Applications and Amendments to Facility Operating Licenses Involving No Significant Hazards Considerations; Biweekly Notice
Federal RegisterJun 6, 1995
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[Federal Register Volume 60, Number 108 (Tuesday, June 6, 1995)]
[Notices]
[Pages 29869-29896]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: 95-13759]
[[Page 29869]]
NUCLEAR REGULATORY COMMISSION
Applications and Amendments to Facility Operating Licenses
Involving No Significant Hazards Considerations; Biweekly Notice
I. Background
Pursuant to Public Law 97-415, the U.S. Nuclear Regulatory
Commission (the Commission or NRC staff) is publishing this regular
biweekly notice. Public Law 97-415 revised section 189 of the Atomic
Energy Act of 1954, as amended (the Act), to require the Commission to
publish notice of any amendments issued, or proposed to be issued,
under a new provision of section 189 of the Act. This provision grants
the Commission the authority to issue and make immediately effective
any amendment to an operating license upon a determination by the
Commission that such amendment involves no significant hazards
consideration, notwithstanding the pendency before the Commission of a
request for a hearing from any person.
This biweekly notice includes all notices of amendments issued, or
proposed to be issued from May 15, 1995, through May 25, 1995. The last
biweekly notice was published on Tuesday, May 23, 1995 (60 FR 27334).
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 July 7, 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 [[Page 29870]] limitations in the order granting leave
to intervene, and have the opportunity to participate fully in the
conduct of the hearing, including the opportunity to present evidence
and cross-examine witnesses.
If a hearing is requested, the Commission will make a final
determination on the issue of no significant hazards consideration. The
final determination will serve to decide when the hearing is held.
If the final determination is that the amendment request involves
no significant hazards consideration, the Commission may issue the
amendment and make it immediately effective, notwithstanding the
request for a hearing. Any hearing held would take place after issuance
of the amendment.
If the final determination is that the amendment request involves a
significant hazards consideration, any hearing held would take place
before the issuance of any amendment.
A request for a hearing or a petition for leave to intervene must
be filed with the Secretary of the Commission, U.S. Nuclear Regulatory
Commission, Washington, DC 20555, Attention: Docketing and Services
Branch, or may be delivered to the Commission's Public Document Room,
the Gelman Building, 2120 L Street, NW., Washington DC, by the above
date. Where petitions are filed during the last 10 days of the notice
period, it is requested that the petitioner promptly so inform the
Commission by a toll-free telephone call to Western Union at 1-(800)
248-5100 (in Missouri 1-(800) 342-6700). The Western Union operator
should be given Datagram Identification Number N1023 and the following
message addressed to (Project Director): petitioner's name and
telephone number, date petition was mailed, plant name, and publication
date and page number of this Federal Register notice. A copy of the
petition should also be sent to the Office of the General Counsel, U.S.
Nuclear Regulatory Commission, Washington, DC 20555, and to the
attorney for the licensee.
Nontimely filings of petitions for leave to intervene, amended
petitions, supplemental petitions and/or requests for a hearing will
not be entertained absent a determination by the Commission, the
presiding officer or the Atomic Safety and Licensing Board that the
petition and/or request should be granted based upon a balancing of
factors specified in 10 CFR 2.714(a)(1)(i)-(v) and 2.714(d).
For further details with respect to this action, see the
application for amendment which is available for public inspection at
the Commission's Public Document Room, the Gelman Building, 2120 L
Street, NW., Washington, DC, and at the local public document room for
the particular facility involved.
Arizona Public Service Company, et al., Docket Nos. STN 50-528, STN 50-
529, and STN 50-530, Palo Verde Nuclear Generating Station, Unit Nos.
1, 2, and 3, Maricopa County, Arizona
Date of amendment requests: March 24, 1995.
Description of amendment requests: The proposed amendments would
make numerous changes to Technical Specification (TS) 3/4.8.1, ``A.C.
Sources,'' and the associated TS Bases, for Palo Verde Units 1, 2, and
3. The proposed amendments would implement recommended changes from
NUREG-1432, ``Standard Technical Specifications: Combustion Engineering
Plants''; Generic Letter (GL) 94-01, ``Removal of Accelerated Testing
and Special Reporting Requirements for Emergency Diesel Generators'';
and GL 93-05, ``Line-Item Technical Specification Improvements to
Reduce Surveillance Requirements for Testing During Power Operation.''
The proposed changes are intended to increase emergency diesel
generator (EDG) reliability by reducing the stresses on the EDGs from
unnecessary testing. Additional changes have also been proposed to TS
3/4.8.1 to further enhance EDG reliability, to achieve consistency with
NUREG-1432, Combustion Engineering Standard TS, and to improve the TS
presentation.
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 to TS 3/4.8.1 and the associated Bases
affect the required actions in response to inoperable offsite and
onsite AC sources, surveillance requirements for the EDG, and
reporting requirements for EDG failures. The majority of the
proposed changes are based on the recommendations of NUREG 1432, GL
94-01, and GL 93-05. These proposed changes have been extensively
reviewed by the NRC during the preparation of these documents, and
by APS during the development of this request for TS amendment. The
proposed changes are expected to result in improvements in EDG
performance and reduce EDG aging due to excessive testing. The
proposed changes will permit the elimination of the unnecessary
mechanical stress and wear on the EDGs while ensuring that the EDGs
will perform their design function. The elimination of mechanical
stress and wear will improve reliability and availability of the
EDGs which will have a positive effect on the ability of the EDGs to
perform their design function. The proposed changes to [do] not
affect the availability or the testing requirements of the offsite
circuits.
Because the proposed changes do not affect the design or
performance of the EDGs or their ability to perform their design
function, the changes are expected to result in a decrease in the
probability or consequences of an accident previously evaluated. The
proposed changes will increase EDG reliability, thereby increasing
overall plant safety. Because these changes do not affect the
probability of accident precursors (EDGs do not initiate any
accidents), the proposed type license amendment 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 accident previously
evaluated.
The proposed changes to TS 3/4.8.1 and the associated Bases do
not introduce any new modes of plant operation or new accident
precursors, involve any physical alterations to plant
configurations, or make any changes to system setpoints which could
initiate a new or different kind of accident. The proposed changes
do not affect the design or performance characteristics of any EDG
or its ability to perform its design function. No new failure modes
have been defined nor new system interactions introduced for any
plant system or component, nor has any new limiting failure been
identified as a result of the proposed changes. The proposed changes
will eliminate unnecessary EDG testing, increasing EDG reliability
and availability, and thereby having an overall positive affect on
plant safety. Accidents concerning loss of offsite power and a
single failure (e.g., loss of an EDG) have previously been
evaluated. These changes are intended to improve plant safety,
decrease equipment degradation, and remove unnecessary burden on
personnel resources by reducing the amount of testing that the TS
requires during power operation. Therefore, the proposed license
amendment 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.
Under the proposed changes to TS 3/4.8.1 and the associated
Bases, the EDGs will remain capable of performing their safety
function. The changes do not affect the design or performance of any
EDG, but will increase EDG reliability and availability by reducing
the stresses and the effects of aging on the EDG by eliminating
unnecessary testing. This will result in an overall increase in
plant safety. Since the ability of the EDGs to perform their safety
function will not be degraded, the proposed license amendment does
not involve a significant reduction in the margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
that [[Page 29871]] review, it appears that the three standards of
50.92(c) are satisfied. Therefore, the NRC staff proposes to determine
that the amendment requests involve no significant hazards
consideration.
Local Public Document Room location: Phoenix Public Library, 12
East McDowell Road, Phoenix, Arizona 85004.
Attorney for licensees: Nancy C. Loftin, Esq., Corporate Secretary
and Counsel, Arizona Public Service Company, P.O. Box 53999, Mail
Station 9068, Phoenix, Arizona 85072-3999.
NRC Project Director: William H. Bateman.
Arizona Public Service Company, et al., Docket Nos. STN 50-528, STN 50-
529, and STN 50-530, Palo Verde Nuclear Generating Station, Unit Nos.
1, 2, and 3, Maricopa County, Arizona
Date of amendment requests: March 31, 1995.
Description of amendment requests: The proposed amendment would
clarify the shutdown margin definition, change the shutdown margin
applicability and surveillance requirements to comply with safety
analysis assumptions for subcritical inadvertent control element
assembly withdrawal (UFSAR Section 15.4), and expand the applicability
for core protection calculator (CPC) operability. In addition, the
proposed amendment would add a reference to the Core Operating Limits
Report (COLR) for the MODE 6 refueling boron concentration limit. The
proposed amendment would also change the power calibration requirements
for the linear power level, the CPC delta T power, and CPC nuclear
power signals to allow more conservative settings than presently
required.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis about the issue of no significant hazards
consideration, which is presented below:
1. The proposed changes do not involve a significant increase in
the probability or consequences of an accident previously evaluated.
These changes are being made to ensure compliance with the
safety analysis assumptions for subcritical inadvertent CEA [control
element assembly] withdrawal. These changes also ensure that the
boron concentration in the reactor is sufficient to prevent
criticality if an inadvertent withdrawal of a shutdown CEA bank were
to occur with all other CEAs inserted. Therefore, the consequences
of the inadvertent CEA withdrawal is no greater than those of the
event previously evaluated. This change also has no affect on the
probability of an accident since it is not introducing or changing
any accident initiating mechanism.
The analysis of uncontrolled CEA withdrawal from MODES 2 and 3
subcritical with four RCPs [reactor coolant pumps] running is
presented in UFSAR Section 15.4.1 as an anticipated operational
occurrence. The consequences of this event are that the acceptable
fuel design limits are not exceeded (General Design Criterion 25 as
specified in the NRC Standard Review Plan). The proposed change to
TS requiring that either the CPCs or Logarithmic Power Level--High
trip (trip setpoint lowered to 10-4% of Rated Thermal Power)
are Operable in MODES 3, 4, and 5, ensures that an inadvertent CEA
withdrawal with less than four pumps operating, results in
consequences no greater than those of the previously evaluated
uncontrolled CEA withdrawal event.
The revised TS will also ensure that the reactivity worth of any
full-length CEAs not capable of being inserted is accounted for in
the determination of the shutdown margin. This change will ensure
the shutdown margin will continue to be within safety analysis
assumptions for previously evaluated accidents.
The proposed changes to TS, replacing the MODE 6 boron
concentration specification with the requirement to maintain the
boron concentration within the limit specified in the COLR, will not
affect the probability or consequences of an accident, because it is
not changing the MODE 6 reactivity requirement of Keff less
than or equal to 0.95, but provides a specific boron concentration
value in the COLR to ensure the MODE 6 required Keff value of
less than or equal to 0.95 is met.
The proposed changes will reduce the amount of non-conservatism
presently allowed for the linear power level, the CPC delta T power
and CPC nuclear power signals. Changing the tolerance range from
plus or minus 2% to between -0.5% and 10% between 15% and 80% RATED
THERMAL POWER, except during initial post refueling power ascension
and restricting recalibration, will allow more conservative settings
than currently required.
2. The proposed changes do not create the possibility of a new
or different kind of accident from any accident previously analyzed.
The changes revising the mode applicabilities are being made to
comply with safety analysis assumptions for subcritical CEA
withdrawal. The SR [surveillance requirement] ensures that the
shutdown margin is within the safety analysis assumptions when the
reactor trip breakers are open and any full-length CEA is not fully
inserted. No new or different kind of accident will be initiated
since this change will ensure that the required shutdown margin is
maintained when the reactor trip breakers are closed.
The proposed change to TS, requiring either the CPCs or
Logarithmic Power Level--High trip to be operable, will provide
protection from inadvertent CEA withdrawal when less than four RCPs
are operating. No new or different kind of accident will be
initiated by this change, since this change incorporates TS
limitations to ensure protection for an existing accident scenario.
The revised TS shutdown margin definition ensures that the
reactivity worth of any full-length CEAs not capable of being
inserted is accounted for in the determination of the shutdown
margin. This ensures the shutdown margin will continue to be within
safety analysis assumptions. Maintaining the shutdown margin within
the safety analyses assumption will not create any new or different
kind of accident.
The proposed changes to TS power calibration tolerance limits
are conservative relative to the current TS requirements and
therefore will not create any new or different kind of accident.
The proposed change to TSs replacing the MODE 6 boron
concentration specification with the requirement to maintain the
boron concentration within the limit specified in the COLR does not
create the possibility of a new or different kind of accident from
any accident previously analyzed. The proposed change is not
changing the MODE 6 reactivity requirements of less than or equal to
0.95 while providing a specific boron concentration value in the
COLR to ensure the MODE 6 required Keff value of less than or
equal to 0.95.
3. The proposed changes do not involve a significant reduction
in a margin of safety.
The proposed change to TS adds an additional requirement for the
CPCs or Logarithmic Power Level--High trip to be operable in MODES
3, 4, and 5. This change maintains the margin of safety in the
safety analysis by providing a TS that will ensure appropriate
protection is provided in the event of an inadvertent CEA withdrawal
with less than four RCPs operating.
The proposed changes to TS (Boration Control, Shutdown Margin),
revising the mode applicabilities, maintains the margin of safety
provided in the TS by ensuring that the safety analysis assumptions
for subcritical CEA withdrawal are met. The new SR does not reduce
the margin of safety since the shutdown margin assumed in the safety
analysis will be maintained by this TS.
The revised TS shutdown margin definition ensures that the
reactivity worth of any full length CEAs not capable of being
inserted is accounted for in the determination of shutdown margin.
This ensures shutdown margin will continue to be within safety
analysis assumptions. This change maintains the margin of safety
that is currently provided by TS.
The proposed changes to TS, reducing the amount of non-
conservatism in the safety system power indications, maintains the
margin of safety for design basis events which take credit for the
linear power level, the CPC delta T power, and CPC nuclear power
signals.
The proposed change to TS moves the specific MODE 6 boron
concentration value to COLR. The proposed change does not change the
MODE 6 reactivity requirement of Keff of less than or equal to
0.95, but provides a specific boron concentration value in the COLR
to ensure the MODE 6 required Keff value of less than or equal
to 0.95 is met. Therefore, the margin of safety is not affected by
the proposed change.
[[Page 29872]] The NRC staff has reviewed the licensees' analysis
and, based on that review, it appears that the three standards of
50.92(c) are satisfied. Therefore, the NRC staff proposes to determine
that the amendment requests involve no significant hazards
consideration.
Local Public Document Room location: Phoenix Public Library, 12
East McDowell Road, Phoenix, Arizona 85004.
Attorney for licensees: Nancy C. Loftin, Esq., Corporate Secretary
and Counsel, Arizona Public Service Company, P.O. Box 53999, Mail
Station 9068, Phoenix, Arizona 85072-3999.
NRC Project Director: William H. Bateman.
Commonwealth Edison Company, Docket Nos. 50-237 and 50-249, Dresden
Nuclear Power Station, Units 2 and 3, Grundy County, Illinois, Docket
Nos. 50-254 and 50-265, Quad Cities Nuclear Power Station, Units 1 and
2, Rock Island County, Illinois
Date of application for amendment request: December 8, 1992, as
supplemented on September 10, 1993, and May 17, 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 Dresden Nuclear Power
Station and sister site Quad Cities Nuclear Power Station, needed
attention focused on the existing custom Technical Specifications (TS)
used.
The licensee made the decision to initiate a Technical
Specification Upgrade Program (TSUP) for both Dresden and Quad Cities.
The licensee evaluated the current TS for both Dresden and Quad Cities
against the Standard Technical Specifications (STS) contained in NUREG-
0123, ``Standard Technical Specifications General Electric Plants BWR/
4.'' The licensee's evaluation identified numerous potential
improvements such as clarifying requirements, changing TS to make them
more understandable and to eliminate interpretation, and deleting
requirements that are no longer considered current with industry
practice. As a result of the evaluation, ComEd has elected to upgrade
both Dresden and Quad Cities TS to the STS contained in NUREG-0123.
The TSUP for Dresden and Quad Cities is not a complete adaptation
of the STS. The TSUP focuses on (1) Integrating additional information
such as equipment operability requirements during shutdown conditions,
(2) clarifying requirements such as limiting conditions for operations
and action statements utilizing STS terminology, (3) deleting
superseded requirements and modifications to the TS based on the
licensee's responses to Generic Letters (GLs), and (4) relocating
specific items to more appropriate TS locations.
The December 8, 1992, application, as supplemented on September 10,
1993, and May 17, 1995, proposed to upgrade only Section 3/4.1 (Reactor
Protection System) of the Dresden and Quad Cities TS.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
(1) The proposed changes do not involve a significant increase
in the probability or consequences of an accident previously
evaluated because:
In general, the proposed amendment represents the conversion of
current requirements to a more generic format, or the addition of
requirements which are based on the current safety analysis.
Implementation of these changes will provide increased reliability
of equipment assumed to operate in the current safety analysis, or
provide continued assurance that specified parameters remain within
their acceptance limits, and as such, will not significantly
increase the probability or consequences of a previously evaluated
accident.
Some of the proposed changes 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
Dresden and Quad Cities Station's Technical Specification Section 3/
4.1 are based on BWR-STS (NUREG-0123, Revision 4 ``Standard
Technical Specifications General Electric Plants BWR/4) guidance or
NRC accepted changes at later operating BWR plants. Any deviations
from BWR-STS and CTS requirements do not significantly increase the
probability or consequences of any previously evaluated accident for
Dresden and Quad Cities Station. These 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, or provide continued assurance that specified parameters
remain within their acceptance limits. As such, these changes will
not significantly increase the probability or consequences of a
previously evaluated accident.
The associated systems that make up the Reactor Protection
System are not assumed in any safety analysis to initiate any
accident sequence for both Dresden and Quad Cities Stations;
therefore, the probability of any accident previously evaluated is
not increased by the proposed amendment. In addition, the proposed
surveillance requirements for the proposed amendments to these
systems are generally more prescriptive than the current
requirements specified within the Technical Specifications. These
more prescriptive surveillance requirements increase the probability
that the Reactor Protection System will perform its intended
function. Therefore, the proposed TS will improve the reliability
and availability of all affected systems and reduce the consequences
of any accident previously evaluated.
(2) Create the possibility of a new or different kind of
accident from any previously evaluated because:
In general, the proposed amendment represents the conversion of
current requirements to a more generic format, or the addition of
requirements which are based on the current safety analysis. Others
represent minor curtailments of the current requirements which are
based on generic guidance or previously approved provisions for
other stations. These changes do not involve revisions to the design
of the station. Some of the changes may involve revision in the
operation of the station; however, these changes provide for
additional restrictions which are in accordance with the current
safety analyses, or are to provide for additional testing or
surveillances which will not introduce new failure mechanisms beyond
those already considered in the current safety analyses. Therefore,
these changes will not create the possibility of a new or different
kind of accident from any accident previously evaluated.
The proposed amendment for Dresden and Quad Cities Station's
Technical Specification Section 3/4.1 is based on BWR-STS guidelines
or NRC accepted changes at later operating BWR plants. The proposed
amendment has been reviewed for acceptability at the Dresden and
Quad Cities Nuclear Power Stations considering similarity of system
or component design versus the BWR-STS or later operating BWRs. Any
deviations from BWR-STS or CTS requirements do not create the
possibility of a new or different kind of accident than previously
evaluated for Dresden and Quad Cities Stations. No new modes of
operation are introduced by the proposed changes. Surveillance
requirements are changed to reflect improvements in technique,
frequency of performance or operating experience at later plants.
Proposed changes to action statements in many places add
requirements that are not in the present technical specifications or
adopt requirements that have been used at other operating BWRs with
design similar to Dresden and Quad Cities. The proposed changes
maintain at least the present level of operability. Therefore, the
proposed changes do not create the possibility of a new or different
kind of accident from any previously evaluated.
The associated systems that make up the Reactor Protection
System are not assumed in any safety analysis to initiate any
accident sequence for Dresden and Quad Cities Stations. In addition,
the proposed surveillance requirements for affected
[[Page 29873]] systems associated with the Reactor Protection System
are generally more prescriptive than the current requirements
specified within the Technical Specifications; therefore, the
proposed changes do not create the possibility of a new or different
kind of accident from any previously evaluated.
(3) Involve a significant reduction in the margin of safety
because:
In general, the proposed amendment represents the conversion of
current requirements to a more generic format, or the addition of
requirements which are based on the current safety analysis. Others
represent minor curtailments of the current requirements which are
based on generic guidance or previously approved provisions for
other stations. Some of the later individual items may introduce
minor reductions in the margin of safety when compared to the
current requirements. However, other individual changes are the
adoption of new requirements which will provide significant
enhancement of the reliability of the equipment assumed to operate
in the safety analysis, or provide enhanced assurance that specified
parameters remain 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 proposed amendment to Technical Specification Section 3/4.1
implements present requirements, or the intent of present
requirements in accordance with the guidelines set forth in the BWR-
STS. Any deviations from BWR-STS and CTS requirements do not
significantly reduce the margin of safety for Dresden and Quad
Cities Stations. The proposed changes are intended to improve
reliability, usability, and the understanding of technical
specification requirements while maintaining acceptable levels of
safe operation. The proposed changes have been evaluated and found
acceptable for use at Dresden and Quad Cities based on system
design, safety analysis requirements and operational performance.
Since the proposed changes are based on NRC accepted provisions at
other operating plants that are applicable at Dresden and Quad
Cities and maintain necessary levels of system or component
readability, the proposed changes do not involve a significant
reduction in the margin of safety.
The proposed amendment for Dresden and Quad Cities Stations will
not reduce the availability of systems associated with the Reactor
Protection System when required to mitigate accident conditions;
therefore, the proposed changes do not involve a significant
reduction in the margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: For Dresden, Morris Area
Public Library District, 604 Liberty Street, Morris, Illinois 60450;
for Quad Cities, Dixon Public Library, 221 Hennepin Avenue, Dixon,
Illinois 61021.
Attorney for licensee: Michael I. Miller, Esquire; Sidley and
Austin, One First National Plaza, Chicago, Illinois 60603.
NRC Project Director: Robert A. Capra.
Detroit Edison Company, Docket No. 50-341, Fermi-2, Monroe County,
Michigan
Date of amendment request: July 29, 1993.
Description of amendment request: The proposed amendment would
extend the instrument calibration intervals for selected plant
instrumentation from 18 months to 36 months.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
The proposed change to extend to 36 months the calibration
interval of selected instrumentation does not:
(1) Involve a significant increase in the probability or
consequences of an accident previously evaluated. The purpose of the
proposed Technical Specification change is to extend calibration
interval testing requirements for selected instrumentation. However,
because of the continued application of redundant Technical
Specification requirements such as channel checks, channel
functional tests, and logic system functional tests, the performance
of these instruments will be maintained within the acceptance limits
assumed in plant safety analyses and required for the successful
mitigation of an initiating event. The proposed Technical
Specification changes do not affect the capability of the associated
systems to perform their intended function within their instrument
settings.
These other tests are sufficient to identify failure modes or
degradations in instrument performance and ensure operation of the
associated systems within acceptance limits. There are no credible
failure modes that can be detected by instrument calibration that
cannot also be detected by other Technical Specification tests.
(2) Create the possibility of a new or different kind of
accident from any accident previously evaluated. As discussed above,
the proposed Technical Specification changes do not affect the
capability of the associated systems to perform their intended
function within the acceptance limits assumed in plant safety
analyses and required for successful mitigation of an initiating
event. All plant systems continue to operate in an identical manner.
No new accident modes are created.
(3) Involve a significant reduction in a margin of safety. The
current Technical Specification allowable values are based on the
maximum analytical limits assumed in the plant safety analyses.
These analyses conservatively establish the margin of safety. The
proposed Technical Specification changes do not affect the
capability of the associated systems to perform their function
within the instrument settings used as the basis for the plant
safety analyses. Plant and system settings to an initiating events
will remain in compliance within the assumptions of the safety
analyses, and therefore the margin of safety is not affected.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: Monroe County Library System,
3700 South Custer Road, Monroe, Michigan 48161.
Attorney for licensee: John Flynn, Esq., Detroit Edison Company,
2000 Second Avenue, Detroit, Michigan 48226.
NRC Project Director: Cynthia A. Carpenter, Acting.
Detroit Edison Company, Docket No. 50-341, Fermi-2, Monroe County,
Michigan
Date of amendment request: December 15, 1994.
Description of amendment request: The proposed amendment would
relocate, revise, or delete various Technical Specification (TS)
provisions. Administrative controls on working hours in TS 6.2.2.f, the
Independent Safety Engineering Group requirements in TS 6.2.3, the unit
staff qualification requirements in TS 6.3, the reportable event
requirement for the Onsite Review Organization (OSRO) in TS 6.6.1.b,
the radiation protection program requirements in TS 6.11, the record
retention requirements in TS 6.10, and the review and audit functions
in TS 6.5 (with the exception of TS 6.5.2.8), would be relocated to
Chapter 13 of the Updated Final Safety Analysis Report (UFSAR). The
review and approval process for temporary changes to each TS 6.8.1
plant procedure listed in TS 6.8.4 would also be relocated to Chapter
13 of the UFSAR.
The requirements of TS 6.5.2.8, the review and approval process for
administrative procedures in TS 6.8.2, and the review and approval
process for plant procedures in TS 6.8.3, would be relocated to the
Fermi 2 Quality Assurance program. The in-plant radiation monitoring
program requirements in TS 6.8.5.b, and the high radiation area
requirements in TS 6.12 would be relocated to Chapter 12 of the
[[Page 29874]] UFSAR. The radiological environmental monitoring program
requirements in TS 6.8.5.f would be relocated to Chapter 11 of the
UFSAR. The Process Control Program (PCP) requirements in TS 6.13 would
be relocated to the PCP.
The requirements for OSRO to review the Security Plan in TS
6.5.1.6.j and to have Security Plan implementing procedures in TS
6.8.1.e would be relocated to the Fermi 2 Security Plan. The
requirements for OSRO to review the Emergency Plan in TS 6.5.1.6.k and
to have Emergency Plan implementing procedures in TS 6.8.1.f would be
relocated to the Fermi 2 Emergency Plan.
The unit staff qualification requirements, as specified in the H.
R. Denton (NRC) letter of March 29, 1980, in TS 6.3, would be deleted.
The licensee states these have been superseded by 10 CFR Part 55 and
Generic Letter (GL) 87-07. The training requirements in TS 6.4 would be
deleted. The licensee states that other Section 6.0 TS and NRC
regulations provide sufficient control of these training requirements.
The submittal requirement of the annual radioactive effluent release
report in TS 6.9.1.8 would be revised from ``within 90 days after
January 1 * * *'' to ``prior to May 1. * * *''
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
(1) The proposed changes do not involve a significant increase
in the probability or consequences of an accident previously
evaluated because the proposed changes are administrative in nature.
None of the proposed changes involve a physical modification to the
plant, a new mode of operation or a change to the UFSAR transient
analyses. No Limiting Condition for Operation, ACTION statement or
Surveillance Requirement is affected by any of the proposed changes.
Also, these proposed changes, in themselves, do not reduce the level
of qualification or training such that personnel requirements would
be decreased. Therefore, this change is administrative in nature and
does not involve a significant increase in the probability or
consequences of an accident previously evaluated. Further, the
proposed changes do not alter the design, function, or operation of
any plant component and therefore, do not affect the consequences of
any previously evaluated accident.
(2) The proposed changes do not create the possibility of a new
or different kind of accident from any accident previously evaluated
because the proposed changes do not introduce a new mode of plant
operation, surveillance requirement or involve a physical
modification to the plant. The proposed changes are administrative
in nature. The changes propose to revise, delete or relocate the
stated administrative control provisions from the TS to the UFSAR,
plant procedures or the QA Program whereby, adequate control of
information is maintained. Further, as stated above, the proposed
changes do not alter the design, function, or operation of any plant
components and therefore, no new accident scenarios are created.
(3) The proposed changes do not involve a significant reduction
in a margin of safety because they are administrative in nature.
None of the proposed changes involve a physical modification to the
plant, a new mode of operation or a change to the UFSAR transient
analyses. No Limiting Condition for Operation, ACTION statement or
Surveillance Requirement is affected. The proposed changes do not
involve a significant reduction in a margin of safety. Additionally,
the proposed change does not alter the scope of equipment currently
required to be OPERABLE or subject to surveillance testing nor does
the proposed change affect any instrument setpoints or equipment
safety functions. Therefore, the change does not involve a
significant reduction in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: Monroe County Library System,
3700 South Custer Road, Monroe, Michigan 48161.
Attorney for licensee: John Flynn, Esq., Detroit Edison Company,
2000 Second Avenue, Detroit, Michigan 48226.
NRC Project Director: Cynthia A. Carpenter, Acting.
Duquesne Light Company, et al., Docket No. 50-412, Beaver Valley Power
Station, Unit 2, Shippingport, Pennsylvania
Date of amendment request: April 26, 1995.
Description of amendment request: The proposed amendment would add
a requirement to Technical Specification (TS) 4.5.2.a to periodically
verify that the High Head Safety Injection (HHSI) pump minimum flow
valve, 2CHS*MOV373, is maintained open during plant operation in Modes
1, 2, and 3. Valve 2CHS*MOV373 must be maintained open to provide a
minimum flowpath for the HHSI pumps and thereby minimize the likelihood
of HHSI pump damage due to operating the pumps with insufficient flow.
The proposed change would allow flexibility for local verification of
valve position or flow indication if the control room indication is not
available. The proposed amendment would also make several editorial
changes to TS 3/4.5.2 for consistent format with other 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:
1. Does the change involve a significant increase in the
probability or consequences of an accident previously evaluated?
Maintaining 2CHS*MOV373 in a de-energized locked open position
ensures charging/High Head Safety Injection pump (HHSI pump) minimum
flow remains available for normal operation and design basis
accidents. It has been determined that with 2CHS*MOV373 in the open
position there is no significant increase in radiation levels and no
change to the existing environmental qualification or personnel
access routes. Sufficient injection flow to the core will be
maintained during events requiring a Safety Injection (SI)
actuation. Potential HHSI pump damage due to low flow will be
prevented during periods of high Reactor Coolant System (RCS)
pressure following a steam line break and SI. It has also been
determined that the HHSI pumps will remain capable of performing
their safety function with a continuous minimum flow. There is no
impact on analysis assumptions or radiological consequences of an
accident.
There are no postulated events in the Updated Final Safety
Analysis Report (UFSAR) which require that 2CHS*MOV373 be closed.
Thus, the decision to de-energize and lock open the valve ensures
adequate minimum flow for the HHSI pumps.
The proposed addition of 2CHS*MOV373 to Technical Specification
3.5.2 enhances the operator's ability to verify the valve position.
The proposed surveillances and footnote will be used to monitor the
valve position, the status of motor operator, and the valve position
indicating lights. Therefore, the proposed change to the technical
specification will ensure that the HHSI pump minimum flow is always
available.
Several editorial changes were also made to Technical
Specification 3.5.2. These changes do not alter the intent of the
technical specification and as such have no impact on previously
evaluated accident scenarios.
Therefore, the proposed change will not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
2. Does the change create the possibility of a new or different
kind of accident from any accident previously evaluated?
The proposed addition of 2CHS*MOV373 to the technical
specifications does not involve changes to the physical plant. The
proposed change adds surveillance requirements and a footnote which
monitor the valve position, the lack of power to the
[[Page 29875]] motor operator, and the valve position indicating
lights. This assures that the minimum flow valve is open to maintain
the HHSI pumps operable under all conditions.
Therefore, the proposed change does not create the possibility
of a new or different kind of accident from any accident previously
evaluated.
3. Does the change involve a significant reduction in a margin
of safety?
The proposed change provides additional action to ensure that
2CHS*MOV373 remains open and minimum HHSI pump flow remains
available. Safety limits and limiting safety system settings are not
affected by this proposed change. There are no changes to the
offsite dose consequences resulting from this request.
Therefore, use of the proposed technical specification would not
involve a significant reduction in the margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: B. F. Jones Memorial Library,
663 Franklin Avenue, Aliquippa, Pennsylvania 15001.
Attorney for licensee: Gerald Charnoff, Esquire, Jay E. Silberg,
Esquire, Shaw, Pittman, Potts & Trowbridge, 2300 N Street, NW.,
Washington, DC 20037.
NRC Project Director: John F. Stolz.
Entergy Operations Inc., Docket No. 50-382, Waterford Steam Electric
Station, Unit 3, St. Charles Parish, Louisiana
Date of amendment request: July 18, 1991, as supplemented by
letters dated March 16, and December 2, 1994, and March 9, 1995.
Description of amendment request: The proposed amendment would
revise the Technical Specifications (TSs) on control Room Air
Conditioning System (CRACS) by separating the current composite
requirements of TS 3.7.6 into four TSs covering three separate
functions; control room emergency air filtration system (two mode
sets), control room air temperature, and control room isolation and
pressurization. The changes also increase the allowed outage time to
identify and correct breaches to the control room envelope, adds
requirements for make-up air flow rate to be used in conjunction with
existing differential pressure requirements, and adds toxic gas
specifications for Modes 5 and 6. The amendment is related to a
revision to the Technical Specification Bases approved by the NRC in a
letter dated August 9, 1988. The March 16, and December 2, 1994, and
March 9, 1995 submittals provided additional information and included
some additional restrictions in proposed changes by original
application dated July 18, 1991. The original notice was published on
September 4, 1991 (56 FR 43808). The additional submittals do not
change the no significant hazard consideration determination previously
made by the licensee.
Basis for proposed no significant hazards consideration
determination: The proposed change would create new Specifications as
follows: 3/4.7.6.1 Emergency Air Filtration, Modes 1-4; 3/4.7.6.2
Emergency Air Filtration, Modes 5 and 6; 3/4.7.6.3 Control Room Air
Temperature; 3/4.7.6.4 Control Room Isolation and Pressurization. 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 limiting accidents against which the CRACS protects are:
All Chapter 15 scenarios involving a release of
radiation to the environment outside the containment,
Toxic gas releases, and
Smoke resulting from control room envelope fires.
Limiting accidents against which the emergency air filtration
system protects are all Chapter 15 scenarios involving release of
radiation to the environment outside the containment.
The probability and consequences of any of the limiting
accidents listed above are unchanged by the specialization of the
plant TSs. As pointed out in the description of the change, TSs 3/
4.7.6.1 and 3/4.7.6.2 have retained all requirements from the
existing TS with the addition of one action statement based on the
inoperability of both trains, and the exception of one action
statement based on one inoperable train in Modes 5 or 6. This action
statement is unnecessary since it is only applicable in a mode
unlikely to experience the limiting design basis accidents against
which this system protects. Therefore, the protection of the
original specification is uncompromised for the function of
emergency air filtration.
There are two differences between the existing TS and the
proposed TS 3/4.7.6.3 regarding control room air temperature. The
first is the three hour outage allowed when both air conditioning
units are inoperable [this was withdrawn by licensee's March 9,
1995, letter].
This corrects most types of failures. Although three hours are
less restrictive than TS 3.0.3, it is not significantly less and
therefore, does not seriously reduce the protection of the original
specification. The other change is the reduction of the surveillance
temperature from 110 deg.F to 80 deg.F. This is more restrictive
than the existing version. All other requirements for air
conditioning are retained in the proposed TS.
Proposed TS 3/4.7.6.4, which concerns control room isolation and
pressurization, allows more limited continued plant operation than
the existing TS. When compared to existing actions required for
continued operation with a known breach, the proposed specification
recognizes the potential consequences that could arise from
operation with an unidentified breach in the envelope and imposes
more restrictive actions.
Engineering analysis also shows that, for most of the time,
toxic chemical concentrations in the control room envelope after a
postulated release are largely the result of in-leakage from the RAB
[reactor auxiliary building] after isolation. This has the effect of
reducing the chemical concentration of gas leaking into the control
room by at least an order of magnitude and ultimately results in a
control room chemical concentration buildup rate slower than
previously assumed. These characteristics make it likely that the
operators would have sufficient time to don the breathing apparatus
installed in the control room. It is also noteworthy that this
emergency breathing apparatus is considered by Regulatory Guide 1.78
to provide sufficient operator protection for those cases where
chemical toxicity limits might be exceeded.
The limited continued operation allowed by the proposed change,
the design characteristics of the control room, and the installed
breathing apparatus provides a reasonable level of protection for
plant personnel. Some new restrictions are identified for the
control room isolation and pressurization. These were not previously
identified and therefore offer enhanced protection to the TS. All
existing requirements specific to the isolation and pressurization
function are retained in the proposed version. As such, the proposed
specification offers more protection than the existing TS.
Based on the above, these revisions to the TS will not adversely
affect the reliability or performance of any installed equipment.
There are no design changes associated with this proposed amendment,
consequently, all aspects of the safety analysis will remain
unchanged and there will be no physical change to the facility, and
operation of Waterford 3 in accordance with these proposed changes
will not involve a significant increase in the probability or
consequence of any accident previously evaluated.
To create a new or different kind of accident, these changes
must introduce a new failure path. In this regard, these revisions
are benign since they do not alter the system or its operation. With
a few exceptions, all existing TS restrictions have been retained.
The exceptions have been shown to have insignificant impact.
Furthermore, several additional restrictions, not in the existing
specification, have been added.
Based on the above information, these changes do not introduce a
new failure path and therefore, cannot create a new, unevaluated
sequence of events. The current plant safety analyses are bounding
and this revision will not create the possibility of a new or
different kind of accident from any accident previously evaluated.
[[Page 29876]]
Safety margins related to the control room envelope air systems
are established for control room temperature and the habitability of
the control room following all credible accidents. This change does
not modify the equipment installed in the plant or its operation.
Therefore, existing margins of safety are retained, and the
operation of Waterford 3 in accordance with this proposed change
will not involve a significant reduction in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room Location: University of New Orleans
Library, Louisiana Collection, Lakefront, New Orleans, LA 70122.
Attorney for licensee: N.S. Reynolds, Esq., Winston & Strawn 1400 L
Street N.W., Washington, D.C. 20005-3502.
NRC Project Director: William D. Beckner.
Entergy Operations Inc., Docket No. 50-382, Waterford Steam Electric
Station, Unit 3, St. Charles Parish, Louisiana
Date of amendment request: May 12, 1995.
Description of amendment request: The proposed change modifies
surveillance requirements associated with containment leakage Technical
Specification (TS) 3.6.1.2 by removing scheduler requirement for Type A
tests to be performed specifically at 40 plus or minus 10 month
intervals and, instead, reference Type A testing in accordance with 10
CFR part 50, appendix J. The proposed change adopts the wording for
primary containment integrated leak rate testing that is consistent
with the requirements of the Combustion Engineering Improved Standard
Technical Specifications (NUREG-1432). The proposed change also
includes several administrative changes. The May 12, 1995, submittal
superseded the November 16, 1993, submittal in its entirety. The
November 16, 1993, submittal was noticed in Federal Register on January
5, 1994
(59 FR 619).
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
The proposed change will not affect the assumptions, design
parameters, or results of any accident previously evaluated. The
proposed change does not add or modify any existing equipment. The
proposed Type A test schedule will continue to be consistent with 10
CFR 50 Appendix J. Therefore, the proposed change will not involve a
significant increase in the probability or consequences of any
accident previously evaluated.
The proposed change does not involve modifications to any
existing equipment. The proposed change will not affect the
operation of the plant or the manner in which the plant is operated.
Therefore, the proposed change will not create the possibility of a
new or different kind of accident from any accident previously
evaluated.
The margin of safety for the containment barrier is, in part,
preserved by compliance with 10 CFR 50 Appendix J. Although the
proposed change will allow greater flexibility in meeting Appendix J
requirements, the TS will continue to preserve compliance with 10
CFR Appendix J. Therefore, the proposed change will not involve a
significant reduction in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room Location: University of New Orleans
Library, Louisiana Collection, Lakefront, New Orleans, Louisiana 70122.
Attorney for licensee: N.S. Reynolds, Esq., Winston & Strawn 1400 L
Street N.W., Washington, D.C. 20005-3502.
NRC Project Director: William D. Beckner.
Houston Lighting & Power Company, City Public Service Board of San
Antonio, Central Power and Light Company, City of Austin, Texas, Docket
Nos. 50-498 and 50-499, South Texas Project, Units 1 and 2, Matagorda
County, Texas
Date of amendment request: May 1, 1995.
Description of amendment request: The proposed amendment would
provide a special test exception that would allow an extension of the
standby diesel generator (SDG) allowed outage time for a cumulative 21
days on each SDG once per fuel cycle, and it would also allow an
extension of the essential cooling water (ECW) loop allowed outage time
for a cumulative 7 days on each ECW loop once per fuel cycle. These
extended allowed outage times will be used to perform required
inspections and maintenance on the SDGs and the ECW system during power
operation.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. The proposed change does not involve a significant increase
in the probability or consequences of an accident previously
evaluated.
The Standby Diesel Generators are not accident initiators,
therefore the increase in Allowed Outage Times for this system does
not increase the probability of an accident previously evaluated.
The three train design of the South Texas Project ensures that even
during the seven days the Essential Cooling Water loop is inoperable
there are still two complete trains available to mitigate the
consequences of any accident. If the Essential Cooling Water loop is
not operable during the 21 days the Standby Diesel Generator is
inoperable, the Standby Diesel Generator's Engineered Safety
Features bus and equipment in the train will be operable. This
ensures that all three redundant safety trains of the South Texas
Project design are operable. In addition the Emergency Transformer
will be available to supply the Engineered Safety Features bus
normally supplied by the inoperable Standby Diesel Generator. These
actions will ensure that the changes do not involve a significant
increase in the consequences of previously evaluated accidents.
2. The proposed change does not create the possibility of a new
or different kind of accident from any accident previously
evaluated.
The proposed changes affect only the magnitude of the Standby
Diesel Generator and Essential Cooling Water Allowed Outage Times
once per fuel cycle as identified by the marked-up Technical
Specification. As indicated above, the proposed change does not
involve the alteration of any equipment nor does it allow modes of
operation beyond those currently allowed. Therefore, implementation
of these proposed changes does not create the possibility of a new
or different kind of accident from any accident previously
evaluated.
3. The proposed change does not involve a significant reduction
in a margin of safety.
The proposed changes result in no significant increase in core
damage or large early release frequencies.
Three sets of PSA results have been presented to the NRC for the
South Texas Project. One submitted in 1989 from the initial Level 1
PSA of internal and external events with a mean annual average CDF
estimate of 1.7 x 10(-4), a second one submitted in 1992 to meet the
IPE requirements from the Level 2 PSA/IPE with a CDF estimate of 4.4
x 10(-5), and an update of the PSA that was reported in the August
1993 Technical Specifications submittal with a variety of CDF
estimates for different assumptions regarding the rolling
maintenance profile and different combinations of modified Technical
Specifications. The South Texas Project PSA was updated in March of
1995 to include the NRC approved Risk-Based Technical
Specifications, Plant Specific Data and incorporate the Emergency
Transformer into the model. This update resulted in a CDF
[[Page 29877]] estimate of 2.07 x 10(-5). When the requested changes
are modeled along with the compensatory actions, the resulting CDF
estimate is 2.30 x 10(-5). While this is slightly higher (approx.
11%) than the updated results, it is still significantly lower
(approx. 46%) than the previous Risk-Based Evaluation of Technical
Specification submitted in 1993. Therefore, it is concluded that
there is no significant reduction in the margin of safety.
Based on the above evaluation, Houston Lighting & Power has
concluded that these changes do not involve any significant hazards
considerations.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
request for amendments involves no significant hazards consideration.
Local Public Document Room location: Wharton County Junior College,
J. M. Hodges, Learning Center, 911 Boling Highway, Wharton, TX 77488.
Attorney for licensee: Jack R. Newman, Esq., Newman & Holtzinger,
P.C., 1615 L Street, N.W., Washington, D.C. 20036.
NRC Project Director: William D. Beckner.
Houston Lighting & Power Company, City Public Service Board of San
Antonio, Central Power and Light Company, City of Austin, Texas, Docket
Nos. 50-498 and 50-499, South Texas Project, Units 1 and 2, Matagorda
County, Texas
Date of amendment request: May 2, 1995.
Description of amendment request: The proposed amendment would
revise Technical Specifications 3.4.2.2 and 3.7.1.1 (Table 3.7-2) by
relaxing the lift setting tolerances of the pressurizer safety valves
from plus or minus 1% to plus or minus 2% and the main steam safety
valves from plus or minus 1% to plus or minus 3%, respectively. In
addition, a footnote would be added to require that the pressurizer
safety valves and main steam safety valves setpoint tolerances be
restored to within plus or minus 1% whenever a lift setting is
determined to be outside plus or minus 1% following valve testing.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. The proposed change does not involve a significant increase
in the probability or consequences of an accident previously
evaluated because:
The proposed changes increase the ``as-found'' setpoint
tolerances for the Pressurizer Safety valves from plus or minus 1%
to plus or minus 2% and the Main Steam Safety valves from plus or
minus 1% to plus or minus 3%. The proposed changes do not involve
any hardware modifications to plant structures, systems, or
components. An evaluation has determined that the proposed changes
do not significantly affect the structural integrity of either the
reactor coolant system or the main steam system.
The proposed setpoint tolerance of plus or minus 2% for the
Pressurizer Safety valves and plus or minus 3% for the Main Steam
Safety valve ``as-found'' condition was previously evaluated as part
of the evaluation for the transition to VANTAGE 5H fuel. The
evaluation was reviewed and approved by the NRC Staff as part of
License Amendment Nos. 61 and 50 to Operating License NPF-76 and
NPF-80. Since the VANTAGE 5H fuel evaluation incorporated these
proposed changes, the calculated radiological release associated
with that evaluation is unaffected. Similarly, this applies to the
radiological dose associated with a steam generator tube rupture.
Additionally, the proposed change [sic] are consistent with the
guidance provided by Section III and XI of the ASME Code.
Based on the above, the proposed changes do not involve a
significant increase in the probability or consequences of an
accident previously evaluated.
2. The proposed change does not create the possibility of a new
or different kind of accident from any accident previously evaluated
because:
Since the lift setting of a Pressurizer Safety valve or Main
Steam Safety valve will be restored to plus or minus 1% whenever it
is determined to be outside plus or minus 1%, the ``as-left''
setpoint tolerances for the Pressurizer Safety valves and Main Steam
Safety valves are unchanged. The ``as-left'' setpoint will continue
to satisfy the current technical specification requirement on lift
setting tolerance. As such, there is no change in plant operation or
equipment performance. Since neither plant operation or equipment
performance is affected by the proposed changes, the proposed
changes do not create the possibility of a new or different kind of
accident from any accident previously analyzed.
3. The proposed change does not involve a significant reduction
in a margin of safety because:
Since the proposed changes are consistent with the guidance
provided by Section III and XI of the ASME Code, and the proposed
lift setting tolerance of plus or minus 2% for the Pressurizer
Safety valves and plus or minus 3% for the Main Steam Safety valves
has been incorporated into the design basis accident analyses, the
proposed changes do not involve a significant reduction in the
margin of safety.
Based on the safety evaluation presented above for the proposed
changes, Houston Lighting & Power has determined that the health and
safety of the public will not be jeopardized. Therefore, the
proposed changes 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 standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
request for amendments involves no significant hazards consideration.
Local Public Document Room location: Wharton County Junior College,
J. M. Hodges, Learning Center, 911 Boling Highway, Wharton, Texas
77488.
Attorney for licensee: Jack R. Newman, Esq., Newman & Holtzinger,
P.C., 1615 L Street, N.W., Washington, D.C. 20036.
NRC Project Director: William D. Beckner.
Indiana Michigan Power Company, Docket No. 50-315, Donald C. Cook
Nuclear Plant, Unit No. 1, Berrien County, Michigan
Date of amendment request: April 13, 1995.
Description of amendment request: The proposed amendment would
modify the Technical Specifications to allow use of laser-welded
sleeves to repair defective steam generator tubes.
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:
Conformance of the proposed amendments to the standards for a
determination of no significant hazard as defined in 10 CFR 50.92
(three factor test) is shown in the following:
(1) Operation of CNP [Cook Nuclear Plant] Unit 1 in accordance
with the proposed license amendment does not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
The TS [tubesheet] or TSP [tube support plate] intersection LWS
[laser-welded sleeve] configuration has been designed and analyzed
in accordance with the requirements of the ASME [American Society of
Mechanical Engineers] Code and RG [Regulatory Guide] 1.121. Fatigue
and stress analyses of the sleeved tube assemblies produced
acceptable results. Mechanical testing has shown that the structural
strength of the Alloy 690 sleeves under normal faulted and upset
conditions is within acceptable limits. Leak testing has
demonstrated that primary to secondary leakage is not expected
during all plant conditions, including the case where the seal weld
is not produced in the lower joint of the TS sleeve. Testing shows
that non-welding TS sleeve lower joints remained leaktight at
temperature and pressure conditions representative of normal and
accident conditions. Since laser welding produces a hermetic seal
between the tube and sleeve, no leak path can be realized under any
condition. Therefore, installation of LWSs will not influence
offsite dose [[Page 29878]] calculation for a postulated steam line
break event.
The proposed technical specification change to support the
installation of Alloy 690 LWSs does not adversely impact any
previously evaluated design basis accident or the results of
accident analyses for the current technical specification minimum
reactor coolant system flow rate. The results of the qualification
testing, analyses, and plant operating experience demonstrate that
the sleeve assembly is an acceptable means of maintaining tube
integrity. These aforementioned analyses and tests demonstrate that
installation of sleeves spanning degraded areas of the tube will
restore the tube to a condition consistent with its original design
basis. Plugging limit criteria are established using the guidance of
RG 1.121. Furthermore per RG 1.83 recommendations, the sleeved tube
can be monitored through periodic inspections with present eddy
current techniques.
Conformance of the sleeve design with the applicable sections of
the ASME Code and results of the leakage and mechanical tests,
support the conclusion that installation of laser-welded tube
sleeves will not increase the probability or consequences of an
accident previously evaluated. Depending upon the break location for
a postulated steam generator tube rupture event, implementation of
tube sleeving could act to reduce the radiological consequences to
the public due to reduced flow rate through a sleeved tube compared
to a non-sleeved tube based on the restriction afforded by the
sleeve wall thickness.
(2) The proposed license amendment does not create the
possibility of a new or different kind of accident from any accident
previously evaluated.
Implementation of laser-welded sleeving will not introduce
significant or adverse changes to the plant design basis. Stress and
fatigue analysis of the repair has shown the ASME Code and RG 1.121
allowable values are met. Implementation of laser-weld sleeving
maintains overall tube bundle structural and leakage integrity
during all plant conditions at a level consistent to that of the
originally supplied tubing. Leak and mechanical testing of sleeves
supports the conclusions of the calculations that the sleeve retains
both structural and leakage integrity during all conditions.
Sleeving of tubes does not provide a mechanism resulting in an
accident outside of the area affected by the sleeves. Any
hypothetical accident as a result of potential tube or sleeve
degradation in the repaired portion of the tube is bounded by the
existing tube rupture accident analysis. Since the sleeve design
does not affect any other component or location of the tube outside
of the immediate area repaired, in addition to the fact that the
installation of sleeves and the impact on current plugging level
analyses is accounted for, the possibility that laser-weld sleeving
creates a new or different type of accident is not supported.
The design of thermally treated Alloy 600 and 690 sleeved tube
assemblies have performed well historically with regard to
corrosion. There are no reported instances of Alloy 600 thermally
treated or Alloy 690 sleeve degradation for the greater than 35,000
sleeves that Westinghouse has installed in the U.S. Accelerated
corrosion test results show the free span laser-weld joint (LWJ)
(with post weld heat treatment) is capable of exhibiting a
resistance to corrosion of greater that 10 times that of rolled tube
transitions. Most LWS corrosion specimens did not experience
degradation and were subsequently removed from the corrosion test
media after a substantial testing period (supporting the 10x factor
compared to roll transitions) was achieved. Several mill annealed
Alloy 600 material heats were used for corrosion specimen
preparation. All were documented by previous test to have been
highly susceptible to PWSCC. The post weld heat treatment process
applied to LWS free span joints is designed to achieve a minimum
tube OD wall temperature of 1400 deg.F adjacent to the weld and
within the laser weld heat affected zone. Since the target
temperature of 1400 deg.F is achieved on the tube OD, a slightly
higher temperature is achieved at the tube ID surface, where the
weld cooling stresses are concentrated. Also, since the axial length
of the laser weld and laser weld heat affected zone are relatively
narrow compared to other sleeve welding processes, a narrower
section of tube is required to be heat treated. Since the length of
tube required to be heat treated is shorter in the LWS process than
with other sleeving processes, lower residual stresses in the tube
can be expected. Accelerated corrosion tests also show that non-heat
treated laser-weld free span joints exhibit resistance to stress
corrosion cracking equal to or greater than rolled tube transitions.
An extensive data base exists on LWS joint performance in foreign
plants in which the free span joints are not heat treated. Of the
approximately 18,000 non-heat treated joints in service, none has
exhibited a rapid corrosion potential. Corrosion testing of the TS
sleeve lower joint LWJs exhibit a resistance to corrosion cracking
of three to four times that of rolled tube transitions. These
factors suggest postulated sleeve/tube assembly degradation would
occur at a rate less than rolled transitions, and the potential for
a sleeve/tube assembly with accelerated degradation rate
characteristics more severe than rolled transitions, and the
potential for a sleeve/tube assembly with accelerated degradation
rate characteristics more severe than roll transitions is
negligible.
Approximately 800 LWSs are currently in operation in the U.S.
Some of these have been in service since April 1992. The plants in
which these sleeves are installed have not experienced any adverse
operational issues (such as primary to secondary leakage) as has
been detected at other plants with sleeves which have experienced
rapid corrosion of the parent tube.
(3) The proposed license amendment does not involve a
significant reduction in a margin of safety.
The laser-welded sleeving repair of degraded steam generator
tubes as identified in WCAP-13088 Rev. 3 has been demonstrated to
restore the integrity of the tube bundle under normal and postulated
accident conditions. The safety factors used in the design of
sleeves for the repair of degraded tubes are consistent with the
safety factors the ASME Boiler and Pressure Vessel Code used in
steam generator design. The plugging limit criteria for the sleeve
has been established using the methodology of RG 1.121. The design
of the sleeve joints have been verified by testing to preclude
leakage during normal and postulated accident conditions.
Implementation of laser-welded sleeving will reduce the potential
for primary to secondary leakage during a postulated steam line
break while maintaining available primary coolant flow area in the
event of a LOCA. By removing from service degraded intersections
through repair, the potential for tube leakage during a steam line
break is reduced. These degraded intersections now are returned to a
condition consistent with the design basis. While the installation
of a sleeve causes a reduction in flow, the reduction is far below
the reduction incurred by plugging. Therefore, far greater primary
coolant flow area is maintained through sleeving. Use of RG 1.121
criteria assures that the margin of safety with respect to
structural integrity is the same for the sleeves as for the original
steam generator tubes.
The portions of the installed sleeve assembly which represent
the reactor coolant pressure boundary can be monitored for the
initiation and progression of sleeve/tube wall degradation, thus
satisfying the requirements of RG 1.83. Portions of the tube bridged
by the sleeve joints are effectively isolated from the pressure
boundary, and the sleeve then forms the pressure boundary in these
areas. The areas of the sleeved tube assembly which require
inspection are defined in Attachment 4 [WCAP-13088, ``Westinghouse
Series 44 and 51 Steam Generator Generic Sleeving Report, Laser
Welded Sleeves,'' January 1994].
In addition, since the installed sleeve represents a portion of
the pressure boundary, a baseline inspection of these areas is
required prior to operation with sleeves installed. As stated
previously, weld fusion zone width is established using UT testing.
The minimum acceptable weld width as determined by UT examination is
approximately 50% wider than the minimum weld width which satisfies
the stress conditions of the ASME Code.
The generic evaluation uses the pressure stress equation of
Section NB 3224.1 of the ASME Code which is used to establish the
minimum required wall thickness for the sleeve design and
subsequently used to determine the level of sleeve wall degradation
(depth by eddy current determination) that would require the sleeve
to be removed from service. Using the [Delta]PNorm. Op. value
of 1530 psi from Attachment 4 [WCAP-13088, ``Westinghouse Series 44
and 51 Steam Generator Generic Sleeving Report, Laser Welded
Sleeves,'' January 1994] the limiting minimum required sleeve wall
thickness is established. The sleeve wall plugging limit (using
Attachment 4 [WCAP-13088, ``Westinghouse Series 44 and 51 Steam
Generator Generic Sleeving Report, Laser Welded Sleeves,'' January
1994]) of 25% is subsequently established, and includes an allowance
of 10% for eddy current uncertainty and 10% for growth, although
sleeve wall degradation has not been observed to date in
Westinghouse [[Page 29879]] sleeves. The generic evaluation used the
ASME Code minimum property values to establish the sleeve plugging
limit. Certified material test reports indicate that the sleeve
material properties are significantly higher than the ASME Code
minimum values. The generic evaluation considered a primary to
secondary pressure differential of 1530 psia, with a steam pressure
of 720 psia, for normal operating conditions. CNP Units 1 can
operate at full power with a reduced Thot value and RCS
pressure of 2250 psi. Steam pressure can be maintained as low as 650
psi (to keep Thot as low as possible), but cannot go lower than
650 psi or the steam generator operating requirement of a primary to
secondary [Delta]P of 1600 psi (max) will be exceeded. At this
[Delta]PNorm. Op. value of 1600 psi, the sleeve minimum wall
thickness requirement (and subsequently sleeve pressure boundary
plugging limit) using ASME Code minimum material properties can be
recalculated. For this condition (normal operating [Delta]P equal to
1600 psi), the sleeve minimum wall plugging limit is defined to be
23%. An allowance for eddy current uncertainty and continued
degradation are included in this value. The minimum required wall
thickness is determined by examining plant conditions at normal,
upset, faulted, and test conditions. For Model 51 steam generators,
the normal operating condition results in the limiting minimum wall
thickness requirement.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: Maud Preston Palenske Memorial
Library, 500 Market Street, St. Joseph, Michigan 49085.
Attorney for licensee: Gerald Charnoff, Esq., Shaw, Pittman, Potts
and Trowbridge, 2300 N Street, NW, Washington, DC 20037.
NRC Project Director: Cynthia A. Carpenter, Acting.
Indiana Michigan Power Company, Docket Nos. 50-315 and 50-316, Donald
C. Cook Nuclear Plant, Unit Nos. 1 and 2, Berrien County, Michigan
Date of amendment requests: December 20, 1993, as supplemented July
19, 1994, and February 28, 1995.
Description of amendment requests: The proposed amendments would
revise the Technical Specifications to change Train A and B emergency
loads from 8 hour to composite 4 hour, delete a load on the Train B
batteries load list, and revise the operational loads on the Train N
batteries. The supplemental submittals, made in response to NRC staff
concerns, would also add surveillance requirements for a battery with
signs of degradation and modify performance testing requirements.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which was published in the Federal Register on February
2, 1994 (59 FR 4939). This analysis was not changed by the supplemental
submittals.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment requests, including the supplemental submittals, involve no
significant hazards consideration.
Local Public Document Room location: Maud Preston Palenske Memorial
Library, 500 Market Street, St. Joseph, Michigan 49085.
Attorney for licensee: Gerald Charnoff, Esq., Shaw, Pittman, Potts
and Trowbridge, 2300 N Street, NW, Washington, DC 20037.
NRC Project Director: Cynthia A. Carpenter, Acting.
Indiana Michigan Power Company, Docket Nos. 50-315 and 50-316, Donald
C. Cook Nuclear Plant, Unit Nos. 1 and 2, Berrien County, Michigan
Date of amendment requests: March 31, 1995.
Description of amendment requests: The proposed amendments would
revise the technical specifications to provide increased flexibility in
the operation of the containment personnel airlocks during core
alterations.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
Per 10 CFR 50.92, a proposed change does not involve a
significant hazards consideration if the change does not:
1. involve a significant increase in the probability or
consequences of an accident previously evaluated,
2. create the possibility of a new or different kind of accident
from any accident previously evaluated, or
3. involve a significant reduction in a margin of safety.
Criterion 1
The design basis fuel handling accident is the rupture of the
highest rated fuel assembly. As discussed previously [in the
application], the consequences of an accident inside containment
(i.e., site boundary dose) with both airlock doors are bounded by
the existing fuel handling accident currently presented in our UFSAR
[Updated Final Safety Analysis Report].
Since the containment airlock doors do not affect the failure
mechanism of a fuel assembly during a fuel handling accident, we
believe that this amendment request does not involve a significant
increase in the probability or consequences of an accident
previously evaluated. Additionally, no credit was taken for
containment closure in the accident analysis. Therefore, based on
these considerations, it is concluded that the proposed change does
not involve a significant increase in the probability or
consequences of an accident previously evaluated.
Criterion 2
As stated in response to criterion one, the position of the
containment airlock doors in no way affects the mechanism by which a
spent fuel assembly is damaged during a fuel handling accident.
Thus, it is concluded that the proposed changes do not create the
possibility of a new or different kind of accident from any accident
previously evaluated.
Criterion 3
The margin for safety as defined in 10 CFR 100 has not been
reduced. As discussed previously, the existing fuel handling
accident analysis for an event inside containment takes no credit
for the isolation of containment. As a result, the position of the
airlock doors has no impact on the analyzed site boundary doses
resulting from such an accident. Based on these considerations, it
is concluded that the changes do not involve a significant reduction
in a margin of safety. The NRC staff has reviewed the licensee's
analysis and, based on this review, it appears that the three
standards of 10 CFR 50.92(c) are satisfied. Therefore, the NRC staff
proposes to determine that the amendment requests involve no
significant hazards consideration.
Local Public Document Room location: Maud Preston Palenske Memorial
Library, 500 Market Street, St. Joseph, Michigan 49085.
Attorney for licensee: Gerald Charnoff, Esq., Shaw, Pittman, Potts
and Trowbridge, 2300 N Street, NW, Washington, DC 20037.
NRC Project Director: Cynthia A. Carpenter, Acting.
Nebraska Public Power District, Docket No. 50-298, Cooper Nuclear
Station, Nemaha County, Nebraska
Date of amendment request: May 2, 1995.
Description of amendment request: This license amendment request
revises Surveillance Requirement (SR) 4.7.A.2.f.1 to allow a one-time
schedular extension of the two year Type B Local Leak Rate Test (LLRT)
interval required for the Drywell Head and Manport (penetrations DWH
and X-4 respectively). This extension will allow [[Page 29880]] the
Type B testing of penetrations DWH and X-4 to be deferred from the
current due date of July 17, 1995, until Refueling Outage No. 16 (RE-
16), which is currently scheduled to commence in October 1995.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
The enclosed Technical Specifications change is judged to
involve no significant hazards based on the following:
1. Does the proposed change involve a significant increase in
the probability or consequences of an accident previously evaluated?
This license amendment request revises Surveillance Requirement
(SR) 4.7.A.2.f.1 to allow the one-time schedular extension of the
two year Type B Local Leak Rate Test (LLRT) interval required for
the Drywell Head and Manport (Penetrations DWH and X-4
respectively). This extension will allow Penetrations DWH and X-4 to
be Type B tested during Refueling Outage No. 16 (RE-16), which is
currently scheduled to commence October 1995. Currently, the two
year maximum interval for these penetrations comes due July 17,
1995. The District has concluded that a one-time extension of
approximately six months beyond the two year limit will not result
in a significant increase in the probability of these penetrations
failing to perform their safety function. This conclusion is based
on the previous LLRT surveillance history of Penetrations DWH and X-
4, which have not failed an LLRT in the last 19 years. The
surveillance history demonstrates that these penetrations are not
subject to leak related failures.
Additionally, the seals associated with these penetrations will
not have experienced significantly more radiation and heat exposure
by the conclusion of the proposed extension than they would have
during the current two year interval. Although some radiation and
heat is present during plant shutdowns, the seal degradation
resulting from these conditions is significantly slower. Because
seal degradation is a function of heat and radiation, and is
generally not a function of time, the District has concluded that
the one-time extension will not result in a significant increase of
seal degradation. Because seal failure for these penetrations is
largely based on the rate of seal degradation, the probability of
the failure of these penetrations is not significantly increased.
Therefore, a significant increase in the probability or consequences
of an accident is not created.
This proposed change does not introduce any new modes of plant
operation, make any physical changes, or alter any operational
setpoints. The change does not degrade the performance of any safety
system assumed to function in the accident analysis. Therefore, this
proposed change does not involve a significant increase in the
probability or consequences of an accident previously evaluated.
2. Does the proposed change create the possibility for a new or
different kind of accident from any accident previously evaluated?
This license amendment request involves the one-time schedular
extension of the LLRT interval requirement for Penetrations DWH and
X-4. SR 4.7.A.2.f.1 is being revised to extend the surveillance test
interval for Penetrations DWH and X-4 to coincide with RE-16,
currently scheduled to commence October 1995. A one-time extension
of the subject surveillance interval does not involve the creation,
deletion, or modification of the function of any structure, system,
or component, nor does this change introduce or change any mode of
plant operation. This proposed change does not create the
possibility for a new or different kind of accident from any
accident previously evaluated.
3. Does the proposed change create a significant reduction in
the margin of safety?
This license amendment request involves the one-time extension
of the two year maximum surveillance test interval for Penetrations
DWH and X-4 from the current due date of July 17, 1995, to instead
coincide with RE-16, which is scheduled to commence October 1995. By
the time these tests are performed, the penetration seals will not
have experienced significantly more radiation and heat than they
would have during the previous test intervals. Therefore, the
penetration seals will not have experienced significant degradation
as a result of the extended interval. Furthermore, Penetrations DWH
and X-4 have not failed an LLRT in the last 19 years. The
surveillance history demonstrates that these penetrations are not
subject to leak related failure. This proposed change does not
involve any change to plant design, equipment instrument setpoints,
or operation. Therefore, this proposed change does not create a
significant reduction in the margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: Auburn Public Library, 118
15th Street, Auburn, NE 68305.
Attorney for licensee: Mr. G.D. Watson, Nebraska Public Power
District, Post Office Box 499, Columbus, NE 68602-0499.
NRC Project Director: William D. Beckner.
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 request will revise Technical
Specification Section 3.10.5 to allow more than one control bank to be
fully withdrawn from the core simultaneously for rod drop time response
testing. Specifically, the change will delete, (1) the limiting
condition for operation (LCO) 3.10.5.a and (2) a reference to the full
length shutdown rods from LCO 3.10.5. The change will also add a
statement that ``The SHUTDOWN MARGIN requirement of Section 3.1.1.1.2
shall be met without credit for withdrawn control rods.'' Other
editorial changes are to be made for consistency.
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 (SHC), 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 evaluated.
The proposed changes cannot initiate an event since the plant
will be maintained shutdown at all times. Thus, there is no increase
in the probability of occurrence of an accident previously
evaluated.
The proposed changes do not degrade the performance of any
safety system nor do they alter any assumptions made in the accident
analyses. Currently, the technical specifications allow the rod
position indication system to be disabled for each control bank
while performing this test. In addition, this system is not a safety
system credited in the accident analyses. Therefore, allowing more
than one bank to have its indication removed during the test does
not degrade any safety system. Since the shutdown margin will be
maintained without crediting these rods, there is no change to the
assumptions made in the accident analyses. Thus, there is no
increase in the consequences of an accident previously evaluated.
2. Create the possibility of a new or different kind of accident
from any previously analyzed.
The proposed changes do not position the control rods into any
new configurations or sequence not previously analyzed. Ejected rod
worths are evaluated for ARI-1 (all rods in with the most reactive
rod out) and, therefore, bound the test configuration. In addition,
the reactivity state of the system is maintained shut down by the
margin required in Technical Specification 3.1.1.1.2 without
crediting the control rods. Therefore, there is no possibility of a
new or different type of accident than previously analyzed.
3. Involve a significant reduction in the margin of safety.
The proposed changes do not impact any of the physical
protective boundaries, safety systems, or operating conditions. The
plant [[Page 29881]] will be maintained shut down without crediting
the control rods. The accident analyses is not impacted and,
therefore, there is 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: Learning Resources Center,
Three Rivers Community-Technical College, 574 New London Turnpike,
Norwich, CT 06360.
Attorney for licensee: Ms. L.M. Cuoco, Senior Nuclear Counsel,
Northeast Utilities Service Company, Post Office Box 270, Hartford, CT
06141-0270.
NRC Project Director: Phillip F. McKee.
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: April 28, 1995.
Description of amendment request: The request will revise the
diesel generator (DG) fuel oil testing that is performed on new fuel
prior to the addition of the new fuel to the storage tank.
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 (SHC), 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 evaluated.
The proposed changes: correct a typographical error by providing
the appropriate range for the Saybolt viscosity; replace the
qualitative clear and bright test with a quantitative water and
sediment test for new fuel prior to adding it to the storage tank;
and clarify that a calculated cetane index may be performed in lieu
of obtaining the cetane number for the fuel. The water and sediment
test provides a quantitative method for evaluating water and
sediment, and will require a more restrictive limit of 0.05 percent
by volume of water and sediment than the 0.10 percent recommended by
the manufacturer. The cetane index has been shown to be
representative of the cetane number for the fuel. The DG capability
to start and operate is enhanced by the proposed changes. Therefore,
the changes have no negative effect on the consequences of the
previously evaluated accidents.
2. Create the possibility of a new or different kind of accident
from any accident previously evaluated.
The proposed changes do not alter or affect the design,
function, failure mode, or operation of the plant. The proposed
changes have no adverse effect on the quality of the fuel oil that
is utilized by the DG. The proposed changes are administrative in
nature and do not involve any physical alteration to any plant
system or change the method by which any safety-related system
performs its function. For these reasons, there is no possibility of
an accident of a different type than previously evaluated.
3. Involve a significant reduction in a margin of safety.
The proposed changes will assure that the DG fuel oil meets DG
manufacturer's quality requirements by the performance of the
recommended testing of the DG fuel oil. The proposed changes will
not impact the margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: Learning Resources Center,
Three Rivers Community-Technical College, 574 New London Turnpike,
Norwich, CT 06360.
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: April 28, 1995.
Description of amendment request: The proposed revision to the
Action Statement of Limiting Condition for Operation (LCO) 3.7.5 would
permit Millstone Unit No. 3 to remain in Modes 1 through 4 with the
average water temperature of the ultimate heat sink (UHS) greater than
75 deg.F (but lower than 77 deg.F) for 12 hours. An additional action
would be added which would require the plant to be placed in at least
HOT STANDBY within 6 hours and in COLD SHUTDOWN within the following 30
hours upon identifying that the UHS temperature is greater than
77 deg.F.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration (SHC), 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 evaluated.
The proposed addition of a 12-hour period to monitor the UHS
temperature to the Technical Specification LCO Action Statement does
not involve an increase in the probability of an accident previously
evaluated. The probability of an accident previously evaluated is
not increased by a short-term increase in the UHS temperature. The
probability of FSAR Chapter 15 Condition IV accidents occurring in
conjunction with the short duration increase in service water inlet
temperature above 75 deg.F is low enough such that they are not risk
significant. Further, an evaluation has been performed that safe
shutdown will be achieved and maintained for a loss of offsite power
event and a steam generator tube rupture event with the additional
consideration of a single failure with service water inlet
temperatures as high as 77 deg.F. There has been no significant
increase in the consequences of these events previously evaluated.
2. Create the possibility of a new or different kind of accident
from any accident previously evaluated.
The proposed technical specification change does not create the
possibility of a new or different kind of accident previously
evaluated. The addition of a 12-hour time period to monitor the UHS
temperature increases the amount of time that is allowed for the
plant to be in HOT STANDBY from 6 to 18 hours should the UHS
temperature increase above 75 deg.F. This extension of the time
allowed for the plant to be in HOT STANDBY does not change the plant
configuration. As such, the change does not create the possibility
of a new or different kind of accident previously evaluated.
3. Involve a significant reduction in a margin of safety.
The proposed technical specification change does not involve a
significant reduction in the margin of safety. The addition of a 12-
hour time period to monitor the UHS temperature increases the time
required for the plant to be in HOT STANDBY from 6 to 18 hours
should the UHS temperature exceed 75 deg.F. An evaluation has been
performed to demonstrate that the risk significance associated with
the increased action time is very low. In addition, safe shutdown
capability has been demonstrated for service water inlet
temperatures as high as 77 deg.F.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: Learning Resources Center,
[[Page 29882]] Three Rivers Community-Technical College, 574 New London
Turnpike, Norwich, CT 06360.
Attorney for licensee: Ms. L.M. Cuoco, Senior Nuclear Counsel,
Northeast Utilities Service Company, Post Office Box 270, Hartford, CT
06141-0270.
NRC Project Director: Phillip F. McKee.
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: May 1, 1995.
Description of amendment request: Technical Specifications that
specify an 18-month surveillance will be changed to state that these
surveillances are to be performed at least once each refueling (i.e.,
24 months).
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration (SHC), which is presented below:
* * * 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 proposed change to Surveillance Requirement 4.6.3.2 of the
Millstone Unit No. 3 Technical Specifications extends the frequency
for verifying that each containment isolation valve actuates to its
required position in response to Phase A and Phase B isolation test
signals, and for verifying that each containment purge supply and
exhaust isolation valve actuates to its required position in
response to a containment high radiation test signal. The proposal
would extend the frequency from at least once per 18 months to at
least once per refueling interval (24 months).
The proposed change to Surveillance Requirement 4.6.3.2 does not
alter the intent or method by which the surveillances are conducted,
does not involve any physical changes to the plant, does not alter
the way any structure, system, or component functions, and does not
modify the manner in which the plant is operated. As such, the
proposed change to the frequency of Surveillance Requirement 4.6.3.2
will not degrade the ability of the containment isolation valves to
perform their safety function. Also, the containment isolation valve
arrangements are not vulnerable to single failures, because they
provide at least two barriers between the atmosphere outside the
containment and the atmosphere within the containment, the reactor
coolant system, or systems that would become connected to the
containment atmosphere or the reactor coolant system as a result of,
or subsequent to, a DBA.
Additional assurance of containment isolation valve operability
is provided by Surveillance Requirements 4.6.3.1 and 4.6.3.3.
Surveillance Requirement 4.6.3.1 requires that a containment
isolation valve will be restored to an operable status following the
performance of work on the containment isolation valve or its
ancillaries. Surveillance Requirement 4.6.3.3 requires the
confirmation of the mechanical operability of the containment
isolation valves by the inservice inspection program. The proposed
change does not modify these requirements.
Additionally, Surveillance Requirements 4.3.2.1 and 4.3.3.1
assure the operability of the automatic isolation logic (Phase A and
Phase B isolation signals and containment high radiation signal) for
the containment isolation valves by performing tests on a monthly
basis. This proposed change does not modify these Surveillance
Requirements.
Equipment performance over the last four operating cycles was
evaluated to determine the impact of extending the frequency of
Surveillance Requirement 4.6.3.2. This evaluation included a review
of surveillance results, preventive maintenance records, and the
frequency and type of corrective maintenance. It has been concluded
that the containment isolation valves are highly reliable, and that
there is no indication that the proposed extension could cause
deterioration in valve condition or performance.
Based on the above, the proposed change to Surveillance
Requirement 4.6.3.2 of the Millstone Unit No. 3 Technical
Specifications does not involve a significant increase in 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 change to Surveillance Requirement 4.6.3.2 of the
Millstone Unit No. 3 Technical Specifications extends the frequency
for verifying that each containment isolation valve actuates to its
required position in response to Phase A and Phase B isolation test
signals, and for verifying that each containment purge supply and
exhaust isolation valve actuates to its required position in
response to a containment high radiation test signal. The proposal
would extend the frequency from at least once per 18 months to at
least once per refueling interval (24 months).
The proposed change does not alter the intent or method by which
the surveillances are conducted, does not involve any physical
changes to the plant, does not alter the way any structure, system,
or component functions, and does not modify the manner in which the
plant is operated. As such, the proposed change in the frequency of
Surveillance Requirement 4.6.3.2 will not degrade the ability of the
containment isolation valves to perform their safety function. Also,
the containment isolation valve arrangements are not vulnerable to
single failures, because they provide at least two barriers between
the atmosphere outside the containment and the atmosphere within the
containment, the reactor coolant system, or systems that would
become connected to the containment atmosphere or the reactor
coolant system as a result of, or subsequent to, a DBA.
Based on the above, the proposed change to Surveillance
Requirement 4.6.3.2 of the Millstone Unit No. 3 Technical
Specifications will not create the possibility of a new or different
kind of accident from any previously evaluated.
3. Involve a significant reduction in the margin of safety.
The proposed change to Surveillance Requirement 4.6.3.2 of the
Millstone Unit No. 3 Technical Specifications extends the frequency
for verifying that each containment isolation valve actuates to its
required position in response to Phase A and Phase B isolation test
signals, and for verifying that each containment purge supply and
exhaust isolation valve actuates to its required position in
response to a containment high radiation test signal. The proposal
would extend the frequency from at least per 18 months to at least
once per refueling interval (24 months).
The proposed change does not alter the intent or method by which
the surveillances are conducted, does not involve any physical
changes to the plant, does not alter the way any structure, system,
or component functions, and does not modify the manner in which the
plant is operated. As such, the proposed change in the frequency of
Surveillance Requirement 4.6.3.2 will not degrade the ability of the
containment isolation valves to perform their safety function. Also,
the containment isolation valve arrangements are not vulnerable to
single failures, because they provide at least two barriers between
the atmosphere outside the containment and the atmosphere within the
containment, the reactor coolant system, or systems that would
become connected to the containment atmosphere or the reactor
coolant system as a result of, or subsequent to, a DBA.
Additional assurance of the operability of the containment
isolation valves is provided by Surveillance Requirements 4.6.3.1
and 4.6.3.2. Also, assurance of the operability of the automatic
actuation logic of the containment isolation valves is provided by
Surveillance Requirements 4.3.2.1 and 4.3.3.1.
Equipment performance over the last four operating cycles was
evaluated to determine the impact of extending the frequency of
Surveillance Requirement 4.6.3.2. This evaluation included a review
of surveillance results, preventive maintenance records, and the
frequency and type of corrective maintenance. It has been concluded
that the containment isolation valves are highly reliable, and that
there is no indication that the proposed extension could cause
deterioration in valve condition or performance.
Based on the above, the proposed change to Surveillance
Requirement 4.6.3.2 of the Millstone Unit No. 3 Technical
Specifications 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
[[Page 29883]] amendment request involves no significant hazards
consideration.
Local Public Document Room location: Learning Resources Center,
Three Rivers Community-Technical College, 574 New London Turnpike,
Norwich, CT 06360.
Attorney for licensee: Ms. L. M. Cuoco, Senior Nuclear Counsel,
Northeast Utilities Service Company, Post Office Box 270, Hartford, CT
06141-0270.
NRC Project Director: Phillip F. McKee
Omaha Public Power District, Docket No. 50-285, Fort Calhoun Station,
Unit No. 1, Washington County, Nebraska
Date of amendment request: May 8, 1995.
Description of amendment request: The proposed amendment would
change Technical Specifications 2.3, 3.1, 3.2, 3.3 and 3.6. These
changes are in accordance with the guidance of Generic Letter 93-05,
``Line Item Technical Specifications Improvements to Reduce
Surveillance Requirements for Testing During Power Operation,'' dated
September 27, 1993.
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.
GL 93-05, Item 4.2, Control Rod Movement Test; Specification 3.2,
Table 3-5, Item 2
Omaha Public Power District (OPPD) proposes to extend the
control element assembly (CEA) partial movement surveillance test of
Specification 3.2, Table 3-5, Item 2 from a biweekly to a quarterly
frequency. This change is based on operating experience and the
recommendation of Generic Letter (GL) 93-05, Item 4.2.1. A review of
previous surveillance tests and interviews with personnel familiar
with the test did not identify any prior surveillance test failures.
Industry experience has shown that this test can cause reactor
trips, dropped rods and unnecessary challenges to safety systems as
stated in NUREG-1366, ``Improvements to Technical Specification
Requirements,'' dated December 1992. Therefore, extending the
frequency of conducting this surveillance test may be beneficial to
plant operations and does not involve a significant increase in the
probability or consequences of an accident previously evaluated.
GL 93-05, Item 5.14, Radiation Monitors; Specification 3.1, Table
3-3, Items 3b, 4 and 5b
OPPD proposes to replace descriptive wording in Specification
3.1, Table 3-3, Items 3a/b and 5a/b with defined terms. OPPD also
proposes to extend surveillance of the area, post-accident and
primary to secondary leak-rate radiation monitors (Specification
3.1, Table 3-3, Items 3b and 5b) from a monthly to a quarterly
frequency as recommended by GL 93-05, Item 5.14. Most of these
monitors are new (i.e., installed within the last two cycles) or
contain many new components. The value of monthly testing is greatly
reduced as the new monitors include self checking circuitry that
will indicate monitor failure, loss of power, or loss of background.
Although post accident radiation monitors RM-091 A/B are not new,
Station operating experience has shown that they are reliable. In
cases where new components interface with older components, the
older components have a history of reliable operation.
Readings and internal test signals are used to verify instrument
operation on a daily basis. In addition, the proposed frequency
(quarterly) is the same frequency currently specified for the
containment radiation high signal (CRHS) monitors (Specification
3.1, Table 3-2, Item 6b), which generate an engineered safeguards
signal. Replacing descriptive words with defined terms ensures
consistency and that the surveillance test accomplishes its purpose.
A quarterly surveillance conserves resources, increases the
availability of the area, post-accident and primary to secondary
leak-rate detection radiation monitors and is consistent with CRHS
monitor testing. These proposed changes do not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
OPPD proposes to delete Specification 3.1, Table 3-3, Item 4 on
surveillance testing of the emergency plan radiation instruments.
These are portable instruments stored in specified locations for use
by emergency response personnel in the event of an accident. The
instruments may be used to survey onsite/offsite areas for
radioactivity or to facilitate the decontamination of personnel
following an accident. No limiting condition for operation (LCO)
action statement is associated with these instruments. As a result,
there is no basis for the TS to contain a surveillance requirement
for them. In addition, retaining this surveillance in the TS is
unnecessary since it does not meet criteria 1 through 4 of the Final
Policy Statement on Technical Specifications Improvements for
Nuclear Power Reactors, dated July 22, 1993. Therefore, since these
instruments are not utilized until after an accident has occurred,
and do not assist in accident mitigation, deleting this surveillance
requirement does not involve a significant increase in the
probability or consequences of an accident previously evaluated.
GL 93-05, Item 6.1, Reactor Coolant System Isolation Valves;
Specification 3.3(2)a
The reactor coolant system (RCS) pressure isolation valves have
proven to be very reliable. Therefore, OPPD proposes to extend the
time that the plant can be in cold shutdown before the test is
required (Specification 3.3(2)a) from 72 hours to 7 days, following
the recommendation of GL 93-05, Item 6.1. A review of previous
surveillance tests and interviews with personnel familiar with the
test did not identify any prior surveillance test failures. This
proposed change will reduce radiation exposure and does not involve
a significant increase in the probability or consequences of an
accident previously evaluated.
GL 93-05, Item 7.4, Accumulator Water Level and Pressure Channel
Surveillance Requirements; Specification 2.3(2)g, Specification
3.1, Table 3-2, Item 14a
OPPD proposes to revise Specification 2.3(2)g following the
recommendation of GL 93-05, Item 7.4. This revision will clarify
that the safety injection tank (SIT) level and/or pressure
instrumentation may be inoperable, which does not alter the intent
of the Specification, but is more accurate in defining when the
Specification applies. This revision also extends the time limit for
inoperability of SIT instrumentation from 1 hour to 72 hours, which
is justified based upon a review of historical data. As stated in
NUREG-1366: ``While technically inoperable, the accumulator [SIT]
would be available to fulfill its safety function during this time,
and thus, this change would have a negligible increase on risk.''
Currently, Specification 2.3(2)g allows only one hour for SIT
level and pressure instrumentation to be inoperable, which is
insufficient time to initiate repairs. A review of historical data
determined that SIT water level stays relatively constant while
pressure decreases slightly over time. It is unlikely that SIT
pressure would decrease below the Specification 2.3(1)c limit of 240
psig during the proposed 72-hour LCO, since SIT pressure is normally
maintained around 255 psig (Updated Safety Analysis Report (USAR),
Section 6.2.3.5).
OPPD's proposal to revise Specification 3.1, Table 3-2, Item 14a
to require shiftly verification that SIT level and pressure are
within limits and remove reference to verifying ``indications are
between independent high and low alarms for level and pressure,'' is
consistent with the guidance of GL 93-05, Item 7.4. As stated in GL
93-05, Item 7.4, the operability of SIT instrumentation is not
directly related to the capability of a SIT to perform its safety
function. OPPD proposes to suspend this surveillance on the affected
SIT while the instrumentation is being repaired, since as stated
above, SIT level and pressure are expected to stay within the limits
of Specification 2.3(1)c during the proposed 72 hour LCO. Therefore,
these proposed changes do not involve a significant increase in the
probability or consequences of an accident previously evaluated.
GL 93-05, Item 8.1, Containment Spray System; Specification 3.6(2)b
OPPD proposes to extend the surveillance frequency for verifying
that the containment spray nozzles are open (Specification 3.6(2)b)
from five to ten years following the recommendation of GL 93-05,
Item 8.1. Minor revisions to statements in the basis of
Specification 3.6 that refer to conducting this test at five year
intervals are proposed also. OPPD has not experienced problems with
[[Page 29884]] obstructions in the containment spray nozzles as
determined by a review of previous surveillance tests and personnel
interviews. Of the three instances reported in NUREG-1366 concerning
obstructions of containment spray nozzles, all were problems related
to construction errors. Any construction errors in the FCS
containment spray system would have been found by previous
surveillance tests.
The problem that occurred at San Onofre Unit 1 (clogging of
several containment spray nozzles following the application of a
coating material to the carbon steel piping) is not a concern at FCS
since the FCS containment spray system piping and valves are
constructed of stainless steel (USAR Table 6.3-2). Thus, extending
the surveillance frequency of Specification 3.6(2)b from five to ten
years 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.
GL 93-05, Item 4.2, Control Rod Movement Test; Specification 3.2,
Table 3-5, Item 2
OPPD's proposal to extend the CEA partial movement surveillance
test (Specification 3.2, Table 3-5, Item 2) to a quarterly frequency
is based on operating experience and the recommendation of GL 93-05,
Item 4.2.1. The proposed change only lengthens the time between
surveillance tests and will not result in any physical alterations
to the plant configuration, changes to setpoint values, or changes
to the application of setpoints or limits. Therefore, the proposed
change does not create the possibility of a new or different kind of
accident from any accident previously evaluated.
GL 93-05, Item 5.14, Radiation Monitors; Specification 3.1, Table
3-3, Items 3b, 4 and 5b
OPPD proposes to replace unnecessary wording in Specification
3.1, Table 3-3, Items 3a/b and 5a/b with defined terms and to extend
the surveillance frequency of Items 3b and 5b from monthly to
quarterly based on the recommendation of GL 93-05, Item 5.14. Most
of the area, post accident and primary to secondary leak-rate
detection radiation monitors are new or contain new components. The
new monitors include self checking circuitry that provides failure
notification. Although post accident radiation monitors RM-091 A/B
are not new, they have an excellent operating history. The proposed
changes introduce consistent use of terminology and lengthen the
time between surveillance tests and will not result in any physical
alterations to the plant configuration, changes to setpoint values,
or changes to the application of setpoints or limits. Therefore,
these proposed changes do not create the possibility of a new or
different kind of accident from any accident previously evaluated.
OPPD's proposal to delete Specification 3.1, Table 3-3, Item 4
on surveillance testing of the emergency plan radiation instruments
will not result in any physical alterations to the plant
configuration, changes to setpoint values, or changes to the
application of setpoints or limits. Since these instruments are not
utilized until after an accident has occurred, and do not assist in
accident mitigation, this change does not create the possibility of
a new or different kind of accident from any accident previously
evaluated.
GL 93-05, Item 6.1, Reactor Coolant System Isolation Valves;
Specification 3.3(2)a
The RCS pressure isolation valves have proven to be very
reliable. As a result, OPPD proposes to extend the time that the
plant can be in cold shutdown before the test is required
(Specification 3.3(2)a) from 72 hours to 7 days following the
recommendation of GL 93-05, Item 6.1. The proposed change will
reduce radiation exposure and does not result in any physical
alterations to the plant configuration, changes to setpoint values,
or changes to the application of setpoints or limits. Therefore,
this change does not create the possibility of a new or different
kind of accident from any accident previously evaluated.
GL 93-05, Item 7.4, Accumulator Water Level and Pressure Channel
Surveillance Requirements; Specification 2.3(2)g, Specification
3.1, Table 3-2, Item 14a
OPPD's proposal to revise Specification 2.3(2)g following the
guidance of GL 93-05, Item 7.4 more accurately states when the
specification should apply and extends the time limit for
inoperability of SIT instrumentation from 1 hour to 72 hours based
upon a review of historical data. The proposed change will not
result in any physical alterations to the plant configuration,
changes to setpoint values, or changes to the application of
setpoints or limits. As stated in NUREG-1366: ``While technically
inoperable, the accumulator [SIT] would be available to fulfill its
safety function during this time, and thus, this change would have a
negligible increase on risk.''
OPPD's proposal to revise Specification 3.1, Table 3-2, Item 14a
to require shiftly verification that SIT level and pressure are
within limits and remove reference to verifying ``indications are
between independent high and low alarms for level and pressure,'' is
consistent with the guidance of GL 93-05, Item 7.4. As stated in GL
93-05, Item 7.4, the operability of SIT instrumentation is not
directly related to the capability of a SIT to perform its safety
function. OPPD proposes to suspend this surveillance on the affected
SIT while the instrumentation is being repaired, since SIT level and
pressure are expected to stay within the limits of Specification
2.3(1)c during the proposed 72 hour LCO. Therefore, since these
proposed changes do not result in any physical alterations to the
plant configuration, changes to setpoint values, or changes to the
application of setpoints or limits, they do not create the
possibility of a new or different kind of accident from any accident
previously evaluated.
GL 93-05, Item 8.1, Containment Spray System; Specification 3.6(2)b
OPPD's proposal to extend the surveillance frequency for
verifying that the containment spray nozzles are open (Specification
3.6(2)b) from five to ten years as recommended by GL 93-05, Item 8.1
is justified by operating experience. OPPD has not experienced
problems with obstructions in the containment spray nozzles as
determined by a review of previous surveillance tests and personnel
interviews. The problem that occurred at San Onofre Unit 1 (clogging
of several containment spray nozzles following the application of a
coating material to the carbon steel piping) is not a concern at FCS
since the FCS containment spray system piping and valves are
constructed of stainless steel (USAR Table 6.3-2).
The proposed change only extends the time between surveillance
tests and revises associated basis statements to support the
extension. The proposed change will not result in any physical
alterations to the plant configuration, changes to setpoint values,
or changes to the application of setpoints or limits. Therefore,
OPPD's proposal to extend the surveillance frequency of
Specification 3.6(2)b from five to ten years does not create the
possibility of a new or different kind of accident from any accident
previously evaluated.
(3) The proposed change does not involve a significant reduction
in a margin of safety.
GL 93-05, Item 4.2, Control Rod Movement Test; Specification 3.2,
Table 3-5, Item 2
OPPD's proposal to extend the CEA partial movement surveillance
test of Specification 3.2, Table 3-5, Item 2 to a quarterly
frequency is based on operating experience and the recommendation of
GL 93-05, Item 4.2.1. A review of previous surveillance tests and
interviews with personnel familiar with the test did not identify
any prior surveillance test failures. Industry experience has shown
that this test can occasionally cause reactor trips, dropped rods
and unnecessary challenges to safety systems as stated in NUREG-
1366. Therefore, extending the frequency of conducting this
surveillance test may be beneficial to plant operations and does not
involve a significant reduction in a margin of safety.
GL 93-05, Item 5.14, Radiation Monitors; Specification 3.1, Table
3-3, Items 3b, 4 and 5b
OPPD proposes to replace descriptive wording in Specification
3.1, Table 3-3, Items 3a/b and 5a/b with defined terms and to extend
the surveillance frequency of Items 3b and 5b from monthly to
quarterly based on the recommendation of GL 93-05, Item 5.14. Most
of the area, post accident and primary to secondary leak-rate
detection radiation monitors are new or contain new components. Post
accident radiation monitors RM-091 A/B are not new but have a
history of reliable operation. The value of monthly testing is
greatly reduced since the new monitors include self checking
circuitry that provides failure notification. The proposed changes
introduce consistent use of terminology and lengthen the time
between surveillance tests and therefore do not involve a
significant reduction in a margin of safety.
OPPD's proposal to delete Specification 3.1, Table 3-3, Item 4
is justified because the [[Page 29885]] emergency plan radiation
instruments are portable instruments that are not utilized until
after an accident has occurred. The instruments are checked for
proper operation before use and since these instruments do not
assist in accident mitigation, the deletion of this surveillance
requirement does not involve a significant reduction in a margin of
safety.
GL 93-05, Item 6.1, Reactor Coolant System Isolation Valves;
Specification 3.3(2)a
The RCS pressure isolation valves have proven to be very
reliable. Therefore, consistent with the guidance of GL 93-05, Item
6.1, OPPD proposes to revise Specification 3.3(2)a and extend the
time that the plant is allowed to be in cold shutdown before this
surveillance test is required from 72 hours to 7 days. This change
will reduce radiation exposure and does not involve a significant
reduction in a margin of safety.
GL 93-05, Item 7.4, Accumulator Water Level and Pressure Channel
Surveillance Requirements; Specification 2.3(2)g, Specification
3.1, Table 3-2, Item 14a
OPPD's proposal to revise Specification 2.3(2)g following the
guidance of GL 93-05, Item 7.4 more accurately states when the
specification applies and extends the time limit for inoperability
of SIT instrumentation from 1 to 72 hours based upon historical
data. As stated in NUREG-1366: ``While technically inoperable, the
accumulator [SIT] would be available to fulfill its safety function
during this time, and thus, this change would have a negligible
increase on risk.''
OPPD's proposal to revise Specification 3.1, Table 3-2, Item 14a
to require shiftly verification that SIT level and pressure are
within limits and remove reference to verifying ``indications are
between independent high and low alarms for level and pressure,'' is
consistent with the guidance of GL 93-05, Item 7.4. As stated in GL
93-05, Item 7.4, the operability of SIT instrumentation is not
directly related to the capability of a SIT to perform its safety
function. OPPD proposes to suspend this surveillance on the affected
SIT while the instrumentation is being repaired, since SIT level and
pressure are expected to stay within the limits of Specification
2.3(1)c during the proposed 72 hour LCO. Therefore, these proposed
changes do not involve a significant reduction in a margin of
safety.
GL 93-05, Item 8.1, Containment Spray System; Specification 3.6(2)b
OPPD's proposal to extend the surveillance frequency for
verifying that the containment spray nozzles are open (Specification
3.6(2)b) from five to ten years as recommended by GL 93-05, Item 8.1
is justified by operating experience. OPPD has not experienced
problems with obstructions in the containment spray nozzles as
determined by a review of previous surveillance tests and personnel
interviews.
The problem that occurred at San Onofre Unit 1 is not a concern
at FCS since the FCS containment spray system piping and valves are
constructed of stainless steel (USAR Table 6.3-2). Therefore, OPPD's
proposal to extend the surveillance frequency of Specification
3.6(2)b from five to ten years and revise associated basis
statements 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: W. Dale Clark Library, 215
South 15th Street, Omaha, Nebraska 68102.
Attorney for licensee: James R. Curtiss, Winston & Strawn, 1400 L
Street, Washington, DC 20005-3502.
NRC Project Director: William H. Bateman.
Pacific Gas and Electric Company, Docket No. 50-133, Humboldt Bay Power
Plant, Unit 3, Humboldt County, California
Date of amendment request: April 10, 1995.
Description of amendment request: The proposed amendment would
revise License No. DPR-7, to permit the provisions of 10 CFR 50.59 to
be applied with respect to changes to the facility or procedures
described in the Decommissioning Plan or changes to the Decommissioning
Plan, and the conduct of tests or experiments not described in the
Decommissioning Plan.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. Does the change involve a significant increase in the
probability or consequences of an accident previously evaluated?
The probability or consequences of an accident previously
evaluated will not be effected by the ability to perform safety
analyses. As outlined in 10 CFR 50.59, the impact of performing
special tests, experiments, and modifications would be evaluated to
verify there would be no impact on previously evaluated accidents or
increase the probability or consequences of an accident occurring.
Therefore, the proposed change does not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
2. Does the change create the possibility of a new or different
kind of accident from any accident previously evaluated?
The proposed change does not create the possibility of a new or
different kind of accident from any accident previously evaluated
because there is no physical alteration to any plant system, nor is
there a change in the method in which any quality-related activities
are performed or any direct change in equipment or system function
or operation. The proposed change is administrative in nature.
Therefore, the proposed change does not create the possibility
of a new or different kind of accident from any accident previously
evaluated.
3. Does the change involve a significant reduction in a margin
of safety?
The proposed change to the HBPP License does not affect the
margin of safety of any accident analysis since it does not affect
the parameters for any accident analysis, and has no effect on the
current operating methodologies or actions that govern plant
performance.
Therefore, the proposed change does not involve a significant
reduction in a margin of safety.
The NRC staff has reviewed the analysis of the licensee 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: Humboldt County Library, 636 F
Street, Eureka, California 95501.
Attorney for licensee: Christopher J. Warner, Esquire, Pacific Gas
& Electric Company, P.O. Box 7442, San Francisco, California 94120.
NRC Project Director: Seymour H. Weiss.
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: January 17, 1995 as
supplemented by letter dated March 30, 1995.
Description of amendment request: The proposed change revises the
Peach Bottom Atomic Power Station, Units 2 and 3 technical
specifications to reflect the replacement of the source range monitor
(SRM) and intermediate range monitor (IRM) systems with a new system
referred to as the wide range neutron monitoring system (WRNMS).
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. The proposed changes do not involve a significant increase in
the probability or consequences of an accident previously evaluated.
The use of the WRNMS as discussed herein will not increase the
probability or consequences of an accident previously evaluated.
[[Page 29886]]
The probability (frequency of occurrence) of design basis
accidents (DBAs) occurring is not affected by the WRNMS. The only
plant safety analysis affected by WRNMS is the Rod Withdrawal Error
(RWE) at low power, and a reanalysis assuming use of WRNMS shows
that the criteria of 170 cal/gm for fuel enthalpy increase under RWE
is satisfied; thus, RWE is not a limiting event. Scram setpoints
(equipment settings that initiate automatic plant shutdowns) will be
established such that there is no increase in scram frequency due to
the WRNMS. No new challenges to safety-related equipment will result
from WRNMS.
2. The proposed changes do not create the possibility of a new
or different kind of accident from any previously evaluated.
As summarized below, this change will not create the possibility
of a new or different kind of accident from any accident previously
evaluated.
The components of the WRNMS will be supplied to equivalent or
better design and qualification criteria than is currently required
for the plant. Equipment that could be affected by WRNMS has been
evaluated. No new operating mode, safety-related equipment lineup,
accident scenario, system interaction, or equipment failure mode was
identified. Therefore, the WRNMS will not adversely affect plant
equipment.
3. The proposed changes do not involve a significant reduction
in a margin of safety.
All the SRM/IRM functions required in the Technical
Specifications are replaced with equivalent (more reliable) WRNMS
functions. The accuracy and response times of the WRNMS are superior
to those of the SRM/IRM subsystems. Implementation of the WRNMS does
not affect any fuel or safety limit. The applicable Bases of the
Technical Specifications have been rewritten, and the new Bases
maintain the equivalent margin of safety as was provided by the SRM/
IRM Bases.
The WRNMS (a) does not decrease a channel trip occurrence beyond
its acceptable limit, (b) does not increase a channel response time
beyond its acceptable limit, (c) increases indicated accuracies, and
(d) does not cause any plant parameter for any analyzed event to
fall outside of its acceptable limit(s).
The surveillance test frequency change of 7 to 31 days is based
on the WRNMS having (1) fixed in-core detectors, (2) greater
reliability than the SRMs and IRMs, and (3) self test features. The
13 second allowable value for the WRNM Period-Short surveillance,
and the surveillance test frequency change of 184 days to 24 months
is based on trip setpoint calculations using GE's standard (NRC
approved) setpoint methodology.
The WRNMS will not involve a reduction in a margin of safety, as
loads on plant equipment will not increase, and reactions to or
results of transients and postulated accidents will not increase
from those presently approved by the NRC.
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.
Pennsylvania Power and Light Company, Docket No. 50-387, Susquehanna
Steam Electric Station, Unit 1, Luzerne County, Pennsylvania
Date of amendment request: April 11, 1995.
Description of amendment request: This amendment would extend on a
one time basis the allowed outage time in the Susquehanna Steam
Electric Station Technical Specification 3.8.1.1 from 3 to 7 days for
one offsite circuit being out of service. This change will provide
additional time if needed to complete modifications to an offsite
circuit.
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:
I. Involve a significant increase in the probability or
consequences of an accident previously evaluated.
The consequences of losing offsite power have been evaluated in
the FSAR [Final Safety Analysis Report] and the Station Blackout
evaluation. Increasing the AOT [allowed outage time] for T-10 [an
offsite power source] from 3 to 7 days does not increase the
consequences of a LOOP [loss of offsite power] event nor change the
evaluation of LOOP events as stated in the FSAR or Station Blackout
evaluation.
Allowing T-10 to be removed from service for an additional 4
days does increase slightly the possibility of a LOOP event as shown
in PP&L's [Pennsylvania Power & Light Company's] engineering study.
However, implementing the following compensatory actions reduces the
probability of a LOOP event:
1. prohibiting high risk activities within the confines of the
plant or the grid system that may result in a loss of T-20 [the
second offsite power transformer] during the T-10 outage,
2. performing the modification during the Fall when the
frequency of grid and weather related LOOPs are reduced,
3. requiring a unit shutdown if the HPCI [high pressure core
injection] system becomes inoperable during the T-10 outage,
4. requiring a unit shutdown if the SLCS [standby liquid control
system] becomes inoperable during the T-10 outage,
5. requiring that within 24 hours prior to taking T-10 out of
service, Surveillance 4.8.1.1.2.a.4 be successfully completed on the
aligned diesel generators, and
6. maintaining the following equipment operable during the T-10
work window and restoring any failed system/component to operable
status as soon as possible (The failed system/component shall be
worked around the clock):
Both CRD [control rod drive] pumps,
Diesel fire pump, yard fire hydrant (1FH122) and
associated hydrant hose station,
RHR [residual heat removal system]/RHRSW [residual heat
removal service water system]/ESW [emergency service water system]
for suppression pool cooling,
RHR/RHRSW cross tie valves,
RCIC [reactor core isolation cooling]
CIG [containment instrument gas] 150 psig header and
bottles,
Turbine Building Closed Cooling Water System (one pump
and heat exchanger),
Portable diesel generator,
HV-141-F019.
Therefore, this change will not involve a significant increase
in the probability or consequences of an accident previously
evaluated.
II. Create the possibility of a new or different kind of
accident from any accident previously evaluated.
Allowing the AOT for T-10 to increase from 3 to 7 days is a one
time exemption in order to install the new T-10 tap and 230 kV
switch yard. The accident analyses affected by this extension are
the LOOP events. The remaining portions of the station and equipment
are not altered by this change. The potential for the loss of other
plant systems or equipment to mitigate the effects of an accident
are not altered. One offsite source of power will be out of service
for an additional 4 days and compensatory actions will be initiated
to lessen the effect of having the offsite power source out of
service for an additional 4 days. Therefore, this change will not
create the possibility of a new or different kind of accident from
any accident previously evaluated.
III. Involve a significant reduction in a margin of safety.
The proposed change allows, on a one time basis, T-10 to be out
of service for an additional 4 days. This increase in AOT for T-10
results in a slight decrease in the margin of safety (defined as
core damage frequency) with respect to having two offsite sources
available per Specification 3.8.1.1. By implementing the
compensatory measures as described in Item 1 above, the margin of
safety is increased to be the equivalent of allowing the offsite
power source (T-10) to be out of service for 3 days as is allowed by
the existing Specification. Therefore, this one time exemption will
not involve a significant reduction in safety margin.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
[[Page 29887]] satisfied. Therefore, the NRC staff proposes to
determine that the amendment request involves no significant hazards
consideration.
Local Public Document Room location: Osterhout Free Library,
Reference Department, 71 South Franklin Street, Wilkes-Barre,
Pennsylvania 18701.
Attorney for licensee: Jay Silberg, Esquire, Shaw, Pittman, Potts
and Trowbridge, 2300 N Street NW., Washington, DC 20037.
NRC Project Director: John F. Stolz.
Pennsylvania Power and Light Company, Docket Nos. 50-387 and 50-388
Susquehanna Steam Electric Station, Units 1 and 2, Luzerne County,
Pennsylvania
Date of amendment request: April 10, 1995.
Description of amendment request: This amendment would relocate
response time limit tables from the Susquehanna Steam Electric Station
Unit 1 and Unit 2 Technical Specifications (TS) to the Final Safety
Analysis Report. This modification is a line item improvement to the TS
as described in Generic Letter 93-08, ``Relocation of the Technical
Specification Tables of Instrument Response Time Limits,'' dated
December 29, 1993.
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:
I. This proposal does not involve a significant increase in the
probability or consequences of an accident previously evaluated.
The purpose of the proposed Tech. Spec. [Technical
Specification] change is to delete and subsequently relocate Tech.
Spec. Table 3.3.1-2, Table 3.3.2-3, and Table 3.3.3-3, to the SSES
FSAR consistent with the guidance provided in Generic Letter 93-08.
This is a line item Tech. Spec. improvement change recommended by
the NRC in Generic Letter 93-08. This change will allow PP&L
[Pennsylvania Power & Light Company] to administratively control
subsequent changes to the response time limits in accordance with
10CFR50.59. The procedures that contain the various response time
limits are also subject to the change control provisions in the
Administrative Controls section of the Tech. Specs. The proposed
change only relocates the existing response time limits; the
surveillance requirements and associated Actions are not affected
and remain in the Tech. Specs. Relocating the response time limit
information does not affect the analysis of any design basis
accident. The response times of these systems will be maintained
within the acceptance limits assumed in SSES [Susquehanna Steam
Electric Station] safety analyses and required for successful
mitigation of an initiating event. Also, since any subsequent
changes to the FSAR or procedures will be evaluated in accordance
with 10 CFR 50.59, no increase in the probability or consequences of
an accident previously evaluated will occur. Therefore, this change
does not involve a significant increase in the probability or
consequences of an accident previously evaluated.
II. This proposal does not increase the possibility of a new or
different kind of accident from any accident previously evaluated.
As discussed above, the proposed Tech. Spec. changes do not
affect the capability of the associated systems to perform their
intended function within the acceptance limits assumed in SSES
safety analyses and required for successful mitigation of an
initiating event. The proposed change does not involve a physical
modification of the plant or changes in methods governing normal
plant operations. The proposed change will not impose any different
operational or surveillance requirements. This change only proposes
to relocate these requirements to other plant documents whereby
adequate control of information will be maintained. No new failure
modes will be introduced. Therefore, the proposed change does not
create the possibility of a new or different kind of accident from
any accident previously evaluated.
III. This change does not involve a significant reduction in a
margin of safety.
The proposed change will not reduce a margin of safety because
it has no impact on any safety analysis assumption. The proposed
change does not alter the scope of equipment currently required to
be OPERABLE or subject to testing, nor does the proposed change
affect any instrument setpoints or equipment safety functions. Since
any future changes to these requirements in the FSAR or procedures
will be evaluated per the requirements of 10 CFR 50.59, no reduction
in a margin of safety will occur. Therefore, the change does not
involve a significant reduction in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: Osterhout Free Library,
Reference Department, 71 South Franklin Street, Wilkes-Barre,
Pennsylvania 18701.
Attorney for licensee: Jay Silberg, Esquire, Shaw, Pittman, Potts
and Trowbridge, 2300 N Street NW., Washington, DC 20037.
NRC Project Director: John F. Stolz.
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: May 2, 1995.
Description of amendment request: The amendments would eliminate
the manual start for auxiliary feedwater from the Technical
Specification for Engineered Safety Feature (ESF) Actuation System
Instrumentation. The manual start will be tested during the quarterly
pump test. This change is consistent with NUREG-1431, ``Standard
Technical Specifications- Westinghouse Plants.''
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 analyzed.
The change to the ESF Actuation Instrumentation specification to
eliminate the requirements for manual initiation of the [Auxiliary
Feedwater] (AFW) Pumps does not change any operating characteristics
of the plant. The change will eliminate unnecessary AFW Pump starts
which increase wear on system components. Manual initiation is not
credited in the Salem safety analyses. Manual initiation is verified
quarterly on a staggered test basis by performance of specification
4.7.1.2.b. 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.
The proposed technical specification modifications do not change
system configurations, plant equipment or safety analyses.
Therefore, the proposed modifications will not increase the
possibility of a new or different kind of accident from any accident
previously identified.
3. Involve a significant reduction in a margin of safety.
The proposed change to the ESF Actuation Instrumentation
Specification does not affect the ability of the AFW System to
perform its design function. The manual initiation of the AFW Pump
is not credited in the Salem safety analyses. Manual initiation is
verified quarterly by performance of specification 4.7.1.2.b.
Therefore, these changes do not involve a significant reduction in
any margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
[[Page 29888]] amendment request involves no significant hazards
consideration.
Local Public Document Room location: Salem Free Public library, 112
West Broadway, Salem, New Jersey 08079.
Attorney for licensee: Mark J. Wetterhahn, Esquire, Winston and
Strawn, 1400 L Street, NW., Washington, DC 20005-3502.
NRC Project Director: John F. Stolz.
Tennessee Valley Authority, Docket Nos. 50-259, 50-260 and 50-296,
Browns Ferry Nuclear Plant, Units 1, 2 and 3, Limestone County, Alabama
Date of amendment request: January 4, 1995 (TS 355).
Description of amendment request: The proposed amendment changes
the applicability and surveillance requirements for the intermediate
range monitor (IRM), average power range monitor (APRM), and APRM
Inoperative Trip functions. The proposed amendment adopts provisions of
the Improved Standard Technical Specifications (NUREG-1433).
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 amendment does not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
The proposed change revises the frequency of functional tests
for the IRM and APRM High Flux (15% Scram) Trip Functions and
eliminates operability requirements for the IRM, APRM High Flux (15%
Scram), and APRM Inoperative Trip Functions in certain modes of
operation. The operation of these trip functions is not a precursor
to any design basis accident or transient analyzed in the Browns
Ferry Updated Final Safety Analysis Report. Therefore, this change
does not increase the probability of any previously evaluated
accident.
The proposed change will eliminate the requirement to re-perform
the functional tests for these trip functions prior to each startup
if the test is within its periodicity (once per 7 days). It will
also eliminate the operability requirement for the IRM High Flux
Trip Function in the Shutdown Mode and IRM, APRM High Flux (15%
Scram), and APRM Inoperative Trip Functions during the Refuel Mode
except when any control rod is withdrawn from a core cell containing
one or more fuel assemblies. The Specifications will still provide
for operability of the equipment in Modes where credit is taken in
the safety analysis. Therefore, this change does not increase the
consequences of any previously evaluated accident.
[2]. The proposed amendment does not create the possibility of a
new or different kind of accident from any accident previously
evaluated.
The propo
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