveillance
requirements for the design feature that trips the feedwater pump
turbines and main turbine on high reactor vessel water level. The
requirements will assure the continued operability of a trip
function that is designed to initiate protective measures in the
event of excessive feedwater flow. Tripping the feedwater pump
turbines and main turbine on high reactor vessel water level,
precludes potential adverse safety implications associated with a
reactor overfill condition. Accordingly, the proposed changes will
enhance the plant safety margin.
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: Reference and Documents
Department, Penfield Library, State University of New York, Oswego, New
York 13126.
Attorney for licensee: Mr. Charles M. Pratt, 1633 Broadway, New
York, New York 10019.
NRC Project Director: Ledyard B. Marsh
Power Authority of the State of New York, Docket No. 50-333, James
A. FitzPatrick Nuclear Power Plant, Oswego County, New York
Date of amendment request: March 2, 1995
Description of amendment request: The proposed changes to the James
A. Fitzpatrick Technical Specifications extend the surveillance test
intervals for the snubber systems to support 24 month operating cycles.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. Involve a significant increase in the probability or
consequences of an accident previously evaluated.
The proposed changes increase the interval between snubber
functional tests. These changes are consistent with the guidance
provided in Generic Letter 91-04. These changes do not involve any
physical changes to the plant, nor do they alter the way snubbers
function. The type of testing and the actions taken if a snubber
fails a functional test remain the same. The review of the snubber
installation and maintenance records will continue to ensure that
the snubbers service life is not exceeded prior to the next
scheduled review. The proposed changes to bases 4.0 and 4.6 clarify
that the snubber functional testing interval is consistent with the
length of the operating cycle. Therefore, the proposed changes do
not involve a significant increase in the probability or
consequences of an accident previously evaluated.
2. Create the possibility of a new or different kind of accident
from any accident previously evaluated.
The proposed changes increase the interval between snubber
functional tests. These changes are consistent with the guidance
provided in Generic Letter 91-04. The proposed changes do not change
the ability of the snubbers to provide dynamic load support during a
design basis accident. Past operating experience indicates that the
snubber program at the FitzPatrick plant adequately identifies
snubber failures. No changes are proposed to the type of testing
performed only to the surveillance interval length. The proposed
changes do not modify the design or operation of plant equipment,
therefore, no new or different failure modes are introduced. The
Technical Specification for snubber testing is self-corrective. If
any snubber fails a functional test, Technical Specifications
require additional testing of a 10% sample of that type of snubber
until no more failures are found. The functional test criteria
remains unchanged and ensures a 95% confidence level that at least
90% of the snubbers are operable. The proposed changes to bases 4.0
and 4.6 clarify that the snubber functional testing interval is
consistent with the length of the operating cycle. Therefore, the
proposed changes do not create the possibility of a new or different
kind of accident from any accident previously evaluated.
3. Involve a significant reduction in a margin of safety.
The proposed changes increase the interval between snubber
functional tests. These changes are consistent with the guidance
provided in Generic Letter 91-04. The proposed changes do not alter
the configuration of the snubbers nor change the manner in which the
snubbers function. Operation of the facility remains unchanged by
the proposed changes. An evaluation of past equipment performance
indicates that snubber operability is not time dependent. The
proposed changes to bases 4.0 and 4.6 clarify that the snubber
functional testing interval is consistent with the length of the
operating cycle. Therefore, a longer surveillance test interval will
not degrade snubber performance and will not involve a significant
reduction in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 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: Reference and Documents
Department, Penfield Library, State University of New York, Oswego, New
York 13126.
Attorney for licensee: Mr. Charles M. Pratt, 1633 Broadway, New
York, New York 10019.
NRC Project Director: Ledyard B. Marsh
Power Authority of the State of New York, Docket No. 50-333, James
A. FitzPatrick Nuclear Power Plant, Oswego County, New York
Date of amendment request: April 12, 1995
Description of amendment request: The proposed changes to the James
A. FitzPatrick Technical Specifications extend the surveillance test
intervals for the nuclear steam supply system to support 24 month
operator cycles.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. Involve a significant increase in the probability or
consequences of an accident previously evaluated.
The proposed changes extend the surveillance test intervals for
nuclear steam supply system components. These changes are consistent
with the guidance provided in Generic Letter 91-04. The proposed
changes do not involve any modification to the plant, nor do they
alter equipment functions. On-line testing will provide a redundant
and early means of demonstrating system [[Page 24917]] operability.
Based on past results, SRV [safety/relief valve] mechanical
performance has been good. No SRV setpoint changes are involved in
this application. The proposed change to bases section 4.6 clarifies
that the nuclear steam supply system surveillance testing interval
is consistent with the length of the operating cycle. Therefore, the
proposed changes do not involve a significant increase in the
probability or consequences of an accident previously evaluated.
2. Create the possibility of a new or different kind of accident
from any accident previously evaluated.
The proposed changes extend the surveillance test intervals for
nuclear steam supply system components. These changes are consistent
with the guidance provided in Generic Letter 91-04. The proposed
changes do not affect the way in which the nuclear steam supply
system operates nor alter the type of surveillance testing
performed. SRV drift analyses indicate that SRV drift with a 3%
tolerance would be acceptable for (i.e., bounded by) a 24 to 30
month interval. Leaking or partially open SRVs are detected by the
acoustic monitoring system. Since the proposed changes do not modify
the design or equipment of the plant, no new failure modes are
introduced. The proposed change to bases section 4.6 clarifies that
the nuclear steam supply system surveillance testing interval is
consistent with the length of the operating cycle. Therefore, the
proposed changes do not create the possibility of a new or different
kind of accident from any accident previously evaluated.
3. Involve a significant reduction in a margin of safety.
The proposed changes extend the surveillance test intervals for
nuclear steam supply system components. These changes are consistent
with the guidance provided in Generic Letter 91-04. The proposed
changes do not alter the configuration of the nuclear steam supply
system nor change the manner in which the system functions.
Operation of the facility remains unchanged by the proposed changes.
An evaluation of past equipment performance indicates that SRV
mechanical performance has been good. In addition, SRV drift has
been analyzed to be within the allowable tolerance for the extended
surveillance interval. The proposed change to bases section 4.6
clarifies that the nuclear steam supply system surveillance testing
interval is consistent with the length of the operating cycle.
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 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: Reference and Documents
Department, Penfield Library, State University of New York, Oswego, New
York 13126.
Attorney for licensee: Mr. Charles M. Pratt, 1633 Broadway, New
York, New York 10019.
NRC Project Director: Ledyard B. Marsh
Power Authority of The State of New York, Docket No. 50-286, Indian
Point Nuclear Generating Unit No. 3, Westchester County, New York
Date of amendment request: March 3, 1995, as supplemented April 12,
1995
Description of amendment request: The licensee commenced operating
on a 24-month fuel cycle, instead of the previous 18-month fuel cycle,
with cycle 9. Fuel cycle 9 started in August 1992; however, the
licensee shut down the facility in February 1993 for a performance
improvement outage. Although a firm restart date has not yet been
established, restart is expected in the spring of 1995. In order to
accommodate operation on a 24-month cycle after the facility restarts,
the licensee requested an amendment to the Technical Specifications
(TSs) to incorporate the indicating instrument calibration frequency
changes listed below:
(1) The licensee proposed changing the calibration frequency for
the containment water level monitor instrumentation (specified in TS
Table 4.1-1) to accommodate operation on a 24-month cycle.
(2) The licensee proposed changing the calibration frequency for
the auxiliary feedwater (AFW) flow rate instrumentation (specified in
TS Table 4.1-1) to accommodate operation on a 24-month cycle.
(3) The licensee proposed changing the calibration frequency for
the containment building ambient temperature sensors (specified in TS
Table 4.1-1) to accommodate operation on a 24-month cycle.
(4) The licensee proposed changing the calibration frequency for
the seismic monitoring instrumentation (specified in TS Table 4.10-2)
to accommodate operation on a 24-month cycle.
In addition, the licensee proposed adding a new surveillance
requirement to TS Table 4.1-1 for testing the core exit thermocouples.
These proposed changes follow the guidance provided in Generic
Letter 91-04, ``Changes in Technical Specification Surveillance
Intervals to Accommodate a 24-Month Fuel Cycle,'' as applicable.
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:
Consistent with the criteria of 10 CFR 50.92, the enclosed
application is judged to involve no significant hazards based on the
following information:
(1) Does the proposed license amendment involve a significant
increase in the probability or consequences of any accident
previously evaluated?
Response:
The proposed changes do not involve a significant increase in
the probability or consequences of any accident previously
evaluated. The proposed changes extend the calibration frequency (to
24 months) for the:
containment temperature channels,
containment water level monitoring system channels,
seismic instrumentation channels, and
auxiliary feedwater flow rate channels.
These changes are being made to accommodate a 24 month operating
cycle. The proposed changes in the calibration frequencies do not
involve any plant hardware changes, nor do they change the way the
systems function.
Extension of the calibration and surveillance test intervals in
question were evaluated and the results documented in [New York
Power Authority (NYPA) Report No. IP3-RPT-MULT-00424, ``Indicating
Instruments Surveillance Test Extensions,'' May 1993]. An Instrument
Drift Analysis for the indicating instruments [NYPA Report No. IP3-
RPT-MULT-00407, ``Instrument Drift Analysis for Indicating Loops,''
April 1993] was performed to evaluate past and future instrument
drift. The results of these evaluations and analyses indicate that
the calibrations in question can safely be extended to accommodate
the 24 month operating cycle.
For containment temperature, auxiliary feedwater flow and
seismic instrumentation, past instrument drift has generally been
within acceptable limits. Some drift exceeding the calibration
tolerance did occur for the triaxial time-history accelographs, but
on-line testing should ensure that instrument drift over the longer
cycle does not degrade system performance. For containment water
level systems (except containment building level), new electronic
transmitters were recently installed. Due to the lack of data, an
instrument drift analysis was not performed. However, the new
containment water level transmitters improved the overall channel
accuracy.
Future instrument drift was predicted and used to update
existing loop accuracy calculations, with the following results. (1)
For the containment temperature channels, the loop accuracy
calculations were revised to incorporate the larger channel
uncertainties. Postulated drift over 30 months should have a
negligible effect on the EOPs [Emergency Operating Procedures] and
plant shutdown. (2) For the containment system sump water levels,
future drift is not a concern because the containment building water
level is used post accident. The larger uncertainties can safely be
accommodated by changing the EOP setpoint for transfer to cold leg
recirculation. (3) For the seismic instrumentation, past drift was
negligible, and future drift is not expected to be cycle length
dependent. (4) For the auxiliary [[Page 24918]] feedwater flow rate
channels, the larger uncertainties can be safely accommodated by
changing the EOP setting for the minimum AFW flow required for heat
removal.
For the containment temperature and seismic instrumentation, on-
line testing provides added assurance that the instrumentation is
functioning as required.
[For the core exit thermocouples, adding a requirement to
conduct testing every 18 months will serve to ensure system
operability. This new testing requirement does not change the way
the plant operates or involve hardware modifications.]
(2) Does the proposed license amendment create the possibility
of a new or different kind of accident from any previously
evaluated?
Response:
The proposed changes do not create the possibility of a new or
different kind of accident from any previously evaluated. The
proposed changes extend the calibration frequency (to 24 months) for
the:
containment temperature channels,
containment water level monitoring system channels,
seismic instrumentation channels, and
auxiliary feedwater flow rate channels.
These changes are being made to accommodate a 24 month operating
cycle. The proposed changes in the calibration frequencies do not
involve any plant hardware changes, nor do they change the way the
systems function.
Extension of the calibration and surveillance test intervals in
question were evaluated and the results documented in [same as
Question (1)]. An Instrument Drift Analysis for the indicating
instruments [same as Question (1)] was performed to evaluate past
and future instrument drift. The results of these evaluations and
analyses indicate that the calibrations in question can safely be
extended to accommodate the 24 month operating cycle. For the
containment temperature and seismic instrumentation, on-line testing
provides added assurance that the instrumentation is functioning as
required.
[For the core exit thermocouples, adding a requirement to
conduct testing every 18 months will serve to ensure system
operability. This new testing requirement does not change the way
the plant operates or involve hardware modifications.]
(3) Does the proposed amendment involve a significant reduction
in a margin of safety?
Response:
The proposed changes do not involve a significant reduction in a
margin of safety. The proposed changes extend the calibration
frequency (to 24 months) for the:
containment temperature channels,
containment water level monitoring system channels,
seismic instrumentation channels, and
auxiliary feedwater flow rate channels.
These changes are being made to accommodate a 24 month operating
cycle. The proposed changes in the calibration frequencies do not
involve any plant hardware changes, nor do they change the way the
systems function.
For containment temperature, auxiliary feedwater flow and
seismic instrumentation, past instrument drift has generally been
within acceptable limits. Some drift exceeding the calibration
tolerance did occur for the triaxial time-history accelographs, but
on-line testing should ensure that instrument drift over the longer
cycle does not degrade system performance. For containment water
level systems (except containment building level), new electronic
transmitters were recently installed. Due to the lack of data, an
instrument drift analysis was not performed. However, the new
containment water level transmitters improved the overall channel
accuracy.
[For the core exit thermocouples, adding a requirement to
conduct testing every 18 months will serve to ensure system
operability. This new testing requirement does not change the way
the plant operates or involve hardware modifications.]
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: White Plains Public Library,
100 Martine Avenue, White Plains, New York 10601.
Attorney for licensee: Mr. Charles M. Pratt, 10 Columbus Circle,
New York, New York 10019.
NRC Project Director: Ledyard B. Marsh
Public Service Electric & Gas Company, Docket Nos. 50-272 and 50-
311, Salem Nuclear Generating Station, Unit Nos. 1 and 2, Salem
County, New Jersey
Date of amendment request: March 30, 1995
Description of amendment request: The proposed change to the
Technical Specifications eliminates the defined term CONTROLLED
LEAKAGE, removes Controlled Leakage flow from the Reactor Coolant
System Operational Leakage Limiting Condition for Operation (LCO), and
establishes a new Seal Injection Flow LCO.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. Do not involve a significant increase in the probability or
consequence of an accident previously evaluated.
Changing the Technical Specification to limit seal injection
flow instead of seal leakoff flow does not affect the probability of
any accident previously evaluated. Maintaining adequate Emergency
Core Cooling System (ECCS) flow during Loss of Coolant Accident
(LOCA) ensures that the consequences of these accidents are
unaffected. The existing Technical Specification allows seal
injection throttle valve positioning that could result in seal
injection flow path resistance values below those used in the Salem
ECCS hydraulic flow analyses. Reduced line resistances could result
in inadequate ECCS flow to the reactor core. Revising the Technical
Specification to limit RCP seal injection flow ensures that the
accident analysis assumptions are maintained, and the previously
evaluated accident consequences remain unchanged.
Therefore, it may be concluded that the proposed changes do not
increase the probability or consequences of an accident previously
evaluated.
2. Do not create the possibility of a new or different kind of
accident from any accident previously evaluated.
The proposed changes do not involve any hardware modifications
or result in any functional changes to system operation. RCP seal
injection flow is used as a limiting parameter in-place of RCP seal
leakoff flow.
Since design requirements continue to be met and the RCS
pressure boundary is not challenged, no new failure mode is created.
Thus, an accident different from any already evaluated is not
created by this change.
Therefore, it may be concluded that the proposed changes do not
create the possibility of a new or different kind of accident from
any previously evaluated.
3. Do not involve a significant reduction in a margin of safety.
The proposed changes do not alter the manner in which Safety
Limits or Limiting Safety System Setpoints are determined.
Controlled Leakage (RCP seal leakoff)is removed from the Reactor
Coolant System Leakage Limiting Condition for Operation (LCO), and a
new seal injection LCO is established. The new LCO continues to
limit seal injection flow during accident conditions. The limiting
parameter is changed from RCP seal leakoff flow to RCP seal
injection flow. These changes ensure that the accident analysis
assumptions and existing margins of safety are maintained. The seal
injection flow specification limit is not applicable in Mode 4 and
lower, because high seal injection flow is less critical due to
lower Reactor Coolant System (RCS) pressure and decay heat removal
requirements in these modes. Reactor coolant pump seal injection
flow must be limited in Modes 1, 2, and 3 to ensure adequate
Emergency Core Cooling System Flow.
Therefore, it may be 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 request involves no significant hazards consideration.
Local Public Document Room location: Salem Free Public library, 112
West Broadway, Salem, New Jersey 08079
Attorney for licensee: Mark J. Wetterhahn, Esquire, Winston and
[[Page 24919]] Strawn, 1400 L Street, NW, Washington, DC 20005-3502
NRC Project Director: John F. Stolz
Tennessee Valley Authority, Docket Nos. 50-327 and 50-328, Sequoyah
Nuclear Plant, Units 1 and 2, Hamilton County, Tennessee
Date of amendment request: April 6, 1995 (TS 95-05)
Description of amendment request: The proposed change would (1)
replace the reference to Table 3.6-2 from Definition 1.7.a.2 for
Containment Integrity with a phrase that will allow the valves to be
opened under administrative control; (2) replace the reference to Table
3.6-2 from Surveillance Requirement 4.6.1.1 with a phrase that will
allow the valves to be opened under administrative control; (3) delete
the reference to Table 3.6-1 from Technical Specification 3.6.1.2; (4)
delete Table 3.6-1, ``Bypass Leakage Paths to the Auxiliary Building --
Secondary Containment Bypass Leakage Paths;'' (5) revise Specification
3.6.3 to delete the reference to Table 3.6-2, add a footnote that
discusses the opening of penetrations intermittently, add the phrase to
take exception to the containment vacuum isolation valves, and add an
action statement to indicate that Specification 3.0.4 does not apply to
the specification; (6) delete Surveillance Requirement 4.6.3.1; (7)
delete references to Table 3.6-2 in Specifications 4.6.3.2 and 4.6.3.3
and additional wording added to indicate that the specifications apply
to automatic containment isolation valves; (8) delete Table 3.6-2,
``Containment Isolation Valves'' and add a note to the page indicated
that the information has been intentionally deleted; (9) revise
Specification 3.8.3.1 to specify that the Limiting Condition for
Operation applies to primary and backup containment penetration
conductor overcurrent protective devices associated with each
containment electrical penetration shall be operable, add a phrase to
indicate that the scope of these protective devices excludes those
circuits for which credible fault currents would not exceed the
electrical penetration design rating, and delete the phrase that
references appropriate plant instructions in the action statement; (10)
delete the phrase that references appropriate plant procedures from
Specification 4.8.3.1; (11) delete the phrase from SR 4.8.3.1.a.3 that
indicates that a complete listing of all fuses to be verified in
accordance with the requirement will be maintained in appropriate plant
instructions; (12) replace the phrase ``appropriate plant instructions
based on'' with ``procedures prepared in conjunction with'' in SR
4.8.3.1.b; (13) replace the reference to Table 3.8-2 in Specification
3.8.3.2 with a phrase that indicates that the Requirement is applicable
to valves used in safety systems; (14) delete Table 3.8-2, ``Motor
Operated Valves Thermal Overload Protection,'' and replace it with a
note that indicates that the pages are intentionally blank; and (15)
incorporate appropriate changes to the Bases to reflect these changes.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
TVA has evaluated the proposed technical specification (TS)
change and has determined that it does not represent a significant
hazards consideration based on criteria established in 10 CFR
50.92(c). Operation of Sequoyah Nuclear Plant (SQN) in accordance
with the proposed amendment will not:
1. Involve a significant increase in the probability or
consequences of an accident previously evaluated.
The removal of the component listings from the SQN TSs will not
create an increase in the probability or consequences of any
accident previously evaluated. Although no longer in the TSs, the
components listed in Tables 3.6-1, 3.6-2, and 3.8-2 will be
contained in administratively controlled documents. This equipment
must be tested at the required intervals and each unit's action
statements must still be adhered to. These procedures are revised
and approved in accordance with requirements of TS Section 6.5.1A.
This review process also requires an evaluation based on 10 CFR
50.59 requirements. As indicated in GL 91-08, this is adequate
control for changes to these components lists.
2. Create the possibility of a new or different kind of accident
from any previously analyzed.
The removal of the component lists from the TSs does not modify
safety-related equipment or systems, nor does it change any safety-
related setpoints used to prevent or mitigate previously analyzed
accidents. The component lists are presently located in separate
documents that are subject to the requirements of 10 CFR 50.59.
Also, the limiting condition of operation requirements remain in
effect and appropriate actions will be taken if any limits are
exceeded. Therefore, the proposed amendment does not create the
possibility of a new or different kind of accident from any accident
previously evaluated.
3. Involve a significant reduction in a margin of safety.
The margin of safety is not affected by the removal of the
previously discussed component lists from the TS. Appropriate
measures presently exist to control the setpoint of the components
listed. Any changes to these setpoints are controlled by the SQN
design change process that is subject to the requirements of 10 CFR
50.59 in which the reduction of the present margin of safety is
addressed. The proposed amendment continues to require operation
within the set values for these components, and appropriate actions
to be taken when or if the limits are exceeded. Based on these
controls, this amendment will not involve a reduction in a 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
The Cleveland Electric Illuminating Company, Centerior Service
Company, Duquesne Light Company, Ohio Edison Company, Pennsylvania
Power Company, Toledo Edison Company, Docket No. 50-440, Perry
Nuclear Power Plant, Unit No. 1, Lake County, Ohio
Date of amendment request: March 24, 1995
Description of amendment request: The licensee has requested a one-
time extension of the performance intervals for certain Technical
Specification Surveillance Requirements (SR). Affected SRs include
penetration leak rate testing, valve operability testing, instrument
calibration, response time testing, and logic system functional tests.
The proposed changes are requested to support refueling outage 5
scheduled to begin no later than February 15, 1996.
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 TS change requests a one-time extension of the
surveillance intervals related to: a) RPS Instrumentation
calibration, LSFTs, and response time testing; b) Isolation
Actuation System Instrumentation calibration, LSFTs, and response
time testing; c) ECCS Actuation Instrumentation calibration, LSFTs,
and response time testing; d) Control Rod Block Instrumentation
calibration and LSFTs; e) Remote Shutdown Instrumentation and
Controls calibration and operability testing; f)
[[Page 24920]] Accident Monitoring Instrumentation calibration; g)
Plant Systems Instrumentation calibration and LSFTs; h) Primary
Containment automatic valve actuation; i) Reactor Coolant System
Pressure Isolation Valve (PIV) testing; j) system automatic
initiation testing; and, k) Emergency Diesel Generator inspection
and testing.
Also proposed is the re-establishment of the baseline for the
``N times 18 months'' cumulative surveillance interval for response
time testing.
The discussion in the License Amendment Request demonstrates the
following:
i) Rosemount transmitter calibration period extension is
acceptable based on Rosemount D8900126, Revision A which supported
extension of the calibration interval from 18 months to 30 months
based on the reduction in the drift allowance;
ii) Extrapolation of plant specific calibration data is
acceptable in supporting the extension of other calibration
surveillance intervals to RFO-5;
iii) LSFT interval extension is acceptable based on the NRC
Safety Evaluation Report (Peach Bottom Atomic Power Plant, Units 2
and 3, dated August 2, 1993) which supported extension of the
interval for LSFT from 18 to 24 months. This was based on the small
probability of relay or contact failure relative to mechanical
component failure probability and, therefore, the increase in LSFT
interval represented no significant change in the overall safety
system unavailability;
iv) Response time testing interval extension for Isolation
Actuation and ECCS Actuation instrumentation channels is acceptable
based on the BWR Owners Group (BWROG) Licensing Topical Report NEDO-
32291 (January 1994) which provided the necessary justification for
elimination of response time testing and, therefore, provides a
suitable argument for extending the interval for a short period of
time. The NRC approved the use of NEDO-32291 as a basis for License
Amendment Requests, with additional conditions specified, in a
letter to the BWROG in December 1994.
v) Response time testing interval extension for RPS
Instrumentation channels is acceptable because: i) there are
redundant sensors that can initiate the scram function; ii) one-out-
of-two redundancy exists in every individual instrument channel
within each trip function; iii) several redundant and diverse
instrument channels are provided which can detect and generate a
scram signal; iv) the failure probability is a small fraction of the
total control rod insertion (scram) failure probability; v) failure
of instrumentation in the sluggish mode is a small fraction of its
overall failure modes; and iv) NRC Safety Evalution Report dated
August 2, 1993 (Peach Bottom Atomic Power Station, Units 2 and 3
docket) has previously provided approval for extension of the RPS
response time testing surveillance interval from 18 to 24 months.
vi) Response time testing interval extension for the Main Steam
Line isolation is acceptable because i) redundancy and diversity
exist in individual instrument channels within a trip function; ii)
instrumentation response time is a small fraction of the overall
response time of the actuating device; iii) instrumentation failure
probability is a very small portion of the total MSIV failure
probability; and, iv) failure of instrumentation in the sluggish
responding mode is a small fraction of its overall failure modes.
vii) Containment Isolation Valve leakage determination and
actuation interval extension is acceptable based on: i) redundancy
provided in the design of the penetrations; ii) the periodic testing
of the valves during power operation; and, iii) the short period of
time the interval is being extended.
viii) Reactor Coolant System PIVs have exhibited low as-found
leak rates as measured during the last refueling outage; there is
substantial margin available for the PIVs from the as-left leakage
to the allowed TS leakage; the requested extension of the
surveillance interval is small; and the conclusion of NUREG-1463,
``Regulatory Analysis for the Resolution of Generic Safety Issue
105: Interfacing System Loss-of-Coolant Accident in Light Water
Reactors'' (July 1993), and the confirmation of the PNPP Individual
Plant Examination that the ISLOCA (for which PIVs are provided to
prevent) is not a risk concern to BWRs or PNPP.
ix) System initiation and actuation testing interval is
acceptable based on the periodic testing of components during power
operation and the short period of time the interval is being
extended.
x) Emergency Diesel Generator testing interval extension is
acceptable based on: i) the past testing results which support
extension for the short period of time; ii) the testing that is done
during power operation; and, iii) the short period of time the
interval is being extended.
xi) The re-establishment of the baseline for the ``N times 18
months'' cumulative surveillance interval for response time testing
is acceptable in that the extension of the cumulative interval would
not be for more than the individual extension requested and
justified herein.
Therefore, from the above it is shown that the proposed change
will not significantly increase the probability of an accident
previously evaluated.
2. The proposed change would not create the possibility of a new
or different kind of accident from any accident previously
evaluated.
The proposed TS change requests a one-time extension of the
surveillance intervals for instrument calibration, instrument
channel LSFT and response time testing, containment isolation valve
leakage determination and actuation, PIV leak rate determination,
system actuation testing, and diesel generator inspection and
testing. The proposed changes do not necessitate a physical
alteration to the plant (no new or different type of equipment will
be installed). The requested extension durations are small as
compared to the overall interval allowed by TS; drift data supports
extension of the calibration intervals; NRC and industry evaluations
support extension of LSFT; industry evaluations and redundancy in
system design support extension of response time testing; past
testing and periodic testing provides confidence of no effect on
equipment availability by extending the confidence of no effect on
equipment availability by extending the surveillance interval.
Therefore, the change does not create the possibility of a new or
different kind of accident from any accident previously evaluated.
In addition, the requested re-establishment of the baseline at
RFO-5 for the ``N time 18 months'' cumulative surveillance interval
for response time testing is acceptable in that the cumulative
surveillance interval will not be extended by more than that which
is proposed for individual response time tests during RFO-5. The
individual response time test surveillance interval extensions have
been justified herein. The justification for individual response
time test surveillance interval extensions applies to the cumulative
surveillance interval extension which is requested and will be
granted by allowing the re-establishment of the baseline of the ``N
times 18 months'' surveillance interval to the response time testing
dates for those response time tests to be performed during RFO-5.
The proposed changes do not necessitate a physical alteration to the
plant (no new or different type of equipment will be installed).
Therefore, the change does not create the possibility of a new or
different kind of accident.
3. The proposed change will not involve a significant reduction
in the margin of safety.
The proposed TS change requests a one-time extension of the
surveillance intervals for instrument calibration, instrument
channel LSFT, and response time testing, containment isolation valve
leakage determination and actuation, PIV leak rate determination,
system actuation testing, and diesel generator inspection and
testing. The proposed changes do not necessitate a physical
alteration to the plant (no new or different type of equipment will
be installed). In that the requested extension durations are small
as compared to the overall interval allowed by TS, drift data
supports extension of the calibration intervals, NRC and industry
evaluations support extension of LSFT, industry evaluations and
redundancy in system design support extension of response time
testing, past testing and periodic testing provides confidence of no
effect on equipment availability by extending the surveillance
interval, the change does not involve a significant reduction in the
margin of safety.
In addition, the requested re-establishment of the baseline at
RFO-5 for the ``N times 18 months'' cumulative surveillance interval
for response time testing is acceptable in that the cumulative
surveillance interval will not be extended by more than that which
is proposed for individual response time tests during RFO-5. The
individual response time test surveillance interval extensions have
been justified herein. The justification for individual response
time test surveillance interval extensions applies to the cumulative
surveillance interval extension which is requested and will be
granted by allowing the re-establishment of the baseline of the ``N
times 18 months'' surveillance interval to the response time testing
dates for those response [[Page 24921]] time tests to be performed
during RFO-5. The proposed changes do not necessitate a physical
alteration to the plant (no new or different type of equipment will
be installed). Therefore, the 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: Perry Public Library, 3753
Main Street, Perry, Ohio 44081
Attorney for licensee: Jay Silberg, Esq., Shaw, Pittman, Potts &
Trowbridge, 2300 N Street, NW., Washington, DC 20037
NRC Project Director: Gail H. Marcus
The Cleveland Electric Illuminating Company, Centerior Service
Company, Duquesne Light Company, Ohio Edison Company, Pennsylvania
Power Company, Toledo Edison Company, Docket No. 50-440, Perry
Nuclear Power Plant, Unit No. 1, Lake County, Ohio
Date of amendment request: April 3, 1995
Description of amendment request: The proposed amendment would add
new programmatic requirements governing radiological effluent into the
Administrative Controls section of the Technical Specifications in
accordance with Generic Letter 89-01, ``Implementation of Programmatic
Controls for Radiological Effluent Technical Specifications in the
Administrative Controls Section of Technical Specifications and the
Relocation of Procedural Details of RETS to the Offsite Dose
Calculation Manual or to the Process Control Program.''
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration which is presented below:
1. The proposed change does not involve a significant increase
in the probability or consequences of an accident previously
evaluated.
The proposed changes are administrative in nature and alter only
the format and location of programmatic controls and procedural
details relative to radioactive effluent, radiological environmental
monitoring, solid radioactive wastes, and associated reporting
requirements. Compliance with applicable regulatory requirements
will continue to be maintained. In addition, the proposed changes do
not alter the conditions or assumptions in any of the Updated Safety
Analysis Report (USAR) accident analyses. Since the USAR accident
analyses remain bounding, the radiological consequences previously
evaluated are not adversely affected by the proposed changes.
Therefore, it can be concluded that 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 previously evaluated.
The proposed changes do not involve any changes to the
configuration or method of operation of any plant equipment.
Accordingly, no new failure modes have been defined for any plant
system or component important to safety nor has any new limiting
single failure been identified as a result of the proposed changes.
Also, there will be no change in types or increase in the amounts of
any radioactive effluent released offsite. Therefore, it can be
concluded that 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 the margin of safety.
The proposed changes do not involve any actual change in the
methodology used in the control of radioactive effluents, solid
radioactive wastes, or radiological environmental monitoring. These
changes are considered administrative in nature, provide for the
relocation of procedural details outside the Technical
Specifications, and add appropriate administrative controls in the
Technical Specifications to provide continued assurance of
compliance with applicable regulatory requirements. These proposed
changes also comply with the guidance contained in Generic Letter
89-01. Therefore, it can be 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 request involves no significant hazards consideration.
Local Public Document Room location: Perry Public Library, 3753
Main Street, Perry, Ohio 44081
Attorney for licensee: Jay Silberg, Esq., Shaw, Pittman, Potts &
Trowbridge, 2300 N Street, NW., Washington, DC 20037
NRC Project Director: Gail H. Marcus
Union Electric Company, Docket No. 50-483, Callaway Plant, Unit 1,
Callaway County, Missouri
Date of amendment request: February 24, 1995
Description of amendment request: The proposed amendment would
revise Technical Specification (TS) Surveillance Requirement 4.6.1.7.4
and its associated Bases to delete the quarterly verification of the
measured leakage rate for containment mini-purge supply and exhaust
isolation 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:
The proposed revision does not involve a significant hazards
consideration because operation of Callaway Plant with this change
would not:
1) Involve a significant increase in the probability or
consequences of an accident previously evaluated.
The proposed revision to the T/S will not adversely impact plant
safety since the requirement to perform the quarterly surveillance
will still be implemented to verify valve leakage and seal
degradation. The mini-purge valves will still perform their intended
safety function to close within 5 seconds after receipt of an
isolation signal.
2) Create the possibility of a new or different kind of accident
from any previously evaluated.
There are no design changes being made that would create a new
type of accident or malfunction and the method and manner of plant
operation remain unchanged. Deletion of the individual leakage rate
for these valves does not affect the severity of any accident
previously evaluated. The consequences of a valve failure or
malfunction are not increased by the removal of the acceptance
criteria, leakage rate will still be measured on a quarterly basis
as is currently done to determine if the seals are degrading.
3) Involve a significant reduction in a margin of safety.
There are no changes being made to the safety limits or safety
system settings that would adversely impact plant safety. The valves
will still be surveilled on a quarterly basis to verify leakage and
seal degradation to assure gross failure will not occur and that
containment integrity is maintained.
Based on the above discussions, it has been determined that the
requested Technical Specification change does not involve a
significant increase in the probability or consequences of an
accident or create the possibility of a new or different kind of
accident or condition over previous evaluations; or involve a
significant reduction in a margin of safety. Therefore, the
requested license amendment 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: Callaway County Public
Library, 710 Court Street, Fulton, Missouri 65251. [[Page 24922]]
Attorney for licensee: Gerald Charnoff, Esq., Shaw, Pittman, Potts
& Trowbridge, 2300 N Street, N.W., Washington, DC 20037
NRC Project Director: Gail H. Marcus
Union Electric Company, Docket No. 50-483, Callaway Plant, Unit 1,
Callaway County, Missouri
Date of amendment request: April 17, 1995
Description of amendment request: The proposed amendment would
revise Technical Specification (TS) Table 2.2-1 and associated Bases to
reduce repeated alarms and partial reactor trips related to the C-4
control system interlock and the Overpower Delta-T (OP[delta]T) reactor
trip setpoint.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
The proposed revision does not involve a significant hazards
consideration because operation of Callaway Plant with this change
would not:
1) Involve a significant increase in the probability or
consequences of an accident previously evaluated.
Overall protection system performance will remain within the
bounds of the accident analyses documented in Final Safety Analyses
Report (FSAR) Chapter 15, WCAP-10961-P for Category 1 plants such as
Callaway, and WCAP-11883 since no hardware changes are proposed.
The OP[delta]T reactor trip function provides protection against
excessive power (fuel rod integrity protection within the fuel
temperature design basis). No credit is taken for the OP[delta]T
trip in the Chapter 15 licensing basis accident analyses. The
[delta]T trip function is credited in non-licensing basis analyses
of various steamline breaks.
The OP[delta]T trip will continue to function in a manner
consistent with the plant design basis. There will be no change to
the OP[delta]T safety analysis limit listed in FSAR Table 15.0-4.
Therefore, there will be no degradation in the performance of or an
increase in the number of challenges to equipment assumed to
function during an accident situation.
The reactor trip system response time, as defined in the
Technical Specifications, will be unaffected.
These Technical Specification revisions do not involve any
hardware changes nor do they affect the probability of any event
initiators. There will be no change to normal plant operating
parameters or accident mitigation capabilities. Therefore, these
changes will not increase the probability or consequences of an
accident or malfunction.
2) Create the possibility of a new or different kind of accident
from any previously evaluated.
As discussed above, there are no hardware changes associated
with these Technical Specification revisions nor are there any
changes in the method by which any safety-related plant system
performs its safety function. Revisions to the OP[delta]T values for
K4 and K6 will require scaling changes for summing
amplifier cards (NSA cards) in the 7300 Process Protection System.
These scaling changes are straightforward and similar in nature to
those performed to implement OL Amendments 72 and 84 associated with
the implementation of relaxed axial offset control (RAOC) and a
revised OT[delta]T f1([delta]I) penalty function. These scaling
changes will not affect the normal manner of plant operation. There
will be a reduction in the incidence of C-4 alarms and partial
reactor trips. There will be less of a need to reduce power during
on-line surveillance testing.
No new accident scenarios, transient precursors, failure
mechanisms, or limiting single failures are introduced as a result
of these changes. There will be no adverse effect or challenges
imposed on any safety-related system as a result of these changes.
Therefore, the possibility of a new or different kind of accident is
not created.
3) Involve a significant reduction in a margin of safety.
There will be no change to the Overpower [delta]T safety
analysis limit listed in FSAR Table 15.0-4. Available setpoint
calculation margin will be used to increase the K4 value,
reflected as a new bias on a summing amplifier card in each of the
four protection loops. This will also require corresponding
decreases in the OP[delta]T Total Allowance and Allowable Value in
Technical Specification Table 2.2-1. Available margin in the
OP[delta]T trip protection function will be used to decrease the
K6 value, reflected as a new gain on a summing amplifier card
in each of the four protection loops.
As discussed above, the response time of the OP[delta]T reactor
trip function will remain unchanged.
It has been confirmed that the Z and S terms currently listed in
Table 2.2-1 for the OP[delta]T trip function will remain
conservative. The change in K4 will result in a decrease in the
Total Allowance and Allowable Value for OP[delta]T; however, this
does not affect any margin of safety since the safety analysis
limit, which preserves the overpower safety margin, is unchanged.
There will be no effect on the manner in which safety limits or
limiting safety system settings are determined nor will there be any
effect on those plant systems necessary to assure the accomplishment
of protection functions. There will be no impact on the overpower
limit, DNBR limits, FQ, F[delta]H, LOCA PCT, peak local power
density, or any other margin of safety.
Based upon the preceding information, it has been determined
that the proposed changes to the Technical Specifications do not
involve a significant increase in the probability or consequences of
an accident previously evaluated, create the possibility of a new or
different kind of accident from any accident previously evaluated,
or involve a significant reduction in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: Callaway County Public
Library, 710 Court Street, Fulton, Missouri 65251.
Attorney for licensee: Gerald Charnoff, Esq., Shaw, Pittman, Potts
& Trowbridge, 2300 N Street, N.W., Washington, DC 20037
NRC Project Director: Gail H. Marcus
Vermont Yankee Nuclear Power Corporation, Docket No. 50-271,
Vermont Yankee Nuclear Power Station, Vernon, Vermont
Date of amendment request: October 28, 1994
Description of amendment request: The proposed amendment would
remove the Neutron Monitoring System (NMS) and Control Rod Position
instrumentation from the Vermont Yankee Technical Specifications for
post-accident monitoring. 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:
1. The proposed change to remove the NMS and Control Rod
Position instrumentation from the Technical Specifications for post-
accident monitoring is consistent with NRC requirements concerning
this instrumentation.
Wide Range Neutron Flux (NMS instrumentation) is presently
included in the [boiling water reactor] BWR Standard Technical
Specifications, but the NRC has recently determined [letter, USNRC
to VYNPC, dated April 29, 1993] that this instrumentation need not
meet R.G. 1.97 Category 1 criteria and that licensees may request
the removal of this instrumentation from their post-accident
monitoring Technical Specifications. Control Rod Position
instrumentation is considered R.G. 1.97 Category 3 which is required
to meet the least stringent design and qualification criteria as
specified in this regulatory guide.
Testing, calibration and maintenance of this instrumentation
will continue to assure operability of instrumentation. The portions
of the NMS and the Control Rod Position instrumentation systems to
be removed from the post-accident monitoring Technical
Specifications do not perform any automatic control or trip
function. In addition, this instrumentation does not provide
information that is required to permit the control room operator to
take manual actions that are required for safety systems to
accomplish their safety functions for design basis accident
events. [[Page 24923]]
At a BWR, when all control rods are inserted, these control rods
cannot be withdrawn without deliberate operator action. The proposed
change does not result in any system hardware modification or new
plant configuration. The requested change to post-accident
monitoring instrumentation does not impact any [Final Safety
Analysis Report] FSAR safety analysis involving the NMS or Control
Rod Position System. These monitoring functions are not contributors
to the initiation of accidents.
The administrative changes to correct a typographical error and
instrument ranges will have no effect on plant hardware, plant
design, safety limit setting or plant system operation and
therefore, do not modify or add any initiating parameters that would
significantly increase the probability or consequences of any
previously analyzed accident.
Therefore, it is concluded that there is not a significant
increase in the probability or consequences of an accident
previously evaluated.
2. The function of the instrumentation to be removed from the
Technical Specifications is for monitoring only. These indications
are not necessary for operators to accomplish any safety functions.
The proposed change does not involve any change in hardware,
Technical Specification setpoints, plant operation, redundancy,
protective function or design basis of the plant. There is no impact
on any existing safety analysis or safety design limits. NMS and
Control Rod Position monitoring functions do not initiate nuclear
system parameter variations which are considered potential
initiating causes of threats to the fuel and the nuclear system
process barrier.
As discussed above, the proposed administrative change only
corrects a typographical error concerning equipment identification
numbers and listed instrument ranges. This change does not affect
any equipment and they do not involve any potential initiating
events that would create any new or different kind of accident.
Therefore, the proposed change does not create the possibility
of a new or different kind of accident from any accident previously
evaluated.
3. The proposed change to remove the NMS and Control Rod
Position instrumentation from the Technical Specifications for post-
accident monitoring does not affect any existing safety margins. The
original NMS design basis for BWRs never required a post-accident
neutron monitoring function since there are no design basis
accidents that rely on operator action to control reactor power.
This is also true for Control Rod Position monitoring.
Existing Technical Specifications requirements for automatic
trip functions are unaffected. Failure of the indication of reactor
power from the NMS or the Control Rod Position System does not
preclude the ability of the reactor operator to determine reactor
power levels. Alternate indications are available to ascertain
reactor power. These include reactor coolant boron concentrations,
flux levels from the Traversing Incore Probe (TIP) System and the
status of plant parameters which are linked to reactor power. In
addition, alternate means of determining reactor power have been
incorporated into the Emergency Operating Procedures (EOPs).
Operation, testing and maintenance of this instrumentation will
remain the same. System functions are the same. Post-accident
functional design criteria as described in [BWR Owners Group Topical
Report NEDO-31558-A, dated March 29, 1993], and approved by the NRC
are satisfied by present equipment installed at VY. NMS
instrumentation is still included in the Technical Specifications
for the [Reactor Protection System] RPS. Control Rod Position
instrumentation does not perform any safety function.
As discussed above, the proposed administrative changes do not
affect any equipment involved in potential initiating events or
safety limits.
Based upon the above, it is concluded that the proposed change
does not involve a significant reduction in a margin of safety.
Based upon the above, we conclude that the proposed change does
not constitute a significant hazards consideration as defined in
10CFR50.92(c).
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: Brooks Memorial Library, 224
Main Street, Brattleboro, VT 05301
Attorney for licensee: John A. Ritsher, Esquire, Ropes and Gray,
One International Place, Boston, MA 02110-2624
NRC Project Director: Phillip F. McKee
Virginia Electric and Power Company, Docket Nos. 50-338 and 50-339,
North Anna Power Station, Units No. 1 and No. 2, Louisa County,
Virginia
Date of amendment request: March 30, 1995
Description of amendment request: The licensee is requesting
temporary changes to Technical Specifications (TS) 3.7.3.1, ``Component
Cooling Water Subsystem - Operating,'' and 3.7.4.1, ``Service Water
System - Operating,'' for NA-1&2. The proposed TS changes will allow
one of the two service water loops to be isolated from the component
cooling water heat exchangers during power operation in order to
refurbish the isolated service water headers.
NA-1&2 is currently pursuing refurbishment of the 18-inch, 20-inch
and 24-inch diameter service water supply and return lines to/from the
NA-1 and NA-2 component cooling heat exchangers (CCHXs). Refurbishment
of this piping presents a challenge in that it is not possible to
isolate and plug or blank the section to be worked in a 7-day time
period. The purpose of the proposed change is to request temporary
changes to the existing servicewater (SW) and component cooling water
(CC) TS to permit orderly and efficient conduct of the pipe
refurbishment project during two-unit power operation. Specifically,
the licensee is proposing to temporarily change TS 3.7.4.1 ``Service
Water System - Operating'' to allow operation of the SW system with one
independent source of SW to/from the NA-1 and NA-2 CCHXs for two
periods of up to 49 days each. This proposed change also allows the
automatic closure feature of the SW valves to/from the CCHXs to be
defeated during the 49-day periods. In addition, the licensee proposes
to temporarily change TS 3.7.3.1 ``Component Cooling Water Subsystem -
Operating'' with a footnote which considers the CC subsystems OPERABLE
with only one independent source of SW provided to/from the CCHXs
during these 49-day periods. Further, the proposed change would allow
that during operation with only one SW header available to/from the
CCHXs, the provisions of Specification 3.0.4 would not be applicable
provided two SW loops are capable of providing cooling for the other
operable plant components.
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:
Specifically, operation of North Anna Power Station in accordance
with the proposed Technical Specifications changes will not:
Involve a significant increase in the probability or
consequences of an accident previously evaluated.
The piping refurbishment project and the proposed temporary
changes to the SW and CC Technical Specifications have been
evaluated to assess their impact on the normal operation of the SW
and CC systems and to ensure that the design basis safety functions
of each system are preserved. The SW system is required to function
during all normal and emergency operating conditions. During normal
plant operation, the SW system provides cooling water to the CCHXs,
charging pump coolers, instrument air compressor coolers, and
control room chiller condensors of both units. During the two 49-day
periods, one header will [operate] with its 24-inch piping to/from
the CCHXs temporarily blanked. To avoid operation of the SW pump at
abnormal conditions (low flow) on this ``partially deadlocked''
header, a temporary cross-connect will be installed to by-pass the
CCHXs. [[Page 24924]]
SW system operation with the cross-connect installed was
evaluated for design basis accident (DBA) conditions. The DBA
condition for the SW system is a loss-of-coolant accident on one
unit with simultaneous loss-of-offsite-power to both units. A SW
system hydraulic analysis has been performed to verify that adequate
flow is provided to the containment recirculation spray heat
exchangers (RSHXs) with the temporary cross-connect installed and
throttled open assuming the occurrence of the most limiting single
failure. Therefore, there is no increase in probability or
consequences of the DBA condition.
Utilizing only one SW header to supply flow to the CCHXs has the
potential to affect the reliability of the CC system and all of the
equipment cooled by CC. The activities to be performed during the
refurbishment project and the various system alignments required
have been evaluated using the Individual Plant Examination (IPE)
Probabilistic Safety Assessment (PSA) model for North Anna Power
Station. This model is used in a manner that is generally consistent
with the Nuclear Energy Institute (NEI)/Electric Power Research
Institute (EPRI) draft PSA Applications Guide (Revision H). The
effect on the PSA model is a slight increase in the frequency of
reactor trips and an increase in the probability of RHR failure.
The increased frequency of reactor trips is due to the decreased
reliability of the CC system to supply cooling to the reactor
coolant pump (RCP) motors. When only one SW header is available to
the CCHXs, the increased frequency of losing this single header can
be conservatively estimated by combining the failure probability of
both SW pumps (approximately 1.5E-4 based on IPE PSA d
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