[Federal Register Volume 59, Number 148 (Wednesday, August 3, 1994)]
[Unknown Section]
[Page 0]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: 94-18741]
[[Page Unknown]]
[Federal Register: August 3, 1994]
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NUCLEAR REGULATORY COMMISSION
Biweekly Notice
Applications and Amendments to Facility Operating
LicensesInvolving No Significant Hazards Considerations
I. Background
Pursuant to Public Law 97-415, the U.S. Nuclear Regulatory
Commission (the Commission or NRC staff) is publishing this regular
biweekly notice. Public Law 97-415 revised section 189 of the Atomic
Energy Act of 1954, as amended (the Act), to require the Commission to
publish notice of any amendments issued, or proposed to be issued,
under a new provision of section 189 of the Act. This provision grants
the Commission the authority to issue and make immediately effective
any amendment to an operating license upon a determination by the
Commission that such amendment involves no significant hazards
consideration, notwithstanding the pendency before the Commission of a
request for a hearing from any person.
This biweekly notice includes all notices of amendments issued, or
proposed to be issued from July 11, 1994, through July 22, 1994. The
last biweekly notice was published on July 20, 1994 (59 FR 37060).
Notice Of Consideration Of Issuance Of Amendments To Facility
Operating Licenses, Proposed No Significant Hazards Consideration
Determination, 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
20555. The filing of requests for a hearing and petitions for leave to
intervene is discussed below.
By September 2, 1994, 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 20555 and at the local
public document room for the particular facility involved. If a request
for a hearing or petition for leave to intervene is filed by the above
date, the Commission or an Atomic Safety and Licensing Board,
designated by the Commission or by the Chairman of the Atomic Safety
and Licensing Board Panel, will rule on the request and/or petition;
and the Secretary or the designated Atomic Safety and Licensing Board
will issue a notice of a hearing or an appropriate order.
As required by 10 CFR 2.714, a petition for leave to intervene
shall set forth with particularity the interest of the petitioner in
the proceeding, and how that interest may be affected by the results of
the proceeding. The petition should specifically explain the reasons
why intervention should be permitted with particular reference to the
following factors: (1) the nature of the petitioner's right under the
Act to be made a party to the proceeding; (2) the nature and extent of
the petitioner's property, financial, or other interest in the
proceeding; and (3) the possible effect of any order which may be
entered in the proceeding on the petitioner's interest. The petition
should also identify the specific aspect(s) of the subject matter of
the proceeding as to which petitioner wishes to intervene. Any person
who has filed a petition for leave to intervene or who has been
admitted as a party may amend the petition without requesting leave of
the Board up to 15 days prior to the first prehearing conference
scheduled in the proceeding, but such an amended petition must satisfy
the specificity requirements described above.
Not later than 15 days prior to the first prehearing conference
scheduled in the proceeding, a petitioner shall file a supplement to
the petition to intervene which must include a list of the contentions
which are sought to be litigated in the matter. Each contention must
consist of a specific statement of the issue of law or fact to be
raised or controverted. In addition, the petitioner shall provide a
brief explanation of the bases of the contention and a concise
statement of the alleged facts or expert opinion which support the
contention and on which the petitioner intends to rely in proving the
contention at the hearing. The petitioner must also provide references
to those specific sources and documents of which the petitioner is
aware and on which the petitioner intends to rely to establish those
facts or expert opinion. Petitioner must provide sufficient information
to show that a genuine dispute exists with the applicant on a material
issue of law or fact. Contentions shall be limited to matters within
the scope of the amendment under consideration. The contention must be
one which, if proven, would entitle the petitioner to relief. A
petitioner who fails to file such a supplement which satisfies these
requirements with respect to at least one contention will not be
permitted to participate as a party.
Those permitted to intervene become parties to the proceeding,
subject to any limitations in the order granting leave to intervene,
and have the opportunity to participate fully in the conduct of the
hearing, including the opportunity to present evidence and cross-
examine witnesses.
If a hearing is requested, the Commission will make a final
determination on the issue of no significant hazards consideration. The
final determination will serve to decide when the hearing is held.
If the final determination is that the amendment request involves
no significant hazards consideration, the Commission may issue the
amendment and make it immediately effective, notwithstanding the
request for a hearing. Any hearing held would take place after issuance
of the amendment.
If the final determination is that the amendment request involves a
significant hazards consideration, any hearing held would take place
before the issuance of any amendment.
A request for a hearing or a petition for leave to intervene must
be filed with the Secretary of the Commission, U.S. Nuclear Regulatory
Commission, Washington, DC 20555, Attention: Docketing and Services
Branch, or may be delivered to the Commission's Public Document Room,
the Gelman Building, 2120 L Street, NW., Washington DC 20555, 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 20555, and at the local public document
room for the particular facility involved.
Carolina Power & Light Company, Docket No. 50-261, H. B. Robinson
Steam Electric Plant, Unit No. 2, Darlington County, South Carolina
Date of amendment request: June 29, 1994
Description of amendment request: The proposed amendment will
delete the requirement to perform alternate train testing to
demonstrate that other, similar, safety-related components are operable
when components are found, or made, inoperable in the safety injection,
residual heat removal, and containment spray systems. The surveillance
requirements, which the licensee refers to as accelerated testing
requirements, affect the following components:
(a)Safety Injection (SI) pumps TS 3.3.1.2.b)
(b) Residual Heat Removal (RHR) Pumps (TS 3.3.1.2.c)
(c) SI and RHR flow paths (TS 3.3.1.2.e)
(d) Containment Spray (CS) (TS 3.3.2.2.a and b)
(e) CS flow paths (TS 3.3.2.2.c)
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 amendment does not involve a significant increase
in the probability of an accident previously evaluated because the
availability of the subject components will not be reduced and the
design and performance of the components are not being changed. The
subject components are provided to mitigate the consequences of
analyzed accidents; therefore their availability has no bearing on
the probability of occurrence of these accidents.
The proposed amendment does not involve a significant increase
in the consequences of an accident previously evaluated. This change
deletes alternate train testing requirements which, if maintained,
could result in loss of the safety function. Elimination of the
requirements will serve to ensure that one train of safety equipment
is always available to mitigate the consequences of an analyzed
accident. The remaining surveillance requirements provide adequate
assurance that the components will be operable when required.
Therefore the consequences of previously evaluated accidents will
not be increase.
2. The proposed amendment does not create the possibility of a
new or different kind of accident from any previously evaluated
because these proposed changes do not introduce any new modes of
operation or testing, and no physical changes are being made to the
plant. Therefore no new or different kind of accident could be
initiated by this amendment.
3. The proposed revisions do not involve a significant reduction
in the margin of safety since the routine testing requirements that
remain in the Technical Specifications provide adequate assurance
that the components will be operable when needed. Since the
elimination of this accelerated testing will decrease component wear
and improve availability, the margin of safety should be increased.
Since accelerated testing may still occur when component problems
involve a potential common mode failure, margins of safety
associated with the components' abilities to perform their design
functions will not be affected. Therefore, the proposed changes do
not involve a significant reduction in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 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: Hartsville Memorial Library,
Home and Fifth Avenues, Hartsville, South Carolina 29550
Attorney for licensee: R. E. Jones, General Counsel, Carolina Power
& Light Company, Post Office Box 1551, Raleigh, North Carolina 27602
NRC Project Director: David B. Matthews
Commonwealth Edison Company, Docket Nos. STN 50-454 and STN 50-455,
Byron Station, Unit Nos. 1 and 2, Ogle County, IllinoisDocket Nos.
STN 50-456 and STN 50-457, Braidwood Station, Unit Nos. 1 and 2,
Will County, Illinois
Date of amendment request: June 13, 1994
Description of amendment request: The proposed amendment would make
several changes to the Administrative Controls in Section 6 of
Technical Specifications (TS) for Byron and Braidwood stations. The
proposed changes include: (1) a change to the submittal frequency of
the Radiological Effluent Release Report, (2) a revision to the Shift
Technical Advisor description, (3) clarification of the Shift
Engineer's responsibilities, and (4) editorial changes. The references
to the Semiannual Radiological Effluent Release Report are also revised
in other sections of the 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:
A. The proposed changes do not involve a significant increase in
the probability or consequences of an accident previously evaluated.
The proposed changes to Section 6 of Technical Specifications do
not affect any accident initiators or precursors and do not change
or alter the design assumptions for the systems or components used
to mitigate the consequences of an accident.
The proposed changes are administrative in nature and provide
clarification. These changes provide consistency with station
procedures, programs, the Code of Federal Regulations, other
Technical Specifications, and Standard Technical Specifications.
These changes do not impact any accident previously evaluated in the
UFSAR.
B. The proposed changes do not create the possibility of a new
or different kind of accident from any accident previously
evaluated.
The proposed changes do not affect the design or operation of
any system, structure, or component in the plant. There are no
changes to parameters governing plant operation; no new or different
type of equipment will be installed. The proposed changes are
considered to be administrative changes. All responsibilities
described in Technical Specifications for management activities will
continue to be performed by qualified individuals.
C. The proposed changes do not involve a significant reduction
in a margin of safety.
The proposed changes do not affect the margin of safety for any
Technical Specification. The initial conditions and methodologies
used in the accident analyses remain unchanged, therefore, accident
analysis results are not impacted.
The proposed changes are administrative in nature and have no
impact on the margin of safety of any Technical Specification. They
do not affect any plant safety parameters or setpoints. The
descriptions for the Shift Technical Advisor and Shift Engineer are
clarified, however, include no reduction to their responsibilities.
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 Byron, the Byron Public
Library, 109 N. Franklin, P.O. Box 434, Byron, Illinois 61010; for
Braidwood, the Wilmington Township Public Library, 201 S. Kankakee
Street, Wilmington, Illinois 60481
Attorney for licensee: Michael I. Miller, Esquire; Sidley and
Austin, One First National Plaza, Chicago, Illinois 60690
NRC Project Director: Robert A. Capra
Commonwealth Edison Company, Docket Nos. STN 50-454 and STN 50-455,
Byron Station, Unit Nos. 1 and 2, Ogle County, IllinoisDocket Nos.
STN 50-456 and STN 50-457, Braidwood Station, Unit Nos. 1 and 2,
Will County, Illinois
Date of amendment request: July 6, 1994
Description of amendment request: The proposed amendments would
revise Technical Specification (TS) 3/4.4.5, ``Steam Generators,'' and
the associated bases. Previously the NRC granted amendments to the TSs
which authorized the use of selected steam generator sleeving
processes. In authorizing use of the processes, the amendments cited
references to specific NRC approved vendor technical reports, including
revision number. The proposed changes reference the reports in generic
terms as those that have been approved by the NRC, subject to
limitations and restrictions as noted by the NRC staff. While the
licensee will still have to request NRC approval for application of the
technologies as referenced in vendor reports, the licenses will not
have to be amended each time.
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 the Steam Generator section of Technical
Specifications do not affect any accident initiators or precursors
and do not alter the design assumptions for the systems or
components used to mitigate the consequences of an accident. These
changes are editorial changes to the requirements currently
identified in the Technical Specifications. The requirements
approved by the NRC will not be reduced by this request. The
proposed change maintains the administrative controls necessary to
ensure safe plant operation.
The original amendment requested tubesheet sleeves and tube
support plate sleeves as an alternate tube repair method for Bryon
and Braidwood Units 1 and 2. The steam generator sleeves approved
for installation use the Westinghouse process (laser welded joints)
or the Babcock & Wilcox Nuclear Technologies (BWNT) process of
kinetically welded joints. The sleeve configuration was designed and
analyzed in accordance with the criteria of Regulatory Guide (RG)
1.121 and the design requirements of Section III of the American
Society of Mechanical Engineers (ASME) Code. Fatigue and stress
analyses of the sleeved tube assemblies for both processes produced
acceptable results documented in the current Westinghouse and BWNT
Technical Reports. The proposed Technical Specifications change to
allow the use of the current NRC approved laser welded or
kinetically welded sleeving process does not adversely impact any
other previously evaluated design basis accident or the results of
these analyses. Therefore, the editorial changes to the referenced
sleeving Technical Reports will not increase the probability of
occurrence of an accident previously evaluated.
2. The proposed change does not create the possibility of a new
or different kind of accident from any accident previously
evaluated.
The proposed changes are considered to be administrative
changes. All the requirements described in Technical Specifications
``Acceptance Criteria'' for the Steam Generators will continue to be
implemented as described in the current Technical Reports.
Referencing the current Westinghouse or BWNT Sleeving Technical
Reports currently approved by the NRC and subject to the limitations
and restrictions as noted by the NRC, has no effect upon any design
transient or accident analyses. The proposed changes do not affect
the design or operation of any system, structure, or component in
the plant. There are no changes to parameters governing plant
operation and no new or different type of equipment will be
installed.
The use of the proposed sleeving processes will not introduce
significant or adverse changes to the plant design basis. Stress and
fatigue analyses of the repair have shown the ASME Code and
RG 1.121 allowable values are met. Implementation of the
currently approved laser welded or kinetically welded sleeving will
continue to maintain the overall tube bundle structural integrity at
a level consistent with that of the originally supplied tubing.
Repair of a tube with a sleeve does not provide a mechanism which
would result in an accident outside of the area affected by the
sleeve. Any hypothetical accident as a result of potential tube or
sleeve degradation in the repaired portion of the tube is bounded by
the existing steam generator tube rupture accident analysis. The
tube rupture accident analysis accounts for the installation of
sleeves and the impact on current plugging level analyses. The
sleeve design does not affect any other component or location on the
tube outside of the immediate area repaired.
Thus, the possibility of a new or different type of accident
from any accident previously evaluated is not created.
3. The proposed change does not involve a significant reduction
in a margin of safety.
The proposed change is administrative in nature and has no
impact on the margin of safety of any Technical Specification.
Specific technical reports are no longer referenced in Technical
Specifications. An editorial change is made to TS referencing the
current NRC approved vendor Technical Report, subject to the
limitations and restrictions noted by the NRC. The initial
conditions and methodologies used in the accident analyses remain
unchanged.
The laser welded and kinetically welded sleeving repair of
degraded steam generator tubes has been shown by analysis to restore
the integrity of the tube bundle to its original design basis
condition. The safety factors used in the design of sleeves for the
repair of degraded tubes are consistent with the safety factors in
the ASME Boiler and Pressure Vessel Code used in steam generator
design. The design of the tube sleeves has been verified by testing
to preclude leakage during normal and postulated accident
conditions. Installation of either type of vendor sleeve using the
current approved process will continue to maintain the structural
integrity of the steam generator tubes.
Thus, these changes do not involve a significant reduction in
the margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: For Byron, the Byron Public
Library, 109 N. Franklin, P.O. Box 434, Byron, Illinois 61010; for
Braidwood, the Wilmington Township Public Library, 201 S. Kankakee
Street, Wilmington, Illinois 60481
Attorney for licensee: Michael I. Miller, Esquire; Sidley and
Austin, One First National Plaza, Chicago, Illinois 60690
NRC Project Director: Robert A. Capra
Commonwealth Edison Company, Docket Nos. 50-373 and 50-374, LaSalle
County Station, Units 1 and 2, LaSalle County, Illinois
Date of amendment request: June 9, 1994
Description of amendment request: The proposed amendments would
revise the LaSalle County Station, Units 1 and 2, Technical
Specifications (TS), Appendix A, in order to facilitate implementation
of the Thermal Limits portion of the General Electric Average Power
Range Monitor (APRM)/Rod Block Monitor (RBM)/TS Improvement Program
(ARTS).
Specifically, the proposed TS change will create power and flow
dependent Minimum Critical Power Ratio (MCPR) and Maximum Average
Planar Linear Heat Generation Rate (MAPLHGR) limits, and other
administrative changes.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
The proposed changes do not involve a significant increase in
the probability or consequences of an accident previously evaluated
because:
The probability of an accident previously evaluated will not
increase as a result of this change, because no changes to plant
systems will occur. All changes are related to core monitoring
software, and there will be no physical changes to equipment.
The consequences of an accident previously evaluated will not
increase as a result of the proposed changes. The power- and flow-
dependent MCPR and MAPLHGR limits incorporate sufficient
conservatism so the safety limit MCPR [SLMCPR] (operating limit MCPR
[OLMCPR] for automatic flow control) and the fuel thermal-mechanical
limits will not be violated for any power and flow condition.
Because these limits are protected during normal operation, the
consequences of any transient will not increase with this change in
limit definition. General Electric has verified in Attachment E that
the introduction of Arts will not cause any change in the Licensing
Basis PCT [Peak Centerline Temperature] resulting from a Loss-Of-
Coolant Accident [LOCA], nor any change in the results satisfying
the other LOCA acceptance criteria of 10 CFR 50.46 and Section
15.6.5 of NUREG-0800 (Standard Review Plan), which are: cladding
oxidation, metal-water reaction (hydrogen generation), coolable
geometry and long-term cooling.
The proposed changes do not create the possibility of a new or
different kind of accident from any accident previously evaluated
because:
Since no physical changes to any plant system are occurring,
there will be no new or different types of accidents created by this
change. No interactions between equipment systems will be changed in
any manner.
The proposed changes do not involve a significant reduction in a
margin of safety because:
The power- and flow-dependent MCPR and MAPLHGR limits will
sufficiently protect the SLMCPR (OLMCPR for automatic flow control)
and the fuel thermal-mechanical limits at all power and flow
conditions. The ARTS limits conservatively assure that all licensing
criteria are satisfied without setdown of the flow referenced APRM
scram and rod block trips. The limits were developed using NRC
approved methods, and satisfy the same NRC approved criteria that
the APRM setdown requirement does.
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: Public Library of Illinois
Valley Community College, Rural Route No. 1, Ogelsby, Illinois 61348
Attorney for licensee: Michael I. Miller, Esquire; Sidley and
Austin, One First National Plaza, Chicago, Illinois 60690
NRC Project Director: Robert A. Capra
Duke Power Company, et al., Docket Nos. 50-413 and 50-414, Catawba
Nuclear Station, Units 1 and 2, York County, South Carolina
Date of amendment request: July 13, 1994
Description of amendment request: The requested amendments would
allow the testing interval in Technical Specification Surveillance
Requirement 4.6.2 for the air or smoke flow test through each
containment spray header to be increased from 5 to 10 years. The
licensee states that the proposed amendments are consistent with NRC
staff guidance contained in NUREG-1366, ``Improvements to Technical
Specifications Surveillance Requirements,'' and Generic Letter 93-05,
``Line-Item Technical Specifications Improvements to Reduce
Surveillance Requirements for Testing During Power Operation.'' In
addition, the amendments would also remove an obsolete footnote related
to the Catawba Unit 1 first refueling. The licensee's application
jointly addressed both its Catawba and McGuire Nuclear Stations. This
notice addresses those aspects applicable only to the Catawba Station.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
Criterion 1
The requested amendments will not involve a significant increase
in the probability or consequences of an accident previously
evaluated. Increasing the surveillance interval of TS [Technical
Specification] 4.6.2d from five to ten years will have no impact
upon the probability of any accident, since the NS [containment
spray] system is not accident initiating equipment. Also, since
Catawba's... flow test histor[y] support[s] making the proposed
change, system response following an accident will not be adversely
affected. Therefore, the requested amendments will not result in
increased accident consequences. Deletion of the obsolete footnote
as indicated in the Catawba TS markup is purely an administrative
change, and therefore will have no impact upon either the
probability or consequences of any accident.
Criterion 2
The requested amendments will not create the possibility of a
new or different kind of accident from any accident previously
evaluated. As stated above, the NS system is not accident initiating
equipment. No new failure modes can be created from an accident
standpoint. The plant will not be operated in a different manner.
Deletion of the Catawba obsolete footnote has no bearing on any
accident initiating mechanisms.
Criterion 3
The requested amendments will not involve a significant
reduction in a margin of safety. Plant safety margins will be
unaffected by the proposed changes. The NS system will still be
capable of fulfilling its required safety function, since plant
operating experience supports the proposed change. Finally, the
proposed amendments are consistent with the NRC position and
guidance set forth in NUREG-1366 and Generic Letter 93-05. Deletion
of the Catawba obsolete footnote will not result in any impact to
plant safety margins.
Based upon the preceding analyses, Duke Power Company concludes
that the requested amendments do not involve a significant hazards
consideration.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: York County Library, 138 East
Black Street, Rock Hill, South Carolina 29730
Attorney for licensee: Mr. Albert Carr, Duke Power Company, 422
South Church Street, Charlotte, North Carolina 28242
NRC Project Director: Herbert N. Berkow
Duke Power Company, Docket Nos. 50-369 and 50-370, McGuire Nuclear
Station, Units 1 and 2, Mecklenburg County, North Carolina
Date of amendment request: July 13, 1994
Description of amendment request: The requested amendments would
allow the testing interval in Technical Specification Surveillance
Requirement 4.6.2 for the air or smoke flow test through each
containment spray header to be increased from 5 to 10 years. The
licensee states that the proposed amendments are consistent with NRC
staff guidance contained in NUREG-1366, ``Improvements to Technical
Specifications Surveillance Requirements,'' and Generic Letter 93-05,
``Line-Item Technical Specifications Improvements to Reduce
Surveillance Requirements for Testing During Power Operation.'' The
licensee's application jointly addressed both its Catawba and McGuire
Nuclear Stations. This notice addresses those aspects applicable only
to the McGuire Station.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
Criterion 1
The requested amendments will not involve a significant increase
in the probability or consequences of an accident previously
evaluated. Increasing the surveillance interval of TS [Technical
Specification] 4.6.2d from five to ten years will have no impact
upon the probability of any accident, since the NS [containment
spray] system is not accident initiating equipment. Also, since...
McGuire's flow test histor[y] support[s] making the proposed change,
system response following an accident will not be adversely
affected. Therefore, the requested amendments will not result in
increased accident consequences. ...
Criterion 2
The requested amendments will not create the possibility of a
new or different kind of accident from any accident previously
evaluated. As stated above, the NS system is not accident initiating
equipment. No new failure modes can be created from an accident
standpoint. The plant will not be operated in a different manner.
...
Criterion 3
The requested amendments will not involve a significant
reduction in a margin of safety. Plant safety margins will be
unaffected by the proposed changes. The NS system will still be
capable of fulfilling its required safety function, since plant
operating experiences supports the proposed change. Finally, the
proposed amendments are consistent with the NRC position and
guidance set forth in NUREG-1366 and Generic Letter 93-05. ...
Based upon the preceding analyses, Duke Power Company concludes
that the requested amendments do not involve a significant hazards
consideration.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: Atkins Library, University of
North Carolina, Charlotte (UNCC Station), North Carolina 28223
Attorney for licensee: Mr. Albert Carr, Duke Power Company, 422
South Church Street, Charlotte, North Carolina 28242
NRC Project Director: Herbert N. Berkow
Duke Power Company, Docket Nos. 50-269, 50-270 and 50-287, Oconee
Nuclear Station, Units 1, 2 and 3, Oconee County, South Carolina
Date of amendment request: December 8, 1993, as supplemented April
20, 1994.
Description of amendment request: The amendments would revise
Technical Specification (TS) 3.4 to address the need to bypass
automatic initiation of the Emergency Feedwater (EFW) system when the
main feedwater pump discharge pressure is below actuation setpoint
during startup and shutdown in order to prevent inadvertent actuation.
The proposed amendment is in response to NRC Inspection Report 50-269,
50-270, 50-287/90-30 (Inspector Followup Item 90-30-02), which
determined that the existing TSs regarding initiation circuitry for the
EFW system were inadequate. The amendments would also delete
operability requirements for the Emergency Condenser Cooling Water
(ECCW) system. The licensee determined that the ECCW system is not
required to remove decay heat following any design basis event.
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 amendment request would not] involve a significant
increase in the probability or consequences of an accident
previously evaluated.
Each accident analysis addressed within the Oconee FSAR [Final
Safety Analysis Report] has been examined with respect to changes
proposed within this amendment request. The design basis of the
Emergency Feedwater (EFW) System is to supply feedwater to the steam
generators in the event Main Feedwater is lost. The EFW system
provides the required flow rate to cool the RCS [reactor coolant
system] down to the point at which the Decay Heat Removal System is
designed to operate. The EFW system is also designed to cool the RCS
following a small break LOCA [loss of coolant accident]. Changes
included within this amendment request are provided to clarify
requirements for the operability of EFW. Specifically, theses
changes clarify that automatic initiation circuitry due to low main
feedwater pump discharge pressure or low hydraulic oil pressure may
be bypassed when the reactor is shutdown to prevent inadvertent
actuation. In addition, these changes provide that if an EFW pump is
inoperable due only to the inoperability of automatic initiation,
cooldown to below 250 deg.F is not necessary after the reactor is
shutdown. Accident analysis for the loss of main feedwater, and
subsequent initiation of EFW, assumes initial conditions of the
reactor at full power operation. The utilization of criticality for
this specific automatic initiation circuitry to be operable ensures
the EFW system is operated within the boundaries of design basis for
Oconee while also providing a reasonable margin to prevent
inadvertent actuation. It is not possible to place this automatic
initiation circuitry in service prior to exceeding 250 deg.F because
the main feedwater pump discharge pressure is well below the
initiation setpoint at this value. Manual initiation circuitry
operability is required prior to exceeding an RCS temperature of
250 deg.F. This change only clarifies existing configuration and
control for the Oconee units and does not increase the probability
or consequences of any accident previously evaluated.
This change also removes the requirement for Emergency Condenser
Cooling Water (ECCW) System operability for the removal of decay
heat using the secondary systems. The ability to provide flow
through the condenser from the ECCW system is a preferred method for
decay heat removal. However, this mode of operation is not necessary
to prevent or mitigate any accident previously evaluated. The
primary success path for decay heat removal following loss of
station power events, and thus loss of normal CCW flow, is the use
of the turbine driven EFW pump providing flow to the steam
generators and heat removal via the main steam safety relief valves
to the atmosphere. Analysis has shown that sufficient inventory
exists in secondary systems, as limited by Technical Specification
3.4.4, to provide for decay heat removal.
Therefore, this proposed change deletes the requirement for ECCW
for secondary systems decay heat removal. The probability or
consequences of any design basis accident are not increased by this
change. As such, this change will not involve a significant increase
in the probability or consequences of an accident previously
evaluated.
(2) [The amendment request would not] create the possibility of
a new or different kind of accident from any kind of accident
previously evaluated.
Changes included within this amendment request are provided to
clarify existing requirements for operability of the EFW System and
remove the requirement for ECCW flow through the condenser for decay
heat removal. Operation of Oconee units in accordance with these
Technical Specifications will not create any failure modes not
bounded by previously evaluated accidents. Previously evaluated
accidents assume an initial condition of power operation for loss of
main feedwater events. Providing for automatic initiation prior to
criticality ensures operation within the bounds of design analysis.
Previously evaluated accidents also assume the removal of decay
heat, following loss of normal CCW flow, to be via the main steam
safety relief valves to the atmosphere which eliminates the need for
ECCW operability. Consequently, this change will not create the
possibility of a new or different kind of accident from any kind of
accident previously evaluated.
(3) [The amendment request would not] involve a significant
reduction in a margin of safety.
The design basis of the EFW system is to supply feedwater to the
steam generators in the event Main Feedwater is lost. By providing
clarification that manual initiation circuitry is operable prior to
exceeding an RCS temperature of 250 deg.F and automatic initiation
circuitry, due to low main feedwater discharge pressure or low
hydraulic oil pressure, is operable prior to criticality, there is
no significant reduction in the margin of safety associated with
this amendment request. The ECCW system is designed to provide a
means to remove decay heat without a loss of secondary side
inventory. However, analysis has shown that sufficient secondary
side inventory exist, as specified by Technical Specification 3.4.4,
to provide for coping with loss of station power events.
Furthermore, even though this method of decay heat removal is
desirable, Oconee PRA [probabilistic risk assessment] studies do not
model the loss of ECCW for accident precursors since it is not
required and margins of safety are not reduced if it is not
available. Changes included within this amendment request clarify
existing requirements for the operability of secondary system for
decay heat removal based on previously evaluated accidents. As such,
all margins of safety are preserved. Therefore, there will be no
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
amendment request involves no significant hazards consideration.
Local Public Document Room location: Oconee County Library, 501
West South Broad Street, Walhalla, South Carolina 29691
Attorney for licensee: J. Michael McGarry, III, Winston and Strawn,
200 17th Street, NW., Washington, DC 20036
NRC Project Director: Herbert N. Berkow
Duquesne Light Company, et al., Docket Nos. 50-334 and 50-412,
Beaver Valley Power Station, Unit Nos. 1 and 2, Shippingport,
Pennsylvania
Date of amendment request: June 2, 1994
Description of amendment request: The proposed amendment would
revise Technical Specification (TS) sections 3.4.6.1 and 3.4.6.2
related to reactor coolant system (RCS) operational leakage and leakage
detection instrumentation. The proposed amendment would revise the TSs
to be in accordance with the standard TSs in NUREG-1431 in so far as
the plant-specific design will allow. The proposed changes relate to
the limiting conditions for operation and the surveillance requirements
for the four primary instruments used to detect RCS leakage. Changes
are also proposed for the index and definition sections. A new TS,
section 3/4.5.4, is proposed to address reactor coolant pump seal
injection flow.
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 of occurrence of a previously evaluated
accident, i.e., loss of coolant accident (LOCA), is not increased
because the ability of the plant operators to detect RCS leakage and
take appropriate corrective action is not changed. The proposed
change will continue to ensure that diverse means for detecting
extremely small leaks are available to plant operators. In
addition, the proposed amendment does not change the operational
leakage limits. The seal injection flow limit is not affected by
this proposed change. Due to these three factors, the probability of
occurrence of a LOCA is not increased. The consequences of an
accident previously evaluated are not significantly increased
because the proposed changes do not affect the ability of the
various safety systems to perform their intended function. The
leakage detection monitors do not initiate any automatic function to
mitigate the consequences of a LOCA. They provide an early
indication of RCS leakage. The operational leakage limits are not
affected by this proposed change and they do not initiate any
automatic function to mitigate the consequences of a LOCA. The
proposed change to the seal injection flow requirement will continue
to ensure that ECCS flow will be as assumed in the accident
analyses.
Therefore, based on the continued ability of the leakage
detection monitors and independent monitoring capabilities to detect
extremely small leaks, the fact that this proposed amendment does
not change the operational leakage limits, the seal injection flow
limit is not affected by this proposed change, and that the proposed
changes do not affect the ability of the various safety systems to
perform their intended functions, this 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 amendment does not change the plant configuration
in a way which introduces a new potential hazard to the plant. Since
design requirement[s] continue to be met and the integrity of the
RCS pressure boundary is not challenged, no new failure mode has
been created. As a result, an accident which is different than any
already evaluated in the Updated Final Safety Analysis Report
(UFSAR) will not be created due to this change.
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 does not involve a significant reduction in
a margin of safety since the operational leakage limits will not be
affected. Continued plant operation will not be permitted if
operational leakage exceeds the current technical specification
limits. The operational leakage limits establish limits which ensure
that any RCS leakage does not compromise safety. The protection of
the RCS pressure boundary from degradation and the core form [from]
inadequate cooling, in addition to preventing the accident analyses
radiation release assumptions from being exceeded, is the main
purpose of the operational leakage limits. The ability to detect and
quantify operational leakage allows plant operators to perform
actions to place the plant in a safe condition when leakage rate
indicates possible RCS pressure boundary degradation. The proposed
change will continue to ensure that diverse measurement means are
available to provide the plant operators with an early indication of
extremely small RCS leakage. Therefore, [the change is] allowing
action to be taken to place the plant in a safe condition when RCS
leakage indicates possible RCS pressure boundary leakage.
The proposed addition of the separate seal injection
specification will not change the flow limit on seal injection. The
new specification will continue to ensure that seal injection flow
is limited. This will ensure that sufficient flow to the reactor
core is provided during accident conditions. The proposed
elimination of the Mode 4 applicability, for seal injection flow
specification, will not involve a significant reduction in the
margin of safety since high seal injection flow is less critical as
a result of the lower initial RCS pressure and decay heat removal
requirements in Mode 4.
Therefore, this proposed change does not involve a significant
reduction in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: 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: Walter R. Butler
Entergy Operations Inc., Docket No. 50-382, Waterford Steam
ElectricStation, Unit 3, St. Charles Parish, Louisiana
Date of amendment request: June 6, 1994
Description of amendment request: The proposed admendment would
revise the technical specifications (TSs) by relocating the seismic and
meteorological monitoring instrumentation and their associated
requirements from the TSs to the Waterford 3 updated final safety
analysis report and plant procedures pursuant to the NRC final policy
statement on TSs improvements for nuclear power reactors. The final
policy statement was published in the Federal Register on Thursday,
July 22, 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:
The proposed change relocates Seismic and Meteorological
Monitoring Instrumentation requirements from the TS to licensee
controlled documents consistent with the NRC Policy Statement on
Technical Specification Improvements. Criterion 1 of the Policy
Statement indicates that the TS should include installed
instrumentation that is used to detect, and indicate in the control
room, a significant abnormal degradation of the reactor coolant
pressure boundary. This criterion is intended to ensure that the TS
control those instruments specifically installed to detect excessive
reactor coolant system leakage. This criterion is not interpreted to
include instrumentation used to detect precursors to reactor coolant
pressure boundary leakage (e.g., loose parts monitor, seismic
instrumentation, valve position indicators). Combustion Engineering
and the NRC have previously determined that relocating Seismic and
Meteorological Monitoring Instrumentation requirements from the TS
does not affect any material condition of the plant that could
directly contribute to causing or mitigating the effects of an
accident.
Therefore, the proposed change will not involve a significant
increase in the probability or consequences of any accident
previously evaluated.
The proposed change will not involve any design change or
modification to the plant. The proposed change will not alter the
operation of the plant or the manner in which it is operated. Any
subsequent change to the Seismic or Meteorological Monitoring
Instrumentation requirements will undergo a review in accordance
with the criteria of 10 CFR 50.59 to ensure that the change does not
involve an unreviewed safety question.
Therefore, the proposed change will not create the possibility
of a new or different kind of accident from any accident previously
evaluated.
The proposed change will relocate Seismic and Meteorological
Monitoring Instrumentation requirements from the TS to licensee
controlled documents subject to the criteria of 10 CFR 50.59. The
proposed change will have no adverse impact on any protective
boundary or safety limit.
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
Entergy Operations Inc., Docket No. 50-382, Waterford Steam
ElectricStation, Unit 3, St. Charles Parish, Louisiana
Date of amendment request: June 22, 1994
Description of amendment request: The proposed amendment would
revise the technical specifications (TSs) to change three plant
protection system (PPS) trip setpoints to be consistent with the
current setpoint/uncertainty methodology being implemented at Waterford
3. The change adjusts the affected TSs values in a more conservative
direction.
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:
Implementing the proposed change will not affect any design
basis accident. The revised trip and actuation setpoints are based
upon the same analytical limits that form the basis for the current
trip and actuation setpoints. The design basis for each trip and
actuation setpoint was verified to be consistent with the
appropriate accident analyses as part of the process of revising the
PPS setpoint analysis. The proposed changes in trip and actuation
setpoints are all in the conservative (away from the analytical
limits) direction. Therefore, the proposed change will not involve a
significant increase in the probability or consequences of any
previously analyzed accident.
Plant operation and the manner in which the plant is operated
will not be altered as a result of implementing the proposed change
since no new system or design change is being implemented.
Therefore, the proposed change will not create the possibility of a
new or different kind of accident from any accident previously
evaluated.
The current safety margins of the affected trip setpoints and
allowable values is preserved by the proposed change. This is
assured by retaining the current analytical limit for the affected
parameters. Since the analytical limits are not affected and the
total channel uncertainty is increased, the margin of safety for the
affected trip setpoints and allowable values is preserved.
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
Entergy Operations Inc., Docket No. 50-382, Waterford Steam
ElectricStation, Unit 3, St. Charles Parish, Louisiana
Date of amendment request: June 22, 1994
Description of amendment request: The proposed amendment would
revise the technical specifications (TSs) to replace the generic
control room outside air intake (CROAI) radiation alarm/trip setpoint
(less than or equal to 2x background) with a specific setpoint (less
than or equal to 4.09E-5). The new setpoint is based on radioactive
material concentrations in the control room not exceeding the derived
air concentrations (DAC) occupational values listed in 10 CFR Part 20,
Appendix B, Table 1, Column 3.
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 replaces the current CROAI radiation monitor
alarm/trip setpoint of less than or equal to 2x background with a
fixed value independent of background radiation. The new setpoint
will continue to provide protection to plant personnel such that
occupational radiation exposure is maintained within the limits of
10 CFR 20 during normal plant operation, anticipated operational
occurrences or design basis accidents.
Therefore, the proposed change will not involve a significant
increase in the probability or consequences of any accident
previously evaluated.
The proposed change will replace the generic CROAI radiation
monitor alarm/trip setpoint with a setpoint derived from a site-
specific calculation. The proposed change will not alter the
operation of the plant or the manner in which it 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 proposed change will replace the current CROAI radiation
monitor setpoint with a new setpoint that will ensure occupational
radiation exposure will not exceed the DAC limits of 10 CFR 20. The
proposed change has no adverse impact on protective boundaries,
safety limits, or margin of safety.
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
Florida Power and Light Company, Docket Nos. 50-250 and 50-251,
Turkey Point Plant Units 3 and 4, Dade County, Florida
Date of amendment request: July 19, 1994
Description of amendment request: The licensee proposes to change
Turkey Point Units 3 and 4 Technical Specifications by relocating
cycle-specific parameter limits from the Technical Specifications to
the Core Operating Limits Report (COLR). Presently, the parameter
limits for Turkey Point Units 3 and 4 are calculated using NRC-approved
methodologies. These limits are evaluated for every reload cycle and
may be revised by a license amendment as appropriate, to reflect
changes to cycle-specific variables.The curves to be relocated include
(a) TS Figure 3.1-2, Rod Bank Insertion Limits versus Thermal Power
curve, and (b) TS Figure 3.2-2, K(Z) Normalized FQ(Z) as a
Function of Core Height curve.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
(1) Operation of the facility in accordance with the proposed
amendments would not involve a significant increase in the
probability or consequences of an accident previously evaluated.
The removal of cycle-specific Rod Bank Insertion limits and the
K(Z) curve from the Turkey Point Units 3 and 4 Technical
Specifications is administrative in nature and has no impact on the
probability or consequences of any Design Bases Event (DBE)
occurrences which was previously evaluated. The determination of the
Rod Bank Insertion limits and K(Z) curve will be performed using
methodology approved by the NRC and poses no significant increase in
the probability or consequences of any accident previously
evaluated.
The Rod Bank Insertion limits and K(Z) curve will be evaluated
every cycle to ensure proper compliance with the Updated Final
Safety Analysis Report (UFSAR). These limits will be evaluated in
accordance with 10 CFR Sec. 50.59, which ensures that the reload
will not involve an increase in the probability of occurrences or
consequences of an accident previously evaluated. 10 CFR Sec. 50.59
(2) states that a proposed change involves an unreviewed safety
question (i) if the probability of occurrence or the consequences of
an accident or malfunction of equipment important to safety
previously evaluated in the safety analysis report may be increased.
Consequently, since any change to the reload core design analysis
must be evaluated relative to the more restrictive evaluation
criterion of 10 CFR Sec. 50.59, then operation of the facility in
accordance with the proposed amendments would not involve a
significant increase in the probability or consequences of an
accident previously evaluated.
(2) Operation of the facility in accordance with the proposed
amendments would not create the possibility of a new or different
kind of accident from any accident previously evaluated.
The removal of the Rod Bank Insertion limits and K(Z) curve from
the Technical Specifications is administrative in nature and has no
impact, nor does it contribute in any way to the possibility of a
new or different kind of accident from any accident previously
evaluated. No new accident scenarios, failure mechanisms or limiting
single failure events are introduced as a result of the proposed
change.
The generation of the Rod Bank Insertion limits and K(Z) curve
will be performed using NRC-approved methodology and are submitted
to the NRC, as a revision to the COLR, to allow the NRC staff to
trend. The Technical Specifications will continue to require
operation within the core operating limits and appropriate actions
will be taken if these limits are exceeded.
10 CFR Sec. 50.59 permits a licensee to make changes in the
facility as described in the safety analysis report without prior
Commission approval, provided that the proposed changes does not
involve an unreviewed safety question. 10 CFR Sec. 50.59 (2) states
that a proposed change involves an unreviewed safety question (ii)
if a possibility for an accident or malfunction of a different type
than any evaluated previously in the safety analysis report may be
created. Consequently, since any change to the reload core design
analysis must be evaluated relative to the more restrictive
evaluation criterion of 10 CFR Sec. 50.59, then operation of the
facility in accordance with the proposed amendments would not create
the possibility of a new or different kind of accident from any
accident previously evaluated.
(3) Operation of the facility in accordance with the proposed
amendments would not involve a significant reduction in a margin of
safety.
The margin of safety is not affected by the removal of the Rod
Bank Insertion limits and K(Z) curve from the Technical
Specifications. The methodology for the reload core design analysis
have been approved by the NRC and does not constitute a significant
reduction in the margin of safety.
The supporting Technical Specification values are defined by the
accident analyses which are performed to conservatively bound the
operating conditions defined by the Technical Specifications. The
development of the limits for future reloads will continue to
conform to the methodology described in NRC approved documentation.
In addition, each future reload will involve a 10 CFR 50.59 review
to assure that operation of the units within the cycle specific
limits will not involve a reduction in a margin of safety. 10 CFR
Sec. 50.59 (2) states that a proposed change involves an unreviewed
safety question (iii) if the margin of safety as defined in the
basis for any technical specification is reduced. Consequently,
since any change to the reload core design analysis must be
evaluated relative to the more restrictive evaluation criterion of
10 CFR Sec. 50.59, then operation of the facility in accordance with
the proposed amendments would not involve a significant reduction in
a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 50.92(c) are
satisfied.Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: Florida International
University, University Park, Miami, Florida 33199
Attorney for licensee: Harold F. Reis, Esquire, Newman and Holtzer,
P.C., 1615 L Street, NW., Washington, DC 20036
NRC Project Director: Victor M. McCree, Acting
Florida Power and Light Company, Docket Nos. 50-250 and 50-251,
Turkey Point Plant Units 3 and 4, Dade County, Florida
Date of amendment request: July 19, 1994
Description of amendment request: The licensee proposes to change
Turkey Point Units 3 and 4 Technical Specifications and its associated
BASES, which address the maximum allowed reactor thermal power
operation with inoperable main steam safety valves (MSSVs).
Westinghouse issued Nuclear Safety Advisory Letter (NSAL) 94-001 which
notified the licensee of a deficiency in the basis of the Turkey Point
Technical Specification 3/4.7.1, which allows the plant to operate at
reduced power levels with a specified number of MSSVs inoperable. This
amendment request corrects the allowable power level with inoperable
MSSVs.
The licensee also proposed changes to the TS 3.7.1.1 applicability
statement to indicate that, for mode 3 only, the actions are required
when the Reactor Trip System breakers are in the closed position and
the Control Rod Drive System is capable of rod withdrawal.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
(1) Operation of the facility in accordance with the proposed
amendments would not involve a significant increase in the
probability or consequences of an accident previously evaluated.
The proposed changes do not involve a significant increase in
the probability or consequences of an accident previously evaluated.
The new power range neutron flux high setpoint values will ensure
that the secondary side steam pressure will remain below 110 percent
of the design value following a Loss of Load/Turbine Trip event,
when one or more main steam safety valves (MSSVs) are declared
inoperable. The proposed change will not impact the classification
of the Loss of Load/Turbine Trip event as a Condition II probability
event (faults of moderate frequency) per ANSI - N18.2, 1973.
Accordingly, since the new power range neutron flux setpoints will
maintain the capability of the MSSVs to perform their pressure
relief function associated with a Loss of Load/Turbine Trip event,
there will be no effect on the probability or consequences of an
accident previously evaluated.
In addition, the proposed change to the applicability statement
of TS 3.7.1.1, will not effect the probability or consequences of an
accident previously evaluated, since the proposed plant condition
with the reactor trip breakers open and the rod control system not
capable of withdrawing rods is an analyzed safe shutdown condition.
(2) Operation of the facility in accordance with the proposed
amendments would not create the possibility of a new or different
kind of accident from any accident previously evaluated.
The proposed changes do not create the possibility of a new or
different kind of accident from any accident previously evaluated.
The proposed changes do not involve any change to the configuration
or method of operation of any plant equipment, and no new failure
modes have been defined for any plant system or component. The new
power range neutron flux high setpoints will maintain the capability
of the MSSVs to perform their pressure relief function to ensure the
secondary side steam design pressure is not exceeded following a
Loss of Load/Turbine Trip event. Therefore, since the function of
the MSSVs is unaffected by the proposed changes, the possibility of
a new or different kind of accident from any accident previously
evaluated is not created.
In addition, the proposed change to the applicability statement
of TS 3.7.1.1, will not create the possibility of a new or different
kind of accident from any accident previously evaluated, since the
proposed plant condition with the reactor trip breakers open and the
rod control system not capable of withdrawing rods is an analyzed
safe shutdown condition.
(3) Operation of the facility in accordance with the proposed
amendments would not involve a significant reduction in a margin of
safety.
The proposed changes to the Technical Specifications [do] not
involve a significant reduction in a margin of safety. The algorithm
methodology used to calculate the new power range neutron flux high
setpoints is conservative and bounding since it is based on a number
of inoperable MSSVs per loop; i.e., if only one MSSV in one loop is
out of service, the applicable power range setpoint would be the
same as if one MSSV in each loop were out of service. Another
conservatism with the algorithm methodology is with the assumed
minimum total steam flow rate capability of the operable MSSVs. The
assumption is that if one or more MSSVs are inoperable per loop, the
inoperable MSSVs are the largest capacity MSSVs, regardless of which
capacity MSSVs are actually inoperable. Therefore, since the power
range neutron flux setpoints calculated for the proposed changes
using the algorithm methodology are more conservative and ensure the
secondary side steam design pressure is not exceeded following a
Loss of Load/Turbine Trip event, this proposed license amendment
will not involve a significant reduction in a margin of safety.
In addition, the proposed change to the applicability statement
of TS 3.7.1.1, does not involve a significant reduction in a margin
of safety, since the proposed plant condition with the reactor trip
breakers open and the rod control system not capable of withdrawing
rods is an analyzed safe shutdown condition.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 50.92(c) are
satisfied.Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: Florida International
University, University Park, Miami, Florida 33199
Attorney for licensee: Harold F. Reis, Esquire, Newman and Holtzer,
P.C., 1615 L Street, NW., Washington, DC 20036
NRC Project Director: Victor M. McCree, Acting
Florida Power and Light Company, Docket Nos. 50-250 and 50-251,
Turkey Point Plant Units 3 and 4, Dade County, Florida
Date of amendment request: July 19, 1994
Description of amendment request: The licensee proposes to change
Turkey Point Units 3 and 4 Technical Specifications (TS) by revising
Surveillance Requirements 4.8.1.1.2e. and 4.8.1.1.2f., to delete the
specific reference in the TS of the American Society for Testing and
Materials (ASTM) testing standard being used to meet TS testing
requirements. The Emergency Diesel Generator (EDG) fuel oil TS
Surveillance Requirements will be replaced with a requirement to test
the EDG fuel oil in accordance with the Turkey Point Units 3 and 4
Diesel Fuel Oil Testing Program.
The licensee proposes the addition of ACTION statements g. and h.
of TS 3.8.1.1 to address the required action in the event the diesel
fuel oil does not meet the Diesel Fuel Oil Testing Program limits.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
(1) Operation of the facility in accordance with the proposed
amendments would not involve a significant increase in the
probability or consequences of an accident previously evaluated.
The proposed changes to the Technical Specifications will permit
the Technical Specification required testing of Emergency Diesel
Generator (EDG) fuel oil in accordance with the Turkey Point Units 3
and 4 Diesel Fuel Oil Testing Program. The proposed change will
permit FPL to use more recent editions of the American Society for
Testing and Materials (ASTM) standards currently listed in Technical
Specification Surveillance Requirements 4.8.1.1.2e. and 4.8.1.1.2f.
Prior to changing the Diesel Fuel Oil Testing Program, the proposed
change will be evaluated pursuant to Title 10 Code of Federal
Regulations Sec. 50.59 (10 CFR Sec. 50.59), ``Changes, tests, and
experiments.'' Title 10 CFR Sec. 50.59 permits a licensee to make
changes in the procedures as described in the safety analysis report
without prior Commission approval, provided that the proposed
changes [do] not involve an unreviewed safety question.
Title 10 CFR Sec. 50.59(a)(2) states that a proposed change
involves an unreviewed safety question (i) if the probability of
occurrence or the consequences of an accident or malfunction of
equipment important to safety previously evaluated in the safety
analysis report may be increased. Consequently, since any change to
the Diesel Fuel Oil Testing Program, including the ASTM standard or
ASTM edition standard to be used to evaluate EDG fuel oil
acceptability, the change must be evaluated relative to the more
restrictive evaluation criterion of 10 CFR Sec. 50.59, then
operation of the facility in accordance with the proposed amendments
would not involve a significant increase in the probability or
consequences of an accident previously evaluated. The EDG fuel oil
TS Surveillance Requirements will be replaced with a requirement to
test the EDG fuel oil in accordance with the Turkey Point Units 3
and 4 Diesel Fuel Oil Testing Program.
ACTION statement g. of TS 3.8.1.1 is added to address the
required action in the event the new fuel oil properties do not meet
the Diesel Fuel Oil Testing Program limits. A failure to meet the
API gravity, kinematic viscosity, flash point or clarity limits is
cause for rejecting the new fuel oil prior to the addition to the
Diesel Fuel Oil Storage Tanks, but does not represent a failure to
meet the Limiting Condition for Operation (LCO) of TS 3.8.1.1, since
the new fuel oil has not been added to the storage tanks. Provided
these new fuel oil properties are met subsequent to the addition of
the new fuel oil to the storage tanks, 30 days is provided to
complete the analyses of the other fuel oil properties specified in
Table 1 of ASTM-D975-81, except sulfur which may be performed in
accordance with ASTM-D1552-79 or ASTM-D2622-82. In the event the
other new fuel oil properties specified in Table 1 of ASTM-D975-81
are not met, ACTION statement g. of TS 3.8.1.1 provides an
additional 30 days to meet the Diesel Fuel Oil Testing Program
limits. This additional 30 day period is acceptable because the fuel
oil properties of interest, even if they are not within limits,
would not have an immediate effect on EDG operation.
ACTION statement h. of TS 3.8.1.1 is added to address the
required action in the event the stored fuel oil total particulates
does not meet the Diesel Fuel Oil Testing Program limits. Fuel oil
degradation during long term storage shows up as an increase in
particulate, due mostly to oxidation. The presence of particulate
does not mean the fuel oil will not burn properly in a diesel
engine. The frequency for performing surveillance on stored fuel oil
is based on stored fuel oil degradation trends which indicate that
particulate concentration is unlikely to change significantly
between surveillances.
Prior to changing the Turkey Point Units 3 and 4 Diesel Fuel Oil
Testing Program, FPL will need to determine if the proposed program
change is at least as, if not more, effective, in detecting
unsatisfactory fuel oil. The EDGs will thus continue to function as
designed and the probability or consequences of previously evaluated
accidents will be unaffected.
(2) Operation of the facility in accordance with the proposed
amendments would not create the possibility of a new or different
kind of accident from any accident previously evaluated.
The proposed changes to the Technical Specifications will permit
the Technical Specification required testing of Emergency Diesel
Generator fuel oil using more recent editions of the American
Society for Testing and Materials standards listed in Technical
Specification Surveillance Requirements 4.8.1.1.2e. and 4.8.1.1.2f.
Prior to changing the edition of the previously approved ASTM
standard being used to evaluate the EDG fuel oil, the proposed
edition standard will be evaluated pursuant to 10 CFR Sec. 50.59,
``Changes, tests, and experiments.'' Title 10 CFR Sec. 50.59 permits
a licensee to make changes in the procedures as described in the
safety analysis report without prior Commission approval, provided
that the proposed changes does not involve an unreviewed safety
question. Title 10 CFR Sec. 50.59(a)(2) states that a proposed
change involves an unreviewed safety question (ii) if a possibility
for an accident or malfunction of a different type than any
evaluated previously in the safety analysis report may be created.
Consequently, since any change to the edition of the ASTM standard
to be used to evaluate EDG fuel oil acceptability must be evaluated
relative to the more restrictive evaluation criterion of 10 CFR
Sec. 50.59, then operation of the facility in accordance with the
proposed amendments would not create the possibility of a new or
different kind of accident from any accident previously evaluated.
ACTION statement g. of TS 3.8.1.1 is added to address the
required action in the event the new fuel oil properties do not meet
the Diesel Fuel Oil Testing Program limits. A failure to meet the
API gravity, kinematic viscosity, flash point or clarity limits is
cause for rejecting the new fuel oil prior to the addition to the
Diesel Fuel Oil Storage Tanks, but does not represent a failure to
meet the Limiting Condition for Operation (LCO) of TS 3.8.1.1, since
the new fuel oil has not been added to the storage tanks. Provided
these new fuel oil properties are met subsequent to the addition of
the new fuel oil to the storage tanks, 30 days is provided to
complete the analyses of the other fuel oil properties specified in
Table 1 of ASTM-D975-81, except sulfur which may be performed in
accordance with ASTM-D1552-79 or ASTM-D2622-82. In the event the
other new fuel oil properties specified in Table 1 of ASTM-D975-81
are not met, ACTION statement g. of TS 3.8.1.1 provides an
additional 30 days to meet the Diesel Fuel Oil Testing Program
limits. This additional 30 day period is acceptable because the fuel
oil properties of interest, even if they are not within limits,
would not have an immediate effect on EDG operation.
ACTION statement h. of TS 3.8.1.1 is added to address the
required action in the event the stored fuel oil total particulates
[do] not meet the Diesel Fuel Oil Testing Program limits. Fuel oil
degradation during long term storage shows up as an increase in
particulate, due mostly to oxidation. The presence of particulate
does not mean the fuel oil will not burn properly in a diesel
engine. The frequency for performing surveillance on stored fuel oil
is based on stored fuel oil degradation trends which indicate that
particulate concentration is unlikely to change significantly
between surveillances.
Prior to changing the Turkey Point Units 3 and 4 Diesel Fuel Oil
Testing Program, FPL will need to determine if the proposed program
change is at least as, if not more, effective, in detecting
unsatisfactory fuel oil. Since the proposed changes do not involve a
change in the design of any plant system or component, and since the
proposed changes will need to evaluate the effect of any ASTM
standard edition change on the level of EDG reliability, the change
proposed will not create the possibility of a new or different kind
of accident from any accident previously evaluated.
(3) Operation of the facility in accordance with the proposed
amendments would not involve a significant reduction in a margin of
safety.
The proposed changes to the Technical Specifications will permit
the Technical Specification required testing of Emergency Diesel
Generator (EDG) fuel oil using more recent editions of the American
Society for Testing and Materials (ASTM) standards listed in
Technical Specification Surveillance Requirements 4.8.1.1.2e. and
4.8.1.1.2f. Prior to changing the edition of the previously approved
ASTM standard being used to evaluate the EDG fuel oil, the proposed
edition standard will be evaluated pursuant to 10 CFR Sec. 50.59,
``Changes, tests, and experiments.'' Title 10 CFR Sec. 50.59 permits
a licensee to make changes in the procedures as described in the
safety analysis report without prior Commission approval, provided
that the proposed changes [do] not involve an unreviewed safety
question. Title 10 CFR Sec. 50.59(a)(2) states that a proposed
change involves an unreviewed safety question (iii) if the margin of
safety as defined in the basis for any technical specification is
reduced. Consequently, since any change to the edition of the ASTM
standard to be used to evaluate EDG fuel oil acceptability must be
evaluated relative to the more restrictive evaluation criterion of
10 CFR Sec. 50.59, then operation of the facility in accordance with
the proposed amendments would not involve a significant reduction in
a margin of safety.
ACTION statement g. of TS 3.8.1.1 is added to address the
required action in the event the new fuel oil properties do not meet
the Diesel Fuel Oil Testing Program limits. A failure to meet the
API gravity, kinematic viscosity, flash point or clarity limits is
cause for rejecting the new fuel oil prior to the addition to the
Diesel Fuel Oil Storage Tanks, but does not represent a failure to
meet the Limiting Condition for Operation (LCO) of TS 3.8.1.1, since
the new fuel oil has not been added to the storage tanks. Provided
these new fuel oil properties are met subsequent to the addition of
the new fuel oil to the storage tanks, 30 days is provided to
complete the analyses of the other fuel oil properties specified in
Table 1 of ASTM-D975-81, except sulfur which may be performed in
accordance with ASTM-D1552-79 or ASTM-D2622-82. In the event the
other new fuel oil properties specified in Table 1 of ASTM-D975-81
are not met, ACTION statement g. of TS 3.8.1.1 provides an
additional 30 days to meet the Diesel Fuel Oil Testing Program
limits. This additional 30 day period is acceptable because the fuel
oil properties of interest, even if they are not within limits,
would not have an immediate effect on EDG operation.
ACTION statement h. of TS 3.8.1.1 is added to address the
required action in the event the stored fuel oil total particulates
[do] not meet the Diesel Fuel Oil Testing Program limits. Fuel oil
degradation during long term storage shows up as an increase in
particulate, due mostly to oxidation. The presence of particulate
does not mean the fuel oil will not burn properly in a diesel
engine. The frequency for performing surveillance on stored fuel oil
is based on stored fuel oil degradation trends which indicate that
particulate concentration is unlikely to change significantly
between surveillances.
Prior to changing the Turkey Point Units 3 and 4 Diesel Fuel Oil
Testing Program, FPL will need to determine if the proposed program
change is at least as, if not more, effective, in detecting
unsatisfactory fuel oil. Since the proposed changes will require a
safety evaluation to assure that the reliability of the EDGs using
fuel oil tested in accordance with the different ASTM standard
edition maintains the current margin of safety, the proposed changes
do not involve a reduction in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 50.92(c) are
satisfied.Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: Florida International
University, University Park, Miami, Florida 33199
Attorney for licensee: Harold F. Reis, Esquire, Newman and Holtzer,
P.C., 1615 L Street, NW., Washington, DC 20036
NRC Project Director: Victor M. McCree, Acting
Georgia Power Company, Oglethorpe Power Corporation, Municipal
Electric Authority of Georgia, City of Dalton, Georgia, Docket Nos.
50-424 and 50-425, Vogtle Electric Generating Plant, Units 1 and 2,
Burke County, Georgia
Date of amendment request: April 28, 1994
Description of amendment request: The proposed amendments would
revise Technical Specification (TS) 3/4.8.1.1, ``AC Sources
Operating,'' and the associated TS Bases for demonstrating the
operability of the diesel generators (DGs), based upon three NRC
guidelines:
A. Generic Letter (GL) 93-05, ``Line-Item Technical Specifications
Improvements to Reduce Surveillance Requirements for Testing During
Power Operation.''
1. Delete from TS action statement a the requirement to test the
DGs in the event of an inoperable offsite circuit.
2. Eliminate from TS action statement b the need to test the
operable DG if the other DG became inoperable due to an inoperable
support system or an independently testable component in addition to
the existing provision excluding preplanned preventive maintenance or
testing. Furthermore, ifthe operable DG must be tested, it would be
tested within 8 hours (rather than 24 hours) unless the absence of any
potential common mode failure for the remaining DG is demonstrated.
3. Also eliminate from TS action statement c the need to test the
operable DG if the other DG became inoperable due to an inoperable
support system or an independently testable component in addition to
the existing provision excluding preplanned preventive maintenance or
testing. In addition, the operable DG would not have to be tested if
the absence of any potential common mode failure for the remaining DG
is demonstrated. A reference to TS action statement a would be deleted
because of the proposed change to TS action statement a described
above.
4. Eliminate from TS action statement e the need to test the DGs
when two offsite circuits are inoperable.
5. Revise TS 4.8.1.1.2.g.2 to allow the DG to be gradually loaded,
as opposed to a fast loading of 60 seconds or less to an indicated
value of 6100-7000 kw. This change would extend gradual loading of DGs
(that GL 93-05 recommends for routine monthly surveillance) to the 6-
month surveillance.
B. Regulatory Guide (RG) 1.9, Revision 3, ``Selection, Design,
Qualification, and Testing of Emergency Diesel Generator Units Used as
Class 1E Onsite Electric Power Systems at Nuclear Power Plants,''
(insofar as this guide relates to reducing DG stress and wear due to
testing and the elimination of certain reporting requirements).
1. Incorporate into TS 4.8.1.1.2.a.4 the provision to perform
routine monthly testing by gradually accelerating the DG to operating
speed, rather than requiring the DG to attain rated voltage and
frequency within 11.4 seconds. As a direct result of this proposed
change, TS action statements b, c, and f would be revised to reference
TS 4.8.1.1.2.g.1 instead of TS 4.8.1.1.2.a.4 in the event that an
operable DG must be tested when the other DG is inoperable. This has
the effect of requiring the operable DG to be fast-started for testing
pursuant to the action statement.
2. In TS 4.8.1.1.2.h.7, separate the 24-hour endurance run from the
hot restart test. As a result, create new TS 4.8.1.1.2.h.8 to require
the hot restart test. The DG would be operated for a minimum of 2 hours
at a load of 6800-7000 kw, and the DG would be shut down. Within 5
minutes of shutdown, the DG would be restarted and required to attain
rated voltage and frequency within 11.4 seconds. Delete existing TS
footnote (which provides for reperforming the hot
restart test without repeating the 24-hour endurance test) which is no
longer required. Renumber existing TSs 4.8.1.1.2.h.8, .9, .10, .11, and
.12 to accommodate the addition of new TS 4.8.1.1.2.h.8.
3. Delete TS 4.8.1.1.3, ``Reports.'' (This is also in accordance
with the Improved Technical Specifications, Revision 0, dated September
28, 1992).
4. In TS 4.8.1.2, ``A. C. Sources Shutdown,'' delete the reference
to deleted TS 4.8.1.1.3.
C. NUMARC 87-00, Revision 1, ``Guidelines and Technical Bases for
NUMARC Initiatives Addressing Station Blackout at Light Water
Reactors,'' (insofar as it relates to the test frequency for a problem
DG). Specifically, TS Table 4.8-1, ``Diesel Generator Test Schedule,''
would be revised to incorporate the test schedule of Section D.2.4.4 of
Appendix D to NUMARC 87-00, Revision 1. Under the proposed schedule,
testing pursuant to TS 4.8.1.1.2.a would be conducted monthly provided
the number of valid failures in the last 25 demands for a given DG is
no more than 3. If the number of valid failures is 4 or more, testing
would be conducted at least once per 7 days (but at intervals of no
less than 24 hours) until 7 consecutive failure-free starts from
standby conditions and load-run demands have been performed. Note that
both NUMARC 87-00, Revision 1, and RG 1.9, Revision 3, use and define
the terms ``start demand, start failure, load-run demand, and load-run
failure'' rather than the old RG 1.108 terminology of valid tests. In
fact, Section D.2.4.4 of Appendix D to NUMARC 87-00 refers to the last
25 ``demands'' rather than tests. Therefore, the proposed change to TS
Table 4.8-1 would count valid failures in terms of demands rather than
valid tests. The criteria for determining the number of valid failures
and demands would be in accordance with RG 1.9, Revision 3.
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 proposed changes affect the required actions in response to
inoperable offsite and onsite ac sources, surveillance requirements
for the emergency diesel generators, and reporting requirements for
diesel generator failures. The proposed changes are based on the
recommendations of Regulatory Guide 1.9, Revision 3, NUMARC 87-00,
Revision 1, and Generic Letter 93-05. They are expected to result in
improvements in diesel generator testing and failure reporting and
reduce diesel generator aging due to excessive testing. As such, the
proposed changes should result in improved diesel generator
reliability, thereby providing additional assurance that the diesel
generators will be capable of performing their safety function.
Therefore, the proposed changes will not significantly increase the
probability or consequences of any accident previously evaluated.
2. The proposed changes do not create the possibility of a new
or different kind of accident from any accident previously
evaluated. The proposed changes affect the action and surveillance
requirements for the onsite and offsite ac sources. Accordingly, the
proposed changes do not involve any change to the configuration or
method of operation of any plant equipment, and no new failure modes
have been defined for any plant system or component nor has any new
limiting failure been identified as a result of the proposed
changes. Therefore, the proposed changes do not create the
possibility of a new or different kind of accident from any
previously evaluated.
3. The proposed changes do not involve a significant reduction
in a margin of safety. The proposed changes are based on existing
regulatory guidance. Under the proposed changes, the emergency
diesel generators will remain capable of performing their safety
function, and the effects of aging on the diesel generators will be
reduced by eliminating unnecessary testing. Therefore, the proposed
changes do not involve a significant reduction in a margin of
safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 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: Burke County Public Library,
412 Fourth Street, Waynesboro, Georgia 30830.
Attorney for licensee: Mr. Arthur H. Domby, Esquire, Troutman
Sanders, Nations Bank Plaza, Suite 5200, 600 Peachtree Street, NE.,
Atlanta, Georgia 30308-2210.
NRC Project Director: Herbert N. Berkow
GPU Nuclear Corporation, Docket No. 50-320, Three Mile Island
Nuclear Station, Unit No. 2, (TMI-2), Dauphin County, Pennsylvania
Date of amendment request: October 9, 1991.
Description of amendment request: Facility Operating License No.
DPR-73, a possession only license for the TMI-2 facility, held by
General Public Utilities Nuclear Corporation (GPU Nuclear), expires
November 4, 2009. The proposed amendment would extend the expiration
date of
Facility Operating License No. DPR-73 for TMI-2 to April 19, 2014.
No other changes to the license or the Technical Specifications are
proposed.
The TMI-2 facility is currently in long term storage. GPU Nuclear,
the licensee, has named this storage period Post Defueling Monitored
Storage or PDMS. The licensee plans to maintain TMI-2 in PDMS until
Three Mile Island Nuclear Station Unit No. 1 (TMI-1) permanently ceases
operation, at which time both TMI-1 and TMI-2 will be decommissioned
simultaneously. The TMI-1 Operating License expires on April 19, 2014.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration which is presented below:
10 CFR 50.92 provides the criteria which the Commission uses to
perform a No Significant Hazards Consideration. 10 CFR 50.92 states
that an amendment to a facility license involves No Significant
Hazards if 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 proposed modification of the expiration date of the TMI-2
License does not involve any physical changes to the facility. All
that is involved is an extension of the time TMI-2 would be in a
monitored storage condition. Based on this, GPU Nuclear concludes
that the proposed change does not:
1. Involve a significant increase in the probability or
consequences of an accident previously evaluated. Accident
evaluations for PDMS are provided in the PDMS Safety Analysis Report
(SAR) and the PDMS Final Programmatic Environmental Impact
Statement, Supplement 3 (PEIS) dated August 1989. These documents
evaluated monitored storage of TMI-2 for extended periods of time
and provide for surveillances to ensure monitored storage conditions
are appropriately maintained. The PDMS PEIS specifically evaluated
monitored storage until 2014. No evaluated accident has a
probability or consequence that are increased significantly during
the period 2009 to 2014 over the period before 2009. Therefore, it
can be concluded that this change does not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
2. Create the possibility of a new or different kind of accident
from any accident previously evaluated. As previously stated, the
act of modifying the expiration date of the TMI-2 License does not
involve any physical changes to the facility and therefore, the
possibility of a new or different kind of accident is not created.
3. Involve a significant reduction in the margin of safety
during PDMS. The surveillances identified in the PDMS SAR will be
performed to ensure that the facility is maintained in the condition
defined by the SAR. These conditions and surveillances will continue
to apply during the extended license. Therefore, there will not be a
reduction in the margin of safety.
Based on the above analysis, it is concluded that the proposed
changes involve No Significant Hazards Consideration as defined by
10 CFR 50.92.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 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, Walnut Street and Commonwealth
Avenue, Box 1601 Harrisburg, Pennsylvania 17105
Attorney for licensee: Ernest L. Blake, Jr., Esquire, Shaw,
Pittman, Potts & Trowbridge, 2300 N Street, NW, Washington, D.C. 20037
NRC Project Director: Seymour H. Weiss
IES Utilities Inc., Docket No. 50-331, Duane Arnold Energy Center,
Linn County, Iowa
Date of amendment request: June 30, 1994
Description of amendment request: The proposed amendment would
clarify the requirement for the audit of conformance to Technical
Specifications, delete the requirement for Safety Committee oversight
of the Emergency Plan and Security Plan and allow designation by the
Plant Superintendent signature authority for procedure approval.
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 amendment does not involve a significant increase
in the probability or consequences of an accident previously
evaluated. No physical changes will result from this amendment.
These changes revise audit requirements and procedure approval
requirements. The subject audits will still be performed to provide
assurance of conformance to the requirements, and the procedures
will still receive adequate technical reviews by the cognizant
departments while relieving the Plant Superintendent-Nuclear of the
administrative burden of signing each procedure revision.
2) The proposed amendment does not create the possibility of a
new or different kind of accident from any accident previously
evaluated. No physical changes will result from this amendment.
These changes revise audit requirements and procedure approval
requirements. The subject audits will still be performed to provide
assurance of conformance to the requirements, and the procedures
will still receive adequate technical reviews by the cognizant
departments while relieving the Plant Superintendent-Nuclear of the
administrative burden of signing each procedure revision.
3) The proposed amendment does not involve a significant
reduction in a margin of safety. No physical changes will result
from this amendment. These changes revise audit requirements and
procedure approval requirements. The subject audits will still be
performed to provide assurance of conformance to the requirements,
and the procedures will still receive adequate technical reviews by
the cognizant departments while relieving the Plant Superintendent-
Nuclear of the administrative burden of signing each procedure
revision.
The NRC staff has reviewed the licensee's analysis and, based on
thisreview, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: Cedar Rapids Public Library,
500 First Street, S.E., Cedar Rapids, Iowa 52401
Attorney for licensee: Jack Newman, Esquire, Kathleen H. Shea,
Esquire, Newman and Holtzinger, 1615 L Street, NW., Washington, DC
20036
NRC Project Director: John N. Hannon
IES Utilities Inc., Docket No. 50-331, Duane Arnold Energy Center,
Linn County, Iowa
Date of amendment request: June 30, 1994
Description of amendment request: The proposed amendment would add
Operability Requirements, Limiting Conditions for Operations (LCO) and
Surveillance Requirements for the Control Building Chillers.
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 will not involve a significant
increase in the probability or consequences of an accident
previously evaluated because the requested revisions do not affect
any FSAR analysis involving these systems.
The proposed revision only adds LIMITING CONDITIONS for
OPERATION (LCO) and Surveillance Requirements (SR) for the Control
Building Chillers. These additions will provide assurance that the
affected systems will be OPERABLE when required.
2) The proposed amendment will not create the possibility of a
new or different kind of accident from any accident previously
evaluated because there is no equipment or design change associated
with this change. The proposed amendment only adds LCOs and SRs for
the Control Building Chillers.
3) The proposed amendment will not involve any reduction in a
margin of safety. The safety function of the Control Building
Chillers is to remove the design basis heat load under all normal
and emergency conditions. The addition of LCOs and SRs for the
Control Building Chillers ensures they will be OPERABLE when
required.
The NRC staff has reviewed the licensee's analysis and, based on
thisreview, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: Cedar Rapids Public Library,
500 First Street, S.E., Cedar Rapids, Iowa 52401
Attorney for licensee: Jack Newman, Esquire, Kathleen H. Shea,
Esquire, Newman and Holtzinger, 1615 L Street, NW., Washington, DC
20036
NRC Project Director: John N. Hannon
IES Utilities Inc., Docket No. 50-331, Duane Arnold Energy Center,
Linn County, Iowa
Date of amendment request: June 30, 1994
Description of amendment request: The proposed amendment would
modify the surveillance testing of the Emergency Service Water (ESW)
system by deleting the flow rate test and the requirement to test the
pumps each week when river water temperature exceeds 80 deg.F and by
adding a surveillance regarding the Cedar River (Ultimate Heat Sink)
water temperature.
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. No physical changes will result from this
amendment. The ESW system will still maintain its ability to support
various safety related equipment which is designed to mitigate the
consequences of certain accidents and transients. These safety
related systems play no part in the probability of these accidents
or transients occurring. Since the ESW system will continue to fully
support the cooling requirements of the safety related equipment
which mitigate the consequences of certain accidents and transients,
this amendment will not affect the consequences of these accidents
and transients. The re-analysis of the component heat loads assumed
worst case conditions and involved conservative assumptions. Our
continuing program for monitoring heat exchanger performance, which
was established in response to Generic Letter 89-13, ``Service Water
System Problems Affecting Safety-Related Equipment,'' will continue
to verify that the individual components are capable of performing
their design function. Therefore, the proposed amendment does not
involve a change in the probability or consequences of an accident
previously evaluated.
2) The proposed license amendment does not create the
possibility of a new or different kind of accident from any
previously evaluated. The safety function of the ESW system is
unchanged. The revised flow requirements for the system have been
established using conservative assumptions and worst case heat loads
and are appropriately documented in the FSAR and plant procedures.
This amendment will result in no physical changes to the ESW system
and therefore, will not affect its ability to continue to provide
reliable cooling water. Consequently, the proposed license amendment
does not create the possibility of a new or different kind of
accident from any previously evaluated.
3) The proposed amendment will not reduce the margin of safety.
The re-analysis of the ESW flow rate requirements and component heat
loads was performed using conservative assumptions and maximum
component heat loads. The actual operation of the ESW system will
not be changed. Any degradation of ESW pump performance would be
detected by the IST program which requires quarterly testing of
these pumps and monitoring of the pump's differential pressure and
flow. Deleting the requirement to perform the surveillance each week
when river water temperature exceeds 80 deg.F will not reduce the
margin of safety because even at a river water temperature of
95 deg.F, the required ESW flow to supply all the branches is well
below the normal system flow rate of approximately 1100 gpm.
Deleting the weekly surveillance will eliminate unnecessary testing
of the ESW pumps, thereby reducing wear on the pumps. Adding a
surveillance requirement for river water temperature will formalize
the recording of water temperature every hour to assure acceptable
ESW performance.
The NRC staff has reviewed the licensee's analysis and, based on
thisreview, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: Cedar Rapids Public Library,
500 First Street, S.E., Cedar Rapids, Iowa 52401
Attorney for licensee: Jack Newman, Esquire, Kathleen H. Shea,
Esquire, Newman and Holtzinger, 1615 L Street, NW., Washington, DC
20036
NRC Project Director: John N. Hannon
Niagara Mohawk Power Corporation, Docket No. 50-220, Nine Mile
Point Nuclear Station Unit No. 1, Oswego County, New York
Date of amendment request: June 30, 1994
Description of amendment request: The proposed amendment would
revise Technical Specification (TS) 3.2.7.1, ``Primary Coolant System
Pressure Isolation Valves.'' Specifically, TS Table 3.2.7.1, ``Primary
Coolant System Pressure Isolation Valves,'' would be revised to add
Shutdown Cooling System (SCS) check valves 38-165, 166, 167, 168, 169,
170, 171, and 172 each with a maximum allowable leakage rate of less
than or equal to 0.375 gpm. The proposed amendment would add the check
valves in lieu of replacing the SCS isolation valves with ones that are
10 CFR Part 50, Appendix J, Type C air testable. The added check valves
would provide high pressure/low pressure interfaces between the high
pressure Reactor Coolant System and the low pressure Core Spray System.
The addition of the check valves will allow utilization of the Core
Spray System as a seal water system for sealing the Shutdown Cooling
isolation valves as permitted by Section III.C.3 of 10 CFR Part 50,
Appendix J.
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 operation of Nine Mile Point Unit 1, in accordance with the
proposed amendment, will not involve a significant increase in the
probability or consequences of an accident previously evaluated.
The proposed change requires the addition of Primary Coolant
System pressure isolation valves for the prevention of an
intersystem LOCA [loss-of-coolant accident]. The proposed addition
does not affect operation of either the Shutdown Cooling or Core
Spray Systems. These changes do not alter any accident initiators or
precursors and therefore does not affect the probability of a
previously evaluated accident.
Testing these valves in accordance with Specification 3.2.7.1
provides assurance that the Core Spray System will not be damaged by
an overpressurization event which could lead to potential loss of
integrity of the system and subsequent release of radioactivity.
Thus, the addition of the valves would not increase the consequences
of any accident. Therefore, the operation of Nine Mile Point Unit 1,
in accordance with the proposed amendment, will not involve a
significant increase in the probability or consequences of an
accident previously evaluated.
The operation of Nine Mile Point Unit 1, in accordance with the
proposed amendment, will not create the possibility of a new or
different kind of accident from any accident previously evaluated.
The proposed addition of Primary Coolant System pressure
isolation valves, although a physical change, does not alter the
initial conditions used for any design basis accident. The check
valves provide the high pressure/low pressure isolation between the
Reactor Coolant and Core Spray Systems. These valves will be subject
to leak rate testing in accordance with Specification 3.2.7.1. This
ensures that an intersystem LOCA is prevented. The proposed change
has no effect on operation of either the Shutdown Cooling or Core
Spray Systems. Therefore, the design capabilities of these systems
are not challenged in a manner previously assessed so as to create
the possibility of a new or different kind of accident. Accordingly,
operation of Nine Mile Point Unit 1, in accordance with the proposed
amendment, will not create the possibility of a new or different
kind of accident from any accident previously analyzed.
The operation of Nine Mile Point Unit 1, in accordance with the
proposed amendment, will not involve a significant reduction in a
margin of safety.
The proposed change which requires the addition of Primary
Coolant System pressure isolation valves, ensures proper isolation
of a high pressure/low pressure interface between the Reactor
Coolant and Core Spray Systems. The pressure isolation valves will
be leak tested in accordance with Specification 3.2.7.1. This
provides assurance that the Core Spray System will not be damaged by
an overpressurization event and will not result in loss of integrity
of the system. Thus, the results of any event previously analyzed
remains unchanged. Therefore, the operation of Nine Mile Point Unit
1, in accordance with the proposed amendment will not involve a
significant reduction in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: Reference and Documents
Department, Penfield Library, State University of New York, Oswego, New
York 13126.
Attorney for licensee: Mark J. Wetterhahn, Esquire, Winston &
Strawn, 1400 L Street, NW., Washington, DC 20005-3502.
NRC Project Director: Michael L. Boyle
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 18, 1994
Description of amendment request: The proposed change will revise
the current surveillance frequency which verifies area temperature
limits at least once per 12 hours. The revised surveillance requirement
will verify area temperature limits at least once per 7 days when the
data-logger alarm is operable, or at least once per 12 hours when the
data-logger alarm is inoperable.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
...The proposed change does not involve an SHC [significant
hazards consideration] because the change would not:
1. Involve a significant increase in the probability or
consequences of an accident previously evaluated.
The proposed change reduces the frequency at which area
temperature monitoring must be verified when the temperature data-
logger alarm function is operable. For conditions where the
temperature data-logger alarm function is inoperable, the frequency
at which normal ambient temperature is verified remains unchanged.
In addition, the proposed change does not affect any system,
equipment, or component credited in any previous accident
evaluation, any environmental qualification or post-accident
profiles. Therefore, the proposed change will not involve a
significant increase in the probability or consequences of an
accident previously evaluated.
2. Create the possibility of a new or different kind of accident
from any previously evaluated.
The proposed change does not alter or affect the design,
function, failure mode, or operation of the plant. There is no
change to the way in which the plant is operated and, therefore, no
increase in the probability of plant operation with any area
temperature outside of its limits. Therefore, this change does not
create the possibility of a new or different kind of accident from
those previously evaluated.
3. Involve a significant reduction in a margin of safety.
The proposed change does not challenge or affect the performance
of any of the protective boundaries, revise temperature limits in
the technical specifications, or perform any modifications that
would increase the likelihood of technical specification temperature
limits being exceeded. The proposed change requires the data-logger
alarm function to be operable in order to relax the surveillance
frequency. This alarm function provides continuous monitoring that
would detect temperature excursions prior to the current
surveillance which does not credit operability of the data-logger
alarm function. Also, the proposed change does not increase the
interval for which temperatures could exceed technical specification
limits. Therefore, the proposed change does not cause a reduction in
the margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: Learning Resource Center,
Three Rivers Community-Technical College, Thames Valley Campus, 574 New
London Turnpike, Norwich, Connecticut 06360.
Attorney for licensee: Gerald Garfield, Esquire, Day, Berry &
Howard, City Place, Hartford, Connecticut 06103-3499.
NRC Project Director: John F. Stolz
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: June 30, 1994
Description of amendment request: The proposed change revises the
Technical Specifications to change the trip setpoint for the 4kV bus
undervoltage relay (for the grid degraded voltage) from its current
value of [greater than or equal to] 3710 volts to its new setting of
[greater than or equal to] 3730 volts.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
...The proposed change does not involve an SHC [significant
hazards consideration] because the change does not:
1. Involve a significant increase in the probability or
consequences of an accident previously evaluated.
The proposed change involves the modification of the
undervoltage relay setpoint from 3710V to 3730V. The protection
provided by this system in unaffected and is still in accordance
with the guidance provided in NRC Branch Technical Position PSB-1.
This refinement increases the Technical Specification minimum trip
setpoint for the degraded voltage relays on the 4kV safety buses. It
does not detrimentally affect the safe operation of the plant, nor
does this proposed modification increase the probability or
consequences of an accident previously evaluated. The actual trip
setpoints of the subject relays do not require any changes and are
currently conservatively set at 3745V. The allowable value of
[greater than or equal to] 3706V remains unchanged. This slightly
higher than required setting was chosen by NNECO to provide added
margin should an undervoltage condition be present. This higher
setting will not cause more actuations of the ESF [engineered safety
feature] systems.
2. Create the possibility of a new or different kind of accident
from any accident previously evaluated.
The undervoltage protection system is provided to address the
concerns identified in NRC Branch Technical Position PSB-1 by
providing a scheme to detect the loss of offsite power at the class
1E buses, and a second level of undervoltage protection to protect
class 1E equipment. The change in the setpoint will not affect the
ability of this circuitry to detect a loss of offsite power or to
respond to an undervoltage condition.
Since the equipment will operate as previously described in the
FSAR [Final Safety Analysis Report], and there are no physical plant
modifications required (the current setting at the undervoltage
relay is 3745V), the proposed amendment will not create the
possibility of a new or different kind of accident from any accident
previously evaluated.
3. Involve a significant reduction in a margin of safety.
These relays do not cause a loss of offsite power, nor do they
cause a degraded voltage condition. These relays react to conditions
that have been placed upon the Plant. In the event that a degraded
voltage condition exists on the 4kV safety buses, alarms in the
control room alert the operators of this condition. In addition, the
Connecticut Valley Electric Exchange (CONVEX), the system dispatch
center for generation and VAR/voltage control, is aware of the
minimum voltage requirements for the three nuclear plants at the
Millstone station and has a minimum target switchyard voltage of
345kV. Under normal operation conditions the switchyard voltage
would have to degrade below 328kV before one of the 4kV safety buses
would start to enter the degraded voltage level and trip the
degraded voltage circuit. These administrative controls help
preclude a degraded voltage condition on the 4kV safety buses prior
to actuation of the degraded voltage protection circuits.
The proposed change of the 4kV degraded voltage minimum trip
setpoint to 3730V from 3710V will not result in any physical relay
setting change. The existing trip setting for the 4kV degraded
voltage relays have been conservatively set at 3745V, while the
existing allowable value remains unchanged at 3706V.
The response times or actuation logic of the degraded voltage
protection circuit remains unaffected, therefore revising the trip
setpoint value in the Technical Specifications will 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: Learning Resource Center,
Three Rivers Community-Technical College, Thames Valley Campus, 574 New
London Turnpike, Norwich, Connecticut 06360.
Attorney for licensee: Gerald Garfield, Esquire, Day, Berry &
Howard, City Place, Hartford, Connecticut 06103-3499.
NRC Project Director: John F. Stolz
Northern States Power Company, Docket Nos. 50-282 and 50-306,
Prairie Island Nuclear Generating Plant, Unit Nos. 1 and 2, Goodhue
County, Minnesota
Date of amendment requests: January 29, 1993, as revised June 15,
1994.
Description of amendment requests: The proposed amendments would
change core exit thermocouple action statements.
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 will not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
The purpose of the post accident monitoring equipment is to
display unit variables that provide information required by the
control room operators during accident situations and as such help
limit the consequences of an accident. The proposed changes, which
will allow continued plant operation with less than four core exit
thermocouples per core quadrant, have no impact on the probability
of an accident because they are only used in response to accident
situations.
Continued plant operation with the core exit thermocouple system
in the degraded condition as allowed by the proposed core exit
thermocouple action statements would not affect the operators
ability to monitor for inadequate core cooling following an
accident. At least two core exit thermocouples would be operable per
core quadrant, a minimum of four thermocouples would be available in
the center region of the core and at least one thermocouple would be
available in each quadrant of the outside core region. The smaller
size of the Prairie Island core, and therefore higher density of
thermocouples per unit of core area, provides additional assurance
that core exit temperatures can be adequately monitored with a
reduced number of core exit thermocouples.
Alternate means of monitoring for inadequate core cooling would
also be available. These include the reactor vessel water level
indication system, the subcooling margin monitors and wide range
reactor coolant system temperature.
The combination of the remaining operable core exit
thermocouples and the alternate monitoring capability will ensure
that the operators ability to identify inadequate core cooling in a
timely manner and take appropriate corrective action will not be
impaired, and therefore; the proposed changes will have no
significant impact on the consequences of an accident.
The core exit thermocouples perform no active role in the
mitigation of an accident. Their inoperability will not affect the
operability of any engineered safety features equipment or that
equipments ability to mitigate the consequences of an accident.
Therefore, for the reasons discussed above, the proposed changes
will not significantly affect the probability or consequences of an
accident previously evaluated.
2. The proposed amendment will not create the possibility of a
new or different kind of accident from any accident previously
analyzed.
There are no new failure modes or mechanisms associated with the
proposed changes. The proposed changes do not involve any
modification of plant equipment or any changes in operational
limits. The proposed changes only modify the requirements for
instrumentation used to monitor plant parameters during an accident.
The core exit thermocouples are passive monitoring devices, their
failure or inoperability cannot result in a plant accident of any
kind.
Therefore, for the reasons discussed above, the proposed changes
do not create the possibility of a new or different kind of accident
from any previously evaluated, and the accident analyses presented
in the Updated Safety Analysis Report will remain bounding.
3. The proposed amendment will not involve a significant
reduction in the margin of safety.
Continued plant operation with the core exit thermocouple system
in the degraded condition as allowed by the proposed core exit
thermocouple action statements would not affect the operators
ability to monitor for inadequate core cooling following an
accident. At least two core exit thermocouples would be operable per
core quadrant, a minimum of four thermocouples would be available in
the center region of the core and at least one thermocouple would be
available in each quadrant of the outside core region. The smaller
size of the Prairie Island core, and therefore higher density of
thermocouples per unit of core area, provides additional assurance
that core exit temperatures can be adequately monitored with a
reduced number of core exit thermocouples.
Alternative means of monitoring for inadequate core cooling
would also be available. These include the reactor vessel water
level indication system, the subcooling margin monitors and wide
range reactor coolant system temperature.
The combination of the remaining operable core exit
thermocouples and the alternate monitoring capability will ensure
that the operators ability to identify inadequate core cooling in a
timely manner and take appropriate corrective action will no be
impaired.
Therefore, for the reasons discussed above, the proposed changes
will not result in any reduction in the plant's 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: Minneapolis Public Library,
Technology and Science Department, 300 Nicollet Mall, Minneapolis,
Minnesota 55401
Attorney for licensee: Jay Silberg, Esq., Shaw, Pittman, Potts, and
Trowbridge, 2300 N Street, NW, Washington, DC 20037
NRC Project Director: L. B. Marsh
PECO Energy Company, Docket Nos. 50-352 and 50-353, Limerick
Generating Station, Units 1 and 2, Montgomery County, Pennsylvania
Date of amendment request: June 30, 1994
Description of amendment request: This amendment would relocate
selected recirculation and control rod block instrumentation setpoints
from Technical Specifications (TS) Table 3.3.6-2, and Section 3/4.4.1
to the Core Operating Limits Report (COLR), thereby revising TS Section
6.9.1.9 to document relocation of these items into the COLR.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. The proposed Technical Specifications (TS) changes do not
involve a significant increase in the probability or consequences of
an accident previously evaluated.
The TS change proposed is the relocation of the recirculation
pump Motor-Generator (MG) set mechanical and electrical stop and
control rod block recirculation flow upscale trip setpoint values to
the COLR. No physical plant equipment change is proposed. The TS
requirements for the setpoints and the associated surveillance
requirements remain unchanged. Only the location of the setpoint
values will be changed. The subject setpoint values will become
cycle depend[e]nt and will be determined by NRC approved methods, as
are the balance of setpoints and thermal limits found in the COLR.
However, the subject setpoint values are not modified as part of
this TS change.
Therefore, the proposed TS change does not involve an increase
in the probability or consequences of an accident previously
evaluated.
2. The proposed TS changes do not create the possibility of a
new or different kind of accident from any accident previously
evaluated.
The TS changes proposed are the relocation of the recirculation
pump MG set mechanical and electrical stop and control rod block
recirculation flow upscale trip setpoint values to the COLR. No
physical plant equipment change is part of the proposed TS changes.
The TS LCOs and surveillance requirements remain unchanged. The only
change proposed is the relocation of the subject setpoint values as
noted above. These setpoint values have been determined in
accordance with previously NRC approved methods and assure
sufficient operating margins in accordance with existing core design
methodology. Therefore, the proposed TS changes do not create the
possibility of a new or different kind of accident from any
previously evaluated.
3. The proposed TS changes do not involve a significant
reduction in a margin of safety.
The following TS BASES were reviewed for potential reduction in
the margin of safety:
3/4.2 Power Distribution Limits
3/4.3.6 Control Rod Block Instrumentation
3/4.4.1 Recirculation System
The margin of safety, as defined in the TS BASES, will not be
reduced. The proposed TS changes do not affect existing accident
analyses or design assumptions, nor do they impact any safety limits
of the plant, since they are administrative in nature.
Therefore, the proposed TS changes do not involve a reduction in
a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: Pottstown Public Library, 500
High Street, Pottstown, Pennsylvania 19464.
Attorney for licensee: J. W. Durham, Sr., Esquire, Sr. V. P. and
General Counsel, PECO Energy Company, 2301 Market Street, Philadelphia,
Pennsylvania 19101
NRC Project Director: Charles L. Miller
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: June 23, 1994
Description of amendment request: This amendment will change the
Technical Specification 4.0.5 for each unit to reflect NRC's policy
with respect to relief requests for the inservice inspection programs.
Specifically, the change would clarify the fact that relief requests
for impracticable testing or surveillance requirements can be
implemented prior to the Commission approval of such requests.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration which is presented below:
1. Involve a significant increase in the probability or
consequences of an accident previously evaluated.
The proposed changes are administrative in nature in that the
changes eliminate any possibility of misinterpretation of the ASME
Code requirements that allow for a utility to submit relief requests
to the Commission within one year and allows for the implementation
of these request[s] prior to Commission review and approval. The
relief requests are based on and provide for alternative testing
based on industry practice that provides an equivalent level of
quality and safety as the Code requirement. The Commission will
still provide acceptance of the relief requests in writing.
Therefore, it can be concluded that the proposed changes do not
involve a significant increase in the probability or consequences of
an accident previously evaluated.
2. Create the possibility or a new or different kind of accident
from any accident previously evaluated.
No new failure modes have been defined for any plant system or
component important to safety nor has any new limiting failure been
identified as a result of the proposed changes. Therefore, it can be
concluded that the proposed changes do not create the possibility of
a new or different kind of accident from those previously evaluated.
3. Involve a significant reduction in a margin of safety.
The proposed changes are administrative in nature and do not
adversely impact the plant's ability to meet applicable regulatory
requirements related to inservice testing or inspection. The
proposed changes eliminate any possible misinterpretation of the
Code requirements regarding relief requests and do not reduce the
protection of public health and safety. Therefore, it can be
concluded that the proposed changes do not involve a significant
reduction in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: 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: Charles L. Miller
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: June 13, 1994
Description of amendment request: The proposed amendments would
permit an out-of-service component to be returned to service under
administrative controls for the purpose of determining operability. The
proposed change is consistent with the method utilized in the new
Westinghouse Standard Technical Specifications (NUREG-1431). In
addition, the proposed amendment corrects a typographical error in the
header information on Page 3/4 0-2 of the current technical
specifications.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. Will not involve a significant increase in the probability or
consequences of an accident previously evaluated.
a. Header Information
This editorial change corrects a typographical error only. As
such, existing accident analyses are unaffected.
b. Specification 3.0.6
The proposed change merely clarifies the intent of Specification
3.0.2. As such, existing accident analyses are unaffected.
2. Will not create the possibility of a new or different kind of
accident from any accident previously evaluated.
a.Header Information
This editorial change corrects a typographical error only. As
such, it does not alter the function of any plant equipment, involve
any design changes, nor does it create any new operating modes or
accident scenarios.
b. Specification 3.0.6
The proposed change merely clarifies the intent of Specification
3.0.2. As such, it does not alter the function of any plant
equipment, involve any design changes, nor does it create any new
operating modes or accident scenarios.
3. Will not involve a significant reduction in a margin of
safety.
a. Header Information
This editorial change corrects a typographical error only. As
such, the present margins of safety are unaffected.
b. Specification 3.0.6
The proposed change merely clarifies the intent of Specification
3.0.2. As such, the present margins of safety are unaffected.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: Salem Free Public library, 112
West Broadway, Salem, New Jersey 08079
Attorney for licensee: Mark J. Wetterhahn, Esquire, Winston and
Strawn, 1400 L Street, NW., Washington, DC 20005-3502
NRC Project Director: Charles L. Miller
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: June 17, 1994 as supplemented July 13,
1994
Description of amendment request: The proposed amendment would
change the requirement to perform the Channel Functional Test of the
Power Operated Relief Valve (PORV) position indication from quarterly
to every 18 months and to exempt the PORV Block Valve position
indication from performance of the channel Function Test if the PORV
Block Valve is shut as required to isolate a PORV that cannot be
manually cycled.
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 not involve a significant increase in the probability or
consequences of an accident previously evaluated.
A change from quarterly (Q) to at least every 18 months (R) may
appear to be non-conservative, at first; however, by extending the
surveillance requirement to cycle the PORV during non-power
conditions, the change eliminates the potential risk and
consequences of having the valve sticking open at power or not fully
closing (leaking). Therefore, by extending the surveillance the
probability and consequences of any previously analyzed accident is
reduced, since the testing would now be conducted in a non-power
condition, and the margin to safety is increased. Consequently, a
net safety gain is realized by eliminating or minimizing these
risks.
The added note for PORV block valve is included for consistency
and alignment between the surveillance requirement under this T/S
(Table 4.3-11) with that of T/S surveillances 4.4.3.2 and 4.4.5.2
for Units 1 and 2 respectively.
Therefore, the proposed amendment does not involve a physical or
procedural change to any structure, component, or system that
significantly affects accident/malfunction probabilities or
consequences previously evaluated in the UFSAR.
2. Does not create the possibility of a new or different kind of
accident from any accident previously evaluated.
The proposed changes do not introduce any design or physical
configuration changes to the facility which could create new
accident scenarios.
3. Does not involve a significant reduction in a margin of
safety.
As stated in response to question number 1 above, the proposed
changes do not eliminate the required T/S surveillance requirements.
The first change eliminates the need to cycle the PORV valves
through one complete cycle of full travel at power. The second
change allows for not having to perform a surveillance on a valve
that it is being used as an isolation point. The valve has been
closed to comply with requirements of another T/S action statement.
Therefore, the probability and consequences of any previously
analyzed accident is reduced, thus increasing the margin to safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: Salem Free Public library, 112
West Broadway, Salem, New Jersey 08079
Attorney for licensee: Mark J. Wetterhahn, Esquire, Winston and
Strawn, 1400 L Street, NW., Washington, DC 20005-3502
NRC Project Director: Charles L. Miller
Tennessee Valley Authority, Docket No. 50-296, Browns Ferry Nuclear
Plant, Unit 3, Limestone County, Alabama
Date of amendment request: March 29,1994 (TS 340)
Description of amendment request: The proposed amendment to the
Unit 3 Technical Specifications adds a limiting condition for operation
and a surveillance requirement for a load shedding logic being added by
a design change to Unit 3. The load shedding logic is being added to
ensure that the maximum capacity of the Unit 3 Emergency Diesel
Generators is not exceeded during a postulated loss of offsite power
event concurrent with a design basis accident.
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.
This proposed change establishes a surveillance testing
requirement and limiting condition for operation for the Unit 3 480-
volt load shedding logic system. This Technical Specification change
will not introduce any new failure mode and will not alter any
assumptions previously made in evaluating the consequences of an
accident. Accordingly, this change does not affect any design
limiting safety system settings or operating parameters.
Furthermore, the change does not modify or add any accident
initiating events or parameters. Therefore, these proposed changes
do not involve an increase in the probability or consequences of an
accident previously evaluated.
2. The proposed amendment does not create the possibility of a
new or different kind of accident from any accident previously
evaluated.
This proposed change establishes a limiting condition for
operation and a surveillance requirement for the Unit 3 480-volt
load shedding logic system. The addition of a limiting condition for
operation and surveillance requirement will not adversely affect the
operation of Unit 3 or the manner in which it is operated.
Furthermore, the change does not create a failure mode that can lead
to an accident of a different type than previously evaluated.
Therefore, the proposed change does not create the possibility of a
new or different kind of accident from any accident previously
evaluated.
3. The proposed amendment does not involve a significant
reduction in a margin of safety.
The addition of a limiting condition for operation and
surveillance requirement will not reduce the margin of safety. The
testing of the 480-volt load shedding logic on an 18-month interval
is consistent with BWR/4 (NUREG-1433) Standard Technical
Specifications. These are based on the guidance set forth in NRC
Regulatory Guide 1.108, ``Periodic Testing of Diesel Generator Units
Used as Onsite Electric Power Systems at Nuclear Power Plants.'' The
addition of a limiting condition for operation establishes a minimum
acceptable level of performance for the 480-volt load shedding logic
system. Thus, the ability of the Emergency Diesel Generators to
supply power during a loss of offsite power coincident with a design
basis accident is assured.
Furthermore, no reductions in the requirements or setpoints of
the equipment supplied by the Emergency Diesel Generators are made
which could result in a reduction in the margin of safety.
Therefore, this proposed change does not involve a reduction in the
margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: Athens Public Library, South
Street, Athens, Alabama 35611
Attorney for licensee: General Counsel, Tennessee Valley Authority,
400 West Summit Hill Drive, ET llH, Knoxville, Tennessee 37902
NRC Project Director: Frederick J. Hebdon
TU Electric Company, Docket Nos. 50-445 and 50-446, Comanche Peak
Steam Electric Station, Units 1 and 2, Somervell County, Texas
Date of amendment request: February 23, 1994
Brief description of amendments: The proposed changes would revise
the Comanche Peak Steam Electric Station (CPSES), Units 1 and 2,
Technical Specifications (TS) to (1) allow a one-hour allowed outage
time (AOT) following discovery of a closed cold leg injection
accumulator discharge isolation in Modes 1, 2, or 3; (2) eliminate the
redundant requirement to reverify accumulator boron concentration
following fill from the refueling water storage tank (RWST); (3)
relocate the accumulator water level and pressure channel analog
channel operational test (ACOT) and channel calibration from the CPSES
Technical Specifications to an administratively controlled program; (4)
change the accumulator limits to analysis values rather than indicated
values; and (5) reduce the inspection frequency following containment
entries.
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 a previously evaluated accident.
The current requirement to immediately open a cold leg
accumulator discharge isolation valve (or shut down the unit) upon
discovery that the valve is closed is modified by the requested
change to provide a one hour allowed outage time (AOT) prior to
requiring a unit shutdown. This change is consistent with NUREG-
1431. The currently required action is more restrictive than that
required by CPSES Technical Specification 3.0.3, that specifies the
action required if an LCO [Limiting Condition for Operation] and its
associated action requirements are not met and which provides a one
hour AOT prior to taking steps to place the plant in Mode 3 within
the following 6 hours. Following this requested change, the required
actions for an accumulator declared inoperable due to a closed
discharge isolation valve will be identical to those actions
required for inoperability for other reasons, with the exception of
the accumulator boron concentration being out of specification that
has an AOT of 72 hours. Changing the AOT from ``immediate'' to one
hour does not affect the probability of an accident. The only
previously evaluated accident that is potentially impacted is the
Loss of Coolant Accident (LOCA). With all valves open and thus all
accumulators available, a potential LOCA is bounded by the existing
accident analyses. With one accumulator discharge isolation valve
closed and thus one accumulator not available, the consequences of a
LOCA could be more severe; however, this requested amendment does
not create this scenario. In other words, although the change in AOT
may slightly increase the probability that, were a LOCA to occur, an
accumulator would not be available (see the response to [number] 3
below), it does not involve a significant increase in the
consequences of an accident previously evaluated.
The requirement to test the accumulator boron concentration
following a 101 gallon or greater solution volume increase is
modified by the requested change to exclude volume additions from
the Refueling Water Storage Tank (RWST). Since the RWST boron
concentration must be confirmed to satisfy the limits for the
accumulators, there is no impact on the probability or consequences
of any accident.
The relocation of the accumulator water level and pressure
channel ACOT and Channel Calibration from CPSES Technical
Specifications to an administratively controlled program is
essentially an administrative change. Because proper tests will
still be performed, there is no impact on the probability or
consequences of any accident.
The requested change to reduce the containment debris
inspections from ``at the completion of every entry'' to ``[a]t
least once daily'' will require fewer inspections and is consistent
with SR [Surveillance Requirement] 4.6.1.3 for the containment air
locks. The accident of concern is a LOCA and these inspections have
no impact on the probability of a LOCA. Performing fewer inspections
would slightly increase the possibility that, should a LOCA occur,
there could be debris in containment which could be transported to
and partially clog the containment sump. However, inspecting at
least daily if containment entries have been made is adequate and is
justified by the reduced total radiation exposure for plant
personnel. The inspections conducted at least daily assures that
there is not a significant increase in the consequence of any
accident.
The requested changes do not modify the existing LCOs for
Technical Specifications 3.5.1 and 3.5.2 with the exception of the
replacement of ``indicated'' values with analysis values in LCO
3.5.1, consistent with the relocation of the SRs for accumulator
instrumentation. The requested changes are consistent with NUREG-
1431 and GL 93-05, and, as such, have already been generically
assessed by the NRC. It is concluded that the requested changes do
not involve a significant increase in the probability or
consequences of an accident previously evaluated.
2. The proposed changes do not create the possibility of a new
or different kind of accident from any previously evaluated
accident.
The only requested change that modifies current operation of the
plant is the requested one hour alowed outage time for action
following discovery of a closed cold leg accumulator discharge
isolation valve. The requested one hour completion time to open the
valve continues to ensure that prompt action will be taken to return
the inoperable accumulator to an operable status, minimizing the
potential for exposure of the plant to a LOCA under this condition.
In addition, as LCO 3.5.1a will continue to require that the
accumulator discharge isolation valve be open with power removed
from the valve operator, the probability of the discharge isolation
valve being closed in Modes 1, 2, or 3 will remain low. This change
in current operation does not create the possibility of a new or
different kind of accident.
The requested slight reduction in the containment inspection
frequencies specified in SR 4.5.2 only serves to reduce the number
of unnecessary inspections. It does not make substantial changes to
the inspection requirements, nor does it change the method of
performing these requirements. Thus, the requested change does not
create the possibility of a new or different kind of accident.
No significant changes to the limiting conditions for operation
of the accumulators or the emergency core cooling system are
requested as part of this amendment request. The requested changes
do not involve any physical changes to the plant. The requested
changes are consistent with NUREG-1431 and GL 93-05, and, as such,
have already been generically assessed by the NRC. Thus, the
requested changes do not create the possibility of a new or
different kind of accident from any accident previously evaluated.
3. The proposed changes do not involve a significant reduction
in the margin of safety.
The only requested change that modifies current operation of the
plant is the requested one hour allowed outage time for action
following discovery of a closed cold leg accumulator discharge
isolation valve. As noted in the response to [number] 1 above, this
requested change in AOT does not significantly affect the
probability or consequences of an accident, but does increase the
possibility that, should a LOCA occur, one of the accumulators may
not be available to help mitigate the consequences of the accident.
However, the requested one hour completion time to open the valve
continues to ensure that prompt action will be taken to return the
inoperable accumulator to an operable status, minimizing the
potential for exposure of the plant to a LOCA under this condition.
In addition, as LCO 3.5.1a will continue to require that the
accumulator discharge isolation valve be open with power removed
from the valve operator, the probability of the discharge isolation
valve being closed in Modes 1, 2, or 3 will remain low. Considering
the controls above and the fact that the requested action statement
is consistent with TS 3.0.3, it is concluded that the requested
change does not involve a significant reduction in the margin of
safety.
The requested slight reduction in the containment inspection
frequencies specified in SR 4.5.2 only serves to reduce the number
of unnecessary inspections conducted and reduce the personnel
exposure associated with the inspections. As adequate inspections
will continue to be conducted, this requested change does not
involve a significant reduction in a margin of safety.
No significant changes to the limiting conditions for operation
of the accumulators or the emergency core cooling systems are
requested as part of this amendment request. The requested changes
are consistent with NUREG-1366, NUREG-1431 and GL 93-05, and, as
such, have already been generically assessed by the NRC. Thus, it is
concluded that the requested 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: University of Texas at
Arlington Library, Government Publications/Maps, 701 South Cooper, P.O.
Box 19497, Arlington, Texas 76019
Attorney for licensee: George L. Edgar, Esq., Newman and
Holtzinger, 1615 L Street, N.W., Suite 1000, Washington, D.C. 20036
NRC Project Director: William D. Beckner
TU Electric Company, Docket Nos. 50-445 and 50-446, Comanche Peak
Steam Electric Station, Units 1 and 2, Somervell County, Texas
Date of amendment request: April 22, 1994
Brief description of amendments: The proposed amendments would
revise the technical specifications by changing the frequency of
auxiliary feedwater pump operational testing from monthly to quarterly.
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 or a previously evaluated
accident.
Because the Auxiliary Feedwater System pumps are provided to
mitigate certain accidents, altering the test frequency of the pumps
will not impact the probability of an accident. The Auxiliary
Feedwater System pumps will continue to be tested quarterly on a
staggered basis to the same standards applied to safety-related
pumps as defined by ASME Section XI. Satisfactory completion of the
testing in accordance with the Code is used as verification that
safety-related pumps will be available to perform their intended
function. Quarterly testing of the Auxiliary Feedwater System pumps
on a staggered basis, therefore, will continue to assure that the
Auxiliary Feedwater System will be capable of performing its
intended function. It is thus concluded that the requested change
will not involve a significant increase in the probability of
occurrence or consequences of an accident previously evaluated.
2. The proposed changes do not create the possibility of a new
or different kind of accident from any accident previously
evaluated.
Changing the surveillance test frequency of the Auxiliary
Feedwater Pumps does not involve any physical modification of the
plant or result in a change in a method of operation. Therefore, it
is concluded that the requested change does not create the
possibility of a new or different kind of accident from any accident
previously evaluated.
3. The proposed changes do not involve a significant reduction
in a margin of safety.
Changing the surveillance testing frequency of the Auxiliary
Feedwater System pumps does not affect any safety limits or any
limiting safety system settings. System operating parameters are
unaffected. The availability of equipment required to mitigate or
assess the consequence of an accident is not reduced; in fact the
availability is increased because the system is rendered inoperable
on a quarterly basis to perform pump testing, rather than a monthly
basis. Further, vibration testing being the most effective early
indication of gradual pump degradation continues to be performed on
the same frequency. Quarterly testing of the Auxiliary Feedwater
pumps on a staggered basis in accordance with the criteria specified
in the ASME Section XI code provides adequate assurance that the
Auxiliary Feedwater System pumps are capable of performing their
intended function. Thus, its [sic] is concluded that the requested
change does not involved [sic] 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 Texas at
Arlington Library, Government Publications/Maps, 701 South Cooper, P.O.
Box 19497, Arlington, Texas 76019
Attorney for licensee: George L. Edgar, Esq., Newman and
Holtzinger, 1615 L Street, N.W., Suite 1000, Washington, D.C. 20036
NRC Project Director: William D. Beckner
TU Electric Company, Docket Nos. 50-445 and 50-446, Comanche Peak
Steam Electric Station, Units 1 and 2, Somervell County, Texas
Date of amendment request: April 25, 1994
Brief description of amendments: The proposed amendments would
revise the technical specifications to reduce the number of fast starts
currently required by surveillance requirements for the emergency
diesel generators (EDGs).
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 a previously evaluated accident.
This change reduces the number of ``fast starts'' required on
the EDGs and allows the EDGs to be tested using ``slow starts.''
Reducing the number of ``fast starts'' (required to start in 10
seconds or less) will reduce the wear on the EDGs primarily by
minimizing coking of fuel in the cylinder and preventing premature
wearing of the turbocharger thrust bearing. This increases engine
reliability and availability. A ``slow start'' may require that the
EDG be taken out of service to perform the test if the EDG start
time is not 10 seconds or less. TU Electric feels that testing the
``fast start'' capability of the EDG every 184 days will maintain
its present level of reliability. The period of time in which the
EDG is actually inoperable due to testing (i.e., may not start and
be ready to load in 10 seconds) is quite short. Overall the
reliability and the availability of the EDG will be increased.
The impact of the EDGs on the postulated accidents is directly
related to their reliability and availability. Therefore, the
proposed reduction in the number of ``fast starts'' does not involve
a significant increase in the probability or 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 previously evaluated
accident.
The revised testing allowed by this Technical Specification
change does not create a new or different kind of accident. The EDGs
are primarily accident mitigation components. The potential failure
of EDGs have already been assessed in the CPSES design.
3. The proposed changes do not involve a significant reduction
in the margin of safety.
The only aspect of this change that could adversely affect the
margin of safety is the potential impact on the start time of the
engine. The start time of the engine is not expected to exceed the
assumption in the accident analyses except possibly in the short
period of time required to perform the ``slow start'' test. If the
EDG does not start in 10 seconds or less under these conditions, the
EDG is declared inoperable, as allowed by the Technical
Specifications, to perform the ``slow start'' test. Because these
periods of inoperability are only implemented as allowed by the
Technical Specifications, there is no impact on the margin of
safety. The margin of safety established by the assumed EDG
availability will be enhanced by the increased reliability and
availability of the EDGs.
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 Texas at
Arlington Library, Government Publications/Maps, 701 South Cooper, P.O.
Box 19497, Arlington, Texas 76019
Attorney for licensee: George L. Edgar, Esq., Newman and
Holtzinger, 1615 L Street, N.W., Suite 1000, Washington, D.C. 20036
NRC Project Director: William D. Beckner
TU Electric Company, Docket Nos. 50-445 and 50-446, Comanche Peak
Steam Electric Station, Units 1 and 2, Somervell County, Texas
Date of amendment request: April 25, 1994
Brief description of amendments: The proposed amendments would
revise the technical specifications by updating the unit staff
qualification requirements to Regulatory Guide 1.8, Revision 2,
``Qualification and Training of Personnel for Nuclear Power Plants,''
and by relocating administrative control of training from the technical
specifications to the Final Safety Analysis Report.
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 a previously evaluated accident.
This proposed amendment involves a consolidation of previous
unit staff qualification requirements into one source document, and
a relocation of training requirements to a more appropriate license
basis document. The qualification requirements of the unit staff
remains the same as the existing requirements, and the relocation of
the training requirements does not change the scope of the program
as it now exists. The relocated training program requirements retain
adequate administrative and regulatory controls to ensure the plant
is not placed in an unanalyzed condition.
These changes are administrative in nature. They remain within
the assumptions of the current accident analysis. As a result, they
do not increase the probability or consequences of an accident
previously evaluated.
2. The proposed changes do not create the possibility of a new
or different kind of accident from any previously evaluated
accident.
These changes are administrative in nature. They merely
consolidate qualification requirements and relocate training
requirements. The relocated program requirements retain adequate
administrative and regulatory controls to ensure the plant is not
introduced to an unreviewed safety question.
These changes are administrative in nature. They do not
introduce any new initiating events. As a result, they do not create
a new or different kind of accident from any accident previously
evaluated.
3. The proposed changes do not involve a signficant reduction in
a margin of safety.
These changes are administrative in nature and have no impact on
actual plant protection or safety actuation systems, or the assumed
actions performed in accordance with normal, abnormal, or emergency
operating procedures. There are adequate regulatory and plant
configuration controls existing to ensure there is no impact on the
plant 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 Texas at
Arlington Library, Government Publications/Maps, 701 South Cooper, P.O.
Box 19497, Arlington, Texas 76019
Attorney for licensee: George L. Edgar, Esq., Newman and
Holtzinger, 1615 L Street, N.W., Suite 1000, Washington, D.C. 20036
NRC Project Director: William D. Beckner
Wisconsin Public Service Corporation, Docket No. 50-305, Kewaunee
Nuclear Power Plant, Kewaunee County, Wisconsin
Date of amendment request: April 21, 1994
Description of amendment request: The proposed amendment would
revise Kewaunee Nuclear Power Plant (KNPP) Technical Specification (TS)
by adding requirements for the steam exclusion system. The new TS
addresses the steam exclusion system (TS 3.15) and the surveillance on
the steam exclusion system (TS 4.15). This system was not previously
addressed by the 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:
A) TS 3.15 Steam Exclusion System (New)
The proposed change was reviewed in accordance with the
provisions of 10 CFR 50.92 to show no significant hazards exist. The
proposed change will not:
1) Involve a significant increase in the probability or
consequences of an accident previously evaluated.
The intent of this new specification is to specify the
operability requirements for the Steam Exclusion System and to
demonstrate the acceptability of removing the Steam Exclusion System
from service for short periods of time.
The proposed change will not significantly increase the
probability of an accident previously evaluated. The accident under
consideration is a high energy line break outside of containment.
Allowing a steam exclusion boundary to be inoperable for a short
period of time has no effect on the probability of occurrence of a
high energy line break outside of containment.
The proposed change will not significantly increase the
consequences of an accident previously evaluated. Again, the
accident under consideration is a high energy line break outside of
containment. Calculations conclude that the core damage frequency
for a high energy line break outside of containment with a non-
redundant steam exclusion boundary open is 2.57E-8 per 12-hour
period. Further conservative assumptions of one non-redundant steam
exclusion boundary being open 12 hours per day, 5 days per week, 52
weeks per year results in a core damage frequency of 6.68E-6 per
year. This analysis was conservatively calculated taking minimal
credit for mitigating the accident, and is considered to be an
acceptable level of risk on an annual basis. A safety factor of five
was applied to NUREG/CR-4550 data to determine the initiating event
frequency of a high energy line break. This calculation supports the
conclusion that this addition to the Technical Specifications will
not result in a significant increase in the probability or
consequences of an high energy line break outside of containment.
Furthermore, calculations conclude that the core damage
frequency for a high energy line break outside of containment with
one of two redundant steam exclusion boundaries open is 4.62E-10 per
72-hour period. Further conservative assumptions of one redundant
steam exclusion damper being open 24 hours per day, 5 days per week,
52 weeks per year results in a core damage frequency of 4.00E-8 per
year. This analysis was conservatively calculated taking minimal
credit for mitigating the accident, and is also considered to be an
acceptable level of risk on an annual basis. Again, a safety factor
of five was applied to NUREG/CR-4550 data to determine the
initiating event frequency of a high energy line break. This
calculation also supports the conclusion that this addition to the
Technical Specifications will not result in a significant increase
in the probability or consequences of an high energy line break
outside of containment.
Specific requirements for the Steam Exclusion System do not
currently exist in the Technical Specifications. Addition of TS 3.15
is an enhancement to the Kewaunee Technical Specifications, and
providing this information for the plant staff and operators will
not significantly increase the probability or consequences of an
accident previously evaluated, nor will it adversely affect the
health and safety of the public.
2) Create the possibility of a new or different kind of accident
from any accident previously evaluated.
The proposed amendment does not alter the plant configuration,
operating setpoints or overall plant performance. Therefore, it
cannot create the possibility of a new or different kind of accident
from any accident previously evaluated.
3) Involve a significant reduction in the margin of safety.
Addition of the specification is an enhancement to the Technical
Specifications and does not alter input to the safety analysis.
Furthermore, the supporting analysis demonstrates an acceptable
level of risk for removing components from service for limited
periods of time. Therefore, it will not involve a significant
reduction in the margin of safety.
Additionally, the proposed change is similar to example
C.2.e(ii) in 51 FR 7751. Example C.2.e(ii) states that changes that
constitute an additional limitation, restriction or control not
presently included in the TS's are not likely to involve a
significant hazard.
B) S 4.15 Steam Exclusion System (New)
The proposed change was reviewed in accordance with the
provisions of 10 CFR 50.92 to show no significant hazards exist. The
proposed change will not:
1) Involve a significant increase in the probability or
consequences of an accident previously evaluated.
The proposed change will not significantly increase the
probability or consequences of an accident previously evaluated. The
accident under consideration is a high energy line break outside of
containment. The performance of periodic surveillance requirements,
testing which verifies that components in the Steam Exclusion System
are operating properly, cannot significantly increase the
probability or consequences of a high energy line break, nor will it
adversely affect the health and safety of the public.
2) Create the possibility of a new or different kind of accident
from any accident previously evaluated.
The proposed amendment does not alter the plant configuration,
operating setpoints or overall plant performance. Therefore, it
cannot create the possibility of a new or different kind of accident
from any accident previously evaluated.
3) Involve a significant reduction in the margin of safety.
Addition of the specification is an enhancement to the Kewaunee
Technical Specifications and does not alter input to the safety
analysis. Therefore, it will not involve a significant reduction in
the margin of safety.
Additionally, the proposed change is similar to example
C.2.e(ii) in 51 FR 7751. Example C.2.e(ii) states that changes that
constitute an additional limitation, restriction or control not
presently included in the TS's are not likely to involve a
significant hazard.
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 Wisconsin
Library Learning Center, 2420 Nicolet Drive, Green Bay, Wisconsin
54301.
Attorney for licensee: Bradley D. Jackson, Esq., Foley and Lardner,
P. O. Box 1497, Madison, Wisconsin 53701-1497.
NRC Project Director: John N. Hannon
Wisconsin Public Service Corporation, Docket No. 50-305, Kewaunee
Nuclear Power Plant, Kewaunee County, Wisconsin
Date of amendment request: May 26, 1994
Description of amendment request: The proposed amendment would
revise Kewaunee Nuclear Power Plant (KNPP) Technical Specification (TS)
Sections 2.3, 3.6, and 4.6 by correcting minor typographical errors and
format inconsistencies. These changes are being proposed as a part of
the licensee's ongoing effort to revise each section of the KNPP TS to
achieve a consistent format and to convert the entire document to Word
Perfect. In addition, changes to the basis for TS Sections 2.3, 3.6,
and 4.6 have also been proposed.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration which is presented below:
The proposed changes were reviewed in accordance with the
provisions of 10 CFR 50.92 to show no significant hazards exist. The
proposed changes will 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.
3) involve a significant reduction in the margin of safety.These
proposed changes involve the conversion of the TS to the Word
Perfect format now being used at WPSC. Minor typographical errors
and format inconsistencies were corrected. These proposed changes
are administrative in nature; accordingly, these proposed changes do
not involve a significant hazards consideration. Additionally, the
proposed changes are similar to example C.2.e.(i) in 51 FR 7751.
Example C.2.e.(i) states that changes which are purely
administrative in nature; i.e., to achieve consistency throughout
the Technical Specifications, correct an error, or a change in
nomenclature, are not likely to involve a significant hazard.
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 Wisconsin
Library Learning Center, 2420 Nicolet Drive, Green Bay, Wisconsin
54301.
Attorney for licensee: Bradley D. Jackson, Esq., Foley and Lardner,
P. O. Box 1497, Madison, Wisconsin 53701-1497.
NRC Project Director: John N. Hannon
Previously Published Notices Of Consideration Of Issuance Of
Amendments To Facility Operating Licenses, Proposed No Significant
Hazards Consideration Determination, And Opportunity For A Hearing
The following notices were previously published as separate
individual notices. The notice content was the same as above. They were
published as individual notices either because time did not allow the
Commission to wait for this biweekly notice or because the action
involved exigent circumstances. They are repeated here because the
biweekly notice lists all amendments issued or proposed to be issued
involving no significant hazards consideration.
For details, see the individual notice in the Federal Register on
the day and page cited. This notice does not extend the notice period
of the original notice.
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 application for amendment: January 4, 1994
Brief description of amendment request: The proposed amendment
would implement recommended changes from Generic Letter (GL) 93-05,
``Line-Item Technical Specification Improvements to Reduce Surveillance
Requirements for Testing During Power Operation.'' Specifically, the
licensee proposed to change their Technical Specifications
corresponding to the following GL 93-05 line numbers: 4.1.2, 5.8, 5.14,
6.1, 7.5, 8.1, 9.1, 12, and 14. Date of individual notice in Federal
Register: July 22, 1994 (59 FR 37513)
Expiration of individual notice: August 22, 1994
Local Public Document Room location: Phoenix Public Library, 12
East McDowell Road, Phoenix, Arizona 85004
Arizona Public Service Company, et al., Docket No. STN 50-529, Palo
Verde Nuclear Generating Station, Unit 2, Maricopa County, Arizona
Date of application for amendment: July 1, 1994
Brief description of amendment request: The proposed amendment
would modify Technical Specification (TS) Figure 3.2-1, ``Reactor
Coolant Cold Leg Temperature vs. Core Power Level.'' Specifically, the
minimum cold leg temperature for core power levels between 90 percent
and 100 percent would be changed to 552 deg.F (which is a reduction of
10 deg.F from the previous TS requirement). This TS change permits
reactor operation at full power with a lower reactor coolant
temperature to minimize potential steam generator tube degradation.
Date of individual notice in Federal Register: July 13, 1994 (59 FR
35767)
Expiration of individual notice: August 12, 1994
Local Public Document Room location: Phoenix Public Library, 12
East McDowell Road, Phoenix, Arizona 85004
Commonwealth Edison Company, Docket No. STN 50-456, Braidwood
Station, Unit 1, Will County, Illinois
Date of amendment request: June 20, 1994
Description of amendment request: The proposed amendment would
revise the Braidwood, Unit 1, Technical Specifications (TSs) to remove
the condition limiting operation of the facility to 100 days during the
present fuel cycle when Thot is greater than 500 deg.F and to
restore the reactor coolant dose equivalent Iodine-131 limit to 1
microcurie per gram of coolant from the present value of 0.35. Both the
limit on permissible operational time and the reduction in the
permissible level of Iodine-131 were incorporated into the TSs by
Amendment No. 50 issued to
Facility Operating License No. NPF-72 for Braidwood Station, Unit
1, on May 7, 1994.
Date of publication of individual notice in Federal Register: July
11, 1994 (59 FR 35389)
Expiration of individual notice: August 10, 1994
Local Public Document Room location: Wilmington Township Public
Library, 201 S. Kankakee Street, Wilmington, Illinois 60481.
Commonwealth Edison Company, Docket Nos. 50-295 and 50-304, Zion
Nuclear Power Station, Units 1 and 2, Lake County, Illinois
Date of amendment request: June 16, 1994
Description of amendment request: The proposed amendments would
consist primarily of an administrative change to the Zion Station's
Technical Specifications (TSs) to reflect an exemption to 10 CFR Part
50, Appendix J, Section III.D.3.
Date of publication of individual notice in Federal Register: June
30, 1994 (59 FR 33798)
Expiration of individual notice: August 1, 1994
Local Public Document Room location: Waukegan Public Library, 128
N. County Street, Waukegan, Illinois 60085
Northeast Nuclear Energy Company, et al., Docket No. 50-336,
Millstone Nuclear Power Station, Unit No. 2, New London County,
Connecticut
Date of amendment request: June 24, 1994
Description of amendment request: The proposed amendment would
revise the Technical Specifications (TS) to change the Administrative
Controls section to require an individual who serves as the Operations
Manager to either hold a Millstone Unit 2 senior reactor operator (SRO)
license or have an SRO license at another pressurized water reactor. If
the Operations Manager does not hold a Millstone Unit 2 SRO license,
then an individual serving as the Assistant Operations Manager would be
required to possess an SRO license at Millstone Unit 2. Date of
publication individual notice in Federal Register: July 7, 1994 (59 FR
34872).
Expiration of individual notice: August 8, 1994
Local Public Document Room location: Learning Resource Center,
Three Rivers Community-Technical College, Thames Valley Campus, 574 New
London Turnpike, Norwich, Connecticut 06360.
Southern Nuclear Operating Company, Inc., Docket No. 50-348,Joseph
M. Farley Nuclear Plant, Unit 1
Date of amendment request: June 17, 1994
Brief description of amendment request: The amendment changes the
Technical Specifications to revise the nuclear enthalpy rise hot
channel factor (F delta H) from equal to or less than 1.65 [1 plus
0.3(1-P)] to equal to or less than l.70 [1 plus 0.3(1-P)] where P is a
fraction of rated power. The amendment also revises the action
statement to reflect guidance contained in the improved standard
technical specifications.
Date of publication of individual notice in Federal Register: June
22, 1994 (59 FR 32249)
Expiration of individual notice: July 22, 1994
Local Public Document Room location: Houston-Love Memorial Library,
212 W. Burdeshaw Street, Post Office Box 1369, Dothan, Alabama 36302
Notice Of Issuance Of Amendments To Facility Opersting Licenses
During the period since publication of the last biweekly notice,
the Commission has issued the following amendments. The Commission has
determined for each of these amendments that the application complies
with the standards and requirements of the Atomic Energy Act of 1954,
as amended (the Act), and the Commission's rules and regulations. The
Commission has made appropriate findings as required by the Act and the
Commission's rules and regulations in 10 CFR Chapter I, which are set
forth in the license amendment.
Notice of Consideration of Issuance of Amendment to Facility
Operating License, Proposed No Significant Hazards Consideration
Determination, and Opportunity for A Hearing in connection with these
actions was published in the Federal Register as indicated.
Unless otherwise indicated, the Commission has determined that
these amendments satisfy the criteria for categorical exclusion in
accordance with 10 CFR 51.22. Therefore, pursuant to 10 CFR 51.22(b),
no environmental impact statement or environmental assessment need be
prepared for these amendments. If the Commission has prepared an
environmental assessment under the special circumstances provision in
10 CFR 51.12(b) and has made a determination based on that assessment,
it is so indicated.
For further details with respect to the action see (1) the
applications for amendment, (2) the amendment, and (3) the Commission's
related letter, Safety Evaluation and/or Environmental Assessment as
indicated. All of these items are available for public inspection at
the Commission's Public Document Room, the Gelman Building, 2120 L
Street, NW., Washington, DC 20555, and at the local public document
rooms for the particular facilities involved.
Arizona Public Service Company, et al., Docket Nos. STN 50-528, STN
50-529, and STN 50-530, Palo Verde Nuclear Generating Station,
Units 1, 2, and 3, Maricopa County, Arizona
Date of application for amendment: August 5, 1993, as supplemented
by letter dated January 19, 1994.
Brief description of amendment: The amendment extends the
surveillance interval of selected channel functional test for the
reactor protection system (RPS) and the engineered safety feature
actuation system (ESFAS) instrumentation from once per month to
quarterly. In addition, the amendment will modify Technical
Specification (TS) 2.2.1, Table 2.2-1 to change selected reactor trip
setpoints and allowable values. This amendment supersedes an amendment
request dated December 28, 1992, published in the Federal Register on
February 3, 1993 (58 FR 6994)
Date of issuance: July 15, 1994
Effective date: July 15, 1994
Amendment Nos.: 78, 64, and 50
Facility Operating License Nos. NPF-41, NPF-51, and NPF-74:
Amendment revised the Technical Specifications.
Date of initial notice in Federal Register: September 15, 1993 (58
FR 48378) The additonal information contained in the January 19, 1994,
letter was clarifying in nature, was within the scope of the initial
notice, and did not affect the NRC staff's proposed no significant
hazards consideration determination. The Commission's related
evaluation of the amendment is contained in a Safety Evaluation dated
July 15, 1994.No significant hazards consideration comments received:
No.
Local Public Document Room location: Phoenix Public Library, 12
East McDowell Road, Phoenix, Arizona 85004
Carolina Power & Light Company, et al., Docket No. 50-400, Shearon
Harris Nuclear Power Plant, Unit 1, Wake and Chatham Counties,
North Carolina
Date of application for amendment: March 25, 1994
Brief description of amendment: The amendment revises TS 3/4.8.4.2,
by eliminating the term ``motor starter'' and replacing it with a more
accurate description of the MOV bypass configuration.
Date of issuance: July 12, 1994
Effective date: July 12, 1994
Amendment No.: 48
Facility Operating License No. NPF-63. Amendment revises the
Technical Specifications.
Date of initial notice in Federal Register: April 28, 1994 (59 FR
22002) The Commission's related evaluation of the amendment is
contained in a Safety Evaluation dated July 12, 1994.No significant
hazards consideration comments received: No
Local Public Document Room location: Cameron Village Regional
Library, 1930 Clark Avenue, Raleigh, North Carolina 27605.
Carolina Power & Light Company, Docket No. 50-261, H. B. Robinson
SteamElectric Plant, Unit No. 2, Darlington County, South Carolina
Date of application for amendment: May 20, 1994
Brief description of amendment: The amendment revises the Technical
Specification requirement for the Manager - Operations to hold a Senior
Reactor Operators (SRO) license at HBR.The revision allows the Manager
- Operations position to be filled by an individual who holds or has
held an SRO license at either HBR or a similar plant. The amendment
also requires the Manager - Shift Operations to hold an SRO license for
HBR.
Date of issuance: July 15, 1994
Effective date: July 15, 1994
Amendment No.: 148
Facility Operating License No. DPR-23. Amendment revises the
Technical Specifications.
Date of initial notice in Federal Register: June 8, 1994 (59 FR
29625) The Commission's related evaluation of the amendment is
contained in a Safety Evaluation dated July 15, 1994.No significant
hazards consideration comments received: No
Local Public Document Room location: Hartsville Memorial Library,
Home and Fifth Avenues, Hartsville, South Carolina 29550
The Cleveland Electric Illuminating Company, Centerior Service
Company, Duquesne Light Company, Ohio Edison Company, Pennsylvania
Power Company, Toledo Edison Company, Docket No. 50-440, Perry
Nuclear Power Plant, Unit No. 1, Lake County, Ohio
Date of application for amendment: March 12, 1993
Brief description of amendment: The amendment revises Technical
Specification Table 3.3.7.1-1 Action 72, to clarify the actions to be
taken if the control room ventilation radiation monitor becomes
inoperable.
Date of issuance: July 18, 1994
Effective date: as of its date of issuance to be implemented within
90 days.
Amendment No. 64
Facility Operating License No. NPF-58. This amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: April 28, 1994 (59 FR
22013) The Commission's related evaluation of the amendment is
contained in a Safety Evaluation dated July 18, 1994. No significant
hazards consideration comments received: No
Local Public Document Room location: Perry Public Library, 3753
Main Street, Perry, Ohio 44081
Commonwealth Edison Company, Docket Nos. 50-254 and 50-265, Quad
Cities Nuclear Power Station, Units 1 and 2, Rock Island County,
Illinois
Date of application for amendments: March 11, 1994
Brief description of amendments: The amendments add new hydraulic
snubbers on the main steamlines to the Technical Specifications (TS)
for Quad Cities, Units 1 and 2, and also change the snubber visual
inspection interval and corrective actions in TS Section 3.6.1 and
4.6.1 to the format and content of the BWRs STSs, as revised by Generic
Letter 90-09, ``Alternative Requirements for Snubber Visual Inspection
Intervals and Corrective Actions,'' dated December 11, 1990.
Date of issuance: July 13, 1994
Effective date: July 13, 1994
Amendment Nos.: 149 and 145
Facility Operating License Nos. DPR-29 and DPR-30. The amendments
revised the Technical Specifications.
Date of initial notice in Federal Register: 59 FR 17595 (April 13,
1994) The Commission's related evaluation of the amendments is
contained in a Safety Evaluation dated July 13, 1994.No significant
hazards consideration comments received: No
Local Public Document Room location: Dixon Public Library, 221
Hennepin Avenue, Dixon, Illinois 61021.
Duke Power Company, et al., Docket Nos. 50-413 and 50-414, Catawba
Nuclear Station, Units 1 and 2, York County, South CarolinaDate of
application for amendments: January 25, 1993, as supplemented May
12, 1993
Brief description of amendments: The amendments revise the
Technical Specifications to allow longer surveillance test intervals
and allowed outage times for the reactor protection system and the
engineered safety features actuation system.
Date of issuance: July 18, 1994
Effective date: July 18, 1994
Amendment Nos.: 122 and 116
Facility Operating License Nos. NPF-35 and NPF-52: Amendments
revised the Technical Specifications.
Date of initial notice in Federal Register: August 4, 1993 (58 FR
41501) The Commission's related evaluation of the amendments is
contained in a Safety Evaluation dated July 18, 1994.No significant
hazards consideration comments received: No
Local Public Document Room location: York County Library, 138 East
Black Street, Rock Hill, South Carolina 29730
Duke Power Company, Docket Nos. 50-369 and 50-370, McGuire Nuclear
Station, Units 1 and 2, Mecklenburg County, North Carolina
Date of application for amendments: February 25, 1994
Brief description of amendments: The amendments revise Technical
Specification (TS) Tables 3.3-10 and 4.3-7 to add four instruments as
part of the accident monitoring instrumentation, and delete five
instruments from the TS Tables that are not part of the accident
monitoring instrumentation.
Date of issuance: July 18, 1994
Effective date: July 18, 1994
Amendment Nos.: 144 and 126
Facility Operating License Nos. NPF-9 and NPF-17: Amendments
revised the Technical Specifications.
Date of initial notice in Federal Register: April 13, 1994 (59 FR
17597) The Commission's related evaluation of the amendments is
contained in a Safety Evaluation dated July 18, 1994.No significant
hazards consideration comments received: No.
Local Public Document Room location: Atkins Library, University of
North Carolina, Charlotte (UNCC Station), North Carolina 28223
Florida Power and Light Company, et al., Docket Nos. 50-335 and 50-
389, St. Lucie Plant, Unit Nos. 1 and 2, St. Lucie County,
FloridaDate of application for amendments: February 22, 1994
Brief description of amendments: These amendments relocate the
instrument response time limits for the Reactor Protective System and
the Engineered Safety Features Actuation System from the Technical
Specifications to be Updated Safety Analysis Report for both units.
Date of issuance: July 12, 1994
Effective date: July 12, 1994
Amendment Nos.: 128 and 67
Facility Operating License Nos. DPR-67 and NPF-16: Amendments
revised the Technical Specifications.
Date of initial notice in Federal Register: April 13, 1994 (59 FR
17598) The Commission's related evaluation of the amendments is
contained in a Safety Evaluation dated July 12, 1994No significant
hazards consideration comments received: No.
Local Public Document Room location: Indian River Junior College
Library, 3209 Virginia Avenue, Fort Pierce, Florida 34954-9003
GPU Nuclear Corporation, et al., Docket No. 50-289, Three Mile
Island Nuclear Station, Unit No. 1, Dauphin County, Pennsylvania
Date of application for amendment: April 11, 1994
Brief description of amendment: The amendment revises the plant
Technical Specifications (TS) to relocate the detailed inspection
criteria, methods and frequencies of the containment tendon
surveillance program to the Final Safety Analysis Report (FSAR) and to
provide a direct reference to the tendon surveillance program in the
TS, and make certain editorial changes.
Date of issuance: July 14, 1994
Effective date: As of the date of issuance to be implemented within
30 days.
Amendment No.: 187
Facility Operating License No. DPR-50. Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: June 8, 1994 (59 FR
29627). The Commission's related evaluation of the amendment is
contained in a Safety Evaluation dated July 14, 1994. No significant
hazards consideration comments received: No.
Local Public Document Room location: Government Publications
Section, State Library of Pennsylvania, Walnut Street and Commonwealth
Avenue, Box 1601, Harrisburg, Pennsylvania 17105.
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: March 21, 1994
Brief description of amendments: The amendments revise Technical
Specifications 3.1.2.3 ``Reactivity Control Systems Charging Pumps -
Shutdown'' and 3.1.2.1 ``Boration Systems Flow Paths - Shutdown.'' The
amendments allow energizing of an inoperable centrifugal charging pump
in preparation for switching of the centrifugal charging pumps,
provided the pump discharge is isolated from the reactor coolant
system. The amendment allows for continued flow to the reactor coolant
pump seals.
Date of issuance: July 12, 1994
Effective date: July 12, 1994, to be implemented within 31 days of
issuance.
Amendment Nos.: Unit 1 - Amendment No. 62; Unit 2 - Amendment No.
51
Facility Operating License Nos. NPF-76 and NPF-80. The amendments
revised the Technical Specifications.
Date of initial notice in Federal Register: April 13, 1994 (59 FR
17602) The Commission's related evaluation of the amendments is
contained in a Safety Evaluation dated July 12, 1994.No significant
hazards consideration comments received: No
Local Public Document Room location: Wharton County Junior College,
J. M. Hodges Learning Center, 911 Boling Highway, Wharton, Texas 77488
Illinois Power Company and Soyland Power Cooperative, Inc., Docket
No. 50-461, Clinton Power Station, Unit No. 1, DeWitt County,
Illinois
Date of application for amendment: April 18, 1994
Brief description of amendment: Technical Specification Sections 3/
4.6.6.3, ``Standby Gas Treatment System,'' and 3/4.7.2, ``Control Room
Ventilation System,'' require periodic testing of the charcoal filter
beds to demonstrate their continuing effectiveness in removing
radioiodine. The specifications, which referenced the testing
methodology of ASTM D3803-79, have been updated to reference the 1989
version of the standard.
Date of issuance: July 22, 1994
Effective date: July 22, 1994
Amendment No.: 91
Facility Operating License No. NPF-62. The amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: May 12, 1994 (59 FR
24750) The June 16, 1994, submittal consisted of revisions/
clarifications which did not change the staff's initial proposed no
significant hazards considerations determination.
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated July 22, 1994.No significant hazards
consideration comments received: No
Local Public Document Room location: The Vespasian Warner Public
Library, 120 West Johnson Street, Clinton, Illinois 61727.
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 application for amendments: November 17, 1993.
Brief description of amendments: The amendments modify the
Technical Specifications to allow a portion of the Waste Gas Holdup
System Explosive Monitoring System to be inoperable for 160 days on a
one-time basis so that the Waste Gas Oxygen Analyzer can be replaced.
These amendments also make an editorial change to the Automatic Gas
Analyzer tag numbers.
Date of issuance: July 7, 1994
Effective date: July 7, 1994
Amendment Nos.: 179 & 163
Facility Operating License Nos. DPR-58 and DPR-74. Amendments
revised the Technical Specifications.
Date of initial notice in Federal Register: February 2, 1994 (59 FR
4938) The Commission's related evaluation of the amendments is
contained in a Safety Evaluation dated July 7, 1994.No significant
hazards consideration comments received: No.
Local Public Document Room location: Maud Preston Palenske Memorial
Library, 500 Market Street, St. Joseph, Michigan 49085.
Public Service Electric & Gas Company, Docket Nos. 50-272 and 50-
311, Salem Nuclear Generating Station, Unit Nos. 1 and 2, Salem
County, New Jersey
Date of application for amendments: August 6, 1993
Brief description of amendments: This amendment request changes the
technical specifications to provide for a separate action if the
Accumulator cannot meet the requirements of the Limiting Condition for
Operation due to boron concentration. The allowed outage time to
restore boron concentration is changed from 1 hour to 72 hours.
Date of issuance: July 20, 1994
Effective date: July 20, 1994
Amendment Nos. 152 and 132
Facility Operating License Nos. DPR-70 and DPR-75. These amendments
revised the Technical Specifications.
Date of initial notice in Federal Register: September 15, 1993 (58
FR 48389) The Commission's related evaluation of the amendments is
contained in a Safety Evaluation dated July 20, 1994.No significant
hazards consideration comments received: No
Local Public Document Room location: Salem Free Public Library, 112
West Broadway, Salem, New Jersey 08079
Public Service Electric & Gas Company, Docket No. 50-311, Salem
Nuclear Generating Station, Unit No. 2, Salem County, New Jersey
Date of application for amendment: June 11, 1993 and supplemented
July 19, August 3, and September 16, 1993.
Brief description of amendment: The amendment reduces the boron
concentration in the boric acid tank from 12 percent by weight to
between 3.75 and 4 percent by weight. The reduced boron concentration
results in eliminating the need for heat tracing in the boric acid tank
piping system.
Date of issuance: July 20, 1994
Effective date: July 20, 1994
Amendment No. 133
Facility Operating License No. DPR-75: This amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: August 18, 1993 (58 FR
43932) The Commission's related evaluation of the amendment is
contained in a Safety Evaluation dated July 20, 1994.No significant
hazards consideration comments received: No
Local Public Document Room location: Salem Free Public Library, 112
West Broadway, Salem, New Jersey 08079
Tennessee Valley Authority, Docket Nos. 50-259, 50-260 and 50-296,
Browns Ferry Nuclear Plant, Units 1, 2 and 3, Limestone County,
AlabamaDate of application for amendments: December 23, 1993 (TS
346)
Brief description of amendment: The amendments revise the Technical
Specification surveillance requirements regarding the visual inspection
of snubbers, consistent with the guidance in Generic Letter 90-09,
``Alternative Requirements for Snubber Visual Inspection Intervals and
Corrective Actions.''
Date of issuance: July 5, 1994
Effective date: July 5, 1994
Amendment Nos.: 210, 225 and 183
Facility Operating License Nos. DPR-33, DPR-52 and DPR-68: This
amendment revised the Technical Specifications.
Date of initial notice in Federal Register: May 25, 1994 (59 FR
27067) The Commission's related evaluation of the amendment is
contained in a Safety Evaluation dated July 5, 1994.No significant
hazards consideration comments received: None
Local Public Document Room location: Athens Public Library, South
Street, Athens, Alabama 35611
Tennessee Valley Authority, Docket Nos. 50-327 and 50-328, Sequoyah
Nuclear Plant, Units 1 and 2, Hamilton County, Tennessee
Date of application for amendments: May 16, 1994 (TS 93-18)
Brief description of amendments: The amendments change the
Electrical Power Systems surveillance requirements wording to reflect
the use of the new common station service transformers with auto load
tap changers as the normal power supply for the 6.9 KV unit boards.
Date of issuance: July 11, 1994
Effective date: July 11, 1994
Amendment Nos.: 184 and 176
Facility Operating License Nos. DPR-77 and DPR-79: Amendments
revise the technical specifications.
Date of initial notice in Federal Register: June 8, 1994 (59 FR
29637) The Commission's related evaluation of the amendments are
contained in a Safety Evaluation dated July 11, 1994. No significant
hazards consideration comments received: None
Local Public Document Room location: Chattanooga-Hamilton County
Library, 1101 Broad Street, Chattanooga, Tennessee 37402
Toledo Edison Company, Centerior Service Company, and The Cleveland
Electric Illuminating Company, Docket No. 50-346, Davis-Besse
Nuclear Power Station, Unit No. 1, Ottawa County, Ohio
Date of application for amendment: March 18, 1994, as supplemented
on June 20, 1994.
Brief description of amendment: The amendment revises TS 2.1.2
(Reactor Core), TS 2.2.1 (Reactor Protection System Setpoints), Bases
2.1.1 and 2.1.2 (Reactor Core), Bases 2.2.1 (Reactor Protection System
Instrumentation Setpoints), TS 3.2.2 and 3.2.3 (Power Distribution
Limits), Bases 3/4 (Power Distribution Limits), and TS 6.9.1.7
(Administrative Controls, Core Operating Limits Report). This amendment
removes cycle-specific limits from TS and relocates them in the Core
Operating Limits Report.
Date of issuance: July 22, 1994
Effective date: July 22, 1994
Amendment No. 189
Facility Operating License No. NPF-3. Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: April 28, 1994 (59 FR
22014) The June 20, 1994, submittal provided supplemental information
that did not change the initial proposed no significant hazards
consideration determination. The Commission's related evaluation of the
amendment is contained in a Safety Evaluation dated July 22, 1994.No
significant hazards consideration comments received: No
Local Public Document Room location: University of Toledo Library,
Documents Department, 2801 Bancroft Avenue, Toledo, Ohio 43606.
Union Electric Company, Docket No. 50-483, Callaway Plant, Unit 1,
Callaway County, Missouri
Date of application for amendment: February 21, 1992, supplemented
by letters dated April 16, 1992, and March 29, 1994
Brief description of amendment: The amendment revises the Technical
Specifications Appendix B, Environmental Protection Plan (Non-
radiological), by removing Sections 2.3 and 4.3, ``Cultural
Resources.'' Union Electric has developed and maintains a management
plan for the protection of cultural resources on the Callaway Plant
site. The amendment request summarizes the plan that provides the
status and disposition of each portion of the current Appendix B
sections related to cultural resources.
Date of issuance: July 13, 1994
Effective date: to be implemented within 30 days of issuance
Amendment No.: 90
Facility Operating License No. NPF-30. Amendment revised the
Technical Specifications Appendix B, Environmental Protection Plan.
Date of initial notice in Federal Register: April 13, 1994 (59 FR
17607) The additional information contained in the April 16, 1992, and
March 29, 1994, letters was supplemental in nature, is within the scope
of the initial notice, and did not affect the NRC staff's proposed no
significant hazards consideration. The Commission's related evaluation
of the amendment is contained in a Safety Evaluation dated July 13,
1994.No significant hazards consideration comments received: No.
Local Public Document Room location: Callaway County Public
Library, 710 Court Street, Fulton, Missouri 65251.
Virginia Ele1ctric and Power Company, et al., Docket Nos. 50-338
and 50-339, North Anna Power Station, Units No. 1 and No. 2, Louisa
County, Virginia
Date of application for amendments: March 1, 1994, as supplemented
by letter dated June 16, 1994
Brief description of amendments: The amendments modify the
requirement for operability testing of an EDG when the alternate safety
buses' EDG is inoperable. Also, the requirement for operability testing
of the EDGs when one or both offsite AC sources are inoperable is
deleted. Finally, the amendments eliminate fast loading of EDGs except
for the Loss of Offsite Power test and separate the hot restart test
from the 24-hour loaded test run of the EDGs. The changes are
consistent with NRC Generic Letter 93-05 dated September 27, 1993.
Date of issuance: July 18, 1994
Effective date: July 18, 1994
Amendment Nos.: 184 and 165
Facility Operating License Nos. NPF-4 and NPF-7. Amendments revised
the Technical Specifications.
Date of initial notice in Federal Register: March 30, 1994 (59 FR
14899) The Commission's related evaluation of the amendments is
contained in a Safety Evaluation dated July 18, 1994No significant
hazards consideration comments received: No.
Local Public Document Room location: The Alderman Library, Special
Collections Department, University of Virginia, Charlottesville,
Virginia 22903-2498.
Washington Public Power Supply System, Docket No. 50-397, Nuclear
Project No. 2, Benton County, Washington
Date of application for amendment: December 6, 1993, supplemented
by letter dated May 6, 1994
Brief description of amendment: The amendment revised Table
4.3.7.5-1 of Technical Specification (TS) 3/4.3.7.5, ``Accident
Monitoring Instrumentation,'' to include a note that requires that
safety/relief valve (SRV) position indicator surveillance testing be
performed within 12 hours after steam pressure and flow are adequate to
do the testing. The amendment also revised TS 3/4.4.2, ``Safety/Relief
Valves,'' and TS 3/4.5.1, ``Emergency Core Cooling Systems,'' to
require that the main steam system and automatic depressurization
system SRVs be surveilled within 12 hours after steam pressure and flow
are adequate to do the testing. Additionally, Bases Section 3/4.4.2 was
revised to reflect the changes in the TS.
Date of issuance: July 8, 1994
Effective date: July 8, 1994
Amendment No.: 128
Facility Operating License No. NPF-21: The amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: May 13, 1994 (59 FR
25131) The Commission's related evaluation of the amendment is
contained in a Safety Evaluation dated July 8, 1994.No significant
hazards consideration comments received: No.
Local Public Document Room location: Richland Public Library, 955
Northgate Street, Richland, Washington 99352
Wisconsin Public Service Corporation, Docket No. 50-305, Kewaunee
NuclearPower Plant, Kewaunee County, Wisconsin
Date of application for amendment: December 7, 1993
Brief description of amendment: The amendment revises Kewaunee
Nuclear Power Plant (KNPP) Technical Specification (TS) 5.3.a.1 to
provide flexibility in the repair of fuel assemblies containing damaged
and leaking fuel rods by reconstituting the assemblies, provided that
an NRC-approved methodology is used. This change is consistent with
guidance provided in Supplement 1 to Generic Letter (GL) 90-02,
``Alternative Requirements for Fuel Assemblies in the Design Features
Section of Technical Specifications,'' dated July 31, 1992. In
addition, administrative changes to KNPP TS Section 5 have been made.
Date of issuance: July 15, 1994
Effective date: to be implemented within 30 days
Amendment No.: 109
Facility Operating License No. DPR-43. Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: February 2, 1994 (59 FR
4951) The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated July 15, 1994.No significant hazards
consideration comments received: No.
Local Public Document Room location: University of Wisconsin
Library Learning Center, 2420 Nicolet Drive, Green Bay, Wisconsin
54301.
Notice Of Issuance Of Amendments To Facility Operating Licenses And
Final Determination Of No Significant Hazards Consideration And
Opportunity For A Hearing (Exigent Public Announcement Or Emergency
Circumstances)
During the period since publication of the last biweekly notice,
the Commission has issued the following amendments. The Commission has
determined for each of these amendments that the application for the
amendment complies with the standards and requirements of the Atomic
Energy Act of 1954, as amended (the Act), and the Commission's rules
and regulations. The Commission has made appropriate findings as
required by the Act and the Commission's rules and regulations in 10
CFR Chapter I, which are set forth in the license amendment.
Because of exigent or emergency circumstances associated with the
date the amendment was needed, there was not time for the Commission to
publish, for public comment before issuance, its usual 30-day Notice of
Consideration of Issuance of Amendment, Proposed No Significant Hazards
Consideration Determination, and Opportunity for a Hearing.
For exigent circumstances, the Commission has either issued a
Federal Register notice providing opportunity for public comment or has
used local media to provide notice to the public in the area
surrounding a licensee's facility of the licensee's application and of
the Commission's proposed determination of no significant hazards
consideration. The Commission has provided a reasonable opportunity for
the public to comment, using its best efforts to make available to the
public means of communication for the public to respond quickly, and in
the case of telephone comments, the comments have been recorded or
transcribed as appropriate and the licensee has been informed of the
public comments.
In circumstances where failure to act in a timely way would have
resulted, for example, in derating or shutdown of a nuclear power plant
or in prevention of either resumption of operation or of increase in
power output up to the plant's licensed power level, the Commission may
not have had an opportunity to provide for public comment on its no
significant hazards consideration determination. In such case, the
license amendment has been issued without opportunity for comment. If
there has been some time for public comment but less than 30 days, the
Commission may provide an opportunity for public comment. If comments
have been requested, it is so stated. In either event, the State has
been consulted by telephone whenever possible.
Under its regulations, the Commission may issue and make an
amendment immediately effective, notwithstanding the pendency before it
of a request for a hearing from any person, in advance of the holding
and completion of any required hearing, where it has determined that no
significant hazards consideration is involved.
The Commission has applied the standards of 10 CFR 50.92 and has
made a final determination that the amendment involves no significant
hazards consideration. The basis for this determination is contained in
the documents related to this action. Accordingly, the amendments have
been issued and made effective as indicated.
Unless otherwise indicated, the Commission has determined that
these amendments satisfy the criteria for categorical exclusion in
accordance with 10 CFR 51.22. Therefore, pursuant to 10 CFR 51.22(b),
no environmental impact statement or environmental assessment need be
prepared for these amendments. If the Commission has prepared an
environmental assessment under the special circumstances provision in
10 CFR 51.12(b) and has made a determination based on that assessment,
it is so indicated.
For further details with respect to the action see (1) the
application for amendment, (2) the amendment to Facility Operating
License, and (3) the Commission's related letter, Safety Evaluation
and/or Environmental Assessment, as indicated. All of these items are
available for public inspection at the Commission's Public Document
Room, the Gelman Building, 2120 L Street, NW., Washington, DC 20555,
and at the local public document room for the particular facility
involved.
The Commission is also offering an opportunity for a hearing with
respect to the issuance of the amendment. By September 2, 1994, 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 20555 and at the local public document room for the
particular facility involved. If a request for a hearing or petition
for leave to intervene is filed by the above date, the Commission or an
Atomic Safety and Licensing Board, designated by the Commission or by
the Chairman of the Atomic Safety and Licensing Board Panel, will rule
on the request and/or petition; and the Secretary or the designated
Atomic Safety and Licensing Board will issue a notice of a hearing or
an appropriate order.
As required by 10 CFR 2.714, a petition for leave to intervene
shall set forth with particularity the interest of the petitioner in
the proceeding, and how that interest may be affected by the results of
the proceeding. The petition should specifically explain the reasons
why intervention should be permitted with particular reference to the
following factors: (1) the nature of the petitioner's right under the
Act to be made a party to the proceeding; (2) the nature and extent of
the petitioner's property, financial, or other interest in the
proceeding; and (3) the possible effect of any order which may be
entered in the proceeding on the petitioner's interest. The petition
should also identify the specific aspect(s) of the subject matter of
the proceeding as to which petitioner wishes to intervene. Any person
who has filed a petition for leave to intervene or who has been
admitted as a party may amend the petition without requesting leave of
the Board up to 15 days prior to the first prehearing conference
scheduled in the proceeding, but such an amended petition must satisfy
the specificity requirements described above.
Not later than 15 days prior to the first prehearing conference
scheduled in the proceeding, a petitioner shall file a supplement to
the petition to intervene which must include a list of the contentions
which are sought to be litigated in the matter. Each contention must
consist of a specific statement of the issue of law or fact to be
raised or controverted. In addition, the petitioner shall provide a
brief explanation of the bases of the contention and a concise
statement of the alleged facts or expert opinion which support the
contention and on which the petitioner intends to rely in proving the
contention at the hearing. The petitioner must also provide references
to those specific sources and documents of which the petitioner is
aware and on which the petitioner intends to rely to establish those
facts or expert opinion. Petitioner must provide sufficient information
to show that a genuine dispute exists with the applicant on a material
issue of law or fact. Contentions shall be limited to matters within
the scope of the amendment under consideration. The contention must be
one which, if proven, would entitle the petitioner to relief. A
petitioner who fails to file such a supplement which satisfies these
requirements with respect to at least one contention will not be
permitted to participate as a party.
Those permitted to intervene become parties to the proceeding,
subject to any limitations in the order granting leave to intervene,
and have the opportunity to participate fully in the conduct of the
hearing, including the opportunity to present evidence and cross-
examine witnesses. Since the Commission has made a final determination
that the amendment involves no significant hazards consideration, if a
hearing is requested, it will not stay the effectiveness of the
amendment. Any hearing held would take place while the amendment is in
effect.
A request for a hearing or a petition for leave to intervene must
be filed with the Secretary of the Commission, U.S. Nuclear Regulatory
Commission, Washington, DC 20555, Attention: Docketing and Services
Branch, or may be delivered to the Commission's Public Document Room,
the Gelman Building, 2120 L Street, NW., Washington, DC 20555, 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 the
factors specified in 10 CFR 2.714(a)(1)(i)-(v) and 2.714(d).
The Cleveland Electric Illuminating Company, Centerior Service
Company, Duquesne Light Company, Ohio Edison Company, Pennsylvania
Power Company, Toledo Edison Company, Docket No. 50-440, Perry
Nuclear Power Plant, Unit No. 1, Lake County, Ohio
Date of application for amendment: July 14, 1994
Brief description of amendment: The amendment revises Technical
Specification 3.6.1.4, main steam isolation valve (MSIV) leakage
control system (LCS), to add a footnote to the APPLICABILITY statement.
The footnote states, ``The provisions of Specification 3.0.4 are not
applicable from the effective date of this amendment until the
completion of Operating Cycle 5.''
Date of issuance: July 15, 1994
Effective date: July 15, 1994
Amendment No. 63
Facility Operating License No. NPF-58. This amendment revised the
Technical Specifications. Public comments requested as to proposed no
significant hazards consideration: No.The Commission's related
evaluation of the amendment, finding of emergencycircumstances, and
final determination of no significant hazards consideration are
contained in a Safety Evaluation dated July 15, 1994.
Local Public Document Room location: Perry Public Library, 3753
Main Street, Perry, Ohio 44081
Attorney for licensee: Jay Silberg, Esq., Shaw, Pittman, Potts &
Trowbridge, 2300 N Street, NW., Washington, DC 20037.
NRC Project Director: John N. Hannon
Patriots Point Development Authority and U.S. Maritime
Administration, Docket No. 50-238, N.S. Savannah
Date of application for amendment: May 19, 1994, as supplemented on
May 24 and 27, 1994, and June 3, 1994.
Brief description of amendment: The amendment (1) deleted the
Patriots Point Development Authority (PPDA) as a co-licensee, (2)
allowed relocation of the N.S. Savannah to the James River Reserve
Fleet, a U.S. Maritime Administration (MARAD) facility, (3) changed the
performance of radiological health physics coverage, surveillance and
response to the U.S. Army Center for Public Works, Humphries
Engineering Center, (4) changed the composition of the Review and Audit
Committee to be consistent with the deletion of PPDA as a co-licensee,
and (5) discontinued public access to the facility and made other minor
changes to the TS.
Date of issuance: June 29, 1994
Effective date: June 29, 1994
Amendment No.: 12Amended Facility License No. NS-1: Amendment
revised the Technical Specifications and license.Public comments
requested as to proposed no significant hazards consideration: The NRC
published a public notice of the proposed amendment, issued a proposed
finding of no significant hazards consideration and requested that any
comments on the proposed no significant hazards consideration be
provided to the staff by the close of business on June 2, 1994. The
notice was published in The Virginian-Pilot/The Ledger-Star, Norfolk,
Virginia on Sunday, May 29, 1994, The Daily Press, Newport News,
Virginia on Friday, May 27, 1994, and The Post and Courier, Charleston,
South Carolina, on Friday, May 27, 1994. No comments have been
received.
The Commission's related evaluation of the amendment, finding of
exigent circumstances, consultation with the States of South Carolina
and Virginia and final no significant hazards consideration
determination are contained in a Safety Evaluation dated June 29, 1994.
Attorney for licensees: M. J. McMorrow, Office of the Chief Council
(MAR 220), U.S. Maritime Administration, Room 7228, 400 Seventh Street,
SW, Washington, D.C. 20590.
Local Public Document Room location: N/A
Washington Public Power Supply System, Docket No. 50-397, Nuclear
ProjectNo. 2, Benton County, Washington
Date of application for amendment: July 8, 1994
Brief description of amendment: The amendment modified the
technical specifications to permit post-maintenance testing of control
rod scram insertion times to be performed at lower reactor coolant
pressures than currently allowed.
Date of issuance: July 14, 1994
Effective date: July 14, 1994
Amendment No.: 129
Facility Operating License No. NPF-21: The amendment revised the
Technical Specifications.Public comments on proposed no significant
hazards consideration comments received: No.The Commission's related
evaluation of the amendment, finding of emergency circumstances, and
final determination of no significant hazards consideration are
contained in a Safety Evaluation dated July 14, 1994.
Local Public Document Room location: Richland Public Library, 955
Northgate Street, Richland, Washington 99352
Attorney for licensee: M. H. Philips, Jr., Esq., Winston & Strawn,
1400 L Street, NW., Washington, DC 20005-3502
NRC Project Director: Theodore R. Quay
Dated at Rockville, Maryland, this 27th day of July 1994.
For the Nuclear Regulatory Commission
Steven A. Varga,
Director, Division of Reactor Projects - I/II,Office of Nuclear Reactor
Regulation.
[FR Doc. 94-18741 Filed 8-2-94; 8:45 am]
BILLING CODE 7590-01-F
