[Federal Register Volume 60, Number 168 (Wednesday, August 30, 1995)]
[Notices]
[Pages 45172-45196]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: X95-10830]
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NUCLEAR REGULATORY COMMISSION
Biweekly Notice
Applications and Amendments to Facility Operating Licenses
Involving No Significant Hazards Considerations
I. Background
Pursuant to Public Law 97-415, the U.S. Nuclear Regulatory
Commission (the Commission or NRC staff) is publishing this regular
biweekly notice. Public Law 97-415 revised section 189 of the Atomic
Energy Act of 1954, as amended (the Act), to require the Commission to
publish notice of any amendments issued, or proposed to be issued,
under a new provision of section 189 of the Act. This provision grants
the Commission the authority to issue and make immediately effective
any amendment to an operating license upon a determination by the
Commission that such amendment involves no significant hazards
consideration, notwithstanding the pendency before the Commission of a
request for a hearing from any person.
This biweekly notice includes all notices of amendments issued, or
proposed to be issued from August 4, 1995, through August 18, 1995. The
last biweekly notice was published on August 16, 1995 (60 FR 42597).
Notice Of Consideration Of Issuance Of Amendments To Facility
Operating Licenses, Proposed No Significant Hazards Consideration
Determination, And Opportunity For A Hearing
The Commission has made a proposed determination that the following
amendment requests involve no significant hazards consideration. Under
the Commission's regulations in 10 CFR 50.92, this means that operation
of the facility in accordance with the proposed amendment would not (1)
involve a significant increase in the probability or consequences of an
accident previously evaluated; or (2) create the possibility of a new
or different kind of accident from any accident previously evaluated;
or (3) involve a significant reduction in a margin of safety. The basis
for this proposed determination for each amendment request is shown
below.
The Commission is seeking public comments on this proposed
determination. Any comments received within 30 days after the date of
publication of this notice will be considered in making any final
determination.
Normally, the Commission will not issue the amendment until the
expiration of the 30-day notice period. However, should circumstances
change during the notice period such that failure to act in a timely
way would result, for example, in derating or shutdown of the facility,
the Commission may issue the license amendment before the expiration of
the 30-day notice period, provided that its final determination is that
the amendment involves no significant hazards consideration. The final
determination will consider all public and State comments received
before action is taken. Should the Commission take this action, it will
publish in the Federal Register a notice of issuance and provide for
opportunity for a hearing after issuance. The Commission expects that
the need to take this action will occur very infrequently.
Written comments may be submitted by mail to the Rules Review and
Directives Branch, Division of Freedom of Information and Publications
Services, Office of Administration, U.S. Nuclear Regulatory Commission,
Washington, DC 20555, and should cite the publication date and page
number of this Federal Register notice. Written comments may also be
delivered to Room 6D22, Two White Flint North, 11545 Rockville Pike,
Rockville, Maryland from 7:30 a.m. to 4:15 p.m. Federal workdays.
Copies of written comments received may be examined at the NRC Public
Document Room, the Gelman Building, 2120 L Street, NW., Washington, DC.
The filing of requests for a hearing and petitions for leave to
intervene is discussed below.
By September 29, 1995, the licensee may file a request for a
hearing with respect to issuance of the amendment to the subject
facility operating license and any person whose interest may be
affected by this proceeding and who wishes to participate as a party in
the proceeding must file a written request for a hearing and a petition
for leave to intervene. Requests for a hearing and a petition for leave
to intervene shall be filed in accordance with the Commission's ``Rules
of Practice for Domestic Licensing Proceedings'' in 10 CFR Part 2.
Interested persons should consult a current copy of 10 CFR 2.714 which
is available at the Commission's Public Document Room, the Gelman
Building, 2120 L Street, NW.,
[[Page 45173]]
Washington, DC and at the local public document room for the particular
facility involved. If a request for a hearing or petition for leave to
intervene is filed by the above date, the Commission or an Atomic
Safety and Licensing Board, designated by the Commission or by the
Chairman of the Atomic Safety and Licensing Board Panel, will rule on
the request and/or petition; and the Secretary or the designated Atomic
Safety and Licensing Board will issue a notice of a hearing or an
appropriate order.
As required by 10 CFR 2.714, a petition for leave to intervene
shall set forth with particularity the interest of the petitioner in
the proceeding, and how that interest may be affected by the results of
the proceeding. The petition should specifically explain the reasons
why intervention should be permitted with particular reference to the
following factors: (1) the nature of the petitioner's right under the
Act to be made a party to the proceeding; (2) the nature and extent of
the petitioner's property, financial, or other interest in the
proceeding; and (3) the possible effect of any order which may be
entered in the proceeding on the petitioner's interest. The petition
should also identify the specific aspect(s) of the subject matter of
the proceeding as to which petitioner wishes to intervene. Any person
who has filed a petition for leave to intervene or who has been
admitted as a party may amend the petition without requesting leave of
the Board up to 15 days prior to the first prehearing conference
scheduled in the proceeding, but such an amended petition must satisfy
the specificity requirements described above.
Not later than 15 days prior to the first prehearing conference
scheduled in the proceeding, a petitioner shall file a supplement to
the petition to intervene which must include a list of the contentions
which are sought to be litigated in the matter. Each contention must
consist of a specific statement of the issue of law or fact to be
raised or controverted. In addition, the petitioner shall provide a
brief explanation of the bases of the contention and a concise
statement of the alleged facts or expert opinion which support the
contention and on which the petitioner intends to rely in proving the
contention at the hearing. The petitioner must also provide references
to those specific sources and documents of which the petitioner is
aware and on which the petitioner intends to rely to establish those
facts or expert opinion. Petitioner must provide sufficient information
to show that a genuine dispute exists with the applicant on a material
issue of law or fact. Contentions shall be limited to matters within
the scope of the amendment under consideration. The contention must be
one which, if proven, would entitle the petitioner to relief. A
petitioner who fails to file such a supplement which satisfies these
requirements with respect to at least one contention will not be
permitted to participate as a party.
Those permitted to intervene become parties to the proceeding,
subject to any limitations in the order granting leave to intervene,
and have the opportunity to participate fully in the conduct of the
hearing, including the opportunity to present evidence and cross-
examine witnesses.
If a hearing is requested, the Commission will make a final
determination on the issue of no significant hazards consideration. The
final determination will serve to decide when the hearing is held.
If the final determination is that the amendment request involves
no significant hazards consideration, the Commission may issue the
amendment and make it immediately effective, notwithstanding the
request for a hearing. Any hearing held would take place after issuance
of the amendment.
If the final determination is that the amendment request involves a
significant hazards consideration, any hearing held would take place
before the issuance of any amendment.
A request for a hearing or a petition for leave to intervene must
be filed with the Secretary of the Commission, U.S. Nuclear Regulatory
Commission, Washington, DC 20555, Attention: Docketing and Services
Branch, or may be delivered to the Commission's Public Document Room,
the Gelman Building, 2120 L Street, NW., Washington DC, by the above
date. Where petitions are filed during the last 10 days of the notice
period, it is requested that the petitioner promptly so inform the
Commission by a toll-free telephone call to Western Union at 1-(800)
248-5100 (in Missouri 1-(800) 342-6700). The Western Union operator
should be given Datagram Identification Number N1023 and the following
message addressed to (Project Director): petitioner's name and
telephone number, date petition was mailed, plant name, and publication
date and page number of this Federal Register notice. A copy of the
petition should also be sent to the Office of the General Counsel, U.S.
Nuclear Regulatory Commission, Washington, DC 20555, and to the
attorney for the licensee.
Nontimely filings of petitions for leave to intervene, amended
petitions, supplemental petitions and/or requests for a hearing will
not be entertained absent a determination by the Commission, the
presiding officer or the Atomic Safety and Licensing Board that the
petition and/or request should be granted based upon a balancing of
factors specified in 10 CFR 2.714(a)(1)(i)-(v) and 2.714(d).
For further details with respect to this action, see the
application for amendment which is available for public inspection at
the Commission's Public Document Room, the Gelman Building, 2120 L
Street, NW., Washington, DC, and at the local public document room for
the particular facility involved.
Arizona Public Service Company, et al., Docket Nos. STN 50-528, STN
50-529, and STN 50-530, Palo Verde Nuclear Generating Station,
Units Nos. 1, 2, and 3, Maricopa County, Arizona
Date of amendments request: August 3, 1995
Description of amendments request: The proposed amendment changes
would add the analytical method supplement entitled ``Fuel Rod Maximum
Allowable Gas Pressure,'' CEN-372-P-A, dated May 1990, and its
associated Nuclear Regulatory Commission Safety Evaluation Report,
dated April 10, 1990, to the list of analytical methods in TS 6.9.1.10
used to determine the PVNGS core operating limits.
Basis for proposed no significant hazards consideration
determination: As requied by 10 CFR 50.91(a), the licensee has provided
its analysis of the issue of no significant hazards consideration,
which is presented below:
1. The proposed change does not involve a significant increase
in the probability or consequences of an accident previously
evaluated.
The proposed change does not involve any change to the
configuration or method of operation of any plant equipment that is
used to mitigate the consequences of an accident. The proposed
change adds an NRC approved methodology and its associated Safety
Evaluation Report (SER), to the list of analytical methods used to
determine the core operating limits. The use of this methodology
ensures that the consequences of an accident remain within the
limits established by existing analyses. They do not alter any of
the assumptions or bounding conditions currently in the UFSAR.
The U3C6 ECCS performance analysis included the analysis of the
impact of the maximum calculated fuel rod gas pressures on the
timing of cladding rupture and on the peak cladding temperature.
This analysis concluded that the peak cladding temperature for Cycle
6 remained below that of the analysis of record and that the peak
cladding temperature continued to occur at
[[Page 45174]]
low burnup, specifically the burnup corresponding to the maximum
initial fuel stored energy.
In addition to the LOCA analysis a DNB propagation analysis was
performed to demonstrate that DNB propagation does not occur during
postulated accidents that experience DNB when pressure in a fuel pin
is higher than the system pressure. This analysis was performed
using the fuel rod strain model described in CEN-372-P-A.
Based on these analyses, there is no increase in the probability
or consequences of an accident previously evaluated.
2. The proposed change does not create the possibility of a new
or different kind of accident from any accident previously
evaluated.
The proposed change does not involve any change to the
configuration or method of operation of any plant equipment that is
used to mitigate the consequences of an accident. Accordingly, 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 change. The intent of the
proposed change is to utilize a new analytical method to ensure that
the consequences of any equipment malfunction remain within the
limits of existing analyses resulting in no impact on radiological
consequences.
The impact of the maximum fuel rod gas pressures calculated for
U3C6 was evaluated as part of the Cycle 6 ECCS performance analysis.
Except for the highest burnup analyzed, the time of cladding rupture
decreased as the initial fuel rod gas pressure increased with
burnup. However, the peak cladding temperature occurred at the
burnup with the maximum initial fuel stored energy. The analysis
also determined that the ECCS performance analysis for U3C6 is
bounded by that of the reference cycle analysis.
An evaluation was conducted to ensure that fuel would not
experience DNB propagation when the pressure in a fuel pin is higher
than the system pressure. DNB was shown not to propagate by
demonstrating that the degree of cladding deformation is no more
than the limit defined by the fuel rod maximum pressure Topical
Report (CEN-372-P-A).
Therefore, it can be concluded that the proposed change to
Section 6.9.1.10 does not create the possibility of a new or
different kind of accident from any accident previously evaluated.
3. The proposed change does not involve a significant reduction
in a margin of safety.
The proposed change adds an NRC approved Topical Report
(methodology) and its associated SER, to the list of analytical
methods used to determine core operating limits. The use of the new
methodology ensures that safety margins are maintained within the
results of existing calculations. Since the core operating limits
will continue to be established by an NRC approved methodology and
will provide adequate core protection, the proposed amendment does
not involve a significant reduction in the margin of safety.
Analyses were conducted to determine the impact of higher fuel
rod pressure on ECCS performance and DNB propagation. The results of
the analyses show that the effects of higher fuel rod pressure are
bounded by previous results.
The NRC staff has reviewed the licensee's analysis and, based on
that review, it appears that the three standards of 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendments request involve no significant hazards consideration. Local
Public Document Room location: Phoenix Public Library, 1221 N. Central
Avenue, Phoenix, Arizona 85004
Attorney for licensee: Nancy C. Loftin, Esq., Corporate Secretary
and Counsel, Arizona Public Service Company, P.O. Box 53999, Mail
Station 9068, Phoenix, Arizona 85072-3999
NRC Project Director: William H. Bateman
Baltimore Gas and Electric Company, Docket Nos. 50-317 and 50-318,
Calvert Cliffs Nuclear Power Plant, Unit Nos. 1 and 2, Calvert
County, Maryland
Date of amendments request: July 13, 1995
Description of amendments request: The proposed amendments would
revise the Calvert Cliffs Nuclear Power Plant, Unit Nos. 1 and 2,
Technical Specifications (TSs) Section 5.2.1, ``Fuel Assemblies.'' The
current TSs only allow fuel that is clad with either zircaloy or ZIRLO.
The proposed change would allow the use of cladding material other than
zircaloy or ZIRLO with an approved exemption. Thus, the proposed change
will eliminate the need for future amendments to allow the use of
different cladding material for which the Commission has issued an
exemption.
Basis for proposed no significant hazards consideration
determination: As requied 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. Would not involve a significant increase in the probability
or consequences of an accident previously evaluated.
Calvert Cliffs Technical Specification 5.2.1, Fuel Assemblies,
states that fuel rods are clad with either zircaloy or ZIRLO. This
reflects the requirements of 10 CFR 50.44, 50.46, and 10 CFR [Part]
50, Appendix K, which also restrict fuel rod cladding materials to
zircaloy or ZIRLO. Baltimore Gas and Electric Company proposes to
insert fuel assemblies into Calvert Cliffs Unit 1 which have some
fuel rods clad in zirconium alloys that do not meet the definition
of zircaloy or ZIRLO for testing purposes and has applied for an
exemption to the regulations to allow that change. The proposed
change to the Calvert Cliffs Technical Specifications will allow the
use of cladding materials that are not zircaloy or ZIRLO with an
approved exemption in accordance with 10 CFR 50.12.
The proposed change to the Unit 1 and Unit 2 Technical
Specifications will allow the use of fuel rod cladding materials
other than zircaloy or ZIRLO as long as those materials have been
approved by an exemption to the regulations. To obtain approval of
new cladding materials, 10 CFR 50.12 requires that the applicant
show that the proposed exemption is authorized by law, is consistent
with the common defense and security, will not present an undue risk
to the public health and safety; and is accompanied by special
circumstances.
Under the proposed change, any fuel rod cladding materials that
are not zircaloy or ZIRLO must still be approved by the Nuclear
Regulatory Commission (NRC) prior to use under 10 CFR 50.12. This
change to the Technical Specifications allows the NRC to approve the
use of cladding materials that are not either zircaloy or ZIRLO
under 10 CFR 50.12 and not require an additional approval under 10
CFR 50.90. As such, the proposed change eliminates a duplicative
regulatory requirement and would have no effect on the probability
or consequences of an accident.
Therefore, the proposed change does not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
2. Would not create the possibility of a new or different type
of accident from any accident previously evaluated.
The proposed change eliminates a duplicated approval requirement
and would have no effect on the possibility of a new or different
type of accident. The proposed change to the Technical
Specifications would allow the NRC to approve the use of fuel rod
cladding materials that are not either zircaloy or ZIRLO under 10
CFR 50.12 and not require an additional approval under 10 CFR 50.90.
Therefore, the proposed change does not create the possibility
of a new or different type of accident from any accident previously
evaluated.
3. Would not involve a significant reduction in a margin of
safety.
The proposed change eliminates a duplicated approval requirement
and will have no effect on the margin of safety. The proposed change
to the Technical Specifications would allow the NRC to approve the
use of fuel rod cladding materials that are not either zircaloy or
ZIRLO under 10 CFR 50.12, and not require an additional approval
under 10 CFR 50.90.
Therefore, the proposed change does not involve a significant
reduction in a margin 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
amendments request involves no significant hazards consideration.
Local Public Document Room location: Calvert County Library, Prince
Frederick, Maryland 20678
Attorney for licensee: Jay E. Silbert, Esquire, Shaw, Pittman,
Potts and
[[Page 45175]]
Trowbridge, 2300 N Street, NW., Washington, DC 20037
NRC Project Director: Ledyard B. Marsh
Commonwealth Edison Company, Docket Nos. STN 50-456 and STN 50-457,
Braidwood Station, Units 1 and 2, Will County, Illinois Docket Nos.
STN 50-454 and STN 50-455, Byron Station, Units 1 and 2, Ogle
County, Illinois Docket Nos. 50-237 and 50-249, Dresden Nuclear
Power Station,Units 2 and 3, Grundy County, Illinois Docket Nos.
50-373 and 50-374, LaSalle County Station, Units 1 and 2, LaSalle
County, Illinois Docket Nos. 50-254 and 50-265, Quad Cities Nuclear
Power Station, Units 1 and 2, Rock Island County, Illinois Docket
Nos. 50-295 and 50-304, Zion Nuclear Power Station, Units 1 and 2,
Lake County, Illinois
Date of application for amendment requests: April 24, 1995
Description of amendment requests: The licensee proposes to amend
Section 6 of the Technical Specifications of all ComEd stations to make
the following changes: (1) delete the ``Review, Investigative and Audit
Functions'' sections, in their entirety, and relocate these
requirements to appropriate sections of the ComEd Quality Assurance
Topical Report, (2) change titles to reflect the reorganization of
ComEd's Nuclear Operations Division, and (3) miscellaneous
administrative and editorial changes.
Basis for proposed no significant hazards consideration
determination: As requied 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.
(1) The proposed relocation of the ``Review, Investigative and
Audit Functions'' sections of Technical Specifications to the QA
Topical Report does not affect any accident initiators or
precursors, and does not change or alter the design assumptions for
the systems and components used to mitigate the consequences of an
accident.
The relocation of these sections is consistent with the
recommended changes specified in the October 25, 1993 letter from W.
T. Russell (USNRC) to the Chairpersons of the Owner Groups'
Technical Specifications Committees, entitled, ``Content of Standard
Technical Specifications, Section 5.0, Administrative Controls''.
Relocating these requirements to the QA Topical Report will
continue to ensure that proposed future changes to these
requirements will receive proper regulatory oversight. NRC review of
the Quality Assurance Program is governed by 10CFR50.54.
10CFR50.54(a)(3) states: ``Changes to the quality assurance program
description that do not reduce the commitments must be submitted to
the NRC in accordance with the requirements of 50.71. Changes to the
quality assurance program description that do reduce the commitments
must be submitted to NRC and receive NRC approval prior to
implementation, ...'' Based on these 10CFR50.54 requirements,
appropriate licensee and regulatory control of the requirements in
the subject relocated Technical Specification sections will be
maintained.
(2) The proposed title and organizational 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.
Commonwealth Edison's organizational changes allow for increased
senior management attention and oversight of station activities.
Position titles and associated responsibilities have changed to
increase the company's efficiency in the management of its nuclear
stations. These administrative changes do not reduce any
requirements or commitments. The proposed changes enhance the
administrative controls necessary to ensure safe plant operation.
(3) Other proposed administrative/editorial changes simply make
corrections or provide needed clarification prompted by the
reorganization. These changes provide consistency with station
procedures, programs, other Technical Specifications, and Standard
Technical Specifications. They are administrative in nature and do
not impact any accident previously evaluated in the UFSAR.
In conclusion, none of the proposed changes involve a
significant increase in the probability or consequences of an
accident previously evaluated.
B. The proposed changes do not create the possibility of a new
or different kind of accident from any accident previously
evaluated.
(1) The proposed relocation of the ``Review, Investigative and
Audit Functions'' sections of Technical Specifications to the QA
Topical Report does 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 that could give rise to a new or
different kind of accident that was previously evaluated.
The proposed changes are considered to be administrative or
programmatic in nature and do not affect equipment or components
that could initiate an accident. All administrative commitments
being relocated to the QA Topical Report will continue to receive
appropriate regulatory oversight pursuant to 10CFR50.54.
(2) The proposed title and organization 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
that will enhance the performance of organizations responsible for
the safe operation of the plant to respond to plant transients or
emergencies. All responsibilities described in Technical
Specifications for management activities will continue to be
performed by qualified individuals.
(3) All other proposed changes are administrative in nature and
do not create the possibility of a new or different kind of accident
from any accident previously evaluated.
In conclusion, the proposed changes do not create the
possibility of a new or different kind of accident from any accident
previously evaluated.
C. The proposed changes do not involve a significant reduction
in a margin of safety.
(1) The proposed changes are administrative or programmatic in
nature and do not affect the margin of safety for any safety
parameters and setpoints addressed in Technical Specifications. The
assumptions, initial conditions and methodologies used in the
accident analyses remain unchanged, therefore, accident analyses
results are not impacted.
Placing these requirements in QA Topical Report will continue to
ensure that proposed future changes to these requirements will
receive proper regulatory oversight pursuant to 10CFR50.54.
(2) The proposed title and organizational changes are
administrative in nature and 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 analyses results are not impacted.
(3) All other proposed changes are administrative in nature and
have no impact on the margin of safety for any Technical
Specification.
In conclusion, 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: for Braidwood, the Wilmington
Public Library, 201 S. Kankakee Street, Wilmington, Illinois 60481; for
Byron, the Byron Public Library District, 109 N. Franklin, P.O. Box
434, Byron, Illinois 61010; for Dresden, Morris Area Public Library
District, 604 Liberty Street, Morris, Illinois 60450; for LaSalle,
Jacobs Memorial Library, Illinois Valley Community College, Oglesby,
Illinois 61348; for Quad Cities, Dixon Public Library, 221 Hennepin
Avenue, Dixon, Illinois 61021; for Zion, Waukegan Public Library, 128
N. County Street, Waukegan, Illinois 60085
[[Page 45176]]
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, Illinois Docket Nos.
STN 50-456 and STN 50-457, Braidwood Station, Unit Nos. 1 and 2,
Will County, Illinois Docket Nos. 50-373 and 50-374, LaSalle County
Station, Units 1 and 2, LaSalle County, Illinois
Date of amendment request: June 8, 1995
Description of amendment request: The proposed amendments would
revise Technical Specifications Section 3/4.8, Electrical Power
Systems, and the associated Bases for LaSalle County, Byron, and
Braidwood Stations. The proposed changes revise surveillance and
administrative requirements associated with emergency diesel generators
(EDGs) in accordance with the guidance of NRC Generic Letter 94-01,
``Removal of Accelerated Testing and Special Reporting Requirements for
Emergency Diesel Generators,'' Generic Letter 93-05, ``Line-Item
Technical Specifications Improvements to Reduce Surveillance
Requirements for Testing During Power Operation,'' and Regulatory Guide
(RG) 1.9, ``Selection, Design, Qualification, and Testing of Emergency
Diesel Generator Units Used as Class 1E Onsite Electric Power Systems
at Nuclear Power Plants.'' The proposed changes include: (1)
eliminating increased testing requirements for EDGs, (2) eliminating
special reporting requirements for EDGs, (3) eliminating the semi-
annual fast load test and replacing it with a requirement to load EDGs
semi-annually in accordance with the vendor recommendations for all
test purposes other than the refueling outage Loss of Offsite Power
(LOOP) tests, (4) de-coupling the 24-hour endurance run and the LOOP/
loss-of-coolant (LOCA) (LOOP only for LaSalle) sequencing requirements
for the hot start test, (5) removing RG 1.108 references to testing
requirements, (6) eliminating testing requirements when an EDG becomes
inoperable due to an inoperable support system, an independently
testable component, or preplanned maintenance or testing, or if there
is not a potential common mode failure for the remaining diesel
generator, (7) deleting the requirement for inspecting the EDGs in
accordance with procedures prepared in conjunction with its
manufacturer's recommendations, and (8) making editorial changes.
Basis for proposed no significant hazards consideration
determination: As requied 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 do not affect accident initiators or
precursors and do not alter the design assumptions affecting the
ability of the EDGs to mitigate the consequences of an accident.
Deleting the special reporting requirements from the Technical
Specifications is administrative. ComEd will continue to notify the
Commission of significant EDG failures in accordance with 10 CFR
50.72 and 50.73 criteria.
Excessive testing requirements have proven to be a contributor
to increased equipment degradation. Removing inappropriate and
redundant requirements increases EDG reliability and enhances the
ability of EDGs to mitigate the consequences of an accident.
Implementing ComEd's alternative to the maintenance rule for the
EDGs provides additional assurance that high EDG performance will be
maintained.
EDG equipment degradation will be reduced by eliminating the
semi-annual fast load test for EDGs in accordance with the vendor
recommendations for test purposes other than the refueling outage
Loss of Offsite Power (LOOP) tests. This improves EDG reliability
and availability and further enhances their ability to mitigate the
consequences of an accident. The LOOP test would still be performed
to provide assurance that the EDG is capable of responding to a LOOP
as assumed in the accident analyses.
De-coupling the 24 hour endurance test and the LOOP/LOCA (for
LaSalle, LOOP) sequencing test requirements for the hot start test
has no effect on accident mitigation. Demonstrating diesel generator
hot restart capability without loading the engine does not
invalidate or reduce the effectiveness of the hot restart test. The
hot restart test can be conducted in any plant condition since its
performance at power will have no adverse effect on plant
operations.
The proposed editorial changes are administrative in nature.
They improve readability and provide consistency with current
industry guidance.
Therefore, the proposed changes do not involve an increase in
the probability or consequences of an accident previously evaluated.
2) Create the possibility of a new or different kind of accident
from any accident previously evaluated:
The proposed changes do not alter the ability of the EDGs to
perform their intended function to mitigate the consequences of an
initiating event within the acceptance limits assumed in plant
safety analyses. The proposed changes have no impact on component or
system interactions, or the plant design basis.
Instrumentation setpoints, starting, sequencing, and loading
functions associated with EDGs are not affected by the proposed
changes. Furthermore, combining the alternate EDG system maintenance
rule implementation program with the proposed amendment will enhance
both the availability and the performance of the EDGS.
Therefore, there is not a potential for creating the possibility
of a new or different type of accident from any accident previously
evaluated.
3) Involve a significant reduction in a margin of safety:
The proposed changes do not increase the probability or
consequences of an accident, and there is no impact on equipment
design or operation. The proposed changes do not affect the results
of accident and transient analyses. Plant and system response to an
initiating event will remain in compliance within the assumptions of
safety analyses. There is no associated change to the type, amount,
or control of radioactive effluents, nor is there an associated
increase in individual or cumulative occupational radiation
exposure. There is no effect upon the capabilities of the associated
systems to perform their intended functions within the allowed
response times assumed in safety analyses.
The proposed changes are compatible with plant operating
experience and are consistent with the guidance provided in NUREG-
1366, Generic Letters 93-05 and 94-01, and Regulatory Guide 1.9. In
two instances ComEd's proposed changes deviate from these guidance
documents. However, the changes are consistent with the intent of
the documents or other NRC guidance documents. Eliminating excessive
testing requirements can improve safety by reducing challenges to
plant systems and reducing equipment wear and degradation. While the
proposed changes affect surveillance intervals; there are no changes
to the methods used to perform the surveillances.
EDG reliability and availability will be improved by the
proposed changes. The surveillances will continue to demonstrate the
ability of the EDGs to perform their intended function of providing
electrical power to the emergency safety systems needed to mitigate
design basis transients. No margin of safety is reduced.
Guidance has been provided in ``Final Procedures and Standards
on No Significant Hazards Considerations,'' Final Rule, 51 FR 7744,
for the application of standards to license change requests for
determination of the existence of significant hazards
considerations. This document provides examples of amendments which
are and are not considered likely to involve significant hazards
considerations. These proposed amendments most closely fit the
example of a change which may either result in some increase to the
probability or consequences of a previously analyzed accident or may
reduce in some way a safety margin, but where the results of the
change are clearly within all acceptance criteria with respect to
the system or component specified in the standard review plan.
This proposed amendment does not involve a significant
relaxation of the criteria used to establish safety limits, a
significant
[[Page 45177]]
relaxation of the bases for the limiting safety system settings, or a
significant relaxation of the bases for the limiting conditions for
operations. The proposed change does not reduce the margin of safety
as defined in the basis for any Technical Specification.
Therefore, based on the guidance provided in the Federal
Register and the criteria established in 10 CFR 50.92(c), ComEd has
concluded that the proposed change does not constitute 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
requested amendments involve no significant hazards consideration.
Local Public Document Room location: For Byron, the Byron Public
Library District, 109 N. Franklin, P.O. Box 434, Byron, Illinois 61010;
for Braidwood, the Wilmington Public Library, 201 S. Kankakee Street,
Wilmington, Illinois 60481; for LaSalle, Jacobs Memorial Library,
Illinois Valley Community College, Oglesby, Illinois 61348
Attorney for licensee: Michael I. Miller, Esquire; Sidley and
Austin, One First National Plaza, Chicago, Illinois 60603
NRC Project Director: Robert A. Capra
Commonwealth Edison Company, Docket Nos. 50-237 and 50-249, Dresden
Nuclear Power Station, Units 2 and 3, Grundy County, Illinois
Docket Nos. 50-254 and 50-265, Quad Cities Nuclear Power Station,
Units 1 and 2, Rock Island County, Illinois
Date of application for amendment requests: August 30, 1994, as
supplemented August 4, 1995.
Description of amendment requests: As a result of findings by a
Diagnostic Evaluation Team inspection performed by the NRC staff at the
Dresden Nuclear Power Station in 1987, Commonwealth Edison Company
(ComEd, the licensee) made a decision that both the Dresden Nuclear
Power Station and sister site Quad Cities Nuclear Power Station needed
attention focused on the existing custom Technical Specifications (TS)
used.
The licensee made the decision to initiate a Technical
Specification Upgrade Program (TSUP) for both Dresden and Quad Cities.
The licensee evaluated the current TS for both Dresden and Quad Cities
against the Standard Technical Specifications (STS) contained in NUREG-
0123, ``Standard Technical Specifications General Electric Plants BWR/
4.'' The licensee's evaluation identified numerous potential
improvements such as clarifying requirements, changing TS to make them
more understandable and to eliminate interpretation, and deleting
requirements that are no longer considered current with industry
practice. As a result of the evaluation, ComEd has elected to upgrade
both the Dresden and Quad Cities TS to the STS contained in NUREG-0123.
The TSUP for Dresden and Quad Cities is not a complete adaption of
the STS. The TSUP focuses on (1) integrating additional information
such as equipment operability requirements during shutdown conditions,
(2) clarifying requirements such as limiting conditions for operation
and action statements utilizing STS terminology, (3) deleting
superseded requirements and modifications to the TS based on the
licensee's responses to Generic Letters (GL), and (4) relocating
specific items to more appropriate TS locations.
The August 30, 1994, and August 4, 1995, applications proposed to
upgrade only Section 3/4.2 (Instrumentation) of the Dresden and Quad
Cities TS.
Basis for proposed no significant hazards consideration
determination: As requied by 10 CFR 50.91(a), the licensee has provided
its analysis of the issue of no significant hazards consideration,
which is presented below:
1) The proposed changes do not involve a significant increase in
the probability or consequences of an accident previously evaluated
because:
In general, the proposed amendment represents the conversion of
current requirements to a more generic format, or the addition of
requirements which are based on the current safety analysis.
Implementation of these changes will provide increased reliability
of equipment assumed to operate in the current safety analysis, or
provide continued assurance that specified parameters remain within
their acceptance limits, and as such, will not significantly
increase the probability or consequences of a previously evaluated
accident.
Some of the proposed changes to the current Technical
Specifications (CTS) represent minor curtailments of the current
requirements which are based on generic guidance or previously
approved provisions for other stations. The proposed amendment for
Dresden and Quad Cities Station's Technical Specification Section 3/
4.2 are based on BWR-STS (NUREG-0123, Revision 4 ``Standard
Technical Specifications General Electric Plants BWR/4) guidance or
NRC accepted changes at later operating BWR plants. Any deviations
from BWR-STS and CTS requirements do not significantly increase the
probability or consequences of any previously evaluated accident for
Dresden and Quad Cities Station. These proposed changes are
consistent with the current safety analyses and have been previously
determined to represent sufficient requirements for the assurance
and reliability of equipment assumed to operate in the safety
analysis, or provide continued assurance that specified parameters
remain within their acceptance limits. As such, these changes will
not significantly increase the probability or consequences of a
previously evaluated accident.
The associated systems that make up the Instrumentation Systems
are not assumed in any safety analysis to initiate any accident
sequence for both Dresden and Quad Cities Stations; therefore, the
probability of any accident previously evaluated is not increased by
the proposed amendment. In addition, the proposed surveillance
requirements for the proposed amendments to these systems are
generally more prescriptive than the current requirements specified
within the Technical Specifications. These more prescriptive
surveillance requirements increase the probability that the
Instrumentation Systems will perform their intended functions.
Therefore, the proposed TS will improve the reliability and
availability of all affected systems and reduce the consequences of
any accident previously evaluated.
2. Create the possibility of a new or different kind of accident
from any previously evaluated because:
In general, the proposed amendment represents the conversion of
current requirements to a more generic format, or the addition of
requirements which are based on the current safety analysis. Others
represent minor curtailments of the current requirements which are
based on generic guidance or previously approved provisions for
other stations. These changes do not involve revisions to the design
of the station, other than technically valid trip setpoint changes.
Some of the changes may involve revision in the operation of the
station; however, these changes provide additional restrictions
which are in accordance with the current safety analyses, or are to
provide for additional testing or surveillances which will not
introduce new failure mechanisms beyond those already considered in
the current safety analyses. Therefore, these changes will not
create the possibility of a new or different kind of accident from
any accident previously evaluated.
The proposed amendment for Dresden and Quad Cities Station's
Technical Specification Section 3/4.2 is based on BWR-STS guidelines
or NRC accepted changes at later operating BWR plants. The proposed
amendment has been reviewed for acceptability at the Dresden and
Quad Cities Nuclear Power Stations considering similarity of system
or component design versus the BWR-STS or later operating BWRs. Any
deviations from BWR-STS or CTS requirements do not create the
possibility of a new or different kind of accident than previously
evaluated for Dresden and Quad Cities Stations. No new modes of
operation are introduced by the proposed changes. Various
surveillance requirements are changed to reflect improvements in
technique, frequency of performance or operating experience at later
plants. Proposed changes to action statements in many places add
requirements that are not in the present technical specifications or
adopt
[[Page 45178]]
requirements that have been used at other operating BWRs with designs
similar to Dresden and Quad Cities. The proposed changes maintain at
least the present level of operability. Therefore, the proposed
changes do not create the possibility of a new or different kind of
accident from any previously evaluated.
The associated systems that make up the Instrumentation Systems
are not assumed in any safety analysis to initiate any accident
sequence for Dresden or Quad Cities Stations. In addition, the
proposed surveillance requirements for affected systems associated
with the Instrumentation Systems are generally more prescriptive
than the current requirements specified within the Technical
Specifications; therefore, the proposed changes do not create the
possibility of a new or different kind of accident from any
previously evaluated.
3. Involve a significant reduction in the margin of safety
because:
In general, the proposed amendment represents the conversion of
current requirements to a more generic format, or the addition of
requirements which are based on the current safety analysis. Others
represent minor curtailments of the current requirements which are
based on generic guidance or previously approved provisions for
other stations. Some of the later individual items may introduce
minor reductions in the margin of safety when compared to the
current requirements. However, other individual changes are the
adoption of new requirements which will provide significant
enhancement of the reliability of the equipment assumed to operate
in the safety analysis, or provide enhanced assurance that specified
parameters remain within their acceptance limits. These enhancements
compensate for the individual minor reductions, such that taken
together, the proposed changes will not significantly reduce the
margin of safety.
The proposed amendment to Technical Specification Section 3/4.2
implements present requirements in accordance with the guidelines
set forth in the BWR-STS. Any deviations from BWR-STS and CTS
requirements do not significantly reduce the margin of safety for
Dresden and Quad Cities Stations. The proposed changes are intended
to improve readability, usability, and the understanding of
technical specification requirements while maintaining acceptable
levels of safe operation. The proposed changes have been evaluated
and found to be acceptable for use at Dresden and Quad Cities based
on system design, safety analysis requirements and operational
performance. Since the proposed changes are based on NRC accepted
provisions at other operating plants that are applicable at Dresden
and Quad Cities and maintain necessary levels of system or component
readability, the proposed changes do not involve a significant
reduction in the margin of safety.
The proposed amendment for Dresden and Quad Cities Stations will
not reduce the availability of systems associated with the
Instrumentation Systems when required to mitigate accident
conditions; therefore, the proposed changes do not involve a
significant reduction in the margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment requests involve no significant hazards consideration.
Local Public Document Room location: for Dresden, Morris Public
Library, 604 Liberty Street, Morris, Illinois 60450; for Quad Cities,
Dixon Public Library, 221 Hennepin Avenue, Dixon, Illinois 61021
Attorney for licensee: Michael I. Miller, Esquire; Sidley and
Austin, One First National Plaza, Chicago, Illinois 60690
NRC Project Director: Robert A. Capra
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: March 8, 1995, as supplemented June 1,
1995
Description of amendment request: The proposed amendments would
revise the secondary undervoltage setpoint.
Basis for proposed no significant hazards consideration
determination: As requied 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 an increase in the
probability of occurrence or consequences of any accident previously
evaluated.
The proposed amendment does not change the fundamental function
or capability of the Secondary Undervoltage protection as described
in UFSAR section 8.3. Inadvertent or spurious operation of the
Secondary Undervoltage protection function will initiate loading of
the safe shutdown loads on the diesel generators and is not assumed
to initiate an accident. The proposed Secondary Undervoltage
setpoints are low enough to prevent spurious actuations given the
expected off site grid voltages.
This change does not affect the initiators or precursors of any
accident previously evaluated. This change will not increase the
likelihood that a transient initiating event will occur because
transients are initiated by equipment malfunction and/or
catastrophic system failure. The change in setpoints for the
Secondary Undervoltage protection system does involve some changes
to existing plant equipment (such as transformer tap changes and
Circulating Water pump excitation circuit changes). However, all
changes to existing plant equipment have been or will be evaluated
in accordance with the requirements of 10CFR50.59 prior to
installation, to determine that no unreviewed safety questions exist
with regard to the plant changes.
Since any design changes have been or will be determined to be
acceptable per 10CFR50.59 prior to installation and no new plant
equipment will be installed, the probability of occurrence of
accidents previously evaluated will not increase.
With Zion Station's new Auxiliary Power System configuration and
the proposed Secondary Undervoltage setpoints, the probability of a
Loss of Off-Site Power (LOOP) is actually reduced since the original
Auxiliary Power System configuration and Secondary Undervoltage
setpoints required a higher grid voltage to ensure that safety
related loads would be powered from Off-Site power sources during a
design basis accident.
The consequences of accidents previously evaluated are not
increased. The proposed change does not affect the required level of
availability or systems required to mitigate the accidents
considered in the Analyses. Administrative controls will be in place
to ensure that the installed setpoints are low enough to ensure that
the Emergency Diesel Generators are not unnecessarily challenged.
The proposed changes will increase the level of confidence that the
ESF equipment will be capable of starting and operating during a
design basis accident with degraded off-site grid voltage. The
increase in the level of confidence is the result of the more
rigorous methodology used to determine limited ESF bus voltages,
given the minimum expected off-site AC voltage. Based on the
previous discussion, it is determined that there will be no
significant increase in the consequences of any accident previously
evaluated.
2. The proposed amendment does not create the possibility of a
new or different kind of accident from any previously analyzed.
The proposed Secondary Undervoltage setpoint change does not
change the design of the Secondary Undervoltage protection system or
its function to protect against degraded offsite power. Actuation of
the Secondary Undervoltage protection system will initiate a
sequence of events that will start the Emergency Diesel Generator
(EDG) for the associated ESF bus, strip all loads from the bus, open
all feed breakers to the bus, close the Emergency feed breaker (thus
energizing the bus from the EDG), and initiate sequenced starting of
the Safe Shutdown equipment supplied by the bus, including a Service
Water pump, Component Cooling Water pump, Auxiliary Feedwater pump,
and Reactor Containment Fan Cooler(s), as applicable.
The proposed change does not involve the addition of any new or
different types of equipment, nor does it involve the operation of
equipment required for safe operation of the facility in a manner
different from those addressed in the Final Safety Analysis Report.
No safety related equipment or function will be altered as a result
of this proposed change. Because no new failure modes are
introduced, the proposed amendment does not create a new or
different kind of accident from any previously analyzed in the
UFSAR.
Based on the above discussion, the proposed amendment does not
create a new or different kind of accident from any previously
analyzed in the UFSAR.
3. The proposed changes do not involve a significant reduction
in a margin of safety.
[[Page 45179]]
The proposed amendment will allow the Secondary Undervoltage
setpoint to be conservatively established based on new engineering
calculations which consider the lowest expected offsite grid voltage
and operation of all required ESF equipment under design basis
accident loading conditions.
The proposed Secondary Undervoltage setpoints will provide
increased confidence that adequate bus voltage will be available to
support starting and operation of all required ESF loads. The
proposed setpoint includes worst case instrument error to ensure
that the lowest possible voltage will not be lower than the degraded
voltage analytical limits. Additionally, the proposed setpoints are
low enough to prevent spurious actuations due to expected
fluctuations in the grid voltage. The new setpoints are based on a
minimum expected grid voltage of 343 kV, with added margin. The
proposed changes will provide an increase in the level of protection
that currently exists and will ensure the margin of safety is
adequately maintained.
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
requested amendments involve no significant hazards consideration.
Local Public Document Room location: Waukegan Public Library, 128
N. County Street, Waukegan, Illinois 60085
Attorney for licensee: Michael I. Miller, Esquire; Sidley and
Austin, One First National Plaza, Chicago, Illinois 60603
NRC Project Director: Robert A. Capra
Connecticut Yankee Atomic Power Company, Docket No. 50-213, Haddam
Neck Plant, Middlesex County, Connecticut
Date of amendment request: August 3, 1995
Description of amendment request: The proposed amendment will add
an one-time footnote to Technical Specification (TS) Section 3/4.7.12,
``Ultimate Heat Sink,'' to increase the allowed outage time from 6
hours to 18 hours for the months of August and September. In addition,
also for the months of August and September, the maximum service water
limit will be elevated from 90 deg.F to 95 deg.F.
Basis for proposed no significant hazards consideration
determination: As requied by 10 CFR 50.91(a), the licensee has provided
its analysis of the issue of no significant hazards consideration,
which is presented below:
1. Involve a significant increase in the probability or
consequences of an accident previously analyzed.
The proposed addition of a 12 hour time period to monitor the
ultimate heat sink temperature to the Technical Specification
Limiting Condition for Operation action statements does not involve
an increase in the probability of an accident previously evaluated.
The probability of an accident previously evaluated is not increased
by a short-term increase in the ultimate heat sink temperature. An
evaluation has been performed that safe shutdown will be achieved
and maintained for a loss of normal AC power event with the
additional consideration of a single failure with service water
inlet temperatures as high as 95 deg.F. In addition, an evaluation
of the credible FSAR Chapter 15 events with AC power available and
no isolation of non-essential service water loads has been performed
that demonstrates that safe shutdown will be achieved and
maintained. There has been no significant increase in the
consequences of these events previously evaluated.
2. Create the possibility of a new or different kind of accident
from any previously analyzed.
The proposed technical specification change does not create the
possibility of a new or different kind of accident previously
analyzed. The addition of a 12 hour time period to monitor the
ultimate heat sink temperature increases the amount of time that is
allowed for the plant to be in Hot Standby from 6 to 18 hours should
the ultimate heat sink temperature increase above 90 deg.F. This
extension of the time allowed for the plant to be in Hot Standby
does not change the plant configuration. As such, the change does
not create the possibility of a new or different kind of accident
previously evaluated.
3. Involve a significant reduction in the margin of safety.
The proposed technical specification change does not involve a
significant reduction in the margin of safety. The addition of a 12
hour time period to monitor the ultimate heat sink temperature
increases the time required for the plant to be in Hot Standby from
6 to 18 hours should the ultimate heat sink temperature exceed
90 deg.F. An evaluation has been performed to demonstrate that the
risk significance associated with the increased action time is very
low. In addition, safe shutdown capability has been demonstrated for
service water inlet temperatures as high as 95 deg.F.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: Russell Library, 123 Broad
Street, Middletown, CT 06457
Attorney for licensee: Lillian M. Cuoco, Esq., Senior Nuclear
Counsel, Northeast Utilities Service Company, P.O. Box 270, Hartford,
CT 06141-0270
NRC Project Director: Phillip F. McKee
Maine Yankee Atomic Power Company, Docket No. 50-309, Maine Yankee
Atomic Power Station, Lincoln County, Maine
Date of amendment request: May 5, 1995
Description of amendment request: The proposed amendment would
change the surveillance frequency of radiation area, and effluent and
process monitors from monthly to quarterly; and the required frequency
for minimum exercise of control element assemblies also from monthly to
quarterly.
Basis for proposed no significant hazards consideration
determination: As requied by 10 CFR 50.91(a), the licensee has provided
its analysis of the issue of no significant hazards consideration. The
NRC staff has reviewed the licensee's analysis against the standards of
10 CFR 50.92(c). The staff's review is presented below:
1. The proposed change does not involve a significant increase
in the probability or consequences of an accident previously
evaluated. Extending surveillance test intervals as proposed will
reduce the probability of inadvertent reactor scrams and ensuing
challenges to safety systems. This is accomplished by reducing the
occasions and thus the total time that the subject systems are
removed from their ``normal'' configuration and placed into the
required ``test'' configuration. In addition, the probability of
test-induced failures, or failures caused by human error, is
likewise decreased. Thus, the proposed change does not involve a
significant increase in the probability or consequences of an
accident previously evaluated.
2. The proposed change does not create the possibility of a new
or different kind of accident from any accident previously
evaluated.
Extending surveillance test intervals as proposed will not
require installation of any new or different equipment, and will not
alter or otherwise modify existing plant equipment. Thus, the
proposed change does not create the possibility of a new or
different kind of accident from any accident previously evaluated.
3. The proposed change does not involve a significant reduction
in a margin of safety.
Independent research has found that equipment failures and
personnel errors during several types of surveillance tests caused a
significant number of reactor scrams and attendant unnecessary
challenges to safety equipment. The results of this research have
been corroborated by the licensee's plant specific operating
experience. The licensee concludes that the reduced test intervals
proposed in this amendment remain sufficient to ensure known
phenomena, such as instrument setpoint drift and random hidden
failures, remain within the assumptions of the safety analysis.
Thus, the proposed change does not involve a significant reduction
in a margin of safety.
Based on this review, it appears that the three standards of 10 CFR
50.92(c)
[[Page 45180]]
are satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: Wiscasset Public Library, High
Street, P.O. Box 367, Wiscasset, ME 04578
Attorney for licensee: Mary Ann Lynch, Esquire, Maine Yankee Atomic
Power Company, 329 Bath Road, Brunswick, ME 04011
NRC Project Director: Phillip F. McKee
North Atlantic Energy Service Corporation, Docket No. 50-443,
Seabrook Station, Unit No. 1, Rockingham County, New Hampshire
Date of amendment request: July 24, 1995.
Description of amendment request: The proposed amendment would
delete Table 3.4-1, ``Reactor Coolant System Pressure Isolation
Valves'' from the Seabrook Station, Unit No. 1 Technical Specification
section 3.4.6.2. Reference to Table 3.4-1 also would be deleted from
Limiting Condition for Operation 3.4.6.2 f and from Surveillance
Requirement 4.4.6.2.2. The information contained in Table 3.4-1 would
be relocated to the Technical Requirements Manual.
Basis for proposed no significant hazards consideration
determination: As requied by 10 CFR 50.91(a), the licensee has provided
its analysis of the issue of no significant hazards consideration. The
NRC staff has reviewed the licensee's analysis against the standards of
10 CFR 50.92(c). The NRC staff's review is presented below.
A. The proposed changes do not involve a significant increase in
the probability or consequences of an accident previously evaluated
(10 CFR 50.92(c)(1)) because they do not in any way alter the
operability or surveillance requirements for pressure isolation
valves. The proposed changes merely delete a listing of valves which
are designated as pressure isolation valves in accordance with the
definition provided in 10 CFR Part 50. Therefore, neither the
probability nor consequences of previously evaluated accidents are
affected.
B. The proposed changes do not create the possibility of a new
or different kind of accident from any accident previously evaluated
(10 CFR 50.92(c)(2)) because they do not affect in any way the
manner by which the facility is operated or make any changes in
structures, systems, or components which could affect the
operational characteristics of the facility.
C. The proposed changes do not involve a significant reduction
in a margin of safety (10 CFR 50.92(c)(3)) because the proposed
changes do not affect the operability requirements or surveillance
testing of any pressure isolation valve and do not affect in any way
the manner by which the facility is operated or involve equipment or
features which affect the operational characteristics of the
facility.
Based on this review, it appears that the three standards of 10 CFR
50.92(c) are satisfied. Therefore, the NRC staff proposes to determine
that the amendment request involves no significant hazards
consideration.
Local Public Document Room location: Exeter Public Library,
Founders Park, Exeter, NH 03833
Attorney for licensee: Lillian M. Cuoco, Esquire, Northeast
Utilities Service Company, Post Office Box 270, Hartford CT 06141-0270
NRC Project Director: Phillip F. McKee
Northeast Nuclear Energy Company (NNECO), Docket No. 50-245,
Millstone Nuclear Power Station, Unit 1, New London County,
Connecticut
Date of amendment request: July 28, 1995
Description of amendment request: The proposed amendment adds
Technical Specifications (TS) to Section 3.10, Refueling and Spent Fuel
Handling. Specifically, the proposed TS (with applicability, action,
and surveillance requirements) will require that: (1) the reactor be
subcritical for at least 100 hours before the start of reactor
refueling operations, (2) the spent fuel pool bulk temperature be
maintained less than or equal to 140 deg.F, and (3) two trains of
shutdown cooling be operable during reactor refueling operations. In
support of the request, NNECO proposes to: (1) use the ORIGEN2 code to
more accurately predict decay heat loads from the spent fuel, (2) use
the ONEPOOL code to credit the effect of evaporative cooling on the
spent fuel pool bulk temperature, and (3) take credit for both trains
of shutdown cooling to assist the spent fuel pool cooling system during
refueling outages. In addition, the proposed amendment modifies the
table of contents and associated Bases section to reflect the changes.
Basis for proposed no significant hazards consideration
determination: As requied by 10 CFR 50.91(a), the licensee has provided
its analysis of the issue of no significant hazards consideration,
which is presented below:
NNECO has reviewed the proposed changes in accordance with
10CFR50.92 and concluded that the changes do not involve a
significant hazards consideration (SHC). The basis for this
conclusion is that the three criteria of 10CFR50.92(c) are not
compromised. The proposed changes do not involve an SHC because the
changes would not:
1. Involve a significant increase in the probability or
consequences of an accident previously analyzed.
The proposed license amendment will allow NNECO to use the
shutdown cooling system (SCS) to assist the spent fuel pool cooling
(SFPC) system to cool the spent fuel pool during refueling outages.
This amendment request does not affect: the number of spent fuel
bundles allowed in the spent fuel pool, spent fuel pool criticality
analysis, structural analysis of the spent fuel pool, or
radiological release scenarios.
The proposed license amendment also allows NNECO to use ORIGEN2
and ONEPOOL codes. The ORIGEN2 code more accurately predicts decay
heat loads from the spent fuel in the spent fuel pool. The ONEPOOL
code credits the effect of evaporative cooling on the spent fuel
pool bulk temperature. The use of these codes will improve the
accuracy of predicting spent fuel pool bulk temperatures during
normal and abnormal refueling scenarios.
The use of the SCS to assist the SFPC system to cool the spent
fuel pool will allow the movement of spent fuel to begin 100 hours
after reactor shutdown. The existing accident analysis for a dropped
spent fuel bundle during refueling bounds this situation as the
analysis assumed a decay time of 24 hours.
The three new proposed technical specifications will provide
sufficient controls on the movement of spent fuel into the spent
fuel pool, bulk temperature of the spent fuel pool and operability
of the shutdown cooling system to operate within analysis
assumptions during refueling operations at Millstone Unit No. 1.
Therefore, based on the above, the use of the SCS to assist the
SFPC system to cool the spent fuel pool during refueling outages,
the use of the ORIGEN2 code, the use of the ONEPOOL code, and the
addition of three technical specifications will not involve a
significant increase in the probability or consequences of an
accident previously analyzed.
2. Create the possibility of a new or different kind of accident
from any previously analyzed.
The proposed license amendment to use the SCS to assist the SFPC
system to cool the spent fuel pool will allow SCS train B to cool
the spent fuel pool in a method similar to train A.
The proposed license amendment to use ORIGEN2 and ONEPOOL codes
to predict spent fuel pool bulk temperatures will increase the
accuracy of analyzing normal and abnormal refueling scenarios.
The three new proposed technical specifications will
sufficiently control refueling operations to support analyzed
accident scenarios.
Therefore, the use of the SCS to assist the SFPC system to cool
the spent fuel pool, the use of the ORIGEN2 code, the use of ONEPOOL
code and the addition of three technical specifications do not
create the possibility of a new or different kind of accident from
any previously analyzed.
3. Involve a significant reduction in the margin of safety.
The proposed license amendment to use the SCS to assist the SFPC
system to cool the spent fuel pool will allow the crediting of the
SCS and SFPC system to remove heat from
[[Page 45181]]
the spent fuel pool during normal refueling scenarios. The analysis
demonstrates that this cooling configuration will maintain the spent
fuel pool bulk temperature below the pool design limit of 140 deg.F
with a postulated single active failure.
The addition of the train B SCS cross-tie does not adversely
affect the existing design basis of the SCS to remove sensible and
decay heat from the reactor water, cool it from 280 deg.F to
125 deg.F within 24 hours, and to maintain the reactor water at
125 deg.F.
The proposed license amendment to use ORIGEN2 and ONEPOOL codes
will improve the accuracy of predicting spent fuel pool bulk
temperatures during normal and abnormal refueling scenarios.
The thermal hydraulic analysis most limiting time to boil
calculation of 5.4 hours for loss of all forced cooling to the spent
fuel pool is consistent with assumed operator response times for
similar scenarios.
The three new proposed technical specifications will ensure that
the margin of safety established by engineering analysis of
refueling operations is maintained.
Therefore, based on the above, the use of the SCS to assist the
SFPC system to cool the spent fuel pool, the use of the ORIGEN2
code, the use of the ONEPOOL code, and the addition of three
technical specifications does not involve a significant reduction in
the margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: Learning Resources Center,
Three Rivers Community-Technical College, 574 New London Turnpike,
Norwich, CT 06360
Attorney for licensee: Lillian M. Cuoco, Esq., Senior Nuclear
Counsel, Northeast Utilities Service Company, P.O. Box 270, Hartford,
CT 06141-0270
NRC Project Director: Phillip F. McKee
Northeast Nuclear Energy Company, et al., Docket Nos. 50-245, 50-
336 and 50-423, Millstone Nuclear Power Station, Unit Nos. 1, 2,
and 3, New London, Connecticut
Date of amendment request: August 4, 1995
Description of amendment request: The proposed license amendments
will modify the Administrative Controls Section (Section 6) of the
Millstone Unit Nos. 1, 2, and 3 Technical Specifications to allow the
Plant Operations Review Committee (PORC) and Site Operations Review
Committee (SORC) to direct its efforts in the review of more critical
safety matters which affect day-to-day operation. This will be
accomplished by the establishment of a Station Qualified Reviewer
Program (SQRP) and the reassignment of certain procedure approvals to
designated managers in lieu of approval by PORC/SORC.
Basis for proposed no significant hazards consideration
determination: As requied by 10 CFR 50.91(a), the licensee has provided
its analysis of the issue of no significant hazards consideration
(SHC), which is presented below:
...These proposed changes do not involve an SHC because the
changes do not:
1. Involve a significant increase in the probability or
consequences of an accident previously evaluated.
These changes are administrative in nature. They do not involve
any modifications to plant systems and do not alter the method of
operation of any plant equipment. The change involves the
establishment of a SQRP for the review of plant procedures, programs
or changes thereto that do not involve a 10CFR50.59 evaluation.
Implementing a SQRP will not result in a degradation of the
current level of procedure review. PORC/SORC will retain the
responsibility for reviewing any document for which a 10CFR50.59
evaluation is required. Personnel selected to be SQRs [Station
Qualified Reviewers] will possess the technical experience and
expertise to provide a thorough technical review as required by
plant procedures. These personnel, and the managers authorized to
approve these procedures, will be designated in writing by the Unit
Director or the Senior Vice President - Millstone Station.
Procedures or classes of procedures that can be reviewed per the
SQRP will be specified in writing by the Unit Director or the Senior
Vice President - Millstone Station. Procedures will receive an
appropriate cross-disciplinary review when necessary.
2. Create the possibility of a new or different kind of accident
from any accident previously evaluated.
The proposed technical specification changes do not change the
design or function of any plant structure, system, or component, nor
do they introduce any new failure modes. As stated above, the
implementation of a SQRP will not degrade the quality of plant
procedures.
There are no modifications to plant structures, systems, or
components associated with these proposed changes, and the operation
of plant equipment and systems remain unchanged. Since the changes
proposed in this license amendment request do not revise existing
plant structures, systems, or components, do not change the manner
in which the plant is operated and, do not change the manner in
which the plant will respond to any design basis accidents, the
proposed changes do not create the possibility of a new or different
kind of accident from any previously analyzed.
3. Involve a significant reduction in a margin of safety.
The changes proposed in this proposed license amendment request
do not affect the ability of any system to perform its safety-
related function. As described above, these proposed changes are
administrative in nature. They do not change any plant operating
parameters or design features and do not reduce the level of
effectiveness of any existing administrative controls. The proposed
change will not result in changes to the bases for any technical
specification. The establishment of the SQRP will continue to
provide for the adequate review of procedures. In addition, another
direct benefit of this program is that the amount of material
presented to PORC/SORC will decrease. The reduction in the amount of
material presented to PORC/SORC for review will allow the PORC/SORC
to focus on safety significant issues. Since none of the assumptions
in the technical specifications bases are affected by the changes
presented in this license amendment request, the margin of safety
which exists in the current technical specifications is not reduced.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: Learning Resources Center,
Three Rivers Community-Technical College, 574 New London Turnpike,
Norwich, CT 06360
Attorney for licensee: Lillian M. Cuoco, Esq., Senior Nuclear
Counsel, Northeast Utilities Service Company, P.O. Box 270, Hartford,
CT 06141-0270
NRC Project Director: Phillip F. McKee
Northern States Power Company, Docket No. 50-263, Monticello
Nuclear Generating Plant, Wright County, Minnesota
Date of amendment request: June 22, 1995
Description of amendment request: The proposed changes modify the
facility requirements for thermal-hydraulic instability avoidance and
protection to address concerns over reactor fuel performance during
instability events. Changes are proposed to the Technical
Specifications to utilize the flow biased Average Power Range Monitor
high neutron flux scram and a power-flow map exclusion region
consistent with one of the NRC approved BWR Owners' Group solutions. In
addition, a change to correct an error in the Average Planar Linear
Heat Generation Rate during single loop operation is also proposed.
Basis for proposed no significant hazards consideration
determination: As requied by 10 CFR 50.91(a), the licensee has provided
its analysis of the issue of no significant hazards
[[Page 45182]]
consideration, which is presented below:
a. The proposed amendment will not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
The implementation of BWR Owner's Group long term stability
solution Option 1-D at Monticello does not modify the assumptions
contained in the existing accident analysis. The use of an exclusion
region and the operator actions required to avoid and minimize
operation inside the region do not increase the possibility of an
accident. Conditions of operation outside of the exclusion region
are within the analytical envelope of the existing safety analysis.
The operator action requirement to exit the exclusion region upon
entry minimizes the probability of an oscillation occurring. The
actions to drive control rods and/or to increase recirculation flow
to exit the region are maneuvers within the envelope of normal plant
evolutions. The flow based scram has been analyzed and will provide
automatic fuel protection in the event of a core wide instability.
Thus, each proposed operating requirement provides defense in depth
for protection from an instability event while maintaining the
existing assumptions of the accident analysis. The proposed change
to the method by which the MAPLHGR [maximum average planar linear
heat-generation rate] is obtained for single loop operation is
consistent with the analysis performed for the Average Power Range
Monitor/Rod Block Monitor Technical Specifications (ARTS) program.
The analysis performed in support of the ARTS program demonstrated
that the limits established assure compliance with fuel limits.
Therefore, this amendment will not cause a significant increase in
the probability or consequences of an accident previously evaluated
for the Monticello plant.
b. The proposed amendment will not create the possibility of a
new or different kind of accident from any accident previously
analyzed.
As stated above, the proposed operating requirements either
mandate operation within the envelope of existing plant operating
conditions or force specific operating maneuvers within those
carried out in normal operation. Since operation of the plant with
all of the proposed requirements is within the existing operating
basis, an unanalyzed accident will not be created through
implementation of the proposed change. Therefore, the proposed
amendment will not create the possibility of a new or different kind
of accident.
c. The proposed amendment will not involve a significant
reduction in the margin of safety.
Each of the proposed requirements for the plant thermal-
hydraulic stability provides a means for fuel protection. The
combination of avoiding possible unstable conditions and the
automatic flow biased reactor scram provides an in-depth means for
fuel protection. Therefore, the individual or combination of means
to avoid and suppress an instability supplements the margin of
safety. The operating limits established for the single loop
operation MAPLHGR provide an acceptable margin of safety as
demonstrated in NEDC-30492, ``Average Power Range Monitor, Rod Block
Monitor and Technical Specification Improvement (ARTS) Program for
Monticello Nuclear Generating Plant-April 1984.'' The proposed
amendment will 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: Minneapolis Public Library,
Technology and Science Department, 300 Nicollet Mall, Minneapolis,
Minnesota 55401
Attorney for licensee: Gerald Charnoff, Esq., Shaw, Pittman, Potts
and Trowbridge, 2300 N Street, NW, Washington, DC 20037
NRC Project Director: John N. Hannon
Northern States Power Company, Docket No. 50-263, Monticello
Nuclear Generating Plant, Wright County, Minnesota
Date of amendment request: July 5, 1995
Description of amendment request: The proposed amendment, part of
the Monticello Surveillance Test Interval/Allowed Outage Time (STI/AOT)
Program, extends the surveillance test intervals and allowable out-of-
service times for selected instrumentation. The proposed changes are
intended to minimize unnecessary testing and remove excessively
restrictive out-of-service times that could potentially degrade overall
plant safety and availability.
Basis for proposed no significant hazards consideration
determination: As requied 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 amendment will not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
The maximum failure frequency change is for the ECCS Actuation
Instrumentation as identified by General Electric topical report
NEDC-30936P-A, and Monticello specific report RE-006. These reports
concluded core damage frequency changed by less than 4% when STIs
were increased to once per 3 months, AOTs for surveillance were
increased to 6 hours, and AOTs for repair were increased to 24
hours. Since this small increase was within the guideline of
acceptability stated in NEDC-30936P-A, and Monticello only proposes
to increase the repair AOT to 12 hours rather than 24 hours, this
amendment will not cause a significant increase in the probability
or consequences of an accident previously evaluated for the
Monticello plant (see RE-006).
The drift analysis determined the associated instrumentation
would not be adversely effected with the longer calibration
intervals. Pertinent process parameters including instrument drift
will still be within acceptance criteria with the longer
surveillance intervals.
The recirculation flow meters and flow instrumentation are not
used in any safety or accident analysis. Therefore, no analysis
would be changed by increasing the calibration interval to once per
cycle.
b. The proposed amendment will not create the possibility of a
new or different kind of accident from any accident previously
analyzed.
These changes only affect the instrument STI and AOT times. No
changes are being made to the functions of the instrumentation.
Therefore, the proposed amendment will not create the possibility of
a new or different kind of accident.
c. The proposed amendment will not involve a significant
reduction in the margin of safety.
These changes will improve the performance of equipment and are
intended to reduce the potential for equipment failures due to
unnecessary testing. The safety limits and the limiting safety
system setpoints will not be affected by these changes. No safety
margins are affected, therefore, the drift will remain within the
margins 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: Minneapolis Public Library,
Technology and Science Department, 300 Nicollet Mall, Minneapolis,
Minnesota 55401
Attorney for licensee: Gerald Charnoff, Esq., Shaw, Pittman, Potts
and Trowbridge, 2300 N Street, NW, Washington, DC 20037
NRC Project Director: John N. Hannon
Omaha Public Power District, Docket No. 50-285, Fort Calhoun
Station,Unit No. 1, Washington County, Nebraska
Date of amendment request: August 4, 1995
Description of amendment request: This proposed amendment would
revise the Technical Specifications (TS) for the requirements for the
containment radiation high signal (CRHS) and the safety injection and
refueling water (SIRW) tank low signal (STLS) contained in TS 2.15,
Tables 2-3 and 2-4. Specification 3.1, Table 3-2 will also be revised
to include administrative changes to the CRHS surveillance
[[Page 45183]]
methods to be consistent with the applicable surveillance functions.
The Basis for Specification 2.15 is being revised to clarify that the
number of installed channels for CRHS is two. The term ``SOURCE CHECK''
is being deleted from the Definitions section.
Basis for proposed no significant hazards consideration
determination: As requied 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 Omaha Public Power District (OPPD) proposes to revise
Technical Specification (TS) 2.15, Table 2-3 by revising the
requirement for placing the Safety Injection Refueling Water (SIRW)
tank low level channel(s) in the tripped condition to placing them
in the bypassed condition. Due to the derived signal, if a channel
was in the tripped condition and a single failure occurred, (that
being one channel of STLS on either A or B circuits), a premature
SIRW tank low signal (STLS) would be generated. During a design
basis accident (DBA) with a valid Containment Pressure High Signal
(CPHS) or Pressurizer Pressure Low Signal (PPLS), this single
failure would prevent the contents of the SIRW tank from being
injected into the reactor coolant system. The resulting logic of
placing the SIRW tank low level channels in BYPASS rather than TRIP
would not cause a premature switchover of the high pressure safety
injection pumps to the containment sump and it would not prevent the
switchover when needed.
OPPD also proposes to revise TS 2.15, Table 2-4, by reducing the
number of minimum operable Containment Radiation High Signal (CRHS)
channels from two to one. This proposed change revises the
requirements of TS 2.15 to coincide with changes to the TS and
Offsite Dose Calculation Manual (ODCM) that were implemented by TS
Amendment 152. The Engineered Safety Feature (ESF) actuation system
supervisory A and B safeguard initiation channels will not be
affected by this proposed TS change. The minimum level of engineered
safeguards performance acceptable for the DBA, (i.e., minimum
safeguards) will continue to be maintained in accordance with IEEE
279 - 1971, ``Criteria for Protection Systems for Nuclear Power
Generating Stations.''
Included in this change are administrative revisions to TS 3.1,
Table 3-2, for replacing the current surveillance methods for
checking and testing the CRHS instrumentation with the defined terms
``CHANNEL CHECK'' and ``CHANNEL FUNCTIONAL TEST,'' respectively.
These proposed revisions are administrative in nature and reflect
TS-defined terminology for the instrumentation surveillance methods
utilized to ensure that the CRHS instrumentation is operable. A
channel check requires a qualitative determination of acceptable
operability by observation of channel behavior during normal plant
operation. A channel functional test requires the injection of a
simulated signal into the channel to verify that it is operable,
including any alarm and/or trip initiating actions. Other proposed
administrative changes include deleting the term ``SOURCE CHECK''
from the TS Definitions section as source check will no longer be
used in the FCS TS and adding verbiage to the TS 2.15 Basis for
clarifying that the number of installed channels for CRHS is two.
Therefore, the proposed change, as described above, would not
increase 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.
There will be no physical alterations to the plant
configuration, changes to setpoint values, or changes to the
implementation of setpoints or limits as a result of the proposed
changes to TS 2.15, Tables 2-3 and 2-4. The proposed revisions to TS
3.1, Table 3-2 are administrative changes to make the TS more
accurately reflect defined terminology and the methods utilized to
ensure that the CRHS instrumentation is operable. The proposed TS
revisions do not require any changes to the present methods of
verifying CRHS instrumentation operability. Therefore, the proposed
change does not create the possibility of a new or different kind of
accident from any previously evaluated.
3. Involve a significant reduction in a margin of safety.
There are no changes to the equipment or plant operations as a
result of the changes being made to the number of minimum operable
CRHS channels. The proposed changes to the STLS will require that
the inoperable channel be placed in BYPASS rather than TRIP. This
action would ensure that a single failure would not cause a
premature safety injection switchover to the containment sump and
would not prevent switchover when needed. Therefore, this proposed
change does not reduce a margin of safety.
The proposed revisions to TS 3.1, Table 3-2 are administrative
changes to make the TS more accurately reflect defined terminology
and the methods utilized to ensure that the CRHS instrumentation is
operable. The proposed TS revisions do not require any changes to
the present methods of verifying CRHS instrumentation operability.
The proposed changes to the Definitions and TS 2.15 Basis sections
are administrative in nature. Therefore, these 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: W. Dale Clark Library, 215
South 15th Street, Omaha, Nebraska 68102
Attorney for licensee: Perry D. Robinson, Winston & Strawn, 1400 L
Street, N.W., Washington, DC 20005-3502
NRC Project Director: William H. Bateman
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 22, 1995
Description of amendment request: The amendments would revise the
Technical Specifications 3.4.1.4 and 3.9.8.2 by deleting footnotes and
associated information regarding Service Water header operation and its
support function for Residual Heat Removal operation. These footnotes
and associated information had been placed in the Technical
Specifications because of the concern about Service Water system piping
integrity in the mid-1980's.
Basis for proposed no significant hazards consideration
determination: As requied by 10 CFR 50.91(a), the licensee has provided
its analysis of the issue of no significant hazards consideration,
which is presented below:
1. Do not involve a significant increase in the probability or
consequences of an accident previously evaluated.
Even though one service water loop will be out for maintenance,
both loops of residual heat removal (RHR) will be kept operable,
consistent with the requirements of STS (NUREG 1431). A minimum of
two RHR, two component cooling (CC), and two service water (SW)
pumps, powered from two different vital busses, will be kept
operable.
Only one component cooling heat exchanger will be operable since
only one service water loop is operable. The CC heat exchangers for
both Units 1 and 2 have a very high reliability. The primary heat
transfer surfaces of the heat exchangers are made of titanium; no
material problems have been experienced in ten years of service.
The remaining active components that, through misoperation,
could potentially defeat RHR capability are, (1) the motor operated
valves in RHR or SW that could develop a ``hot short'' and
subsequently close and (2) the air operated temperature/ flow
control valves of the CC heat exchangers. Additional actions will be
taken to effectively eliminate the possibility of these single point
valves from failing and defeating RHR capability. The motor operator
breakers will be tagged open during MODES 5 and 6, except for
flooding the cavity, when the RHR suction valves must be closed. The
CC Heat Exchanger air operated temperature/flow control valves fail
open, or as is, on loss of air which is the safe position. Operators
will monitor critical temperatures; this equipment is accessible if
any corrective action is required. Thus, with one service water
header out of service, the intent of the
[[Page 45184]]
technical specifications as defined in the bases section (to have a
single failure proof RHR system) is met with the proposed system
configuration. Therefore, the proposed changes do not involve a
significant increase in the probability or consequences of an
accident previously evaluated.
2. Do not create the possibility of a new or different kind of
accident from any previously evaluated.
The catastrophic failure of a moderate energy Class 3 piping
system is not a credible event, based on the upgraded reliability of
the system, the redundancy of active components, the elimination of
single failure points, and on the industry and regulatory positions
established for this type of system. Since SW is a Class 3 moderate
energy system, the only postulated passive failure mode is a leakage
crack. In accordance with Generic Letter (GL) 91-18 and GL 90-05, a
leak in the SW system, following acceptable evaluation, does not
constitute a failure that causes the loss of capability to perform
it's intended safety function. A moderate energy Class 3 piping leak
does not cause the system to be declared inoperable. Therefore, the
proposed changes do not create the possibility of a new or different
type of accident from any previously evaluated.
3. Do not involve a significant reduction in a margin of safety.
RHR redundancy is maintained; no credible single failure point
exists that could cause a nonrecoverable loss of SW. 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: Salem Free Public library, 112
West Broadway, Salem, New Jersey 08079
Attorney for licensee: Mark J. Wetterhahn, Esquire, Winston and
Strawn, 1400 L Street, NW, Washington, DC 20005-3502
NRC Project Director: John F. Stolz
Rochester Gas and Electric Corporation, Docket No. 50-244, R. E.
Ginna Nuclear Power Plant, Wayne County, New York
Date of amendment request: September 15, 1992, as supplemented
April 20, 1993, April 26, 1995, and July 27, 1995.
Description of amendment request: The proposed amendment would
revise Technical Specifications (TSs) 3.1.1.4, 3.1.1.6, and 4.3.4, and
add a Basis to address Generic Letter (GL) 90-06. GL 90-06 represents
the technical resolution of Generic Issue (GI) 70, ``Power Operated
Relief Valve and Block Valve Reliability,'' and GI 94, ``Additional Low
Temperature Overpressure Protection for Light Water Reactors.'' The
resolution of these issues proposes new requirements and TS changes
that enhance the reliability of power-operated relief valves (PORVs)
and block valves along with TS changes that will provide additional
low-temperature overpressure protection (LTOP).
Basis for proposed no significant hazards consideration
determination: As requied by 10 CFR 50.91(a), the licensee has provided
its analysis of the issue of no significant hazards consideration,
which is presented below:
There is no significant increase in the probability or
consequences of an accident previously evaluated because the
accident conditions and assumptions are not significantly affected
by the proposed change.
The proposed change to action statement 3.1.1.4a(i) [proposed to
be renumbered to 3.1.1.6c] to include the removal of power from a
closed block valve will provide additional assurance to preclude any
inadvertent opening of the block valve at a time in which the PORV
may not be operable to assure RCS [reactor coolant system]
integrity.
The provision of the generic letter requires, with one or both
PORV(s) inoperable to initiate shutdown actions if PORV operability
is not restored within 72 hours or 1 hour respectively. RG&E
[Rochester Gas and Electric Corporation] does not address these
shutdown actions, but rather will concentrate on re-establishing
valve operability. If the block valve(s) and power are not removed
within 1 hour shutdown provisions must be initiated. [***].
Proposed action statement 3.1.1.4a(ii) [proposed to be
renumbered to 3.1.1.6d] includes a provision to place the block
valves associated PORV(s) switch in manual control due to an
inoperable block valve(s). This requirement precludes the automatic
opening for an overpressure event to avoid the potential for a
stuck-open PORV at a time that the block valve is open and
inoperable. [***].
The proposed change of maintaining power to closed block valves
could potentially increase the probability of an inadvertent opening
of a block valve. The safety impact is, however, not significant
since the proposed changes are only applicable if the PORV is
inoperable due to excessive seat leakage (proposed action 3.1.1.6b).
[***].
Proposed action statement 3.1.1.6b establishes reactor coolant
pressure boundary integrity for a PORV that has excessive seat
leakage and is therefore considered operable to perform its intended
safety function. [***].
Proposed Surveillance Requirement 4.3.4.3 addresses operability
of the Nitrogen System by demonstration of the PORVs at least once
per 18 months by operating the PORVs through a complete cycle of
full travel. [***].
Based on the above efforts, the proposed amendment does not
involve a significant increase in the probability or consequences of
any accident previously evaluated.
The possibility of a new or different kind of accident from any
previously evaluated is not created. In matters related to nuclear
safety, all accidents continue to bound previous analyses. The
proposed changes do not add or modify any equipment design nor do
the proposed changes involve any significant operational changes to
any plant systems.
The proposed amendment does not involve a significant reduction
in the margin of safety as defined in the basis for any technical
specification because the results of the accident analyses which are
documented in the UFSAR [Updated Final Safety Analysis Report]
continue to bound operation under the proposed changes so that there
is no safety margin reduction. [***].
Therefore, the proposed changes do not involve a significant
reduction in 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: Rochester Public Library, 115
South Avenue, Rochester, New York 14610
Attorney for licensee: Nicholas S. Reynolds, Winston & Strawn, 1400
L Street, NW., Washington, DC 20005
NRC Project Director: Ledyard B. Marsh
Sacramento Municipal Utility District (SMUD), Docket No. 50-312,
Rancho Seco Nuclear Station, Sacramento County, California
Date of amendment request: June 20, 1995 and as amended August 14,
1995
Description of amendment request: The proposed amendment (PA-191)
would permit SMUD to change the Fuel Storage Building load handling
limits to allow placing the shield plugs on the dry shielded cannisters
in order to permit transfer of spent fuel assemblies from the spent
fuel pool (SFP) to the Rancho Seco Independent Spent Fuel Storage
Installation.
Basis for proposed no significant hazards consideration
determination: As requied by 10 CFR 50.91(a), the licensee has provided
its analysis of the issue of no significant hazards consideration,
which is presented below:
PA-191 will not create a significant increase in the probability
or consequences of an accident previously evaluated in the Safety
Analysis Report (SAR), because dropping the dry shielded canister
(DSC) top shield plug over a DSC loaded with 24 spent fuel
assemblies is not considered a credible event. Also, the gantry
crane is designed such
[[Page 45185]]
that it can only handle loads over the SFP cask pit area and can not
move a load over the SFP fuel storage racks.
PA-191 will not create the possibility of a new or different
type of accident than previously evaluated in the SAR, because the
proposed Permanently Defueled Technical Specification heavy load
handling exceptions do not create a new credible accident scenario.
Dropping the DSC top shield plug and damaging spent fuel assemblies
is not considered a credible event.
PA-191 will not involve a significant reduction in the margin of
safety, because the proposed heavy load handling exceptions do not
create a credible accident scenario.
The NRC staff has reviewed the licensee's analyses of June 20, 1995
and August 14, 1995. The August 14 submittal enhanced these analyses by
providing design details regarding the significant safety factors built
into the crane and other lifting hardware. Based on this review, it
appears that the three standards of 50.92(c) are satisfied. Therefore,
the NRC staff proposes to determine that the amendment request involves
no significant hazards consideration.
Local Public Document Room location: Central Library, Government
Documents 828 I Street, Sacramento, CA 95814
Attorney for licensee: Dana Appling, Esq. Sacramento Municipal
Utility District, P. O. Box 15830, Sacramento, CA 95852-1830
NRC Project Director: Seymour H. Weiss
Southern California Edison Company, et al., Docket Nos. 50-361 and
50-362, San Onofre Nuclear Generating Station, Unit Nos. 2 and 3,
San Diego County, California
Date of amendment requests: July 17, 1995
Description of amendment requests: The licensee proposes to revise
surveillance requirements associated with Technical Specifications 3/
4.3.1, ``Reactor Protective Instrumentation,'' and 3/4.3.2,
``Engineered Safety Feature Actuation System Instrumentation.'' The
surveillance interval is to be increased to 120 days for performance of
channel functional tests for certain reactor protective system and
engineered safety feature actuation system instrumentation. The
proposed change also revises Bases 3/4.3.1, ``Reactor Protective and
Engineered Safety Features Actuation System Instrumentation,'' to
reflect the new interval.
Basis for proposed no significant hazards consideration
determination: As requied by 10 CFR 50.91(a), the licensee has provided
its analysis of the issue of no significant hazards consideration,
which is presented below:
1. The proposed change does not involve a significant increase
in the probability or consequences of an accident previously
evaluated.
The proposed change would extend the current sequential Channel
Functional Test (CFT) surveillance interval for Plant Protective
System (PPS) instrumentation and Nuclear Instrumentation (NI). This
change does not involve any changes to plant equipment or operation.
The proposed change actually maintains or decreases the PPS system
unavailability. PPS uncertainty and setpoint modifications will
account for the new surveillance interval. Therefore, the proposed
change will not involve a significant increase in the probability or
consequences of any 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.
This amendment request does not involve any change to plant
equipment or operation. The PPS system is used for monitoring and
mitigation of evaluated accidents. Increasing the availability of
the PPS system, as proposed in this amendment request, will not
create the possibility of a new or different kind of accident from
any previously evaluated.
3. The proposed change does not involve a significant reduction
in a margin of safety.
This amendment does not change the manner in which safety
limits, limiting safety settings, or limiting conditions for
operation are determined. This amendment request will increase
Reactor Protective System and Engineered Safety Features Actuation
System availability. Therefore, this 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 requests involve no significant hazards consideration.
Local Public Document Room location: Main Library, University of
California, P. O. Box 19557, Irvine, California 92713
Attorney for licensee: T. E. Oubre, Esquire, Southern California
Edison Company, P. O. Box 800, Rosemead, California 91770
NRC Project Director: William H. Bateman
Tennessee Valley Authority, Docket Nos. 50-327 and 50-328, Sequoyah
Nuclear Plant, Units 1 and 2, Hamilton County, Tennessee
Date of amendment request: August 7, 1995 (TS 95-12)
Description of amendment request: The proposed change would correct
various errors of an editorial nature that have been identified in the
technical specifications and remove the provisions that have exceeded
their allowed time interval for implementation or the required
conditions no longer exist.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
TVA has evaluated the proposed technical specification (TS)
change and has determined that it does not represent a significant
hazards consideration based on criteria established in 10 CFR
50.92(c). Operation of Sequoyah Nuclear Plant (SQN) in accordance
with the proposed amendment will not:
1. Involve a significant increase in the probability or
consequences of an accident previously evaluated.
The proposed revisions do not change the TS requirements, plant
setpoints or functions, or plant operating practices. These changes
provide clarifications to the existing TSs by correcting editorial
errors and removing provisions that no longer apply in the
specifications. The probability or consequences of an accident will
not be increased by providing the proposed verbiage corrections that
are editorial and nonintent.
2. Create the possibility of a new or different kind of accident
from any previously analyzed.
No plant functions or compliance activities associated with the
TS requirements have been affected by the proposed editorial
changes. Therefore, the possibility of a new or different kind of
accident is not created.
3. Involve a significant reduction in a margin of safety.
The proposed changes will not alter TS setpoint values or
functions. The proposed corrections will enhance the application of
TS requirements and will support the margin of safety provided by
the TSs. Therefore, the margin of safety will not be reduced by the
proposed revisions.
The NRC has reviewed the licensee's analysis and, based on
thisreview, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: Chattanooga-Hamilton County
Library, 1101 Broad Street, Chattanooga, Tennessee 37402
Attorney for licensee: General Counsel, Tennessee Valley Authority,
400 West Summit Hill Drive, ET 11H, Knoxville, Tennessee 37902
NRC Project Director: Frederick J. Hebdon
[[Page 45186]]
Tennessee Valley Authority, Docket Nos. 50-327 and 50-328, Sequoyah
Nuclear Plant, Units 1 and 2, Hamilton County, Tennessee
Date of amendment request: August 7, 1995 (TS 95-17)
Description of amendment request: The proposed change would
relocate the heat flux hot channel factor penalty of two percent from
Surveillance Requirement 4.2.2.2.e.1 to the Core Operating Limits
Report and add a reference to the factor to Specification 6.9.1.14.5.
Also, Specification 6.9.1.14.a.2 would be revised to reference Revision
1A of Westinghouse Commercial Atomic Power (WCAP) 10216-P-A,
``Relaxation of Constant Axial Offset Control - FQ Surveillance
Technical Specifications,'' dated February 1994.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
TVA has evaluated the proposed technical specification (TS)
change and has determined that it does not represent a significant
hazards consideration based on criteria established in 10 CFR
50.92(c). Operation of Sequoyah Nuclear Plant (SQN) in accordance
with the proposed amendment will not:
1. Involve a significant increase in the probability or
consequences of an accident previously evaluated.
The proposed change involves only the manner in which the
penalty factors for FQ(Z) would be specified (i.e, a burnup-
dependent factor specified in the Core Operating Limits Report
[COLR] versus a constant factor specified in the TS). This is simply
used to account for the fact that FQ(Z) may increase between
surveillance intervals. These penalty factors are not assumed in any
of the initiating events for the accident analyses. Therefore, the
proposed change will have no effect on the probability of any
accidents previously evaluated. The penalty factors specified in the
COLR will be calculated using NRC-approved methodology and will
therefore continue to provide an equivalent level of protection as
the existing TS requirement. Therefore, the proposed change will not
affect the consequences of any accident previously evaluated.
2. Create the possibility of a new or different kind of accident
from any previously analyzed.
The proposed change does not involve a physical alteration to
the plant (no new or different kind of equipment will be installed)
or alter the manner in which the plant would be operated. Thus, this
change does not create the possibility of a new or different kind of
accident from any accident previously evaluated.
3. Involve a significant reduction in a margin of safety.
The proposed change will continue to ensure that potential
increases in FQ(Z) over a surveillance interval will be
properly accounted for. The penalty factors will be calculated using
NRC-approved methodology. Therefore, the proposed change will not
involve a reduction in margin of safety.
The NRC has reviewed the licensee's analysis and, based on
thisreview, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: Chattanooga-Hamilton County
Library, 1101 Broad Street, Chattanooga, Tennessee 37402
Attorney for licensee: General Counsel, Tennessee Valley Authority,
400 West Summit Hill Drive, ET 11H, Knoxville, Tennessee 37902
NRC Project Director: Frederick J. Hebdon
Tennessee Valley Authority, Docket Nos. 50-327 and 50-328, Sequoyah
Nuclear Plant, Units 1 and 2, Hamilton County, Tennessee
Date of amendment request: August 7, 1995 (TS 95-18)
Description of amendment request: The proposed change would revise
the titles of various administrative positions found in Section 6.0 of
the 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:
TVA has evaluated the proposed technical specification (TS)
change and has determined that it does not represent a significant
hazards consideration based on criteria established in 10 CFR
50.92(c).
Operation of Sequoyah Nuclear Plant (SQN) in accordance with the
proposed amendment will not:
1. Involve a significant increase in the probability or
consequences of an accident previously evaluated.
The proposed changes only involve the administrative titles of
management positions in TVA [Tennessee Valley Authority]. Plant
equipment and operating practices are not affected by the proposed
administrative changes. Therefore, there is no increase in the
probability or consequences of an accident.
2. Create the possibility of a new or different kind of accident
from any previously analyzed.
Plant features are not impacted by the proposed revision;
therefore, this revision can not create the possibility of a new or
different accident.
3. Involve a significant reduction in a margin of safety.
Plant setpoints and features that establish and maintain the
margin of safety for SQN are not involved in the proposed
administrative TS change. Therefore, the margin of safety is not
reduced by the proposed change.
The NRC has reviewed the licensee's analysis and, based on
thisreview, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.Local
Public Document Romm location: Chattanooga-Hamilton County Library,1101
Broad Street, Chattanooga, Tennessee 37402
Attorney for licensee: General Counsel, Tennessee Valley Authority,
400 West Summit Hill Drive, ET 11H, Knoxville, Tennessee 37902
NRC Project Director: Frederick J. Hebdon
Tennessee Valley Authority, Docket Nos. 50-327 and 50-328, Sequoyah
Nuclear Plant, Units 1 and 2, Hamilton County, Tennessee
Date of amendment request: August 7, 1995 (TS 95-03)
Description of amendment request: The proposed change would modify
Technical Specifications (TS) 3/4.1.3, ``Movable Control Assemblies,''
and Bases 3/4.1.3. The proposed change addresses operation with a rod
urgent failure condition (the control rods are out-of-service because
of failures external to the individual rod drive mechanisms; i.e.,
programming circuitry, but the control rods remain operable), including
limited operation with one control or shutdown bank inserted up to 18
steps below its insertion point. In addition, the surveillance interval
for rod movement verifications would be increased from 31 days to 92
days.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
TVA has evaluated the proposed technical specification (TS)
change and has determined that it does not represent a significant
hazards consideration based on criteria established in 10 CFR
50.92(c).
Operation of Sequoyah Nuclear Plant (SQN) in accordance with the
proposed amendment will not:
1. Involve a significant increase in the probability or
consequences of an accident previously evaluated.
Allowing for continued operation during diagnosis and repair as
a result of electronic or electrical malfunctions of the rod control
system is acceptable, since the design safety function of the
control rods (reactor trip will remain unaffected during the
diagnosis and repair period. During the extended
[[Page 45187]]
troubleshooting and repair period, the requirements for control rod
alignment, insertion limits (except for a small allowed deviation
for one bank) and shutdown margin will be maintained. The small
deviation from the control rod insertion limits allowed for one
bank, for up to 72 hours, will not adversely impact the current TS
requirements for normal operation core power distributions. The
proposed changes do not affect the ability of the control rods to
perform their intended safety function (rods remain trippable) when
a safety system setting is reached. No new or unique accident
precursors be introduced by the proposed changes. Therefore, the
probability and consequences of accidents related to or dependent on
control rod operation will remain unaffected.
The proposed change will result in a small increase in the
probability, that at any given time, a control or shutdown bank will
be inserted slightly below (i.e., up to 18 steps) its insertion
limit. However, by design, the control and shutdown banks will
continue to meet the safety analysis criterion for steady state and
American Nuclear Society (ANS) Condition II (moderate frequency)
transients. The allowed insertion is not a malfunction of equipment
important to safety in this case; therefore, the probability of such
a malfunction is not increased. Limiting the allowed time for
operation with the rod control system out-of-service, but with the
rods trippable and with a control or shutdown bank below the
insertion limit, eliminates the need for consideration of this
condition coincident with any of the low frequency (ANS Condition
III or IV) design basis accidents.
2. Create the possibility of a new or different kind of accident
from any previously analyzed.
There are no new failure mechanisms associated with plant
operation for an extended period to perform diagnosis and repair on
the rod control system. Limited periods of operation with immovable,
but trippable control rods, does not involve any modification to the
operational limits or physical design of the involved systems. There
are no new accident precursors created because of the allowed
diagnosis and repair period.
3. Involve a significant reduction in a margin of safety.
The results of the current accident analyses are not impacted by
the change. In addition, the margin of safety as defined in the
basis of the TS has not been reduced because current core design
limits continue to be met for the accidents of concern. Therefore,
the margin of safety is not impacted.
The NRC has reviewed the licensee's analysis and, based on
thisreview, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: Chattanooga-Hamilton County
Library, 1101 Broad Street, Chattanooga, Tennessee 37402
Attorney for licensee: General Counsel, Tennessee Valley Authority,
400 West Summit Hill Drive, ET 11H, Knoxville, Tennessee 37902
NRC Project Director: Frederick J. Hebdon
Union Electric Company, Docket No. 50-483, Callaway Plant, Unit 1,
Callaway County, Missouri
Date of amendment request: June 23, 1995
Description of amendment request: The proposed amendment would
revise Technical Specification (TS) Surveillance Requirements
4.1.3.1.2, 4.4.6.2.2.b, 4.4.3.2, 4.6.2.1.d, 4.6.4.2, and Table 4.3-3 in
accordance with guidance provided in NRC Generic Letter (GL) 93-05,
``Line Item Technical Specification Improvements to Reduce Surveillance
Requirements for Testing During Power Operations.'' Additionally, the
proposed amendment would revise TS 4.1.1.1.1, 4.1.1.2, 3/4.1.3.1 and
associated Bases to implement portions of the Standard Technical
Specifications - Westinghouse Plants, NUREG-1431.
Basis for proposed no significant hazards consideration
determination: As requied 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 Technical Specification changes do not involve a
significant hazards consideration per 10 CFR 50.92 because operation
of Callaway Plant with the changes would not:
1. Involve a significant increase in the probability or
consequences of an accident previously evaluated.
All changes are in accordance with the recommendations of NRC
Generic Letter 93-05, Line-Item Technical Specifications
Improvements to Reduce Surveillance Requirements for Testing During
Power Operation or NUREG 1431, Standard Technical Specifications -
Westinghouse Plants. None of the changes affects accident initiators
and each has been evaluated against Callaway Plant operating
experience.
2. Create the possibility of a new or different kind of accident
from any previously evaluated.
The proposed Technical Specification changes do not modify any
equipment nor create any potential accident initiators. The changes
per GL 93-05 involve Technical Specification surveillance
frequencies and do not alter the methodology nor associated
acceptance criteria. The changes per NUREG-1431 do not create any
accident initiators and are consistent with Callaway design and
operation.
3. Involve a significant reduction in a margin of safety.
The surveillance frequency changes were recommended via GL 93-05
and are compatible with Callaway Plant experience. The changes per
NUREG-1431 do not impact the margin of safety. The Shutdown margin
requirements and associated safety margins are unaffected by these
changes.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: Callaway County Public
Library, 710 Court Street, Fulton, Missouri 65251
Attorney for licensee: Gerald Charnoff, Esq., Shaw, Pittman, Potts
& Trowbridge, 2300 N Street, NW., Washington, DC 20037
NRC Project Director: Gail H. Marcus
Union Electric Company, Docket No. 50-483, Callaway Plant, Unit 1,
Callaway County, Missouri
Date of amendment request: June 26, 1995
Description of amendment request: The proposed amendment would
revise the allowed outage time for component cooling water motor
operated containment isolation valves, remove the list of containment
isolation valves, and allow containment penetration check valves to be
used as isolation devices.
Basis for proposed no significant hazards consideration
determination: As requied by 10 CFR 50.91(a), the licensee has provided
its analysis of the issue of no significant hazards consideration,
which is presented below:
The proposed revision to TS 3/4.6 to remove the listing of
containment isolation valves, revise the ACTION Statement for the
CCW MOVs, and credit penetration check valves as isolation devices
does not involve a significant hazards consideration because
operation of Callaway Plant with this change would not:
1. Involve a significant increase in the probability or
consequences of an accident previously evaluated.
The proposed changes simplify the TS, meet the regulatory
requirements for control of containment isolation and are consistent
with the guidelines of GL 91-08. The information contained in Table
3.6-1 has not been changed, but only relocated to a different
controlling document. This is an administrative change which should
result in improved plant practices and have no impact on plant
operations. Addition of the footnote to allow up to 12 hours for
valve testing does not affect the severity of any accident
previously evaluated. The proposed revision to the TS will not
adversely impact plant safety since the second barrier of the two
required is still available to provide isolation between the
containment atmosphere or the reactor coolant system and the outside
atmosphere.
[[Page 45188]]
2. Create the possibility of a new or different kind of accident
from any previously evaluated.
There are no design changes being made that would create a new
type of accident or malfunction and the method and manner of plant
operation remain unchanged. Addition of the footnote to allow up to
12 hours for valve testing does not affect the severity of any
accident previously evaluated. The additional time provides
assurance that the inoperable valve is in proper working order prior
to returning it to OPERABLE condition.
3. Involve a significant reduction in a margin of safety.
There are no changes being made to the safety limits or safety
system settings that would adversely impact plant safety.
Containment isolation will still be maintained as provided by the
second isolation valve to ensure that the release of radioactive
material to the environment will be consistent with the assumptions
used in the analyses for a LOCA. This will assure that containment
integrity is maintained.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: Callaway County Public
Library, 710 Court Street, Fulton, Missouri 65251
Attorney for licensee: Gerald Charnoff, Esq., Shaw, Pittman, Potts
& Trowbridge, 2300 N Street, NW., Washington, DC 20037
NRC Project Director: Gail H. Marcus
Union Electric Company, Docket No. 50-483, Callaway Plant, Unit 1,
Callaway County, Missouri
Date of amendment request: July 25, 1995
Description of amendment request: The proposed amendment would
revise Technical Specification (TS) 3/4.8.1 and its associated Bases to
improve overall emergency diesel generator reliability and
availibility.
Basis for proposed no significant hazards consideration
determination: As requied 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 hazards
consideration because operation of Callaway Plant with these changes
would not:
1. Involve a significant increase in the probability or
consequences of an accident previously evaluated.
These proposed changes do not involve a change in the
operational limits or physical design of the emergency power system.
Emergency diesel generator operability and reliability will continue
to be assured while minimizing the number of required emergency
diesel generator starts. Also, emergency diesel generator
reliability will be enhanced by minimizing severe test conditions
which can lead to premature failures.
2. Create the possibility of a new or different kind of accident
from any previously evaluated.
These proposed changes do not involve a change in the
operational limits or physical design of the emergency power system.
The performance capability of the emergency diesel generator will
not be affected. Emergency diesel generator reliability and
availability will be improved by the implementation of the proposed
changes. There is no actual impact on any accident anaiysis.
3. Involve a significant reduction in the margin of safety.
These proposed changes do not involve a change in the
operational limits or physical design of the emergency power system.
The performance capability of the emergency diesel generator will
not be affected. Emergency diesel generator reliability and
availability will be improved by the implementation of the proposed
changes. No margin of safety is reduced.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: Callaway County Public
Library, 710 Court Street, Fulton, Missouri 65251
Attorney for licensee: Gerald Charnoff, Esq., Shaw, Pittman, Potts
& Trowbridge, 2300 N Street, NW., Washington, DC 20037
NRC Project Director: Gail H. Marcus
Virginia Electric and Power Company, Docket Nos. 50-338 and 50-339,
North Anna Power Station, Units No. 1 and No. 2, Louisa County,
Virginia
Date of amendment request: November 29, 1994
Description of amendment request: The proposed change would revise
and update the NA-1&2 Environmental Protection Plan (EPP) to reflect
current obligations to the Commonwealth of Virginia, revise portions of
the transmission corridor rights-of-way erosion control program for
clarification and to be consistent with the state regulations,
eliminate inconsistencies, and delete obsolete material. Specifically,
references to National Pollutant Discharge Elimination System (NPDES)
permits are changed to reflect the correct permit title, Virginia
Pollutant Discharge Elimination System (VPDES). Vegetation and aquatic
biota studies referred to in the EPP were satisfactorily completed on
or before June 24, 1986. The discussion of the detailed subject matter
in these studies is removed because it is extraneous information. A
reference to 10 CFR 51.5(b)(2) (which does not exist) is corrected to
10 CFR 51.60(b)(2). The explicit reporting requirements for unusual or
important environmental events are replaced with the reporting
requirement which the NRC has required pursuant to 10 CFR 50.72
(b)(2)(vi). Therefore, the reporting inconsistency (EPP requires report
to NRC within 24 hours, whereas the 10 CFR 50.72 requires a four hour
report to the NRC) is resolved. The description of the audit program to
be utilized for auditing the EPP is replaced by referring to the Audit
Program established in accordance with 10 CFR 50, Appendix B. Another
inconsistency is eliminated by revising the two year records retention
requirement for erosion control inspection field logs to five years.
This makes the requirement consistent with EPP Section 5.2, Records
Retention. References to the State Water Control Board are updated to
that agency's successor, the Department of Environmental Quality.
Additionally, the licensee's obligation to comply with Virginia
regulations concerning erosion and sediment control within the
transmission corridor rights-of-way are recognized to eliminate
redundancy with previous EPP commitments. The Virginia Soil and Water
Conservation Board is recognized as the regulatory authority concerning
erosion within the transmission corridor rights-of-way. The Virginia
Soil and Water Conservation Board reviews and approves erosion and
sediment control specifications submitted by utilities on an annual
basis.
Basis for proposed no significant hazards consideration
determination: As requied by 10 CFR 50.91(a), the licensee has provided
its analysis of the issue of no significant hazards consideration,
which is presented below:
Specifically, operation of the North Anna Power Station in
accordance with the EPP changes will not:
(1) Involve a significant increase in the probability or
consequences of an accident previously evaluated. The likelihood
that an accident will occur is neither increased or decreased by the
proposed changes to the EPP. Sufficient controls are established to
ensure that environmental controls impacting safety-related
structures, systems, and components are maintained current and
accurate. The only potentially credible accident which might be
affected is the Loss of Offsite Power (if erosion were severe
[[Page 45189]]
enough to undermine the bases of a transmission tower). Each of the
three 500 KV transmission lines connected to North Anna Power
Station can supply sufficient power to the site. This limits the
effect that one transmission tower has on safe operation of the
nuclear facility. However, the erosion noted to date has not been
severe enough to make such an accident credible. Additionally, each
of the 500 KV transmission lines are inspected for material
condition annually. Although the intent of this inspection is not
soil erosion (the annual erosion inspections are currently conducted
by another group who specializes in land management), evidence of
severe erosion would be noted and addressed as appropriate.
Therefore, this EPP change will not impact the function or method of
operation of plant equipment. Thus, a significant increase in the
probability of a previously analyzed accident does not result due to
this change. Nuclear station systems, equipment, or components are
not affected by the proposed changes. Thus, the consequences of a
malfunction of equipment important to safety previously evaluated in
the UFSAR [Updated Final Safety Analysis Report] are not increased
by this change.
(2) Create the possibility of a new or different kind of
accident from any accident previously evaluated. The proposed
changes do not involve changes to the physical plant or operations.
... the proposed EPP changes do not contribute to accident
initiation and therefore do not produce a new accident scenario or
produce a new type of equipment malfunction. Also, this EPP change
does not alter any existing accident scenarios. The proposed changes
do not affect nuclear plant equipment or its operation, and thus do
not create the possibility of a new or different kind of accident.
Therefore, the proposed changes do not create the possibility of a
new or different kind of accident.
(3) Involve a significant reduction in a margin of safety. The
EPP does not have a formal basis description other than the
discussion in the FES-OL [Final Environmental Statement-Operating
License]. The FES-OL discusses the non-radiological impacts of
facility construction and operation on the environment. The
discussion indicates that the environment will be managed to a
stabilized condition during the operations phase, and a program will
be implemented to maintain the environment in a stabilized
condition. This intent is not altered by the proposed changes to the
EPP. The proposed changes do not affect nuclear plant equipment or
its operation, and thus do not involve any reduction in the margin
of safety.
Therefore, use of the proposed EPP would not involve any
reduction in the margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: The Alderman Library, Special
Collections Department, University of Virginia, Charlottesville,
Virginia 22903-2498
Attorney for licensee: Michael W. Maupin, Esq., Hunton and
Williams, Riverfront Plaza, East Tower, 951 E. Byrd Street, Richmond,
Virginia 23219
NRC Project Director: David B. Matthews
Virginia Electric and Power Company, Docket Nos. 50-338 and 50-339,
North Anna Power Station, Units No. 1 and No. 2, Louisa County,
Virginia
Date of amendment request: July 26, 1995
Description of amendment request: The proposed changes would revise
the Technical Specifications (TS) for the North Anna Power Station,
Units No. 1 and No. 2 (NA-1&2). Specifically, the proposed changes
would increase the pressurizer safety valve lift setpoint tolerance as
well as reduce the pressurizer high pressure reactor trip setpoint and
allowable value.
The licensee has prepared a safety evaluation which justifies
increasing the current TS pressurizer safety valve (PSV) at-power
(Modes 1-3) lift setpoint tolerance from plus or minus 1% as-found and
plus or minus 1% as-left to +2%/-3% average as-found with no single
valve outside plus or minus 3% as-found and plus or minus 1% per valve
as-left. The as-found value is based on testing, the results of which
are expressed as an error (i.e., positive or negative percentage
deviation from the nominal lift setpoint). The errors of the tested
valves are summed and the result divided by the number of valves
tested. This result is compared to the acceptable range of +2% to -3%.
No single valve is allowed to be outside of the plus or minus 3%
tolerance.
The safety evaluation also supports an increase to the Hot Shutdown
(Mode-4) required PSV lift setpoint tolerance from plus or minus 1% as-
found and plus or minus 1% as-left to plus or minus 3% per valve as-
found and plus or minus 1% per valve as-left. These proposed changes
will provide greater operational flexibility in meeting periodic test
requirements established by the safety analyses.
A concurrent reduction in the pressurizer high pressure reactor
trip setpoint and allowable value of TS Table 2.2-1 are also proposed.
These changes ensure that the analysis results for the loss of external
load accident continue to meet the acceptance criteria with the higher
PSV tolerance.
The Loss of Load, Locked Rotor, and Rod Withdrawal event analyses
demonstrate that increasing the at-power PSV lift setpoint tolerance to
+2%/-3% average as-found with no single valve outside plus or minus 3%
as-found and plus or minus 1% per valve as-left does not result in a
transient pressure in excess of the overpressure safety limit. Further,
the increased setpoint tolerance does not adversely impact the DNBR
[departure from nucleate boiling ratio] results of any North Anna UFSAR
[Updated Final Safety Analysis Report] Chapter 15 transient analysis.
Mode 4 overpressure protection is adequate with one PSV with a
tolerance of plus or minus 3%.
Finally, the increased PSV setpoint tolerances and reduction of the
high pressurizer pressure reactor trip setpoint do not present any
operational considerations which would significantly impact the
performance of the plant during normal operation or during postulated
accident conditions. In summary, each pertinent safety criterion was
evaluated for the proposed TS changes, and all were found to be
acceptable.
Basis for proposed no significant hazards consideration
determination: As requied by 10 CFR 50.91(a), the licensee has provided
its analysis of the issue of no significant hazards consideration,
which is presented below:
Specifically, operation of North Anna Power Station in
accordance with the proposed Technical Specifications changes will
not:
1. Involve a significant increase in the probability or
consequence of an accident previously evaluated.
Affected safety related parameters were analyzed for a change to
North Anna 1 and 2 Technical Specifications 3.4.2 and 3.4.3 and
Table 2.2-1 item 10. It was determined that the overpressure safety
limits would not be exceeded in the most limiting overpressure
transients (Loss of Load, Locked Rotor, and Rod Withdrawal events)
with the as-found pressurizer safety valve lift setpoint tolerance
increased to an average of +2%/-3%, no single valve outside of [plus
or minus] 3%, and the 25 psi reduction in the Pressurizer High
Pressure Reactor Trip setpoint. The DNBR results of transients
impacted by the proposed setpoint tolerance increase meet the
acceptance criterion after accounting for the impact of the proposed
changes. The increased setpoint tolerance will not result in an
inadvertent opening of the pressurizer safety valves. Mode 4
overpressure protection is adequate with one PSV with a tolerance of
[plus or minus] 3%.
2. Create the possibility of a new or different kind of accident
from any accident previously identified.
The proposed change to North Anna 1 and 2 Technical
Specifications 3.4.2 and 3.4.3 and Table 2.2-1 item 10 does not
involve any changes which would introduce any new or unique
operational modes or accident precursors. Only the allowable
tolerance about the existing PSV lift setpoint will be changed,
along with a reduction in the
[[Page 45190]]
pressurizer high pressure reactor trip setpoint.
3. Involve a significant reduction in a margin of safety.
It was determined that the most limiting overpressure transients
do not result in maximum pressures in excess of the overpressure
safety limits. The DNBR results of transients impacted by the
proposed setpoint tolerance increase meet the acceptance criterion
after accounting for the impact of the proposed changes. Therefore,
the margin of safety is unchanged by the proposed increase in the
safety valve setpoint tolerances.
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: The Alderman Library, Special
Collections Department, University of Virginia, Charlottesville,
Virginia 22903-2498
Attorney for licensee: Michael W. Maupin, Esq., Hunton and
Williams, Riverfront Plaza, East Tower, 951 E. Byrd Street, Richmond,
Virginia 23219
NRC Project Director: David B. Matthews
Virginia Electric and Power Company, Docket Nos. 50-338 and 50-339,
North Anna Power Station, Units No. 1 and No. 2, Louisa County,
Virginia
Date of amendment request: July 26, 1995
Description of amendment request: The proposed change would revise
the Technical Specifications (TS) for the North Anna Power Station,
Units No. 1 and No. 2 (NA-1&2). Specifically, the change would clarify
the TS to allow switching of charging and low-head safety injection
pumps during unit shutdown conditions. The proposed changes would also
allow additional methods of rendering these same pumps incapable of
injecting into the reactor coolant system (RCS) when required for low-
temperature conditions. NA-1&2 is equipped with three charging pumps.
These charging pumps provide inventory control, normal boration to the
RCS, and flow to the reactor coolant pump seals. They also act as the
high-head safety injection pumps during accident conditions. During
certain shutdown conditions, it is necessary to render two of the three
charging pumps inoperable to maintain the low-temperature overpressure
protection (LTOP) design bases assumptions. This provides assurance
that a mass addition pressure transient can be relieved by the
operation of a single pressurizer power-operated relief valve (PORV).
Low-temperature overpressure protection for each NA-1&2 unit is
provided by two pressurizer PORVs.
During shutdown conditions, periodic surveillance testing of the
charging pumps is required by the NA-1&2 TS. Also during shutdown
conditions, it may be desirable to switch from one charging pump to
another to allow for other activities such as maintenance or testing.
The current NA-1&2 TS associated with charging pumps during
shutdown conditions are very restrictive and do not allow sufficient
latitude for surveillance testing or pump switching. The current NA-1&2
TS specifically state in the surveillance requirements that the method
used to render a charging pump inoperable is to place the pump control
switch in the pull-to-lock position. This requirement would not allow
for surveillance or post-maintenance testing of the inoperable charging
pumps since this switch is used to start those pumps.
Therefore, the licensee proposes to modify NA-1&2 TS to allow more
than one charging pump to be operable and capable of injecting into the
RCS for pump switching operations. Additionally, the methods used to
render charging pumps inoperable will be expanded to allow for post-
maintenance and surveillance testing.
Basis for proposed no significant hazards consideration
determination: As requied by 10 CFR 50.91(a), the licensee has provided
its analysis of the issue of no significant hazards consideration,
which is presented below:
Specifically, operation of North Anna Power Station in
accordance with the proposed Technical Specifications changes will
not:
1. Involve a significant increase in the probability or
consequences of an accident previously evaluated.
Allowing more than one charging pump to be operable and capable
of injecting into the RCS during RCS low temperature operation for
pump switching for post-maintenance and surveillance testing does
not increase the probability of occurrence or the consequences of
any previously analyzed accident. Pump switching operations will be
under the direct administrative control of a licensed operator and
will only be for a short duration of time. Any situation that could
result in an excessive RCS mass addition would be immediately
recognized by the operator and remedial action would be taken to
prevent challenges to RCS integrity. Using methods such as opening
the charging pump power supply breaker or closing the charging pump
discharge valve(s) to render a charging pump inoperable will ensure
that these pumps will not be capable of injecting water into the
RCS. These alternate methods are as effective as placing the control
switches in the pull-to-lock position.
2. Create the possibility of a new or different kind of accident
from any accident previously evaluated.
Allowing more than one charging pump to be operable and capable
of injecting into the RCS during low-temperature operation for pump
switching for post-maintenance and surveillance testing does not
involve any physical modifications of the plant nor result in a
change in a method of operation. Licensed operator control of
charging pump switching operations will continue to ensure that the
RCS will not be challenged by excessive mass addition events. Using
methods other than placing charging pump control switches in the
pull-to-lock position to render the pump inoperable will still
ensure that only one pump will be capable of injecting into the RCS
during low temperature operations. Therefore, a new or different
type of accident is not made possible.
3. Involve a significant reduction in a margin of safety.
Allowing more than one charging pump to be operable and capable
of injecting into the RCS during RCS low temperature operation for
pump switching for post-maintenance and surveillance testing does
not affect any safety limits or limiting safety system settings. The
alternate methods of rendering pumps inoperable provide the same
level of assurance that the pump is incapable of flowing into the
RCS as placing the pump control switch in the pull-to-lock position.
System operating parameters remain unaffected. The availability of
equipment required to mitigate or assess the consequence of an
accident is not reduced. Safety margins are, therefore, not
decreased.
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: The Alderman Library, Special
Collections Department, University of Virginia, Charlottesville,
Virginia 22903-2498
Attorney for licensee: Michael W. Maupin, Esq., Hunton and
Williams, Riverfront Plaza, East Tower, 951 E. Byrd Street, Richmond,
Virginia 23219
NRC Project Director: David B. Matthews
Virginia Electric and Power Company, Docket Nos. 50-280 and 50-281,
Surry Power Station, Unit Nos. 1 and 2, Surry County, Virginia
Date of amendment request: July 20, 1995
Description of amendment request: The proposed amendments would: 1)
revise three Reactor Protection System/Engineered Safety Features
Actuation Systems channel trip setpoint limits, 2)
[[Page 45191]]
add a new setpoint limit for high high steam generator water level, and
3) incorporate editorial changes to revise the measurement units of one
setpoint limit and to delete certain references to two-loop operation.
Basis for proposed no significant hazards consideration
determination: As requied by 10 CFR 50.91(a), the licensee has provided
its analysis of the issue of no significant hazards consideration,
which is presented below:Specifically, operation of Surry Power Station
with the proposed change will not:
(1) Involve a significant increase in either the probability of
occurrence or consequences of any accident or equipment malfunction
scenario which is important to safety and which has been previously
evaluated in the Updated Safety Analysis Report (UFSAR). The effect
of the proposed change is to ensure that actual plant setpoints
remain conservative consistent with respect to accident analysis
assumptions. The proposed change requires safety system actuation
limits that are more conservative than those currently in Technical
Specifications. The change does not invalidate currently implemented
station setpoints or currently applicable accident analysis
assumptions regarding these setpoints. Consequently, the results and
conclusions of the current UFSAR accident analyses are not affected
by these changes. The proposed Technical Specifications change
revises setpoints used to mitigate accidents and therefore has no
bearing on the probability of an accident. Further, the change
ensures that the setpoints used to mitigate an accident bound the
setpoints used in the accident analyses. Therefore, the probability
of an accident or consequences of an accident is not adversely
affected as a result of this change.
(2) Create the possibility of a new or different type of
accident than those previously evaluated in the UFSAR. Implementing
the proposed Technical Specifications setpoint limits cannot create
the possibility of an accident of a different type than was
previously evaluated in the UFSAR. Since actual plant setpoints are
not being affected, new accident precursors will not be introduced.
Furthermore, spurious challenges to safety systems are also not
expected to increase in frequency as a result of these changes since
actual setpoints installed in the plant are not being changed.
Consequently, no new accident precursors are created as a result of
the new Technical Specifications setpoint limits.
(3) Involve a significant reduction in a margin of safety. Since
the results of the existing UFSAR accident analyses remain bounding,
safety margins are not impacted. The proposed Technical
Specifications setpoint limits ensure plant setpoints remain
conservative and consistent with design base accident analysis
assumptions including appropriate instrument channel uncertainties
due to harsh environmental conditions. Therefore, the margin of
safety as defined in the Technical Specifications bases is
unaffected.
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: Swem Library, College of
William and Mary, Williamsburg, Virginia 23185
Attorney for licensee: Michael W. Maupin, Esq., Hunton and
Williams, Riverfront Plaza, East Tower, 951 E. Byrd Street, Richmond,
Virginia 23219
NRC Project Director: David B. Matthews
Wolf Creek Nuclear Operating Corporation, Docket No. 50-482, Wolf
Creek Generating Station, Coffey County, Kansas
Date of amendment request: July 25, 1995
Description of amendment request: This license amendment request
proposes to revise Technical Specification 4.0.5a and Bases Section 3/
4.4.10 to delete the clause ``(g), except where specific written relief
has been granted by the Commission pursuant to 10 CFR Part 50, Section
50.55a(g)(6)(i).'' This proposed change is consistent with NUREG-1482,
``Guidelines for Inservice Testing and Nuclear Power Plants.''
Basis for proposed no significant hazards consideration
determination: As requied by 10 CFR 50.91(a), the licensee has provided
its analysis of the issue of no significant hazards consideration,
which is presented below:
1. The proposed change does not involve a significant increase
in the probability or consequences of an accident previously
evaluated.
This proposed change would remove the wording ''...(g), except
where specific written relief has been granted by the Commission
pursuant to 10 CFR Part 50, Section 50.55a(g)(6)(i).'' The Inservice
Inspection and Testing Programs are described in the technical
specifications pursuant to 10 CFR 50.55a. In addition, the proposed
change, in accordance with NUREG-1431 and NUREG-1482, would provide
relief to the ASME Code requirement in the interim between the time
of submittal of a relief request until the NRC has issued a safety
evaluation and granted the relief. The change being proposed is
administrative in nature and does not affect assumptions contained
in plant safety analyses, the physical design and/or operation of
the plant, nor does it affect any technical specification that
preserves safety analysis assumptions. Any relief from the approved
ASME Section XI Code requirements will require a 10 CFR 50.59
evaluation to ensure no technical specification changes or
unreviewed safety questions exist. Therefore, operation of the
facility in accordance with the proposed change would not affect the
probability or consequences of an accident previously analyzed.
2. The proposed change does not create the possibility of a new
or different kind of accident from any accident previously
evaluated.
This proposed change would remove the wording ''...(g), except
where specific written relief has been granted by the Commission
pursuant to 10 CFR Part 50, Section 50.55a(g)(6)(i).'' The Inservice
Inspection and Testing Programs are described in the technical
specifications pursuant to 10 CFR 50.55a. In addition, the proposed
change, in accordance with NUREG-1431 and NUREG-1482, would provide
relief to the ASME Code requirement in the interim between the time
of submittal of a relief request until the NRC had issued a safety
evaluation and granted the relief. The change being proposed is
administrative in nature and will not change the physical plant or
the modes of operation defined in the facility license. The change
does not involve the addition or modification of equipment nor does
it alter the design or operation of plant systems. Any relief from
the approved ASME Section XI Code requirements will require a 10 CFR
50.59 evaluation to ensure no technical specification changes or
unreviewed safety questions exist. Therefore, operation of the
facility in accordance with the proposed change would not create the
possibility of a new or different kind of accident from any accident
previously evaluated.
3. The proposed change does not involve a significant reduction
in a margin of safety.
This proposed change would remove the wording ''...(g), except
where specific written relief has been granted by the Commission
pursuant to 10 CFR Part 50, Section 50.55a(g)(6)(i).'' The Inservice
Inspection and Testing Programs are described in the technical
specifications pursuant to 10 CFR 50.55a. In addition, the proposed
change, in accordance with NUREG-1431 and NUREG-1482, would provide
relief to the ASME Code requirement in the interim between the time
of submittal of a relief request until the NRC has issued a safety
evaluation and granted the relief. The change being proposed is
administrative in nature and will not alter the bases for assurance
that safety-related activities are performed correctly or the basis
for any technical specification that is related to the establishment
or maintenance of a safety margin. Any relief from the approved ASME
Section XI Code requirements will require a 10 CFR 50.59 evaluation
to ensure no technical specification changes or unreviewed safety
questions exist. Therefore, operation of the facility in accordance
with the proposed change would not involve a significant reduction
in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room locations: Emporia State University,
[[Page 45192]]
William Allen White Library, 1200 Commercial Street, Emporia, Kansas
66801 and Washburn University School of Law Library, Topeka, Kansas
66621
Attorney for licensee: Jay Silberg, Esq., Shaw, Pittman, Potts and
Trowbridge, 2300 N Street, N.W., Washington, D.C. 20037
NRC Project Director: William H. Bateman
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.
GPU Nuclear Corporation, et al., Docket No. 50-289, Three Mile
Island Nuclear Station, Unit No. 1, Dauphin County, Pennsylvania
Date of amendment request: August 11, 1995
Description of amendment request: The proposed amendment would
remove Technical Specification Section 3.2, ``Makeup and Purification
and Chemical Addition Systems,'' and its bases. The pertinent
requirements and bases applicable to these systems are being
incorporated in the TMI-1 Updated Final Safety Analysis Report (UFSAR).
Date of publication of individual notice in Federal Register:
August 18, 1995 (60 FR 43172)
Expiration date of individual notice: September 18, 1995
Local Public Document Room location: Law/Government Publications
Section, State Library of Pennsylvania (REGIONAL DEPOSITORY) Walnut
Street and Commonwealth Avenue, Box 1601, Harrisburg, PA 17105
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: April 6, 1995, and superseded
on August 7, 1995
Description of amendments request: Amend the Sequoyah Nuclear
Plant, Units 1 and 2 Technical Specification (TS) to revise the
numerical values for the overtemperature and overpower delta-
temperature equation constants in TS Table 2.2-1, Reactor Trip System
Instrumentation Trip Setpoints.
Date of publication of individual notice in the Federal Register:
August 15, 1995 (60 FR 42187)
Expiration date of individual notice: September 14, 1995
Local Public Document Room location: Chattanooga-Hamilton County
Library, 1101 Broad Street, Chattanooga, Tennessee 37402
Notice Of Issuance Of Amendments To Facility Operating Licenses
During the period since publication of the last biweekly notice,
the Commission has issued the following amendments. The Commission has
determined for each of these amendments that the application complies
with the standards and requirements of the Atomic Energy Act of 1954,
as amended (the Act), and the Commission's rules and regulations. The
Commission has made appropriate findings as required by the Act and the
Commission's rules and regulations in 10 CFR Chapter I, which are set
forth in the license amendment.
Notice of Consideration of Issuance of Amendment to Facility
Operating License, Proposed No Significant Hazards Consideration
Determination, and Opportunity for A Hearing in connection with these
actions was published in the Federal Register as indicated.
Unless otherwise indicated, the Commission has determined that
these amendments satisfy the criteria for categorical exclusion in
accordance with 10 CFR 51.22. Therefore, pursuant to 10 CFR 51.22(b),
no environmental impact statement or environmental assessment need be
prepared for these amendments. If the Commission has prepared an
environmental assessment under the special circumstances provision in
10 CFR 51.12(b) and has made a determination based on that assessment,
it is so indicated.
For further details with respect to the action see (1) the
applications for amendment, (2) the amendment, and (3) the Commission's
related letter, Safety Evaluation and/or Environmental Assessment as
indicated. All of these items are available for public inspection at
the Commission's Public Document Room, the Gelman Building, 2120 L
Street, NW., Washington, DC, and at the local public document rooms for
the particular facilities involved.
Arizona Public Service Company, et al., Docket Nos. STN 50-528, STN
50-529, and STN 50-530, Palo Verde Nuclear Generating Station,
Units 1, 2, and 3, Maricopa County, Arizona
Date of application for amendments: May 2, 1995
Brief description of amendments: The amendments remove from the
technical specifications (TS) plant elevations for the minimum water
volume required in the spent fuel pool and relocate them to site
procedures. The TS amendment also includes two changes to correct
administrative errors in the TS.
Date of issuance: August 7, 1995
Effective date: August 7, 1995
Amendment Nos.: Unit 1 - Amendment No. 97 ; Unit 2 - Amendment No.
85; Unit 3 - Amendment No. 68
Facility Operating License Nos. NPF-41, NPF-51, and NPF-74: The
amendments revised the Technical Specifications.
Date of initial notice in Federal Register: July 5, 1995 (60 FR
35060) The Commission's related evaluation of the amendments is
contained in a Safety Evaluation dated August 7, 1995.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 Nos. 50-325 and 50-
324, Brunswick Steam Electric Plant, Units 1 and 2, Brunswick
County, North Carolina
Date of application for amendments: January 25, 1993, as
supplemented on December 28, 1993, September 13, 1994, January 13,
1995, and May 25, 1995. The supplemental submittals did not expand the
scope of the original Federal Register notice or change the no
significant hazards determination.
Brief description of amendments: The amendments allow unit entry
into Operational Condition 1 (Power Operation) from Operational
Condition 2 (Startup) with up to eight inoperable control rods,
provided those control rods are not inoperable due to being immovable
or untrippable.
Date of issuance: August 11, 1992
Effectove date: August 11, 1992
[[Page 45193]]
Amendment Nos.: 178 and 209
Facility Operating License Nos. DPR-71 and DPR-62.
Date of initial notice in Federal Register: July 7, 1993 (58 FR
36428) The Commission's related evaluation of the amendments is
contained in a Safety Evaluation dated August 11, 1995.Significant
hazards consideration comments received: No.
Local Public Document Room location: University of North Carolina
at Wilmington, William Madison Randall Library, 601 S. College Road,
Wilmington, North Carolina 28403-3297
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: April 5, 1995, as supplemented
July 31, 1995
Brief description of amendment: The amendment revises various
portions of TS 3/4.9, Refueling Operations, to be consistent with
NUREG-1431, ``Standard Technical Specifications, Westinghouse Plants,''
and allows the relocation of applicable sections from the TS that do
not meet the Commission screening criteria for retention.
Date of issuance: August 9, 1995
Effectove date: August 9, 1995
Amendment No.: 61
Facility Operating License No. NPF-63. Amendment revises the
Technical Specifications
Date of initial notice in Federal Register: May 10, 1995 (60 FR
24906) The July 31, 1995 letter provided clarifying information that
did not change the initial proposed no significant hazards
consideration determination. The Commission's related evaluation of the
amendment is contained in a Safety Evaluation dated August 9, 1995.No
significant hazards consideration comments received: No
Local Public Document Room location: Cameron Village Regional
Library, 1930 Clark Avenue, Raleigh, North Carolina 27605
Commonwealth Edison Company, Docket Nos. 50-373 and 50-374, LaSalle
County Station, Units 1 and 2, LaSalle County, Illinois
Date of application for amendments: January 13, 1995
Brief description of amendments: The amendments revise the pressure
alarm setpoint allowable values for the emergency core cooling system
(ECCS) and reactor core isolation cooling (RCIC) system ``keep filled''
pressure instrumentation channels. The purpose of the change is to
lower the setpoint allowable values for these parameters to more
realistic values based upon calculations performed by the licensee
reflecting design changes and system performance. Also, the term
``setpoint'' is being changed to ``setpoint allowable value'' to
clarify the use of the values. Additionally, two administrative/
editorial changes are included to delete technical specification
footnotes which are no longer applicable.
Date of issuance: August 15, 1995
Effectove date: Immediately, to be implemented within 90 days.
Amendment Nos.: 105 and 91
Facility Operating License Nos. NPF-11 and NPF-18: The amendments
revised the Technical Specifications.
Date of initial notice in Federal Register: March 1, 1995 (60 FR
11128) The Commission's related evaluation of the amendments is
contained in a Safety Evaluation dated August 15, 1995.No significant
hazards consideration comments received: No
Local Public Document Room location: Jacobs Memorial Library,
Illinois Valley Community College, Oglesby, Illinois 61348
Duke Power Company, Docket Nos. 50-269, 50-270, and 50-287, Oconee
Nuclear Station, Units 1, 2, and 3, Oconee County, South Carolina
Date of application of amendments: February 24, 1994, as
supplemented by letters dated April 19, May 25, August 25, 1994,
January 4, January 27, February 22, March 15, April 19, and May 31,
1995
Brief description of amendments: The amendments provide
surveillance requirements for a planned modification to the Keowee
emergency power generators' underground power path breaker closing
logic.
Date of issuance: August 15, 1995
Effectove date: As of the date of issuance to be implemented within
30 days
Amendment Nos.: 210, 210, and 207
Facility Operating License Nos. DPR-38, DPR-47, and DPR-55: The
amendments revised the Technical Specifications.
Date of initial notice in Federal Register: March 30, 1994 (59 FR
14887) The April 19, May 25, August 25, 1994, January 4, January 27,
February 22, March 15, April 19, and May 31, 1995, letters provided
clarifying information that did not change the scope of the February
24, 1994, application and initial no proposed significant hazards
consideration determination. The Commission's related evaluation of the
amendments is contained in a Safety Evaluation dated August 15, 1995.No
significant hazards consideration comments received: No
Local Public Document Room location: Oconee County Library, 501
West South Broad Street, Walhalla, South Carolina 29691
Florida Power and Light Company, Docket Nos. 50-250 and 50-251,
Turkey Point Plant Units 3 and 4, Dade County, Florida
Date of application for amendments: March 30, 1995, as supplemented
May 5, 1995 and June 19, 1995
Brief description of amendments: These amendments relate to
separation of the 24-hour emergency diesel generator test and hot
restart test from the loss of offsite power test.
Date of issuance: August 8, 1995
Effectove date: August 8, 1995
Amendment Nos.: 175 and 169Facility Operating Licenses Nos. DPR-31
and DPR-41: Amendments revised the Technical Specifications.
Date of initial notice in Federal Register: May 23, 1995 (60 FR
27339), and July 5, 1995 (60 FR 35072) The Commission's related
evaluation of the amendments is contained in a Safety Evaluation dated
August 8, 1995.No significant hazards consideration comments received:
No
Local Public Document Room location: Florida International
University, University Park, Miami, Florida 33199
GPU Nuclear Corporation, et al., Docket No. 50-219, Oyster Creek
Nuclear Generating Station, Ocean County, New Jersey
Date of application for amendment: April 15, 1995, as supplemented
by letters on May 20, 1994, and March 8, 1995
Brief description of amendment: The amendment revises Technical
Specification Section 6.5.3, ``AUDITS,'' by removing the specified
frequency for internal audits. These frequency specifications will now
be located in Appendix E of the GPU Nuclear Operational Quality
Assurance Plan (1000-PLN-7200.01). A minor editorial change has been
incorporated into TS 6.5.1.14 correcting a reference in response to a
finding in the Operational Safety Team Inspection (OSTI) report of
December 23, 1993.
Date of issuance: August 7, 1995
Effectove date: As of the date of issuance to be implemented within
30 days
Amendment No.: 181
Facility Operating License No. DPR-16. Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: May 25, 1994 (59 FR
27056)
[[Page 45194]]
The letters of May 20, 1994, and March 8, 1995, provided clarifying
information that did not change the initial proposed no significant
hazards consideration determination. The Commission's related
evaluation of this amendment is contained in a Safety Evaluation dated
August 7, 1995.No significant hazards consideration comments received:
No.
Local Public Document Room location: Ocean County Library,
Reference Department, 101 Washington Street, Toms River, NJ 08753
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 19, 1994, supplemented
March 8, 1995
Brief description of amendment: The amendment revises the TMI-1
Technical Specification (TS) Section 6.5.3 to remove the specified
frequency of various licensee-conducted audits, including those related
to quality assurance, fire protection, security, emergency
preparedness, and offsite dose calculations. The frequencies for
conduct of these audits will now be specified in the licensee's
Operational Quality Assurance Plan, which requires NRC approval for
significant changes. The Commission has determined that these audit
frequencies need not be in the TS to assure public health and safety.
Date of issuance: August 14, 1995
Effectove date: August 14, 1995
Amendment No.: 195
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 March 8, 1995, submittal provided clarifying information
that did not change the initial proposed no significant hazards
consideration determination. The Commission's related evaluation of
this amendment is contained in a Safety Evaluation dated August 14,
1995.No significant hazards consideration comments received: No.
Local Public Document Room location: Law/Government Publications
Section, State Library of Pennsylvania, (REGIONAL DEPOSITORY) Walnut
Street and Commonwealth Avenue, Box 1601, Harrisburg, PA 17105
Niagara Mohawk Power Corporation, Docket No. 50-410, Nine Mile
Point Nuclear Station, Unit 2, Oswego County, New York
Date of application for amendment: December 13, 1994, as
supplemented April 3, 1995
Brief description of amendment: The amendment revises Table
3.6.1.2-1 to allow a maximum leakage of 24.0 scfh for each of the 8
main steam isolation valves instead of the current 6.0 scfh.
Date of issuance: August 10, 1995
Effectove date: As of the date of issuance to be implemented within
60 days
Amendment No.: 67
Facility Operating License No. NPF-69: Amendment revises the
Technical Specifications.
Date of initial notice in Federal Register: January 18, 1995 (60 FR
3675) The April 3, 1995, letter provided clarifying information that
did not change the initial no proposed significant hazards
consideration determination. The Commission's related evaluation of the
amendment is contained in a Safety Evaluation dated August 10, 1995.No
significant hazards consideration comments received: No
Local Public Document Room location: Reference and Documents
Department, Penfield Library, State University of New York, Oswego, New
York 13126
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 application for amendments: March 15, 1995 (published in
Federal Register as March 15, 1994) as supplemented by letter dated
August 5, 1995
Brief description of amendments: These amendments modify the
Susquehanna Steam Electric Station Technical Specification Table 3.6.3-
1, Primary Containment Isolation Valves, concerning the scope of Type C
testing on specified emergency core cooling system and reactor core
isolation cooling containment isolation valves. Specifically, the
subject valves on systems which terminate below the minimum water level
of the suppression pool will no longer require Type C testing but will
instead be tested using requirements of the American Society of
Mechanical Engineers' Section XI Code.
Date of issuance: August 15, 1995
Effectove date: August 15, 1995
Amendment Nos.: 149 and 119
Facility Operating License Nos. NPF-14 and NPF-22. The amendments
revised the Technical Specifications.The supplemental letter did not
change the proposed no significant hazards consideration determination
nor the Federal Register notice.
Date of initial notice in Federal Register: April 26, 1995 (60 FR
20521) The Commission's related evaluation of the amendments is
contained in a Safety Evaluation dated August 15, 1995.No significant
hazards consideration comments received: No
Local Public Document Room location: Osterhout Free Library,
Reference Department, 71 South Franklin Street, Wilkes-Barre,
Pennsylvania 18701
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 application for amendments: March 31, 1995, as supplemented
by letter dated June 22, 1995
Brief description of amendments: These amendments delete from the
Technical Specifications of each unit, the operational condition
restriction in Surveillance Requirement 4.8.1.1.2.d.7, which requires
that 24-hour emergency diesel generator testing be performed with at
least one unit in operational condition 4 or 5 (cold shutdown or
refueling).
Date of issuance: August 15, 1995
Effectove date: Units 1 and 2, effective as of the date of issuance
and shall be implemented within 60 days
Amendment Nos.: 150 and 120
Facility Operating License Nos. NPF-14 and NPF-22. The amendments
revised the Technical Specifications.
Date of initial notice in Federal Register: April 26, 1995 (60 FR
20523) The Commission's related evaluation of the amendments is
contained in a Safety Evaluation dated August 15, 1995.No significant
hazards consideration comments received: No
Local Public Document Room location: Osterhout Free Library,
Reference Department, 71 South Franklin Street, Wilkes-Barre,
Pennsylvania 18701
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 application for amendments: November 21, 1994, as
supplemented by letters dated February 21, 1995, March 28, 1995, April
10, 1995, May 24, 1995, and June 23, 1995
Brief description of amendments: These amendments change the
Technical Specifications for the two units by deleting reference to the
main steamline isolation valve (MSIV) leakage control system and its
associated primary containment isolation valves, and increase the
allowable leakage rate for any MSIV and the total maximum
[[Page 45195]]
pathway leakage for all four main steam lines.
Date of issuance: August 15, 1995
Effectove date: Units 1 and 2 as of date of issuance and shall be
implemented within 30 days
Amendment Nos.: 151 and 121
Facility Operating License Nos. NPF-14 and NPF-22. The amendments
revised the Technical Specifications.
Date of initial notice in Federal Register: January 4, 1995 (60 FR
503) The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated August 15, 1995.No significant hazards
consideration comments received: No
Local Public Document Room location: Osterhout Free Library,
Reference Department, 71 South Franklin Street, Wilkes-Barre,
Pennsylvania 18701
Power Authority of the State of New York, Docket No. 50-333, James
A. FitzPatrick Nuclear Power Plant, Oswego County, New York
Date of application for amendment: March 2, 1995
Brief description of amendment: The amendment extends the
surveillance test intervals for the snubber systems to support 24-month
operating cycles. Surveillance test interval extensions are denoted as
being performed ``every 24 months'' or ``at least once per 24 months''
consistent with the guidance provided in Generic Letter (GL) 91-04,
``Changes in Technical Specification Surveillance Intervals to
Accommodate 24-Month Fuel Cycle,'' dated April 2, 1991. The NRC staff
has determined that the proposed Technical Specification changes are in
accordance with GL 91-04, and are, therefore, acceptable.
Date of issuance: August 8, 1995
Effectove date: As of the date of issuance to be implemented within
30 days
Amendment No.: 226
Facility Operating License No. DPR-59: Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: May 10, 1995 (60 FR
24916) The Commission's related evaluation of the amendment is
contained in a Safety Evaluation dated August 8, 1995.No significant
hazards consideration comments received: No
Local Public Document Room location: Reference and Documents
Department, Penfield Library, State University of New York, Oswego, New
York 13126
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: February 5, 1993, supplemented
April 13, June 11 and November 17, 1993
Brief description of amendments: The amendment eliminates the
Steam/Feedwater Flow Mismatch and Low Steam Generator Water Level
Reactor Trip due to the installation of the digital feedwater control
system incorporating a median signal selector.
Date of issuance: August 7, 1995
Effectove date: Unit 1, as of the date of issuance, to be
implemented by the startup following the twelfth refueling outage, Unit
2, as of the date of issuance, to be implemented by the startup
following the current outage
Amendment Nos.: 173 and 154
Facility Operating License Nos. DPR-70 and DPR-75: The amendments
revised the Technical Specifications.
Date of initial notice in Federal Register: April 28, 1993 (58 FR
25864) The April 13, June 11, and November 17, 1993 submittals provided
clarifying information that did not change the initial proposed no
significant hazards consideration determination.The Commission's
related evaluation of the amendment is contained in a Safety Evaluation
dated August 7, 1995.No significant hazards consideration comments
received: No
Local Public Document Room location: Salem Free Public Library, 112
West Broadway, Salem, New Jersey 08079
South Carolina Electric & Gas Company, South Carolina Public
Service Authority, Docket No. 50-395, Virgil C. Summer Nuclear
Station, Unit No. 1, Fairfield County, South Carolina
Date of amendment request: March 11, 1994
Description of amendment request: The amendment decreases the
allowable time for operation with one inoperable residual heat removal
(RHR) relief valve from 7 days to 72 hours. This amendment request has
been submitted in response to Generic Issue 94 as discussed in Generic
Letter 90-06.
Date of issuance: August 11, 1995
Effectove date: August 11, 1995
Amendment No.: 125
Facility Operating License No. NPF-12: Amendment revises the
Technical Specifications.
Date of initial notice in Federal Register: June 22, 1994 (59 FR
32236) The Commission's related evaluation of the amendment is
contained in a Safety Evaluation dated August 11, 1995.No significant
hazards consideration comments received: No
Local Public Document Room location: Fairfield County Library, 300
Washington Street, Winnsboro, South Carolina 29180
TU Electric Company, Docket Nos. 50-445 and 50-446, Comanche Peak
Steam Electric Station, Unit Nos. 1 and 2, Somervell County, Texas
Date of amendment request: February 14, 1994 (TXX-94045), as
supplemented by letter dated May 23, 1995 (TXX-95147)
Brief description of amendments: The amendments incorporated
appropriate references to and provisions of the new 10 CFR Part 20
regulations. These changes revised a definition and aspects of
radiological effluent technical specifications, clarified the
administrative specification for reporting individual annual exposures
greater than 100 mrem by work/job function, and revised the
administrative specifications for providing alternative measures for
control of access to high radiation areas and designating record
retention for radioactive shipments.
Date of issuance: August 11, 1995
Effectove date: August 11, 1995
Amendment Nos.: Unit 1 - Amendment No. 42; Unit 2 - Amendment No.
28
Facility Operating License Nos. NPF-87 and NPF-89. The amendments
revised the Technical Specifications.
Date of initial notice in Federal Register: April 28, 1994 (59 FR
22016) The additional information contained in the supplemental letter
dated May 23, 1995, was clarifying in nature and thus, within the scope
of the initial notice and did not affect the staff's proposed no
significant hazards consideration determinations. The Commission's
related evaluation of the amendments is contained in a Safety
Evaluation dated August 11, 1995.No significant hazards consideration
comments received: No.
Local Public Document Room location: University of Texas at
Arlington Library, Government Publications/Maps, 702 College, P.O. Box
19497, Arlington, TX 76019
Vermont Yankee Nuclear Power Corporation, Docket No. 50-271,
Vermont Yankee Nuclear Power Station, Vernon, Vermont
Date of application for amendment: March 31, 1994, as supplemented
by letters dated September 9, 1994, and June 22, 1995
Brief description of amendment: The amendment modifies the
requirements for avoidance and protection from thermal hydraulic
instabilities to be
[[Page 45196]]
consistent with the Boiling Water Reactor (BWR) Owners Group long-term
solution Option I-D described in the Licensing Topical Report, ``BWR
Owners Group Long-Term Stability Solutions Licensing Methodology, NEDO-
31960 June 1991'' and NEDO-31960, Supplement 1, Dated March 1992. NEDO-
31960 and NEDO-31960, Supplement 1, were accepted by the NRC staff in a
letter to L.A. England (BWR Owners Group) dated July 12, 1993.
Date of issuance: August 9, 1995
Effectove date: As of the date of issuance to be implemented within
30 days
Amendment No.: 146
Facility Operating License No. DPR-28. Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: January 4, 1995 (60 FR
507) The September 9, 1994, and June 22, 1995, submittals provided
clarifying information that did not change the initial proposed no
significant hazards consideration determination. The Commission's
related evaluation of the amendment is contained in a Safety Evaluation
dated August 9, 1995. No significant hazards consideration comments
received: No
Local Public Document Room location: Brooks Memorial Library, 224
Main Street, Brattleboro, VT 05301
Dated at Rockville, Maryland, this 23rd day of August.
For The Nuclear Regulatory Commission
Elinor G. Adensam,
Acting Director, Division of Reactor Projects III/IV, Office of Nuclear
Reactor Regulation
[Doc. 95-21389 Filed 8-29-95; 8:45 am]
BILLING CODE 7590-01-F