[Federal Register Volume 64, Number 154 (Wednesday, August 11, 1999)]
[Notices]
[Pages 43764-43785]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: 99-20545]
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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 July 17, 1999, through July 30, 1999. The
last biweekly notice was published on July 28, 1999 (64 FR 40903).
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 Chief, Rules and
Directives Branch, Division of Administration Services, Office of
Administration, U.S. Nuclear Regulatory Commission, Washington, DC
20555-
[[Page 43765]]
0001, 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 10, 1999, the licensee may file a request for a
hearing with respect to issuance of the amendment to the subject
facility operating license and any person whose interest may be
affected by this proceeding and who wishes to participate as a party in
the proceeding must file a written request for a hearing and a petition
for leave to intervene. Requests for a hearing and a petition for leave
to intervene shall be filed in accordance with the Commission's ``Rules
of Practice for Domestic Licensing Proceedings'' in 10 CFR part 2.
Interested persons should consult a current copy of 10 CFR 2.714 which
is available at the Commission's Public Document Room, the Gelman
Building, 2120 L Street, NW., Washington, DC and at the local public
document room for the particular facility involved. If a request for a
hearing or petition for leave to intervene is filed by the above date,
the Commission or an Atomic Safety and Licensing Board, designated by
the Commission or by the Chairman of the Atomic Safety and Licensing
Board Panel, will rule on the request and/or petition; and the
Secretary or the designated Atomic Safety and Licensing Board will
issue a notice of a hearing or an appropriate order.
As required by 10 CFR 2.714, a petition for leave to intervene
shall set forth with particularity the interest of the petitioner in
the proceeding, and how that interest may be affected by the results of
the proceeding. The petition should specifically explain the reasons
why intervention should be permitted with particular reference to the
following factors: (1) The nature of the petitioner's right under the
Act to be made a party to the proceeding; (2) the nature and extent of
the petitioner's property, financial, or other interest in the
proceeding; and (3) the possible effect of any order which may be
entered in the proceeding on the petitioner's interest. The petition
should also identify the specific aspect(s) of the subject matter of
the proceeding as to which petitioner wishes to intervene. Any person
who has filed a petition for leave to intervene or who has been
admitted as a party may amend the petition without requesting leave of
the Board up to 15 days prior to the first prehearing conference
scheduled in the proceeding, but such an amended petition must satisfy
the specificity requirements described above.
Not later than 15 days prior to the first prehearing conference
scheduled in the proceeding, a petitioner shall file a supplement to
the petition to intervene which must include a list of the contentions
which are sought to be litigated in the matter. Each contention must
consist of a specific statement of the issue of law or fact to be
raised or controverted. In addition, the petitioner shall provide a
brief explanation of the bases of the contention and a concise
statement of the alleged facts or expert opinion which support the
contention and on which the petitioner intends to rely in proving the
contention at the hearing. The petitioner must also provide references
to those specific sources and documents of which the petitioner is
aware and on which the petitioner intends to rely to establish those
facts or expert opinion. Petitioner must provide sufficient information
to show that a genuine dispute exists with the applicant on a material
issue of law or fact. Contentions shall be limited to matters within
the scope of the amendment under consideration. The contention must be
one which, if proven, would entitle the petitioner to relief. A
petitioner who fails to file such a supplement which satisfies these
requirements with respect to at least one contention will not be
permitted to participate as a party.
Those permitted to intervene become parties to the proceeding,
subject to any limitations in the order granting leave to intervene,
and have the opportunity to participate fully in the conduct of the
hearing, including the opportunity to present evidence and cross-
examine witnesses.
If a hearing is requested, the Commission will make a final
determination on the issue of no significant hazards consideration. The
final determination will serve to decide when the hearing is held.
If the final determination is that the amendment request involves
no significant hazards consideration, the Commission may issue the
amendment and make it immediately effective, notwithstanding the
request for a hearing. Any hearing held would take place after issuance
of the amendment.
If the final determination is that the amendment request involves a
significant hazards consideration, any hearing held would take place
before the issuance of any amendment.
A request for a hearing or a petition for leave to intervene must
be filed with the Secretary of the Commission, U.S. Nuclear Regulatory
Commission, Washington, DC 20555-0001, Attention: Rulemakings and
Adjudications Staff, or may be delivered to the Commission's Public
Document Room, the Gelman Building, 2120 L Street, NW., Washington DC,
by the above date. A copy of the petition should also be sent to the
Office of the General Counsel, U.S. Nuclear Regulatory Commission,
Washington, DC 20555-0001, and to the attorney for the licensee.
Nontimely filings of petitions for leave to intervene, amended
petitions, supplemental petitions and/or requests for a hearing will
not be entertained absent a determination by the Commission, the
presiding officer or the Atomic Safety and Licensing Board that the
petition and/or request should be granted based upon a balancing of
factors specified in 10 CFR 2.714(a)(1)(i)-(v) and 2.714(d).
For further details with respect to this action, see the
application for amendment which is available for public inspection at
the Commission's Public Document Room, the Gelman Building, 2120 L
Street, NW., Washington, DC, and at the local public document room for
the particular facility involved.
Carolina Power & Light Company, et al., Docket No. 50-400, Shearon
Harris Nuclear Power Plant, Unit 1, Wake and Chatham Counties, North
Carolina
Date of amendment request: July 9, 1999.
Description of amendment request: The proposed amendment would
revise Harris Nuclear Plant (HNP) Technical Specification (TS) 3/4.2.2,
``Heat Flux Hot Channel Factor--FQ(Z),'' TS 3/4.2.3, ``RCS
Flow Rate And Nuclear Enthalpy Rise Hot Channel Factor,'' TS 3/4.2.5,
``DNB Parameters,'' an associated note in TS Table 2.2-1, and
associated Bases. Specifically, the proposed amendment would: (1)
Remove the allowance for reduced power operation for reduced Reactor
Coolant System (RCS) flow rate conditions; (2) separate the
requirements for F delta H and RCS flow rate in the format prescribed
by NUREG-1431, Revision 1, ``Standard Technical Specifications,
Westinghouse Plants,'' dated April 1995; and, (3) implement the
guidance of NUREG-1431, Revision 1, and NRC Generic Letter (GL) 88-16,
dated October 4, 1988 for TS 3/4.2.2, TS 3/4.2.3, TS 3/4.2.5 and
associated Bases by removing cycle specific parameters and placing that
[[Page 43766]]
information into the Core Operating Limits Report (COLR).
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. The proposed amendment does not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
The proposed amendment will not introduce any new equipment or
require existing equipment to function different from that
previously evaluated in the Final Safety Analysis Report (FSAR) or
TS.
As described in HNP TS Bases, the limits on heat flux hot
channel factor, RCS flow rate, and enthalpy rise hot channel factor
ensure that: (1) the design limits on peak local power density and
minimum DNBR [departure from nucleate boiling ratio] are not
exceeded and (2) in the event of a LOCA the peak fuel clad
temperature will not exceed the 2200 degree Fahrenheit ECCS
[emergency core cooling system] acceptance limit.
Removing the allowance for reduced power operation for reduced
RCS flow conditions is more restrictive than that currently allowed
by TS. Power Distribution Limiting Conditions for Operation for heat
flux hot channel factor and enthalpy rise hot channel factor are not
affected by this change. Therefore, the consequences of an accident
will not increase because of this change. Power Distribution limits
place administrative restrictions on reactor core parameters and as
such do not initiate nor mitigate accidents.
Power Distribution limits at HNP are developed using NRC
approved methodologies. Changing power distribution limits to be
consistent with NUREG-1431, Revision 1 will not increase the
probability or consequences of an accident that has been previously
evaluated.
Relocating cycle specific information from TS to the COLR will
not impact the ability of structures, systems, or components to
mitigate accidents. Future changes to relocated requirements in the
COLR will be submitted to the NRC for review in accordance with HNP
TS Section 6.
Therefore, the proposed change does not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
2. The proposed amendment does not create the possibility of a
new or different kind of accident from any accident previously
evaluated.
The proposed amendment will not introduce any new equipment or
require existing equipment to function different from that
previously evaluated in the Final Safety Analysis Report (FSAR) or
TS. The changes are consistent with NUREG-1431, Revision 1 and the
Commission's Final Policy Statement on Technical Specification
improvements. The proposed amendment will not create any new
accident scenarios, because the change does not introduce any new
single failures, adverse equipment or material interactions, or
release paths.
Therefore, the proposed change does not create the possibility
of a new or different kind of accident from any accident previously
evaluated.
3. The proposed amendment does not involve a significant
reduction in the margin of safety.
The LCO limit for RCS flow rate at 100.0% reactor power has not
changed. The previous capability to operate with reduced RCS flow
rate has been eliminated. This aspect of the proposed change is more
restrictive than current plant TS in that continued reactor
operation greater than 5% is not allowed if RCS flow rate is less
than the LCO limit at 100% power.
Changes to TS 3/4.2.2, TS 3/4.2.3, TS
3/4.2.5 and associated Bases are in accordance with NUREG-1431,
Revision 1. The completion times for TS Actions are acceptable
because the plant is not allowed to remain in an unacceptable
condition for an extended period of time. Sufficient time to reduce
reactor power in an orderly manner or perform other required actions
is also provided. The surveillance intervals established by NUREG-
1431, Revision 1 have been determined to be adequate for monitoring
the change in power distribution.
Relocating cycle specific information from HNP TS to the COLR is
in accordance with NRC GL 88-16. HNP does not intend to alter the
methodologies for any parameter limit calculation as a result of
this change. The proposed change is in accordance with the plant
safety analysis. Therefore, the proposed change does not involve a
significant reduction in the margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: Cameron Village Regional
Library, 1930 Clark Avenue, Raleigh, North Carolina 27605.
Attorney for licensee: William D. Johnson, Vice President and
Senior Counsel, Carolina Power & Light Company, Post Office Box 1551,
Raleigh, North Carolina 27602.
NRC Section Chief: Sheri R. Peterson.
Carolina Power & Light Company, et al., Docket No. 50-400, Shearon
Harris Nuclear Power Plant, Unit 1, Wake and Chatham Counties, North
Carolina
Date of amendment request: July 9, 1999.
Description of amendment request: The proposed amendment would
relocate Harris Nuclear Plant (HNP) Technical Specification (TS) 3/
4.3.3.3, ``Seismic Instrumentation,'' TS
3/4.3.3.4, ``Meteorological Instrumentation,'' TS 3/4.3.3.9, ``Metal
Impact Monitoring System,'' and TS
3/4.3.3.11, ``Explosive Gas Monitoring Instrumentation,'' to plant
procedure PLP-114, ``Relocated Technical Specifications and Design
Basis Requirements.'' The proposed change is in accordance with
guidance provided by NRC Generic Letter 95-10, ``Relocation of Selected
Technical Specification Requirements Related to Instrumentation.''
Changes to relocated requirements would be performed in accordance with
10 CFR 50.59. Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. The proposed amendment does not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
Seismic Instrumentation, Meteorological Instrumentation, Metal
Impact Monitoring System, and Explosive Gas Monitoring
Instrumentation are not accident initiating components as described
in the Final Safety Analysis Report. Seismic Instrumentation,
Meteorological Instrumentation, Metal Impact Monitoring System, and
Explosive Gas Monitoring Instrumentation are not accident mitigating
components. There are no modifications being made to plant systems
as a result of this change. Additionally, there are no changes being
made to the way in which systems are being operated as a result of
this change. Therefore, the proposed change does not involve a
significant increase in the probability or consequences of an
accident previously evaluated.
2. The proposed amendment does not create the possibility of a
new or different kind of accident from any accident previously
evaluated.
Seismic Instrumentation, Meteorological Instrumentation, Metal
Impact Monitoring System, and Explosive Gas Monitoring
Instrumentation are not accident initiating components as described
in the Final Safety Analysis Report (FSAR). The proposed change
relocates the TS requirements for Seismic Instrumentation,
Meteorological Instrumentation, Metal Impact Monitoring System, and
Explosive Gas Monitoring Instrumentation to plant procedure PLP-114.
Plant systems and components are not modified as a result of this
change. Future changes in these systems will be controlled in
accordance with 10 CFR 50.59.
Therefore, the proposed change does not create the possibility
of a new or different kind of accident from any accident previously
evaluated.
3. The proposed amendment does not involve a significant
reduction in the margin of safety.
The proposed change to Seismic Instrumentation, Meteorological
Instrumentation, Metal Impact Monitoring System, and Explosive Gas
Monitoring Instrumentation does not affect any of the
[[Page 43767]]
parameters that relate to the margin of safety as described in the
Bases of the TS or the FSAR. Accordingly, NRC Acceptance Limits are
not affected by this change. The proposed change relocates the TS
requirements for Seismic Instrumentation, Meteorological
Instrumentation, Metal Impact Monitoring System, and Explosive Gas
Monitoring Instrumentation to plant procedure PLP-114. Plant systems
and components are not modified as a result of this change. Future
changes in these systems will be controlled in accordance with 10
CFR 50.59. Generic Letter 95-10 states that the staff has concluded
that these provisions are not related to dominant contributors to
plant risk.
Therefore, the proposed change does not involve a significant
reduction in the margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: Cameron Village Regional
Library, 1930 Clark Avenue, Raleigh, North Carolina 27605.
Attorney for licensee: William D. Johnson, Vice President and
Senior Counsel, Carolina Power & Light Company, Post Office Box 1551,
Raleigh, North Carolina 27602.
NRC Section Chief: Sheri R. Peterson.
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.
Date of amendment request: June 30, 1999
Description of amendment request: The proposed amendment would
clarify that the source of DC electrical power required for a unit in
Mode 5 or 6 or during the movement of irradiated fuel assemblies may be
cross-tied to the opposite unit. An administrative change would also
delete reference to AT&T batteries since all AT&T batteries have been
replaced with Charter Power Systems, Inc. (C&D) batteries. The
amendment would also remove the Allowed Outage Time (AOT) extension
approved for Braidwood Station by Amendment No. 99. The activity
addressed by Amendment No. 99 is complete and the extension no longer
applies.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. Does the proposed change involve a significant increase in
the probability or consequences of an accident previously evaluated?
The proposed change will allow one DC bus on a shutdown unit to
be supplied via the DC bus cross-tie to the opposite unit. The other
DC bus on the shutdown unit will at all times be required to be
fully operable, supplied by the associated battery and charger, and
the associated cross-ties open. The DC electrical system is not
considered an initiator of any accident previously evaluated, and
therefore the probability of a previously analyzed accident is
unchanged.
The consequences of a previously analyzed event are dependent on
the initial conditions assumed for the analysis, the availability
and successful functioning of the equipment assumed to operated in
response to the analyzed event, and the setpoints at which these
actions are initiated. Sufficient equipment remains available to
mitigate the consequences of previously analyzed events. The Updated
Final Safety Analysis Report (UFSAR) section 8.3.2.1.1 clearly
allows operation with the DC cross-tie closed on one DC bus between
a unit that is operating and a unit that is shutdown, or between two
shutdown units, in the manner proposed by this amendment. The TS in
effect prior to the implementation of the Improved TS also allowed
operation in the manner proposed by this amendment. If DC buses are
cross-tied due to an inoperable DC source on a shutdown unit, both
the previous TS and the change proposed by this amendment limit the
time in this condition to seven days, and if the inoperable source
is a battery, the current on the cross-tie is limited to 200 amps.
These actions protect both the operating unit, and the shutdown
unit. If a shutdown unit's DC bus is cross-tied to an operating
unit's DC bus due to an inoperable charger on the operating unit,
both the previous TS and the change proposed by this amendment limit
the time in this condition to 24 hours. The limitations imposed by
both the previous TS and the change proposed by this amendment
ensure that operation in this configuration is within the design
bases of the plant. Thus the consequences of accidents previously
analyzed are unchanged between the previous TS and the change
proposed by this amendment. In the worst case scenario, assuming a
single failure, one DC bus on the shutdown unit will always be
operable, and the ability to mitigate the consequences of any
accident previously analyzed is preserved.
The change to delete all references in the Braidwood TS to AT&T
batteries and the AOT extension granted under TS Amendment Number 99
is administrative only, and has no impact on the probability or
consequences of accidents previously evaluated.
Therefore this proposed amendment does not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
2. Does the proposed change create the possibility of a new or
different kind of accident from any accident previously evaluated?
The proposed change does not involve a physical change to the
plant. No new equipment is being introduced, and installed equipment
is not being operated in a new or different manner. There is no
change being made to the parameters within which the plant is
operated. There are no setpoints affected by this change at which
protective or mitigative actions are initiated. This change will not
alter the manner in which equipment operation is initiated, nor will
the function demands on credited equipment be changed. No alteration
in the procedures which ensure the plant remains within analyzed
limits in being proposed, and no change is being made to the
procedures relied upon to respond to an off-normal event. As such,
no new failure modes are being introduced. The change does not alter
assumptions made in the safety analysis and licensing basis.
Therefore, the change does not create the possibility of a new or
different kind of accident from any accident previously evaluated.
The change to delete all references in the Braidwood TS to AT&T
batteries and the AOT extension granted under TS Amendment Number 99
is administrative only, and cannot 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 from any previously evaluated.
3. Does the proposed change involve a significant reduction in a
margin of safety?
The margin of safety is established through equipment design,
operating parameters, and the setpoints at which automatic actions
are initiated. Sufficient equipment remains available to actuate
upon demand for the purpose of mitigating an analyzed event. The
proposed change, which will allow one DC bus on a shutdown unit to
be supplied via the DC bus cross-tie to the opposite unit, is
acceptable because of the limitations imposed on operation in this
configuration, and because the other DC bus on the shutdown unit
will at all times be required to be fully operable, supplied by the
associated battery and charger, and the associated cross-ties open.
The TS in effect prior to the implementation of the Improved TS
allowed operation in the manner proposed by this amendment. In the
worst case scenario, assuming a single failure, one DC bus on the
shutdown unit will always be operable. Thus, there is no detrimental
impact on any equipment design parameter, and the plant will still
be required to operate within prescribed limits. Therefore, the
change does not reduce the margin of safety.
The change to delete all references in the Braidwood TS to AT&T
batteries and the AOT extension granted under TS Amendment Number 99
is administrative only, and does not reduce 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
[[Page 43768]]
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.
Attorney for licensee: Ms. Pamela B. Stroebel, Senior Vice
President and General Counsel, Commonwealth Edison Company, P.O. Box
767, Chicago, Illinois 60690-0767.
NRC Section Chief: Anthony J. Mendiola.
Commonwealth Edison Company, Docket Nos. 50-237 and 50-249, Dresden
Nuclear Power Station, Units 2 and 3, Grundy County, Illinois
Date of amendment request: May 3, 1999.
Description of amendment request: The proposed amendments would
relocate Technical Specifications (TS) Section 3/4.6.I to the Updated
Final Safety Analysis Report (UFSAR). TS Section 3/4.6.I contains
reactor coolant chemistry limiting conditions for operation (LCO) and
surveillance requirements (SR) for conductivity, chloride
concentration, and pH.
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:
Does the change involve a significant increase in the
probability or consequences of an accident previously evaluated?
The proposed changes simplify the TS, meet regulatory
requirements for relocated TS's, and implement the recommendations
of the NRC Final Policy Statement on TS improvements. The Chemistry
requirements will be relocated to the Updated Final Safety Analysis
Report (UFSAR) and to applicable station procedures. Future changes
to these requirements will be controlled by 10 CFR 50.59. The
proposed changes are administrative in nature and do not involve any
modification to any plant equipment or affect plant operation.
Therefore, the proposed changes do not involve a significant
increase in the probability or consequences of any previously
evaluated accident.
Consequently, this proposed amendment does not involve a
significant increase in the probability or consequences of any
accident previously evaluated.
Does the change create the possibility of a new or different
kind of accident from any accident previously evaluated?
The proposed changes are administrative in nature, do not
involve any physical alterations to any plant equipment, and cause
no change in the method by which any safety related system performs
its function. Therefore, this proposed TS amendment will not create
the possibility of a new or different kind of accident from any
accident previously evaluated.
Does the change involve a significant reduction in a margin of
safety?
The proposed amendment represents the relocation of current
requirements, which are based on generic guidance or previously
approved provisions for other stations. The proposed changes are
administrative in nature and do not adversely affect existing plant
safety margins or the reliability of the equipment assumed to
operate in the safety analysis. The proposed changes have been
evaluated and found to be acceptable for use at Dresden Nuclear
Power Station. Since the proposed changes are administrative in
nature, and are based on NRC accepted provisions which have been
adopted at other nuclear facilities, and maintain the necessary
levels of system reliability, 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
requested amendments involve no significant hazards consideration.
Local Public Document Room location: Morris Area Public Library
District, 604 Liberty Street, Morris, Illinois 60450.
Attorney for licensee: Ms. Pamela B. Stroebel, Senior Vice
President and General Counsel, Commonwealth Edison Company, P.O. Box
767, Chicago, Illinois 60690-0767.
NRC Section Chief: Anthony J. Mendiola.
Commonwealth Edison Company, Docket No. 50-373, LaSalle County Station,
Unit 1, LaSalle County, Illinois
Date of amendment request: July 7, 1999.
Description of amendment request: The proposed amendments would (1)
revise Technical Specification Section 2.1, Safety Limits, to reflect a
change to the LaSalle, Unit 1, Minimum Critical Power Ratio Safety
Limit; and (2) revise Technical Specification Section 6.6.A.6 to add an
NRC-approved Siemens Power Corporation methodology to the list of
topical reports used to determine the core operating limits.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. Does the change involve a significant increase in the
probability or consequences of an accident previously evaluated?
The probability of an evaluated accident is derived from the
probabilities of the individual precursors to that accident. The
consequences of an evaluated accident are determined by the
operability of plant systems designed to mitigate those
consequences. Limits have been established consistent with NRC-
approved methods to ensure that fuel performance during normal,
transient, and accident conditions is acceptable. These changes do
not affect the operability of plant systems, nor do they compromise
any fuel performance limits.
Changing the MCPR Safety Limit for LaSalle Unit 1 will not
increase the probability or the consequences of an accident
previously evaluated. This change implements the MCPR Safety Limit
resulting from the SPC ANFB critical power correlation methodology
using the approved ATRIUM-9B additive constant uncertainty. For each
cycle, cycle specific MCPR Safety Limit calculations will be
performed, consistent with SPC's approved methodology, to confirm
the appropriateness of the MCPR Safety Limit. Additionally,
operational MCPR limits will be applied that will ensure the MCPR
Safety Limit is not violated during all modes of operation and
anticipated operational occurrences. The MCPR Safety Limit ensures
that less than 0.1% of the rods in the core are expected to
experience boiling transition. Therefore the probability or
consequences of an accident will not increase.
Adding EMF-85-74, Revision 0, Supplement 1 (P)(A) and Supplement
2 (P)(A) to Section 6 does not increase the probability or
consequences of an accident previously evaluated. The NRC-approved
burnup extension for RODEX2A applications has been demonstrated to
meet all applicable design criteria. Therefore adding this
methodology to Technical Specification Section 6 does not increase
the probability or consequences of an accident previously evaluated
.
Therefore, this proposed amendment does not involve a
significant increase in the probability or consequences of any
accident previously evaluated.
2. Does the change create the possibility of a new or different
kind of accident from any accident previously evaluated?
Creation of the possibility of a new or different kind of
accident would require the creation of one or more new precursors of
that accident. New accident precursors may be created by
modifications to the plant configuration, including changes in
allowable modes of operation. This Technical Specification submittal
does not involve any modifications to the plant configuration or
allowable modes of operation. No new precursors of an accident are
created and no new or different kinds of accidents are created.
Therefore, the proposed changes do not create the possibility of a
new or different kind of accident from any accident previously
evaluated.
Changing the MCPR Safety Limit does not create the possibility
of a new accident from any accident previously evaluated. This
[[Page 43769]]
change does not alter or add any new equipment or change modes of
operation. The MCPR Safety Limit is established to ensure that 99.9%
of the rods avoid boiling transition.
The MCPR Safety Limit is changing for LaSalle Unit 1 to support
Cycle 9 operation. This change does not introduce any physical
changes to the plant, alter the processes used to operate the plant,
or change allowable modes of operation. Therefore, no new accidents
are created that are different from any accident previously
evaluated.
The addition of RODEX2A (EMF-85-74, Revision 0, Supplement 1
(P)(A) and Supplement 2 (P)(A)) does not create the possibility of a
new accident from an accident previously evaluated. This change does
not alter or add any new equipment or change modes of operation.
This change does not introduce any physical changes to the plant,
alter the processes used to operate the plant, or change allowable
modes of operation. Therefore, no new accidents are created that are
different from any accident previously evaluated.
Therefore, the proposed changes do not create the possibility of
a new or different kind of accident from any previously evaluated.
3. Does the change involve a significant reduction in the margin
of safety?
Changing the MCPR Safety Limit for LaSalle Unit 1 will not
involve any reduction in margin of safety. The MCPR Safety Limit
provides a margin of safety by ensuring that less than 0.1% of the
rods are calculated to be in boiling transition. The proposed
Technical Specification amendment request reflects the MCPR Safety
Limit results from evaluations by SPC using NRC-approved
methodology.
The revised MCPR Safety Limit will ensure the same level of fuel
protection. Additionally, operational limits will be established
based on the proposed MCPR Safety Limit to ensure that the MCPR
Safety Limit is not violated during all modes of operation including
anticipated operation[al] occurrences. This will ensure that the
fuel design safety criterion of more than 99.9% of the fuel rods
avoiding transition boiling during normal operation as well as
during an anticipated operational occurrence is met.
The addition of EMF-85-74, Revision 0, Supplement 1 (P)(A) and
Supplement 2 (P)(A) to Section 6 does not decrease the margin of
safety. The burnup limit extension for RODEX2A applications has been
reviewed and approved by the NRC. The data supporting the burnup
extension demonstrates that all applicable design criteria are met.
Therefore, since the burnup extension is acceptable and within the
design criteria, using the approved burnup extension will not affect
the margin of safety.
Therefore, these changes do not involve a significant reduction
in the margin of safety.
Therefore, based upon the above evaluation, ComEd has concluded
that these changes involve no significant hazards consideration.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 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: Jacobs Memorial Library, 815
North Orlando Smith Avenue, Illinois Valley Community College, Oglesby,
Illinois 61348-9692.
Attorney for licensee: Ms. Pamela B. Stroebel, Senior Vice
President and General Counsel, Commonwealth Edison Company, P.O. Box
767, Chicago, Illinois 60690-0767.
NRC Section Chief: Anthony J. Mendiola.
Commonwealth Edison Company, Docket Nos. 50-373 and 50-374, LaSalle
County Station, Units 1 and 2, LaSalle County, Illinois
Date of amendment request: November 9, 1998, as supplemented on
July 7, 1999.
Description of amendment request: The proposed amendments would
revise Technical Specification Table 3.3.3-2, ``Emergency Core Cooling
System Actuation Instrumentation Setpoints,'' to modify the degraded
voltage second level undervoltage relay setpoint and allowable value.
These proposed amendments were originally noticed on January 13, 1999
(64 FR 2245), and are being renoticed to include the revised setpoints
that were included in the July 7, 1999, supplement.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
(1) Does the change involve a significant increase in the
probability or consequences of an accident previously evaluated?
The setpoint change does not change the logic or function of the
degraded voltage protection circuits as described in the UFSAR
[Updated Final Safety Analysis Report] Section 8.2.3. They also do
not reduce the reliability of these circuits. The increase in the
degraded voltage protection circuit setpoint is conservative
compared to the existing setpoint. There is no change as a result of
this amendment to the underlying accident and transient analyses
that support operations of LaSalle County Station. Inadvertent or
spurious operation of the degraded voltage protection function will
initiate loading of the safe shutdown loads on the diesel generators
and is not assumed to initiate an accident. The proposed degraded
voltage setpoints are low enough to prevent spurious actuations
given the expected offsite grid voltages. After implementation of
this amendment, no operator actions are required for equipment
operations in response to degraded voltage conditions.
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 consequences of accidents previously evaluated are not
increased. The proposed change does not affect the required level of
availability of systems required to mitigate the accidents
considered in the analyses. The proposed changes will ensure that
the Class 1E equipment will be capable of starting and operating
during a design basis accident with degraded offsite grid voltage.
The increase in the level of confidence is the result of more
rigorous methodology used to determine limiting Class 1E bus
voltages at the minimum expected offsite AC voltage. These
calculations demonstrate that the degraded voltage relays will not
actuate following a block start of the electrical loads that are
automatically actuated by or as a consequence of the LOCA [loss-of-
coolant accident] signal if the switchyard voltage remains above 352
kV.
If the grid voltage drops below 352 kV, then the analytical
limit of 3814 volts for proper operation of class 1E loads connected
to each 4.16 kV Class 1E bus is assured by transfer to the
respective onsite power sources (Emergency Diesel Generators (EDGs))
by the degraded voltage logic.
Therefore this proposed amendment does not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
(2) Create the possibility of a new or different kind of
accident from any accident previously evaluated because:
Setpoint methodology established the bases to ensure that, with
known errors, the relays will detect degraded voltage conditions and
transfer safety loads to the EDGs at a voltage level adequate to
ensure proper safety equipment performance and to prevent equipment
damage.
The trip setpoint of greater than or equal to 3863 volts and
less than or equal to 3877 volts and the allowable value of greater
than or equal to 3814 volts and less than or equal to 3900 volts,
include adequate tolerance to calibrate the relay trip units while
ensuring that the Class 1E bus voltage will remain above the
analytical limits.
These setpoint changes will ensure that adequate voltages will
be available for the continuous operation of safety-related
equipment required to function during a LOCA. These proposed changes
will also ensure that adequate voltages will be available for
starting any Class 1E equipment.
The proposed degraded voltage setpoint change does not change
the design of the degraded voltage protection system or its function
to protect against degraded offsite power. Actuation of the degraded
voltage protection system will initiate a sequence of events that
will start the EDG for the associated Class 1E bus, strip loads from
the Class 1E bus, open all feed breakers to the Class 1E bus, close
the Emergency feed breaker (thus energizing the Class 1E bus
[[Page 43770]]
from the respective EDG), and initiate starting of the Safe Shutdown
equipment supplied by the Class 1E bus.
Since the scope of this change does not affect the operation of
auxiliary power system or any actions necessary to mitigate the
consequences of accidents or achieve safe shutdown, the change does
not involve a new or different accident scenario.
Therefore, these proposed changes do not create the possibility
of a new or different kind of accident from any accident previously
evaluated.
(3) Involve a significant reduction in the margin of safety
because:
The proposed amendment will allow the degraded voltage setpoint
to be conservatively established based on new engineering
calculations which consider the lowest expected offsite grid voltage
and operation of required Class 1E equipment under design basis
accident loading conditions.
The proposed degraded voltage setpoints will ensure that
adequate Class 1E bus voltage will be available to support starting
and operation of required Class 1E loads. The proposed setpoint
includes 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 also set with margin to the minimum
Class 1E bus voltage, which is based on a minimum grid voltage of
352 kV, which is less than the expected grid voltage of 354 kV. 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.
Therefore, these changes do not involve a significant reduction
in the margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
requested amendments involve no significant hazards consideration.
Local Public Document Room location: Jacobs Memorial Library, 815
North Orlando Smith Avenue, Illinois Valley Community College, Oglesby,
Illinois 61348-9692.
Attorney for licensee: Ms. Pamela B. Stroebel, Senior Vice
President and General Counsel, Commonwealth Edison Company, P.O. Box
767, Chicago, Illinois 60690-0767.
NRC Section Chief: Anthony J. Mendiola.
Detroit Edison Company, Docket No. 50-16, Enrico Fermi Atomic Power
Plant, Unit 1, Monroe County, Michigan
Date of amendment request: April 20, 1999 (Reference NRC-99-0035).
Description of amendment request: The proposed amendment will
revise the Technical Specifications by deleting Specification D.3.c.
Specification D.3.c requires the licensee to perform weekly
observations of the nitrogen cover gas pressure within the sodium
storage tanks located in the Sodium Building Complex. Removing this
surveillance requirement would allow the licensee to remove the
nitrogen cover gas system from service for these sodium storage tanks.
This action is necessary for the licensee to begin work on removing the
remaining residual sodium from these tanks. The licensee also requested
an editorial change to delete the words ``STORAGE TANK'' from the title
of Specification D.
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 using the standards in 10 CFR 50.92(c). The licensee's
analysis is presented below:
(1) The proposed change does not involve a significant increase
in the probability or consequences of an accident previously
evaluated.
Removing the primary cover gas supply from the storage tanks
will not significantly increase the probability of an accident
occurring as long as the probability of an uncontrolled water
reaction with residual sodium is not significantly increased. This
is ensured by sealing the storage tanks after the nitrogen cover gas
system is removed except when controlled activities such as sampling
are performed. The consequences of an accident would not be affected
by removing the nitrogen cover gas supply from service as the
previously analyzed primary sodium accident already involves release
of all the radioactive material in the primary sodium. Removing the
cover gas will not increase the amount of radioactive material
available to be released.
(2) The proposed change does not create the possibility of a new
or different accident from any previously evaluated.
A sodium accident has been previously evaluated. No other type
of accident could be caused by removing the primary sodium tanks
cover gas or opening the tanks since no other system or mode of
operation of any other system will be affected.
(3) The proposed change does not involve a significant reduction
in a margin of safety.
Currently, only a small amount of residual sodium remains in the
primary sodium storage tanks. Some of this residual sodium may have
been converted to sodium carbonate. This conversion of sodium to
sodium carbonate would have left even less sodium remaining in these
tanks. The cover gas is a good precaution, especially for tanks
sitting unattended for many years. It prevents moisture from
intruding into the tanks and reacting with the sodium residues. It
also prevents oxygen from entering these tanks and reacting with any
hydrogen formed from reactions of water and sodium. Discontinuing
the use of cover gas slightly reduces the margin of safety, but not
significantly. Removing the cover gas does not, in itself, introduce
water into the tank in an uncontrolled manner. Even if slight
amounts of moisture from humidity in the air enter these tanks over
the next year or two, until the sodium is removed while the tanks
are either opened or sealed, the volume of each tank (15,000
gallons) is large enough that the tank should be able to dissipate
any small reactions that could occur. The design pressure for the
primary sodium storage tanks is from vacuum to 50 pounds per square
inch based on the vendor's drawing.
Even if sufficient water entered the tank, generated hydrogen,
and sufficient oxygen entered the tank to cause a reaction that
released the contents of the tank, there would be no significant
release of radioactivity from the tank. The release of all residual
primary sodium would result in concentration levels well below the
values in 10 CFR 20, Appendix B, Table II for releases to
unrestricted areas. Since there is less sodium in the primary sodium
storage tanks than in the secondary sodium storage tanks, potential
hazard consequences of releasing the contents of a primary sodium
tank are bounded by the hypothetical secondary sodium scenario
evaluated in the Fermi 1 Safety Analysis Report. For these reasons,
the proposed change does not involve a significant reduction in the
margin of safety.
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, NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: Monroe County Library System,
3700 South Custer Road, Monroe, Michigan 48161.
Attorney for licensee: John Flynn, Esquire, Detroit Edison Company,
2000 Second Avenue, Detroit, Michigan 48226.
NRC Branch Chief: Larry W. Camper.
Duke Energy Corporation, et al., Docket Nos. 50-413 and 50-414, Catawba
Nuclear Station, Units 1 and 2, York County, South Carolina
Date of amendment request: July 22, 1998, supplemented by October
22, 1998, January 28, May 6 and June 24, 1999.
Description of amendment request: By the referenced submittals the
licensee requested the Catawba Technical Specifications be changed to
permit the licensee's planned use of fuel supplied by Westinghouse,
which has different design characteristics from the fuel currently in
use. The staff has previously published two Notices of Consideration of
Issuance of Amendments and Proposed No Significant Hazards
Consideration of Issuance of Amendments. The first notice, dated
November 18, 1998 (63 FR 64108), covers the submittals dated July
[[Page 43771]]
22 and October 22, 1998. The second notice, dated May 19, 1999 (64 FR
27317), covers the submittal dated May 6, 1999. The June 24, 1999,
submittal actually requested an amendment separate from that described
above, but nevertheless conveyed a revised proposed Figure 2.1.1-1,
``Reactor Core Safety Limits--Four Loops in Operation'', superseding
what was originally proposed in the licensee's previous submittals.
Hence, this Notice only covers the revised proposed Figure 2.1.1-1. The
Notices referenced above are unaffected.
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 for the June 24, 1999, submittal. The staff has reviewed
the licensee's analysis and has performed its own analysis as follows:
First Standard
No. The proposed changes to Figure 2.1.1-1 will not affect the
safety function and will not involve any change to the design or
operation of any plant system or component. The revised Figure 2.1.1-1
restricts reactor coolant flow to within previously analyzed
temperature and pressure conditions. Therefore, no accident
probabilities or consequences will be impacted.
Second Standard
No. The proposed changes will not lead to any hardware or operating
procedure change. Hence, no new equipment failure modes or accidents
from those previously evaluated will be created.
Third Standard
No. Margin of safety is associated with confidence in the design
and operation of the plant; specifically, the ability of the fission
product barriers to perform their design functions during and following
an accident. The proposed changes to Figure 2.1.1-1 do not involve any
change to plant design, operation, or analysis. Thus, the margin of
safety previously analyzed and evaluated is maintained.
Based on this analysis, it appears that the three standards of 10
CFR 50.92(c) are satisfied for the proposed change to Figure 2.1.1-1.
Therefore, the NRC staff proposes to determine that the amendment
request involves no significant hazards consideration.
Local Public Document Room location: York County Library, 138 East
Black Street, Rock Hill, South Carolina.
Attorney for licensee: Ms. Lisa F. Vaughn, Legal Department
(PB05E), Duke Energy Corporation, 422 South Church Street, Charlotte,
North Carolina.
NRC Section Chief: Richard L. Emch, Jr.
Duke Energy Corporation, et al., Docket Nos. 50-413 and 50-414, Catawba
Nuclear Station, Units 1 and 2, York County, South Carolina
Date of amendment request: June 24, 1999.
Description of amendment request: The proposed amendments would
change the Technical Specifications (TS) as follows: (1) Revise Figure
2.1.1-1, ``Reactor Core Safety Limits--Four Loops in Operation,'' which
defines the current limits of reactor coolant system (RCS) flow under
different combinations of pressure and temperature; (2) revise the
Actions associated with Limiting Condition of Operation (LCO) 3.4.1 and
Table 3.4.1-1 to reflect the updated assumptions for reactor coolant
flow, temperature and pressure; and (3) delete Figure 3.4.1-1, ``RCS
Total Flow Rate Versus Rated Thermal Power--Four Loops in Operation,''
since these requirements are being relocated to LOC 3.4.1 and Table
3.4.1-1.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration for the June 24, 1999, submittal, which is presented
below:
First Standard
No component modification, system realignment, or change in
operating procedure will occur which could affect the probability of
any accident or transient. The increase in RCS total flow rate limit
will not change the probability of actuation of any Engineered
Safety Feature or other device. In order to provide more margin in
the core design limits and allow more flexibility for future cycle-
specific core design, the analyses that establish these limits were
reanalyzed at the proposed TS minimum RCS total flow rate limit. The
impact of the power/flow tradeoff is determined for each reanalyzed
event either by qualitative evaluation or by explicit reanalysis.
An increase in the Technical Specification minimum RCS total
flow rate limit and the revised power/flow tradeoff will not
adversely affect the steady-state or transient analyses documented
in Chapters 3, 4, 6, and 15 of the McGuire and Catawba Nuclear
Station UFSARs [Updated Final Safety Analysis Reports]. The reduced
RCS low flow reactor trip setpoint and allowable value will not
increase the consequences of the partial loss of forced reactor
coolant flow and reactor coolant pump shaft seizure accidents. In
these transient reanalyses, the minimum DNBR and peak primary system
pressure acceptance criteria are not adversely affected. Therefore,
the proposed changes will not involve an increase in the probability
or consequences of an accident previously evaluated.
Second Standard
No component modification, system realignment, or change in
operating procedure will occur which could create the possibility of
a new or different kind of accident. As described in Attachment 3,
the proposed increase in Technical Specification minimum RCS total
flow rate limit and revised power/flow tradeoff will not adversely
affect the steady-state or transient analyses documented in Chapters
3, 4, 6, and 15 of the McGuire and Catawba Nuclear Station UFSARs.
Therefore, the proposed changes will not create the possibility of a
new or different kind of accident from any accident previously
evaluated.
Third Standard
These amendments will not involve a significant reduction in a
margin of safety. As described in Attachment 3, the increase in
minimum RCS total flow rate limit and revised power/flow tradeoff
will not adversely affect the steady-state or transient analyses
documented in Chapters 3, 4, 6, and 15 of the McGuire and Catawba
Nuclear Station UFSARs. DNBR, fuel clad intergrity, reactor vessel
integrity and containment integrity will not be adversely affected
by the proposed changes. Therefore, the proposed changes will not
involve any 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: York County Library, 138 East
Black Street, Rock Hill, South Carolina.
Attorney for licensee: Ms. Lisa F. Vaughn, Legal Department
(PB05E), Duke Energy Corporation, 422 South Church Street, Charlotte,
North Carolina.
NRC Section Chief: Richard L. Emch, Jr.
Duke Energy Corporation, et al., Docket Nos. 50-369 and 50-370, McGuire
Nuclear Station, Units 1 and 2, Mecklenburg County, North Carolina
Date of amendment request: July 22, 1998, supplemented by October
22, 1998, January 28, May 6 and June 24, 1999.
Description of amendment request: By the referenced submittals the
licensee requested the McGuire Technical Specifications be changed to
permit the licensee's planned use of fuel supplied by Westinghouse,
which has different design characteristics from the fuel currently in
use. The staff has previously published two Notices of
[[Page 43772]]
Consideration of Issuance of Amendments and Proposed No Significant
Hazards Consideration of Issuance of Amendments. The first notice,
dated December 16, 1998 (63 FR 69338), covers the submittals dated July
22 and October 22, 1998. The second notice, dated May 19, 1999 (64 FR
35202), covers the submittal dated May 6, 1999. The June 24, 1999,
submittal actually requested an amendment separate from that described
above, but nevertheless conveyed a revised proposed Figure 2.1.1-1,
``Reactor Core Safety Limits--Four Loops in Operation,'' superseding
what was originally proposed in the licensee's previous submittals.
Hence this Notice only covers the revised proposed Figure 2.1.1-1. The
Notices referenced above are unaffected.
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 for the June 24, 1999, submittal. The staff has reviewed
the licensee's analysis, and has performed its own analysis as follows:
First Standard
No. The proposed changes to Figure 2.1.1-1 will not affect the
safety function and will not involve any change to the design or
operation of any plant system or component. The revised Figure 2.1.1-1
restricts reactor coolant flow to within previously analyzed
temperature and pressure conditions. Therefore, no accident
probabilities or consequences will be impacted.
Second Standard
No. The proposed changes would not lead to any hardware or
operating procedure change. Hence, no new equipment failure modes or
accidents from those previously evaluated will be created.
Third Standard
No. Margin of safety is associated with confidence in the design
and operation of the plant; specifically, the ability of the fission
product barriers to perform their design functions during and following
an accident. The proposed changes to Figure 2.1.1-1 do not involve any
change to plant design, operation or analysis. Thus, the margin of
safety previously analyzed and evaluated is maintained.
Based on this analysis, it appears that the three standards of 10
CFR 50.92(c) are satisfied for the proposed change to Figure 2.1.1-1.
Therefore, the NRC staff proposes to determine that the amendment
request involves no significant hazards consideration.
Local Public Document Room location: J. Murrey Atkins Library,
University of North Carolina at Charlotte, 9201 University City
Boulevard, Charlotte, North Carolina.
Attorney for licensee: Mr. Albert Carr, Duke Energy Corporation,
422 South Church Street, Charlotte, North Carolina.
NRC Section Chief: Richard L. Emch, Jr.
Duke Energy Corporation, Docket Nos. 50-369 and 50-370, McGuire Nuclear
Station, Units 1 and 2, Mecklenburg County, North Carolina
Date of amendment request: June 24, 1999.
Description of amendment request: The proposed amendments would
change the Technical Specifications (TS) as follows: (1) Revise Figure
2.1.1-1, ``Reactor Core Safety Limits--Four Loops in Operation,'' which
defines the current limits of reactor coolant system (RCS) flow under
different combinations of pressure and temperature; (2) revise Table
3.3.1-1 to provide values for the trip setpoint and allowable value for
RCS Flow-Low; (3) revise Table 3.3.1-1 to make a typographical
correction for T, the nominal T-average at Rated Thermal Power; (4)
revise the Actions associated with Limiting Condition of Operation
(LCO) 3.4.1 and Table 3.4.1-1 to reflect the updated assumptions for
reactor coolant flow, temperature and pressure; and (5) delete Figure
3.4.1-1, ``RCS Total Flow Rate Versus Rated Thermal Power--Four Loops
in Operation,'' since these requirements are being relocated to LCO
3.4.1 and Table 3.4.1-1.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
First Standard
No component modification, system realignment, or change in
operating procedure will occur which could affect the probability of
any accident or transient. The increase in RCS total flow rate limit
will not change the probability of actuation of any Engineered
Safety Feature or other device. In order to provide more margin in
the core design limits and allow more flexibility for future cycle-
specific core design, the analyses that establish these limits were
reanalyzed at the proposed TS minimum RCS total flow rate limit. The
impact of the power/flow tradeoff is determined for each reanalyzed
event either by qualitative evaluation or by explicit reanalysis.
An increase in the Technical Specification minimum RCS total
flow rate limit and the revised power/flow tradeoff will not
adversely affect the steady-state or transient analyses documented
in Chapters 3, 4, 6, and 15 of the McGuire and Catawba Nuclear
Station UFSARs [Updated Final Safety Analysis Reports]. The reduced
RCS low flow reactor trip setpoint and allowable value will not
increase the consequences of the partial loss of forced reactor
coolant flow and reactor coolant pump shaft seizure accidents. In
these transient reanalyses, the minimum DNBR and peak primary system
pressure acceptance criteria are not adversely affected. Therefore,
the proposed changes will not involve an increase in the probability
or consequences of an accident previously evaluated.
Second Standard
No component modification, system realignment, or change in
operating procedure will occur which could create the possibility of
a new or different kind of accident. As described in Attachment 3,
the proposed increase in Technical Specification minimum RCS total
flow rate limit and revised power/flow tradeoff will not adversely
affect the steady-state or transient analyses documented in Chapters
3, 4, 6, and 15 of the McGuire and Catawba Nuclear Station UFSARs.
Therefore, the proposed changes will not create the possibility of a
new or different kind of accident from any accident previously
evaluated.
Third Standard
These amendments will not involve a significant reduction in a
margin of safety. As described in Attachment 3, the increase in
minimum RCS total flow rate limit and revised power/flow tradeoff
will not adversely affect the steady-state or transient analyses
documented in Chapters 3, 4, 6, and 15 of the McGuire and Catawba
Nuclear Station UFSARs. DNBR, fuel clad integrity, reactor vessel
integrity and containment integrity will not be adversely affected
by the proposed changes. Therefore, the proposed changes will not
involve any 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: J. Murrey Atkins Library,
University of North Carolina at Charlotte, 9201 University City
Boulevard, Charlotte, North Carolina.
Attorney for licensee: Mr. Albert Carr, Duke Energy Corporation,
422 South Church Street, Charlotte, North Carolina.
NRC Section Chief: Richard L. Emch, Jr.
[[Page 43773]]
Entergy Operations, Inc. (EOI), Docket Nos. 50-313 and 50-368, Arkansas
Nuclear One, Units 1 and 2, Pope County, Arkansas
Date of amendment request: July 14, 1999.
Description of amendment request: The proposed amendments delete
requirements from the Technical Specifications to maintain a Post
Accident Sampling System (PASS). Licensees were required to implement
PASS upgrades as a result of NUREG-0737, ``Clarification of TMI [Three
Mile Island] Action Plan Requirements,'' and Regulatory Guide 1.97,
Revision 3, ``Instrumentation for Light-Water-Cooled Nuclear Power
Plants to Access Plant and Environs Conditions During and Following an
Accident.'' Implementation of these upgrades were an outcome of the
NRC's lessons learned from the accident that occurred at Three Mile
Island, Unit 2. EOI has stated that the information obtained using PASS
can be readily obtained through other means or is of little use in the
assessment and mitigation of accident conditions.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
Criterion 1--[The Proposed Change] Does Not Involve a Significant
Increase in the Probability or Consequences of an Accident
Previously Evaluated
The PASS was originally designed to perform many sampling and
analysis functions. These functions were designed and intended to be
used in post accident situations and were put into place as a result
of the Three Mile Island Unit 2 (TMI-2) accident. The specific
intent of the PASS was to provide a system that has the capability
to obtain and analyze samples of plant fluids containing potentially
high levels of radioactivity, without exceeding plant personnel
radiation exposure limits. Analytical results of these samples would
be used largely for verification purposes in aiding the plant staff
in assessing the extent of core damage and subsequent offsite
radiological dose projections.
In the 20 years since the TMI-2 accident and the consequential
promulgation of post accident sampling requirements, operating
experience has demonstrated that the actual benefits afforded by a
PASS provide little benefit to post accident mitigation. Past
experience has indicated that there exists in-plant instrumentation
and methodologies available in lieu of a PASS for collecting and
assimilating information needed to assess core damage following an
accident. Furthermore, the implementation of Severe Accident
Management Guidance (SAMG) emphasizes accident management strategies
based on in-plant instruments. These strategies provide guidance to
the plant staff for mitigation and recovery from a severe accident.
Based on current severe accident management strategies and
guidelines, it is determined that the PASS provides no benefit to
the plant staff in coping with an accident. The use of the PASS may
be counter productive to plant operations since its operation will
divert resources away from accident management, the sample results
may be ambiguous and may be misinterpreted, and the use of PASS may
restrict personnel movements in certain areas of the plant while
resulting in additional fission product release points outside the
containment.
The regulatory requirements for the PASS can be eliminated
without degrading the plant emergency response. The emergency
response, in this sense, refers to the methodologies used in
ascertaining the condition of the reactor core, mitigating the
consequences of an accident, assessing and projecting offsite
releases of radioactivity, and establishing protective action
recommendations to be communicated to offsite authorities.
Additionally, preliminary discussions with the State of Arkansas
have indicated that the elimination of the PASS will not adversely
impact actions taken by the State during an emergency event. The
elimination of the PASS will not prevent an accident management
strategy that meets the initial intent of the post-TMI-2 [accident]
guidance through the use of the SAMGs, the emergency plan (EP), the
emergency operating procedures (EOP), and site survey monitoring
that support modification of emergency plan PARs [protective action
recommendations].
Therefore, the elimination of PASS requirements of the ANO-1 and
ANO-2 [Arkansas Nuclear One, Unit 1 and Unit 2] Technical
Specifications (TS) and subsequent requested relief from the
requirements of NUREG-0737 and Regulatory Guide 1.97, Revision 3,
does not involve a significant increase in the probability or
consequences of any accident previously evaluated.
Criterion 2--[The Proposed Change] Does Not Create the Possibility
of a New or Different Kind of Accident from any Previously
Evaluated
The relief from PASS related NUREG-0737 and Regulatory Guide
1.97 requirements in addition to the proposed TS changes will not
result in any failure mode not previously analyzed. The PASS was
intended to allow for verification of the extent of reactor core
damage and also to provide an input to offsite dose projection
calculations. The PASS is not considered an accident precursor, nor
does its existence or elimination have any adverse impact on the
pre-accident state of the reactor core or post accident confinement
of radionuclides within the containment building.
Therefore, this change does not create the possibility of a new
or different kind of accident from any previously evaluated.
Criterion 3--[The Proposed Change] Does Not Involve a Significant
Reduction in the Margin of Safety
The elimination of the PASS, in light of existing plant
equipment, instrumentation, procedures, and programs that provide
effective mitigation of and recovery from reactor accidents, results
in a neutral impact to the margin of safety at ANO-1 and ANO-2. Non-
PASS methodologies are designed to provide rapid assessment of
current reactor core conditions and the direction of degradation
while effectively responding to the event in order to mitigate the
consequences of the accident. The use of a PASS is redundant and
does not provide quick recognition of core events nor rapid response
to events in progress. The intent of the requirements established as
a result of the TMI-2 accident can be adequately met without
reliance on a PASS.
Therefore, this change does not involve a significant reduction
in the margin of safety.
Therefore, based upon the reasoning presented above and the
previous discussion of the amendment request, Entergy Operations,
Inc. has determined that the requested change does not involve a
significant hazards consideration.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: Tomlinson Library, Arkansas
Tech University, Russellville, Arkansas 72801.
Attorney for licensee: Nicholas S. Reynolds, Esquire, Winston and
Strawn, 1400 L Street, NW., Washington, DC 20005-3502.
NRC Section Chief: Robert A. Gramm.
Indiana Michigan Power Company, Docket Nos. 50-315, Donald C. Cook
Nuclear Plant, Unit 1, Berrien County, Michigan.
Date of amendment requests: December 3, 1998.
Description of amendment requests: The proposed amendments would
revise Technical Specification (TS)
3/4.7.7, ``Sealed Source Contamination,'' and the associated bases to
address testing requirements for fission detectors. The proposed
changes would provide consistency between the unit 1 and Unit 2 TS
requirements and with NUREG-0452, ``Standard 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:
[[Page 43774]]
Criterion 1
Does the change involve a significant increase in the
probability or consequences of an accident previously evaluated?
The proposed changes clarify testing requirements for fission
detectors. When the fission detectors are tested for surface
contamination, they do not interfere with plant equipment and they
do not affect plant operation. The detectors are not assumed to
initiate an accident; therefore, the probability of an accident
previously evaluated is not changed.
Conducting tests prior to using a new fission detector provides
assurance that intake limits will not be exceeded. There is no
change to the nuclear material contained in the detector. The
fission detectors are not used to mitigate the consequences of
postulated accidents. Therefore, the consequences of an accident
remain the same as previously evaluated.
Therefore, it is concluded that the proposed changes do not
involve a significant increase in the probability or consequences of
an accident previously evaluated.
Criterion 2
Does the change create the possibility of a new or different
type of accident from any accident previously evaluated?
The proposed changes do not affect the design or operation of
systems, structures, or components in the plant. There are no
changes to parameters governing plant operation, and no new or
different types of equipment will be installed. Therefore, it is
concluded that the proposed changes do not create the possibility of
a new or different kind of accident from any previously evaluated.
Criterion 3
Does the change involve a significant reduction in a margin of
safety?
The proposed changes do not introduce new equipment, equipment
modifications, or new or different modes of plant operation. These
changes do not affect the operational characteristics of any
equipment or systems.
Therefore, it is concluded that these changes do not involve a
significant reduction in the margin of safety.
Conclusion
In summary, based upon the above evaluation, the Licensee has
concluded that these changes involve no significant hazards
consideration.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment requests involve no significant hazards consideration.
Local Public Document Room location: Maud Preston Palenske Memorial
Library, 500 Market Street, St. Joseph, MI 49085.
Attorney for licensee: Gerald Charnoff, Esq., Shaw, Pittman, Potts
and Trowbridge, 2300 N Street, NW., Washington, DC 20037.
NRC Section Chief: Claudia M. Craig.
PECO Energy Company, Public Service Electric and Gas Company, Delmarva
Power and Light Company, and Atlantic City Electric Company, Dockets
Nos. 50-277 and 50-278, Peach Bottom Atomic Power Station, Units Nos. 2
and 3, York County, Pennsylvania
Date of application for amendments: March 29, 1999.
Description of amendment request: The proposed amendments would
revise Technical Specification Surveillance Requirement 3.9.1.1 and the
associated Bases 3.9.1 to delete the requirement for the refuel
platform fuel grapple fully retracted position interlock.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. The proposed TS changes do not involve a significant increase
in the probability or consequences of an accident previously
evaluated.
This change removes a redundant interlock and will not impact
the functionality of associated interlocks. The removal of the
``refuel platform fuel grapple fully retracted position'' refueling
interlock will not affect the ability of the remaining refueling
interlocks to produce a rod block during fuel moves. The
administrative controls in place do not allow control rod
withdrawals while fuel is being moved or fuel movement while rods
are withdrawn. The fuel grapple full up interlock is a redundant and
diverse interlock and its removal has no impact on plant safety. The
interlock's intent, to provide a backup to the load sensor, is not
required since the setpoint is currently low enough to provide
adequate protection therefore not significantly increasing the
probability of an accident previously evaluated.
The refueling interlocks are not used to prevent or to mitigate
the fuel handling accident as discussed in the PBAPS [Peach Bottom
Atomic Power Station], Units 2 and 3, UFSAR [Updated Final Safety
Analysis Report], Section 14.6.4 (``Refueling Accident''). The
``refuel platform fuel grapple fully retracted position'' interlock
and the ``refuel platform fuel grapple, fuel loaded'' interlock both
provide rod blocks during fuel movement over the core. Additionally,
the refueling interlocks are not assumed as an initial condition in
the control rod drop accident as discussed in the PBAPS, UFSAR,
Section 14.6.2 (``Control Rod Drop Accident''). The control rod drop
accident is only analyzed when the reactor is critical and not
during refueling operations.
The refueling interlocks associated with the refueling platform
provide rod blocks to ensure that control rods can not be withdrawn
when fuel is being moved over the core (PBAPS, Units 2 and 3, UFSAR
Section 14.5.3.3, ``Control Rod Removal Error During Refueling'').
They are also used to prevent refueling bridge motion towards the
core if a control rod is withdrawn during fuel movements (PBAPS,
Units 2 and 3, UFSAR Section 14.5.3.4, ``Fuel Assembly Insertion
Error During Refueling''). These interlocks prevent the possibility
of an inadvertent criticality during refueling. However, removal of
the ``refuel platform fuel grapple fully retracted position''
interlock, which is a redundant and diverse interlock, will not
prevent the remaining interlocks from performing their intended
safety functions. The refueling interlocks are active with the mode
switch in refuel, and are only designed to reinforce administrative
procedures for moving fuel. Therefore, the proposed TS changes will
not involve a significant increase in the probability of an accident
previously evaluated.
The fuel or core loading characteristics are not altered by the
removal of this interlock. The dose resulting from a potential
control rod withdrawal or fuel bundle error event is not increased
as a result of eliminating this redundant and diverse interlock.
Therefore, the removal of the ``refuel platform fuel grapple fully
retracted position'' interlock will not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
2. The proposed TS changes do not create the possibility of a
new or different kind of accident from any accident previously
evaluated.
The refueling interlocks are not accident initiators. Nor will
any new failure mode be introduced by the removal of the ``refuel
platform fuel grapple fully retracted position'' interlock. The
interlocks are used to reinforce administrative controls which
prevent fuel movement over the core with control rods withdrawn and
preclude withdrawal of control rods when the fuel is being moved
over the core. The interlock for ensuring the fuel grapple is fully
up, is a redundant and diverse interlock since a load sensor
determines if the main hoist is loaded with a fuel bundle. This
redundant and diverse interlock prevents the withdrawal of a control
rod while moving fuel during refueling. The setpoint is low enough
to ensure a rod block will be received if the main hoist is being
used to move fuel over the core and to prevent movement of the
refueling bridge. The remaining refueling interlocks, in combination
with the refueling procedures, will still prevent an inadvertent
criticality during refueling operations. Fuel handling procedures
require that interlocks be verified by observing the rod withdraw
permissive light in the control room, and by monitoring the rod
block interlock light on the refuel bridge. Therefore, the proposed
TS changes do not create the possibility of a new or different kind
of accident from any accident previously evaluated.
3. The proposed TS changes do not involve a significant
reduction in a margin of safety.
This change will not involve a significant reduction in a margin
of safety. The ``refuel platform fuel grapple fully retracted
position'' interlock is redundant and diverse to the ``refuel
platform fuel grapple, fuel
[[Page 43775]]
loaded'' interlock on the main hoist. The other two hoists on the
bridge have the fuel loaded interlock but do not have the backup
full up position interlock. The margin of safety of the refueling
interlocks will not be significantly reduced by this change since
redundant interlocks are not required (this a nonsafety-related
function) and the original justification for using it, a high load
weight setpoint, is no longer applicable. The system consists of a
single channel, and no current design basis for using redundant and
diverse interlocks to provide the rod block. Additionally, the
Reactor Manual Control System will not be affected by this change.
The system's ability to provide a rod block is not affected by this
change.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: Government Publications
Section, State Library of Pennsylvania, (REGIONAL DEPOSITORY) Education
Building, Walnut Street and Commonwealth Avenue, Box 1601, Harrisburg,
PA 17105.
Attorney for Licensee: J. W. Durham, Sr., Esquire, Sr. V.P. and
General Counsel, PECO Energy Company, 2301 Market Street, Philadelphia,
PA 19101.
NRC Section Chief: James W. Clifford.
Power Authority of the State of New York, Docket No. 50-333, James A.
FitzPatrick Nuclear Power Plant, Oswego County, New York
Date of amendment request: April 5, 1999
Description of amendment request: The proposed changes would revise
Appendix A (Section 6.1) and Appendix B (Section 7.1) of the James A.
FitzPatrick Technical Specifications. The proposed changes would remove
the position title of General Manager from these sections and would
state that if the Site Executive Officer (SEO) is unavailable, he will
delegate his responsibilities to another staff member, in writing. In
addition the position title of Resident Manager, used in Apendix B,
Section 7.1, would be replaced by Site Executive Officer.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
Consistent with the criteria of 10 CFR 50.92, the proposed
application is judged to involve no significant hazards based on the
following information:
(1) Does the proposed license amendment involve a significant
increase in the probability or consequences of an accident
previously analyzed?
Response: The proposed changes to Appendix A (Section 6.1) and
Appendix B (Section 7.1) are administrative in nature in that they
do not change the intent of the Technical Specifications. If the SEO
is unavailable, he will still delegate his responsibilities to a
qualified personnel member, such as the Plant Manager or one of the
General Managers. These changes can not cause an accident or
contribute to the probability or consequences of one.
The replacement of the position title of Resident Manager with
Site Executive Officer in Appendix B, Section 7.1, was already
approved by the NRC in Amendment 228.
Therefore, the proposed changes do not involve a significant
increase in the probability or consequences of an accident
previously analyzed.
(2) Does the proposed license amendment create the possibility
of a new or different kind of accident from any accident previously
evaluated?
Response: The proposed changes to Appendix A (Section 6.1) and
Appendix B (Section 7.1) are administrative in nature as they do not
affect the function of plant equipment or the way the equipment
operates. The changes do not change the intent of the current TS, in
that if the SEO is unavailable, he will delegate his
responsibilities to another personnel member such as the Plant
Manager or one of the General Managers. Appendix A (Section 6.1) and
Appendix B (Section 7.1) are being revised to eliminate the need for
future TS changes to these sections resulting solely from the
creation of new or revised management positions (such as the Plant
Manager), title changes to the position of General Manager, or a
change to the number of General Managers. These types of
organizational changes will be evaluated using the criteria of 10
CFR 50.59.
The replacement of the position title of Resident Manager with
Site Executive Officer in Appendix B, Section 7.1, was already
approved by the NRC in Amendment 228.
Therefore, the proposed license amendment does not create the
possibility of a new or different kind of accident from any accident
previously evaluated.
(3) Does the proposed amendment involve a significant reduction
in a margin of safety?
Response: The proposed changes to Appendix A (Section 6.1) and
Appendix B (Section 7.1) are administrative changes associated with
the delegation of the SEO's responsibilities when he is unavailable.
These changes do not change the intent of the current TS, in that in
the SEO's absence, he will still delegate his responsibilities to
other personnel members such as the Plant Manager or General
Managers.
The replacement of the position title of Resident Manager with
Site Executive Officer in Appendix B, Section 7.1, was already
approved by the NRC in Amendment 228.
Therefore, the proposed changes do not involve a significant
reduction in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: Reference and Documents
Department, Penfield Library, State University of New York, Oswego, New
York 13126.
Attorney for licensee: Mr. David E. Blabey, 1633 Broadway, New
York, New York 10019.
NRC Section Chief: S. Singh Bajwa.
Power Authority of the State of New York, Docket No. 50-333, James A.
FitzPatrick Nuclear Power Plant, Oswego County, New York
Date of amendment request: June 22, 1999.
Description of amendment request: The proposed amendment would
revise the Technical Specifications by changes to the Pressure and
Temperature (P-T) limits. As part of this proposed change the licensee
is proposing to add separate bottom head curves ABH and
BBH for in-service hydrostatic and leak tests and non-
nuclear heatup and cooldown, respectively. In addition, a non-beltline
curve (i.e., ANB) for in-service hydrostatic and leak tests
is being proposed.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
Operation of the FitzPatrick plant in accordance with the
proposed amendment would not involve a significant hazards
consideration as defined in 10 CFR 50.92, since it would not:
1. Involve a significant increase in the probability or
consequences of an accident previously evaluated.
The changes to the P-T curves are being proposed to preclude
brittle fracture of RPV [Reactor Pressure Vessel] materials for up
to 32 EFPY [effective full-power years]. In addition to the P-T
curve for up to 32 EFPY, a P-T curve has been prepared for exposures
up to 24 EFPY to shorten outage time for startups conducted prior to
reaching this exposure. Safety margins specified in 10 CFR 50,
Appendix G and Appendix G to Section XI of the ASME [American
Society of Mechanical Engineers Boiler and Pressure Vessel Code]
will continue to be met for each of these curves. Therefore, there
is not a significant increase in the probability of an accident
previously evaluated.
The RPV, as part of the reactor coolant system, provides a
barrier to the release of reactor coolant. Operation in accordance
[[Page 43776]]
with the proposed amendment will preclude brittle fracture of the
RPV consistent with current requirements, and consequently, does not
significantly increase the consequences of an accident previously
evaluated.
Based on the above, operation of the FitzPatrick plant in
accordance with the proposed amendment will not involve a
significant increase in the probability or consequences of an
accident previously evaluated.
2. Create the possibility of a new or different kind of accident
from any accident previously evaluated.
The proposed change does not involve any physical alterations to
plant configurations or introduce any new accident precursors which
could initiate a new or different kind of accident. The proposed
change does not affect the intended function of the RPV nor does it
affect the operation of the RPV in a way which would create a new or
different kind of accident. The changes to the P-T curves are being
proposed to preclude brittle fracture of RPV materials for up to 32
EFPY. Safety margins specified in 10 CFR 50, Appendix G and Appendix
G to Section XI of the ASME Code will continue to be met. Therefore,
operation of the FitzPatrick plant in accordance with the proposed
amendment will not create the possibility of a new or different kind
of accident from any accident previously evaluated.
3. Involve a significant reduction in a margin of safety.
The existing FitzPatrick P-T curves were developed using safety
margins for brittle fracture found in 10 CFR 50 Appendix G. The
proposed FitzPatrick P-T curves, which are valid for up to 32 EFPY
of operation, were also developed using safety margins for brittle
fracture found in 10 CFR 50 Appendix G. Based on this, operation of
the FitzPatrick plant in accordance with the proposed amendment will
continue to preclude brittle fracture of the RPV materials during
in-service hydrostatic and leak tests, non-nuclear heatup and
cooldown, and core critical operation without a significant
reduction in a margin of safety. Therefore, operation of the
FitzPatrick plant in accordance with the proposed amendment will not
involve a significant reduction in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: Reference and Documents
Department, Penfield Library, State University of New York, Oswego, New
York 13126.
Attorney for licensee: Mr. David E. Blabey, 1633 Broadway, New
York, New York 10019.
NRC Section Chief: S. Singh Bajwa.
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: July 2, 1999.
Description of amendment request: The proposed amendments would
relocate the requirements from Technical Specification 3/4.3.4,
``Instrumentation, Turbine Overspeed Protection,'' and the associated
bases to licensee-controlled documents in accordance with Generic
Letter 95-10, ``Relocation of Selected Technical Specifications
Requirements Related to Instrumentation.''
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. Will not involve a significant increase in the probability or
consequences of an accident previously evaluated.
The requested amendments will not involve an increase in the
probability or consequences of an accident previously evaluated.
Relocation of the affected Technical Specification sections and
their Bases to the Salem UFSAR [Updated Final Safety Analysis
Report] will have no affect on the probability that any accident
will occur. Additionally, the consequences of an accident will not
be affected because the Turbine Overspeed Protection system will
continue to be utilized in the same manner as before. No impact on
the plant response to accidents will be created.
2. Will not create the possibility of a new or different kind of
accident from any previously evaluated.
The proposed amendments will not create the possibility of a new
or different kind of accident from any accident previously
evaluated. No new accident causal mechanisms will be created as a
result of the relocation of the Turbine Overspeed Protection system
Technical Specification requirements and their Bases to the Salem
UFSAR. Plant operation will not be affect by the proposed amendments
and no new failure modes will be created.
3. Will not involve a significant reduction in a margin of
safety.
The proposed amendments will not involve a reduction in the
margin of safety. Relocation of the affected Technical Specification
requirements to the Salem UFSAR is consistent with NUREG 1431,
Standard Technical Specifications--Westinghouse Plants which do not
include Technical Specification requirements for the Turbine
Overspeed Protection system. The proposed amendments are consistent
with the NRC philosophy of encouraging utilities to propose
amendments that are consistent with NUREG 1431.
Based on the above, the proposed changes will not involve a
significant reduction in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: Salem Free Public Library, 112
West Broadway, Salem, NJ 08079.
Attorney for licensee: Jeffrie J. Keenan, Esquire, Nuclear Business
Unit--N21, P.O. Box 236, Hancocks Bridge, NJ 08038.
NRC Section Chief: James W. Clifford.
Northeast Nuclear Energy Company, et al., Docket No. 50-336, Millstone
Nuclear Power Station, Unit No. 2, New London County, Connecticut
Date of amendment request: July 16, 1999.
Description of amendment request: Proposed Technical Specifications
(TS) change to increase the action requirement time to be in Mode 3 if
the temperature of the ultimate heat sink (UHS) exceeds the TS limit of
75 deg.F. The increased time will only apply if the UHS temperature is
between 75 and 77 deg.F. The Bases for the associated TS will also be
revised.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. Involve a significant increase in the probability or
consequences of an accident previously evaluated.
The proposed changes will allow plant operation to continue for
an additional 12 hours with the temperature of the Ultimate Heat
Sink (UHS) up to 2 deg.F above the Technical Specification limit of
75 deg.F. This increase in UHS temperature will not affect the
normal operation of the plant to the extent which would make any
accident more likely to occur. In addition, there exists adequate
margin in the safety systems and heat exchangers to assure the
safety functions are met at the higher temperature. An evaluation
has confirmed that safe shutdown will be achieved and maintained for
a loss of coolant accident (LOCA) with a loss of normal power (LNP)
and a single active failure with a UHS water temperature as high as
77 deg.F.
The proposed changes will have no adverse effect on plant
operation, or the availability or operation of any accident
mitigation equipment. The plant response to the design basis
accidents will not change. In addition, the proposed changes can not
cause an accident. Therefore, there will be no significant increase
in the probability or consequences of an accident previously
evaluated.
[[Page 43777]]
2. Create the possibility of a new or different kind of accident
from any accident previously evaluated.
The proposed changes will allow plant operation to continue for
an additional 12 hours with the temperature of the UHS up to 2
deg.F above the Technical Specification limit of 75 deg.F. This
will not alter the plant configuration (no new or different type of
equipment will be installed) or require any new or unusual operator
actions. The proposed changes will not alter the way any structure,
system, or component functions and will not significantly alter the
manner in which the plant is operated. There will be no adverse
effect on plant operation or accident mitigation equipment. The
proposed changes do not introduce any new failure modes. Also, the
response of the plant and the operators following these accidents is
unaffected by the changes. In addition, the UHS is not an accident
initiator. Therefore, the proposed changes will 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 proposed changes will allow plant operation to continue for
an additional 12 hours with the temperature of the UHS up to 2
deg.F above the Technical Specification limit of 75 deg.F.
Evaluations have been performed which demonstrate that the safety
systems have adequate margin to ensure their safety functions can be
met with a UHS temperature of 77 deg.F. In addition, safe shutdown
capability has been demonstrated for a UHS water temperature as high
as 77 deg.F.
The proposed changes will have no adverse effect on plant
operation or equipment important to safety. The plant response to
the design basis accidents will not change and the accident
mitigation equipment will continue to function as assumed in the
design basis accident analysis. Therefore, there will be no
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: Learning Resources Center,
Three Rivers Community-Technical College, 574 New London Turnpike,
Norwich, Connecticut, and the Waterford Library, ATTN: Vince Juliano,
49 Rope Ferry Road, Waterford, Connecticut.
Attorney for licensee: Lillian M. Cuoco, Esq., Senior Nuclear
Counsel, Northeast Utilities Service Company, P.O. Box 270, Hartford,
Connecticut.
NRC Section Chief: James W. Clifford.
PECO Energy Company, Public Service Electric and Gas Company, Delmarva
Power and Light Company, and Atlantic City Electric Company, Docket No.
50-278, Peach Bottom Atomic Power Station, Unit No. 3, York County,
Pennsylvania
Date of application for amendment: July 12, 1999.
Description of amendment request: The proposed change will revise
Technical Specifications (TSs) TS 2.1.1.2, ``Reactor Core [Safety
Limits] SLs,'' and Section 5.6.5, ``Core Operating Limits Report.''
These Sections will be revised to: (1) Incorporate revised Safety Limit
Minimum Critical Power Ratios (SLMCPRs) due to the use of a cycle-
specific analysis performed by General Electric Nuclear Energy (GENE)
for Peach Bottom Atomic Power Station, Unit 3, (PBAPS, Unit 3) Cycle
13, (2) delete previously added footnotes which are no longer
necessary, and (3) update a reference contained in TS 5.6.5.b.2 which
documents an analytical method used to determine the core operating
limits.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. The proposed TS changes do not involve a significant increase
in the probability or consequences of an accident previously
evaluated.
The derivation of the cycle specific SLMCPRs for incorporation
into the TS, and its use to determine cycle specific thermal limits,
has been performed using the methodology discussed in ``General
Electric Standard Application for Reactor Fuel,'' NEDE-24011-P-A-13,
and U.S. Supplement, NEDE-24011-P-A-13-US, August 1996, and
Amendment 25. Amendment 25 was approved by the NRC in a March 11,
1999 safety evaluation report. This change in SLMCPRs cannot
increase the probability or severity of an accident.
The basis of the SLMCPR calculation is to ensure that greater
than 99.9% of all fuel rods in the core avoid transition boiling if
the limit is not violated. The new SLMCPRs preserve the existing
margin to transition boiling and fuel damage in the event of a
postulated accident. The fuel licensing acceptance criteria for the
SLMCPR calculation apply to PBAPS, Unit 3, Cycle 13 in the same
manner as they have applied previously. The probability of fuel
damage is not increased. Therefore, the proposed TS changes do not
involve a significant increase in the probability or consequences of
an accident previously evaluated.
In addition to the change to the SLMCPR, the footnotes to TS
2.1.1.2 and TS 5.6.5.b.1 are being deleted. The footnote associated
with TS 2.1.1.2 was originally included to ensure that the SLMCPR
value was only applicable for the identified cycle. The footnote was
added to TS 5.6.5.b.1 because Amendment 25 and the R-factor
calculation methodology were not yet NRC approved. Amendment 25 and
the R-factor methodology have subsequently been approved. Therefore,
these footnotes are no longer necessary. The footnotes were for
information only, and have no impact on the design or operation of
the plant. The deletion of the footnotes associated with TS 2.1.1.2
and TS 5.6.5.b.1 is an administrative change that does not involve a
significant increase in the probability or consequences of an
accident previously evaluated.
The Revision 1 ARTS/MELLLA [Maximum Extended Load Line Limit and
ARTS Improvement Program Analysis for Peach Bottom Atomic Power
Station Unit 2 and 3,] analysis contained in TS 5.6.5.b.2 is being
updated to a Revision 2 analysis, to reflect changes that were
previously approved by the NRC as documented in the safety
evaluation report dated August 10, 1994 (Amendment No. 192 for
PBAPS, Unit 2). This is an administrative change which will ensure
that the references contained in the PBAPS Technical Specifications
are accurate and consistent with other licensing documents. No
technical changes are occurring which have not been previously
approved by the NRC. Therefore, this change does not involve a
significant increase in the probability or consequences of an
accident previously evaluated.
2. The proposed TS changes do not create the possibility of a
new or different kind of accident from any accident previously
evaluated.
The SLMCPR is a TS numerical value, designed to ensure that
transition boiling does not occur in 99.9% of all fuel rods in the
core during the limiting postulated accident. The new SLMCPRs are
calculated using NRC approved methodology discussed in ``General
Electric Standard Application for Reactor Fuel,'' NEDE-24011-P-A-13
(GESTAR-II), and U.S. Supplement, NEDE-24011-P-A-13-US, August 1996,
and Amendment 25. The SLMCPR is not an accident initiator, and its
revision will not create the possibility of a new or different kind
of accident from any accident previously evaluated.
Additionally, this proposed change will delete footnotes
contained in TS 2.1.1.2 and TS 5.6.5.b.1 as the result of the NRC
approval of analysis associated with Amendment 25 and the R-factor
methodology. The proposed change also updates the ARTS/MELLLA
analysis contained in TS 5.6.5.b.2. This revision contains
information which was previously approved by the NRC. Therefore, the
deletion of the footnotes associated with TS 2.1.1.2 and TS
5.6.5.b.1, and the updating of the reference contained in TS
5.6.5.b.2 are administrative changes that do not create the
possibility of a new or different kind of accident from any
previously evaluated.
3. The proposed TS changes do not involve a significant
reduction in a margin of safety.
There is no significant reduction in the margin of safety
previously approved by the NRC as a result of: (1) the proposed
changes
[[Page 43778]]
to the SLMCPRs, (2) the proposed change that will delete the
footnotes to TS 2.1.1.2 and TS 5.6.5.b.1, and (3) updating the
reference to the ARTS/MELLLA analysis contained in TS 5.6.5.b.2. The
new SLMCPRs are calculated using methodology discussed in ``General
Electric Standard Application for Reactor Fuel,'' NEDE-24011-P-A-13
(GESTAR-II), and U.S. Supplement, NEDE-24011-P-A-13-US, August 1996,
and Amendment 25. The fuel licensing acceptance criteria for the
calculation of the SLMCPR apply to PBAPS, Unit 3 Cycle 13 in the
same manner as they have applied previously. The SLMCPRs ensure that
greater than 99.9% of all fuel rods in the core will avoid
transition boiling if the limit is not violated when all
uncertainties are considered, thereby preserving the fuel cladding
integrity. Therefore, the proposed TS changes will not involve a
significant reduction in the margin of safety previously approved by
the NRC.
Additionally, the proposed changes that delete the footnotes to
TS 2.1.1.2 and TS 5.6.5.b.1, and update the revision to the ARTS/
MELLLA analysis contained in TS 5.6.5.b.2, are administrative
changes that will not significantly reduce the margin of safety
previously approved by the NRC.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: Government Publications
Section, State Library of Pennsylvania, (REGIONAL DEPOSITORY) Education
Building, Walnut Street and Commonwealth Avenue, Box 1601, Harrisburg,
PA 17105.
Attorney for Licensee: J. W. Durham, Sr., Esquire, Sr. V.P. and
General Counsel, PECO Energy Company, 2301 Market Street, Philadelphia,
PA 19101.
NRC Section Chief: James W. Clifford.
Rochester Gas and Electric Corporation, Docket No. 50-244, R. E. Ginna
Nuclear Power Plant, Wayne County, New York
Date of amendment request: June 28, 1999.
Description of amendment request: The proposed amendment would
revise the Improved Technical Specifications (ITS) associated with the
Reactor Coolant System (RCS) Leakage Detection Instrumentation.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. Operation of Ginna Station in accordance with the proposed
changes does not involve a significant increase in the probability
or consequences of an accident previously evaluated. The changes add
further requirements for redundancy and a requirement to perform
either an RCS water inventory balance or analyses of containment
atmosphere grab samples once within 12 hours and every 12 hours
thereafter when the particulate containment atmosphere radioactivity
monitor is unavailable while in Modes 1, 2, 3, and 4. This does not
increase the probability of an accident previously evaluated since
the compensatory actions are either a calculation utilizing
installed indication or the measurement of a sample drawn downstream
from the containment atmosphere sample isolation valves and are of
themselves not an accident initiator. The proposed compensatory
actions are based on the NUREG-1431 guidance and the proposed
frequencies are more conservative, which gives a higher assurance
that the RCS leakage rate can be adequately monitored.
Therefore, the probability or consequences of an accident
previously evaluated is not significantly increased.
2. Operation of Ginna Station in accordance with the proposed
changes does not create the possibility of a new or different kind
of accident from any accident previously evaluated. The proposed
changes add further requirements for redundancy and the proposed
change for compensatory actions when the particulate containment
atmosphere radioactivity monitor is inoperable does not of itself
involve a physical alteration of the plant (ie. no new or different
type of equipment will be added to perform the required actions) or
changes in the methods governing normal plant operation. The changes
only involve implementing currently approved alternate methods to
determine the RCS leak rate on an increased frequency. Therefore,
the possibility for a new or different kind of accident from any
accident previously evaluated is not created.
3. Operation of Ginna Station in accordance with the proposed
changes does not involve a significant reduction in a margin of
safety. The proposed changes only add conservatism in the number of
required RCS leakage detection instrumentation and add more
conservative compensatory actions that are to be taken when the
containment atmosphere particulate radioactivity monitor is
inoperable. The compensatory actions are based on the guidance of
NUREG-1431. Therefore, this change does not involve a significant
reduction in a margin of safety.
Based upon the preceding information, it has been determined that
the proposed changes do not involve a significant increase in the
probability or consequences of an accident previously evaluated, create
the possibility of a new or different kind of accident from any
accident previously evaluated, or involve a significant reduction in a
margin of safety. Therefore, it is concluded that the proposed changes
meets the requirements of 10 CFR 50.92(c) and do not involve a
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 Section Chief: S. Singh Bajwa.
Southern California Edison Company, et al., Docket Nos. 50-361 and 50-
362, San Onofre Nuclear Generating Station, Units 2 and 3, San Diego
County, California
Date of amendment requests: September 10, 1998 (PCN-496), as
supplemented July 19, 1999.
Description of amendment requests: The proposed amendments would
modify the Technical Specifications for the San Onofre Nuclear
Generating Station (SONGS) Units 2 and 3 to delete the requirements for
equipment used to control hydrogen in the containment structure.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. Will operation of the facility in accordance with this
proposed change involve a significant increase in the probability or
consequences of an accident previously evaluated?
Response: No
The containment hydrogen control system is currently classified
as an engineered safety feature that serves as the combustible gas
control system in the containment. The hydrogen control system is
composed of a hydrogen recombiner subsystem and a hydrogen purge
subsystem. Hydrogen control subsystem components are not considered
to be accident initiators.
Therefore, this change does not increase the probability of an
accident previously evaluated.
The hydrogen control system is provided to ensure that the
hydrogen concentration is maintained below the flammability limit of
4% so that containment integrity is not challenged following a
design basis Loss Of Coolant Accident (LOCA). Existing analysis
show[s] that the hydrogen concentration will not reach the
flammability limit of 4% for at least 13.5 days after a design basis
LOCA. The time available will be extended to over 30 days using more
realistic hydrogen generation rates. The containment peak pressure
will remain below the San Onofre Nuclear Generating Station Units 2
and 3 (SONGS 2 & 3) containment design pressure of 60 psig [pounds
per square inch gauge] during this time. Beyond 30 days, hydrogen
concentration may reach the flammability limit. However, containment
failure due to hydrogen combustion is unlikely based on the results
of the SONGS 2 & 3 IPE [indvidual plant examination] study. The
detailed
[[Page 43779]]
SONGS 2 & 3-specific containment integrity analysis indicates that
containment rupture pressure is approximately 139 psig with 95%
confidence. Therefore, this change does not increase the
consequences of accidents previously evaluated.
Removal of the existing requirements for hydrogen control will
eliminate the Emergency Operating Instruction (EOI) steps for
hydrogen control and hence simplify the EOls. This would have a
positive impact on public health risk by reducing the probability of
operator error during potential accidents and hence reduce the core
damage frequency. As proposed in this change request, these changes
will allow the operators to address all hydrogen control issues as
part of the proposed Accident Management Guidelines which cover
operator actions at long time frames following accidents.
Removal of the existing requirements for hydrogen control will
eliminate the EOI steps to initiate the containment hydrogen purge.
This will result in a lower probability of a failed open containment
purge valve. Consequently, the offsite doses would be reduced due to
the reduction of the probability of a failed-open containment purge
valve. The changes described in this request result in a ``risk
positive'' change.
Therefore, this change does not involve a significant increase
in the probability or consequences of any accident previously
evaluated.
2. Will operation of the facility in accordance with this
proposed change create the possibility of a new or different kind of
accident from any accident previously evaluated?
Response: No
This proposed change does not change the design or configuration
of the plant beyond the hydrogen control system. Hydrogen generation
following a design basis LOCA has been evaluated in accordance with
regulatory requirements. Deletion of the hydrogen control system
from the Technical Specifications does not alter the hydrogen
generation processes post-LOCA. The consideration of hydrogen
generation will no longer be included in the design basis of SONGS 2
& 3. Therefore, this change does not create the possibility of a new
or different kind of accident from any previously evaluated.
3. Will operation of the facility in accordance with this
proposed change involve a significant reduction in a margin of
safety?
Response: No
The changes described in this change request result in a ``risk
positive'' change. Removal of the existing requirement for a
hydrogen control system will, by eliminating the EOI steps for
hydrogen control, result in lower operator error probabilities.
Elimination of the EOI steps to initiate the containment hydrogen
purge will result in a lower probability of a failed-open
containment purge valve, resulting in lower large early release
probabilities.
Therefore, this change involves an increase in safety, not a
reduction in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment requests involve no significant hazards consideration.
Local Public Document Room location: Main Library, University of
California, Irvine, California 92713.
Attorney for licensee: Douglas K. Porter, Esquire, Southern
California Edison Company, 2244 Walnut Grove Avenue, Rosemead,
California 91770.
NRC Section Chief: Stephen Dembek.
Southern Nuclear Operating Company, Inc., et al., Docket Nos. 50-424
and 50-425, Vogtle Electric Generating Plant, Units 1 and 2, Burke
County, Georgia
Date of amendment request: April 13, 1999.
Description of amendment request: Southern Nuclear Operating
Company (SNC) proposes to revise the Vogtle Electric Generating Plant
(VEGP) Unit 1 and Unit 2 Technical Specifications (TS) Limiting
Condition for Operation (LCO) Applicability LCO 3.0.4 and Surveillance
Requirement (SR) Applicability SR 3.0.4. The proposed changes would
update the versions of LCO 3.0.4 and SR 3.0.4 that appear in the
existing VEGP TS to be consistent with the versions of LCO 3.0.4 and SR
3.0.4 as they appear in Revision 1 to NUREG-1431. The proposed change
would add the words ``or that are part of a shutdown of the unit,'' to
LCO 3.0.4 to allow reactor shutdowns that are not necessarily required
by other TS Required Actions.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. Do the proposed changes involve a significant increase in the
probability or consequences of an accident previously evaluated?
No. The proposed change has impact on what equipment is required
to be OPERABLE or demonstrated OPERABLE via surveillance prior to
unit shutdowns or entry into MODES 5 and 6. This change could
increase the probability or consequences of an accident previously
evaluated if applied without consideration to all applicable
transitions. However, as part of the change, an evaluation is
attached in the form of a matrix that identifies those
specifications to which LCO 3.0.4 and SR 3.0.4 must continue to
apply. Therefore, only those specifications that do not impact
safety for these plant conditions are afforded this relaxation. As
such, there is no increase in the probability or consequences of an
accident previously evaluated as this assessment has been performed
and documented with the submittal.
2. Do the proposed changes create the possibility of a new or
different kind of accident from any accident previously evaluated?
No. The proposed change administratively changes when equipment
is required to be OPERABLE or demonstrated OPERABLE via surveillance
prior to unit shutdown or entry into MODES 5 and 6. However, as no
changes in equipment function or operation are included, there is no
increase in the probability of a new or different kind of accident
from those previously evaluated.
3. Do the proposed changes involve a significant reduction in a
margin of safety?
No. The proposed change has impact on what equipment is required
to be OPERABLE or demonstrated OPERABLE via surveillance prior to
unit shutdown or entry into MODES 5 and 6. This change could impact
the margin of safety of some accidents if applied without
consideration to all applicable transitions. However, as part of the
change, an evaluation is attached in the form of a matrix, that
identifies those specifications to which LCO 3.0.4 and SR 3.0.4 must
continue to apply. Therefore, only those specifications that do not
impact safety for these plant conditions, which includes any impact
on margin of safety are afforded this relaxation. As such, there is
no reduction in the margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: Burke County Public Library,
412 Fourth Street, Waynesboro, Georgia.
Attorney for licensee: Mr. Arthur H. Domby, Troutman Sanders,
NationsBank Plaza, Suite 5200, 600 Peachtree Street, NE., Atlanta,
Georgia.
NRC Section Chief: Richard L. Emch, Jr.
Southern Nuclear Operating Company, Inc., et al., Docket Nos. 50-424
and 50-425, Vogtle Electric Generating Plant, Units 1 and 2, Burke
County, Georgia
Date of amendment request: April 28, 1999.
Description of amendment request: The amendments revise Vogtle's
licensing basis to allow the licensee to establish containment hydrogen
monitoring within 90 minutes of initiation of a safety injection
following a loss-of-coolant accident, compared to the current 30
minutes requirement.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
[[Page 43780]]
consideration, which is presented below:
(1) Operation of the facility in accordance with the proposed
amendment would not involve a significant increase in the
probability or consequences of an accident previously evaluated.
Containment hydrogen concentration is not an input parameter to
the FSAR Chapter 15 accident analyses for a loss of reactor or
secondary coolant accidents; nor is it used as an initial assumption
for the containment response analysis. Control room operators use
the containment hydrogen monitors to establish hydrogen control
measures should it become necessary. However, the actions required
to establish containment hydrogen monitoring are a distraction for
the operators from more important tasks during the early phases of
an accident. Hydrogen production occurs over a long period and a
significant accumulation is not expected for several hours into the
event. This function is more appropriately included as a part of the
long-term core damage assessment process. The one-hour extension
will have a positive impact on the ability of the operators to
concentrate on their more immediate actions while having no negative
impact on the long-term assessment efforts. Therefore, the proposed
license amendment will not involve a significant increase in the
probability or consequences of an accident previously evaluated.
(2) Operation of the facility in accordance with the proposed
amendment would not create the possibility of a new or different
kind of accident from any previously evaluated.
Operation of the containment hydrogen monitors is not an
initiator of any design basis accident. Control room operators use
the containment hydrogen monitors following a LOCA to establish
hydrogen control measures should it become necessary. Accurate
indication of containment hydrogen concentration is needed prior to
initiating recombiner operation or containment venting and for long-
term core damage assessment. The proposed license amendment would
not eliminate the requirement to establish hydrogen monitoring, but
would permit it to be delayed until those actions required to
diagnose the event and verify proper operation of essential safety
equipment have been completed. The one-hour extension maintains the
requirement to establish hydrogen monitoring well before calculated
conditions inside the containment indicate any need to initiate
hydrogen control measures. Therefore, the proposed license amendment
will not create a new or different kind of accident from any
previously evaluated.
(3) Operation of the facility in accordance with the proposed
amendment would not involve a significant reduction in a margin of
safety.
The need to establish hydrogen control measures will not be
present within the first 90 minutes following a LOCA since there
will not be significant hydrogen accumulation. By extending the time
allowed to establish containment hydrogen monitoring, the operators
can remain focused on the actions necessary to assess and mitigate
the accident before redirecting their attention to long-term
recovery actions. The one-hour extension maintains the requirement
to establish hydrogen monitoring well before calculated conditions
inside the containment indicate any need to initiate hydrogen
control measures. Therefore, the proposed license amendment will not
involve a significant reduction in a margin of safety, but will
instead result in an overall enhancement to safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: Burke County Public Library,
412 Fourth Street, Waynesboro, Georgia.
Attorney for licensee: Mr. Arthur H. Domby, Troutman Sanders,
NationsBank Plaza, Suite 5200, 600 Peachtree Street, NE., Atlanta,
Georgia.
NRC Section Chief: Richard L. Emch, Jr.
Southern Nuclear Operating Company, Inc., et al., Docket Nos. 50-424
and 50-425, Vogtle Electric Generating Plant, Units 1 and 2, Burke
County, Georgia
Date of amendment request: May 18, 1999.
Description of amendment request: The proposed change would revise
Surveillance Requirements (SRs) 3.8.1.3 and 3.8.1.13 to reduce the
loading requirements for the diesel generators (DGs). Presently, SR
3.8.1.3 requires that the DGs be loaded and operated for greater than
or equal to 60 minutes between 6800 kW and 7000 kW at least once every
31 days. The proposed change would revise the lower end of the load
band in SR 3.8.1.3 to 6500 kW from 6800 kW. Revised SR 3.8.1.3 would
require that the DGs be loaded and operated for greater than or equal
to 60 minutes at a load greater than or equal to 6500 kW and less than
or equal to 7000 kW at least once every 31 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:
1. Does the proposed change involve a significant increase in
the probability or consequences of an accident previously evaluated?
No. The proposed change affects only the DG loading requirements
(kW and kVAR) specified in SRs 3.8.1.3 and 3.8.1.13. These loading
requirements have no impact on or relationship to the probability of
any of the initiating events assumed for the accidents previously
evaluated. Therefore, the proposed change does not involve a
significant increase in the probability of any accident previously
evaluated. Furthermore, since the proposed loading requirements
bound the maximum expected loading for the DGs, SRs 3.8.1.3 and
3.8.1.13 will continue to demonstrate that the DGs are capable of
performing their safety function. Since the proposed change does not
adversely affect the capability of the DGs to perform their safety
function, the outcomes of the accidents previously evaluated (i.e.,
radiological consequences) will not be affected. Therefore, the
proposed change does not involve a significant increase in the
consequences of any accident previously evaluated.
2. Does the proposed change create the possibility of a new or
different kind of accident from any previously evaluated?
No. The proposed change affects only the DG loading requirements
(kW and kVAR) specified in SRs 3.8.1.3 and 3.8.1.13. The proposed
change will not introduce any new equipment or create new failure
modes for existing equipment. Other than the reduced loading
requirements for the DGs, the proposed change will not affect or
otherwise alter plant operation. The DGs will remain capable of
performing their safety function. No other safety related or
important to safety equipment will be affected by the proposed
change. Therefore, the proposed change will not create the
possibility of a new or different kind of accident from any
previously evaluated.
3. Does the proposed change involve a significant reduction in a
margin of safety?
No. The proposed change reduces the loading requirements of SRs
3.8.1.3 and 3.8.1.13. With one exception, the new loading
requirements are consistent with the latest regulatory guidance
found in Regulatory Guide (RG) 1.9, Revision 3, ``Selection, Design,
and Qualification of Diesel-Generator Units Used as Standby (Onsite)
Electric Power Systems at Nuclear Power Plants,'' July 1993. The one
exception to RG 1.9, the loading requirements for the 2-hour portion
of the endurance and margin test (SR 3.8.1.13), will require testing
at loads in excess of 105 percent of the maximum expected load as
opposed to 105 percent of the continuous duty rating. Testing for at
least 2 hours at 105 percent of the maximum expected load will
continue to demonstrate adequate margin, and it will reduce wear and
tear on the DGs due to testing. Reduction in wear and tear should
inherently increase the reliability of the DGs. Therefore, the
proposed change does not involve a significant reduction in a margin
of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: Burke County Public Library,
412 Fourth Street, Waynesboro, Georgia.
[[Page 43781]]
Attorney for licensee: Mr. Arthur H. Domby, Troutman Sanders,
NationsBank Plaza, Suite 5200, 600 Peachtree Street, NE., Atlanta,
Georgia.
NRC Section Chief: Richard L. Emch, Jr.
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: June 24, 1999 (TS 99-05)
Brief description of amendments: The proposed amendments would
change the Sequoyah Units 1 and 2 Technical Specification (TS)
requirements by clarifying and changing the surveillance requirements
for the ice weight in the ice condenser baskets. This request is a
lead-plant change for all Westinghouse-designed ice condenser plants
and will be incorporated into the Improved Standard Technical
Specifications (ISTSs), if approved.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), Tennessee Valley
Authority, the licensee, has provided its analysis of the issue of no
significant hazards consideration, which is presented below:
A. The proposed amendment does not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
The proposed TS amendments discussed below cannot increase the
probability of occurrence of any analyzed accident because they are
not the result or cause of any physical modification to the ice
condenser structures, and for the current design of the ice
condenser, there is no correlation between any credible failure and
the initiation of any previously analyzed accident.
Regarding the consequences of analyzed accidents, the proposed
amendment provides for consistency with the ISTSs by: (1) requiring
the actions if one or more ice condenser ice baskets are determined
to weigh below the minimum specified value to be made a part of the
TS surveillance requirement (SR) instead of being located in the
bases, and (2) relocating the ice basket selection methodology into
the bases. This ensures consistent interpretation of the
requirements of the TS in accordance with the ISTSs. The
clarification of the response required if one or more ice baskets in
a given bay are determined to be underweight ensures sufficient ice
is maintained in each bay to prevent early meltout in a local zone
following a design basis accident (DBA) and that the required
overall ice weight is maintained in the ice condenser. The
relocation of the ice basket selection methodology to the bases does
not result in any change to the intent or implementation of this
portion of the TSs since plant procedures ensure the requirements of
the bases of the TSs are correctly implemented. Additionally, the
clarification that the weight requirement is applicable to the
beginning of the cycle does not change the present intent of the TS,
but ensures there is no confusion, since the weight at the end of
the operating cycle may be less than that specified in the SR due to
sublimation. This does not result in a change to the intent or
implementation of the TS since a sublimation allowance was provided
in the original SR weight requirement. These clarifications do not
result in any [effect] on plant equipment or operation and the
actions taken during the implementation of the revised TS will be
the same as prior to the revision. Therefore, the clarification of
these requirements will not increase the consequences of any
accident previously evaluated.
B. The proposed amendment does not create the possibility of a
new or different kind of accident from any accident previously
evaluated.
The inclusion of the action required for an underweight ice
basket in the TS SR, instead of in the bases of the TS, provides for
the consistent interpretation of the requirement. The clarification
of the response required if one or more ice baskets in a given bay
are determined to be underweight ensures sufficient ice is
maintained in each bay to prevent early meltout in a local zone
following a DBA and that the required overall ice weight is
maintained in the ice condenser. The relocation of the ice basket
selection methodology to the bases does not result in any change to
the intent or implementation of this portion of the TSs since plant
procedures ensure the requirements of the bases of the TSs are
correctly implemented. Additionally, the clarification that the
weight requirement is applicable to the beginning of the cycle does
not change the present intent of the TS, but ensures there is no
confusion, since the weight at the end of the operating cycle may be
less than that specified in the SR due to sublimation. This does not
result in a change to the intent or implementation of the TS since a
sublimation allowance was provided in the original SR weight
requirement. The operation, design and maintenance of the ice
condenser and its associated equipment will not change as a result
of these clarifications. Therefore, the implementation of these
clarifications will not create the possibility of accidents or
equipment malfunctions of a new or different kind from any
previously evaluated.
C. The proposed amendment does not involve a significant
reduction in a margin of safety.
The proposed amendment allows for the consistent interpretation
of the required actions if an ice basket is determined to weigh less
than the required minimum. The inclusion of these actions in the TS
SR instead of in the TS bases assures the correct actions will be
taken as intended by the TSs. The clarification of the response
required if one or more ice baskets in a given bay are determined to
be underweight ensures sufficient ice is maintained in each bay to
prevent early meltout in a local zone following a DBA and that the
required overall ice weight is maintained in the ice condenser. The
relocation of the ice basket selection methodology to the bases does
not result in any change to the intent or implementation of this
portion of the TSs since plant procedures ensure the requirements of
the bases of the TSs are correctly implemented. Additionally, the
clarification that the weight requirement is applicable to the
beginning of the cycle does not change the present intent of the TS,
but ensures there is no confusion, since the weight at the end of
the operating cycle may be less than that specified in the SR due to
sublimation. This does not result in a change to the intent or
implementation of the TS since a sublimation allowance was provided
in the original SR weight requirement. The proposed clarifications
do not result in or have any [effect] on the operation, design, or
maintenance of any plant equipment. Thus the design limits for the
continued safe function of the containment structure following a DBA
are not exceeded due to this change; therefore, the proposed
amendment does not involve a reduction in a margin of safety.
The NRC 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
amendments request involves no significant hazards consideration.
Local Public Document Room location: Chattanooga-Hamilton County
Library, 1001 Broad Street, Chattanooga, Tennessee 37402.
Attorney for licensee: General Counsel, Tennessee Valley Authority,
400 West Summit Hill Drive, ET 10H, Knoxville, Tennessee 37902.
NRC Section Chief: Sheri R. Peterson.
Tennessee Valley Authority, Docket No. 50-390 Watts Bar Nuclear Plant,
Unit 1, Rhea County, Tennessee
Date of amendment request: June 25, 1999 (TS 99-004).
Description of amendment request: The proposed amendment would
revise the Watts Bar Nuclear Plant Unit 1 Technical Specifications (TS)
and associated TS Bases for Limiting Condition for Operation (LCO)
3.7.1, Main Steam Safety Valves, to provide a new requirement to reduce
the Power Range Neutron Flux-High reactor trip setpoints when two or
more main steam safety valves (MSSVs) per steam generator are
inoperable. This proposal is based on a generic change developed by the
Westinghouse Owners Group (WOG), TSTF-235, Revision 1, which has been
approved by the NRC staff.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
[[Page 43782]]
A. The proposed amendment does not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
The proposed change to TS LCO 3.7.1 requires a reduction of the
Power Range Neutron Flux-High reactor trip setpoints to a
corresponding power level depending on the number of inoperable
MSSVs. The change is based on and consistent with an industry
sponsored change (TSTF-235, Revision 1) which has been reviewed and
accepted by the NRC staff.
Although plant procedures currently require resetting the high
flux trip, it is not a TS requirement. The proposed amendment will
provide a more appropriate barrier to prevent the plant from being
operated under a non-conservative technical specification action
statement in a region where multiple inoperable MSSVs coincident
with a reactivity insertion event such as an inadvertent rod cluster
control assembly (RCCA) bank withdrawal could result in
overpressurization of the secondary system.
No change is made in the probability of initiating accident,
i.e., RCCA bank withdrawal, and by requiring the reactor trip
setpoint reduction, a potential mismatch between core power and
turbine load without sufficient steam relief capacity is eliminated.
Therefore, the change requested by this amendment actually decreases
the consequences of an accident previously evaluated (without credit
for procedure actions to reduce the trip setpoints).
B. The proposed amendment does not create the possibility of a
new or different kind of accident from any accident previously
evaluated.
Without crediting existing plant procedures, the addition of the
proposed TS change prevents the plant from being operated in a
region where an overpressurization of the main steam system is
postulated to potentially occur. The proposed change assures that
the existing FSAR [Final Safety Analysis Evaluation Report] accident
analysis remains bounding for events that challenge the relieving
capacity of the MSSVs. Since the addition of the TS action adds a
more appropriate administrative barrier to prevent operation in an
undesired region and because the change is bounded by the current
accident analysis described in the FSAR, a new or different kind of
accident has not been created as a result of this license amendment.
C. The proposed amendment does not involve a significant
reduction in a margin of safety.
The proposed TS change eliminates a non-conservative TS action
to prevent the plant from being operated in a region where an
overpressurization of the main steam system is postulated to
potentially occur. Since the addition of the TS action adds a more
effective administrative barrier to prevent operation in an
undesired region and because the change is bounded by the existing
FSAR accident analysis, the margin of safety has actually increased
for the proposed change. For these reasons, the proposed amendment
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: Chattanooga-Hamilton County
Library, 1001 Broad Street, Chattanooga, TN 37402.
Attorney for licensee: General Counsel, Tennessee Valley Authority,
400 West Summit Hill Drive, ET l0H, Knoxville, Tennessee 37902.
NRC Section Chief: Sheri R. Peterson.
Wolf Creek Nuclear Operating Corporation, Docket No. 50-482, Wolf Creek
Generating Station, Coffey County, Kansas
Date of amendment request: July 8, 1999.
Description of amendment request: The amendment request proposes to
increase the allowable values for the engineered safety features
actuation system (ESFAS) loss-of-power 4 kV undervoltage trips in the
current Technical Specifications (TSs) Table 3.3-4 (functional units
8.a and 8.b) and in Surveillance Requirement (SR) 3.3.5.3 of the
improved TSs. The word ``nominal'' is also being added to describe the
trip setpoint in SR 3.3.5.3 and in the Bases of the improved TSs. The
improved TSs were issued in Amendment 123 dated March 31, 1999, but
have not yet been implemented.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration. The staff has reviewed the licensee's analysis against
the standards of 10 CFR 50.92(c). The NRC staff's review is presented
below.
1. The proposed change does not involve a significant increase in
the probability or consequences of an accident previously evaluated.
The reactor protection system performance will remain within the
bounds of the previously performed accident analysis. The protection
systems will continue to function in a manner consistent with the plant
design basis. The proposed changes will not affect any of the analysis
assumptions for any of the accidents previously evaluated. The proposed
changes will not affect the probability of any event initiators nor
will the proposed changes affect the ability of any safety related
equipment to perform its intended function. There is no change to the
technical specification trip setpoints; therefore, there is no
degradation in the performance of nor an increase in the number of
challenges imposed on safety related equipment assumed to function
during an accident situation and be no change to normal plant operating
parameters or accident mitigation capabilities. The allowable values
and the trip setpoints in the protection system proposed to be changed
are not initiators of accidents previously evaluated.
Based on the above evaluation, these proposed changes do not
involve a significant increase in the probability or consequences of an
accident previously evaluated.
2. The proposed change does not create the possibility of a new or
different kind of accident from any accident previously evaluated.
There are no changes in the method by which any safety related
plant system performs its safety function. The normal manner of plant
operation remains unchanged because the methodology to determine the
allowable value and the trip setpoints remains unchanged. The increase
in allowable value for the trip setpoints still provides margin between
the nominal trip setpoint and allowable value while taking into account
worst case 4.16 kV Class 1E system (NB) bus voltages that could be
possible during steady state loss-of-coolant accident (LOCA)
conditions. The change in allowable value for the undervoltage
protection functions does not impact the systems capability to:
a. Trip the 4.16 kV preferred normal and alternate bus feeder
breakers to remove the deficient power source to protect the Class 1E
equipment from damage;
b. Shed all loads from the bus except the Class 1E 480 Vac load
centers and centrifugal charging pumps to prepare the buses for re-
energization by the load shedder and emergency load sequencer (LSELS);
and
c. Generate a emergency diesel generator (EDG) start signal.
No new accident scenarios, transient precursors, failure
mechanisms, or limiting single failures are introduced as a result of
the proposed changes. The allowable values and the trip setpoints in
the protection system proposed to be changed are not initiators of
accidents. Therefore, the proposed change does not create the
possibility of a new or different kind of accident from any previously
evaluated.
3. The proposed change does not involve a significant reduction in
a margin of safety.
The undervoltage protection functions are to:
[[Page 43783]]
a. Trip the 4.16 kV preferred normal and alternate bus feeder
breakers to remove the deficient power source to protect the Class 1E
equipment from damage;
b. Shed all loads from the bus except the Class 1E 480 Vac load
centers and centrifugal charging pumps to prepare the buses for re-
energization by the load shedder and emergency load sequencer (LSELS);
and
c. Generate a EDG start signal.
The proposed changes do not affect the acceptance criteria for any
analyzed event nor is there a change in the safety analysis limit.
There will be no effect on the manner in which safety limits or
engineered safety features actuation system settings are determined nor
will there be any affect on those plant systems necessary to assure the
accomplishment of the above protection functions. Therefore, there will
not be a significant reduction in a margin of safety.
Based on this review, it appears that the three standards of 10 CFR
50.92(c) are satisfied. Therefore, the NRC staff proposes to determine
that the amendment request involves no significant hazards
consideration.
Local Public Document Room locations: Emporia State University,
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 Section Chief: Stephen Dembek.
Previously Published Notice of Consideration of Issuance of
Amendment to Facility Operating Licenses, Proposed No Significant
Hazards Consideration Determination, And Opportunity for a Hearing
Wolf Creek Nuclear Operating Corporation, Docket No. 50-482, Wolf Creek
Generating Station, Coffey County, Kansas
Date of amendment request: June 30, 1999.
Brief description of amendment: The proposed amendment would revise
Technical Specification (TS) 3/4.7.5 of the current TSs by adding a
temporary action statement that would allow the plant to operate for up
to 12 hours with an inlet temperature up to but less than 95 deg.F.
Date of individual notice in Federal Register: July 15, 1999 (64 FR
38221).
Expiration date of individual notice: August 16, 1999.
Local Public Document Room locations: Emporia State University,
William Allen White Library, 1200 Commercial Street, Emporia, Kansas
66801 and Washburn University School of Law Library, Topeka, Kansas
66621.
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.
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: February 26, 1999
Brief description of amendment: This amendment changes the Table
Notations for Technical Specification (TS) Table 3.3-4, ``Engineered
Safety Features Actuation System Instrumentation Trip Setpoints.''
Specifically, the time constants used in the lead-lag controller for
Steam Line Pressure--Low (Table item 1.e) and in the rate-lag
controller for Negative Steam Line Pressure Rate--High (Table item 4.e)
have been revised.
Date of issuance: July 28, 1999.
Effective date: July 28, 1999.
Amendment No.: 89.
Facility Operating License No. NPF-63. Amendment revises the
Technical Specifications.
Date of initial notice in Federal Register: March 24, 1999 (64 FR
14280).
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated July 28, 1999.
No significant hazards consideration comments received: No.
Local Public Document Room location: Cameron Village Regional
Library, 1930 Clark Avenue, Raleigh, North Carolina 27605.
Duke Energy Corporation, 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: October 2, 1998, as supplemented
November 20, 1998, December 21, 1998, and May 13, 1999.
Brief description of amendments: The amendments revised the Updated
Final Safety Analysis Report related to an unreviewed safety question
regarding the use of a small amount of containment overpressure to
ensure sufficient net positive suction head for the reactor building
spray and low pressure injection pumps during the post loss of coolant
accident recirculation phase.
Date of Issuance: July 19, 1999.
Effective date: As of the date of issuance and shall be implemented
within 30 days from the date of issuance.
Amendment Nos.: Unit 1--305; Unit 2--305; Unit 3--305.
Facility Operating License Nos. DPR-38, DPR-47, and DPR-55:
Amendments revised the Updated Final Safety Analysis Report.
Date of initial notice in Federal Register: June 16, 1999 (64 FR
32288).
The November 20, 1998, December 21, 1998, and May 13, 1999, letters
provided clarifying information that did not change the scope of the
October 2, 1998, application and the initial proposed no significant
hazards consideration determination.
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated July 19, 1999.
No significant hazards consideration comments received: No.
Local Public Document Room location: Oconee County Library, 501
[[Page 43784]]
West South Broad Street, Walhalla, South Carolina.
Florida Power Corporation, et al., Docket No. 50-302, Crystal River
Nuclear Generating Plant, Unit 3, Citrus County, Florida
Date of application for amendment: May 28, 1998.
Brief description of amendment: Changes the Crystal River Unit 3
(CR-3) licensing bases to incorporate Generic Letter 87-11,
``Relaxation in Arbitrary Intermediate Pipe Rupture Requirements,'' and
NUREG/CR-2913, ``Two-Phase Jet Loads,'' as part of the licensing basis
for CR-3.
Date of issuance: July 27, 1999.
Effective date: As of the date of issuance, to be incorporated into
the Final Safety Analysis Report at the time of its next update.
Amendment No.: 181.
Facility Operating License No. DPR-72: Amendment approves changes
to the Final Safety Analysis Report.
Date of initial notice in Federal Register: July 15, 1998 (63 FR
38200).
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated July 27, 1999.
No significant hazards consideration comments received: No.
Local Public Document Room location: Coastal Region Library, 8619
W. Crystal Street, Crystal River, Florida 34428.
Nebraska Public Power District, Docket No. 50-298, Cooper Nuclear
Station, Nemaha County, Nebraska
Date of amendment request: March 1, 1999, as supplemented by
letters dated March 10, 1999, June 8, 1999, and June 23, 1999.
Brief description of amendment: The amendment changes the Cooper
Nuclear Station Technical Specifications to revise the calibration
frequency of the reactor recirculation flow transmitters from once
every 184 days to once every 18 months.
Date of issuance: July 26, 1999.
Effective date: July 26, 1999, to be implemented within 30 days.
Amendment No.: 179.
Facility Operating License No. DPR-46: Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: April 7, 1999 (64 FR
17027) The March 10, June 8, and June 23, 1999, letters provided
additional clarifying information and updated TS pages. This
information was within the scope of the original Federal Register
notice and did not change the staff's initial proposed no significant
hazards consideration determination.
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated July 26, 1999.
No significant hazards consideration comments received: No.
Local Public Document Room location: Auburn Memorial Library, 1810
Courthouse Avenue, Auburn, NE 68305.
Omaha Public Power District, Docket No. 50-285, Fort Calhoun Station,
Unit No. 1, Washington County, Nebraska
Date of amendment request: September 28, 1998, as supplemented by
letter dated March 12, 1999.
Brief description of amendment: The amendment authorizes the
revision to the licensing basis as described in the Updated Safety
Analysis Report (USAR) to incorporate the modification for overriding
the containment isolation actuation signal to the reactor coolant
system letdown flow containment isolation valves.
Date of issuance: July 22, 1999.
Effective date: July 22, 1999, and shall be implemented in the next
periodic update to the USAR in accordance with 10 CFR 50.71(e).
Amendment No.: 191.
Facility Operating License No. DPR-40. The amendment revised the
Updated Safety Analysis Report.
Date of initial notice in Federal Register: November 18, 1998 (63
FR 64119) The March 12, 1999, supplemental letter provided additional
clarifying information, did not expand the scope of the application as
originally noticed, and did not change the staff's initial no
significant hazards consideration determination.
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated July 22, 1999.
No significant hazards consideration comments received: No.
Local Public Document Room location: W. Dale Clark Library, 215
South 15th Street, Omaha, Nebraska 68102.
Omaha Public Power District, Docket No. 50-285, Fort Calhoun Station,
Unit No. 1, Washington County, Nebraska
Date of amendment request: January 29, 1999.
Brief description of amendment: The amendment revises Technical
Specifications 2.10.2(1) and 2.10.2(3) and deletes Figure 2-11 to
relocate three cycle specific parameters to the Core Operating Limits
Report.
Date of issuance: July 27, 1999.
Effective date: July 27, 1999.
Amendment No.: 192.
Facility Operating License No. DPR-40. Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: February 24, 1999 (64
FR 9193) The Commission's related evaluation of the amendment is
contained in a Safety Evaluation dated July 27, 1999.
No significant hazards consideration comments received: No.
Local Public Document Room location: W. Dale Clark Library, 215
South 15th Street, Omaha, Nebraska 68102.
PECO Energy Company, Docket Nos. 50-352 and 50-353, Limerick Generating
Station, Units 1 and 2, Montgomery County, Pennsylvania
Date of application for amendments: January 25, 1999.
Brief description of amendments: Revise technical specifications
surveillance requirement frequencies for the emergency diesel generator
maintenance inspection outages, the 24-hour endurance run and the hot
restart test from 18 to 24 months.
Date of issuance: July 29, 1999.
Effective date: Both units, as of date of issuance and shall be
implemented within 30 days.
Amendment Nos.: 136 and 101.
Facility Operating License Nos. NPF-39 and NPF-85. The amendments
revised the Technical Specifications.
Date of initial notice in Federal Register: February 24, 1999 (64
FR 9196).
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated July 29, 1999.
No significant hazards consideration comments received: No.
Local Public Document Room location: Pottstown Public Library, 500
High Street, Pottstown, PA 19464.
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 2, 1999, as
supplemented on April 26, 1999.
Brief description of amendments: The amendments revise Technical
Specification (TS) 5.6, ``Fuel Storage, Criticality,'' to change the
maximum unirradiated fuel assembly enrichment value for new fuel
storage from 4.5 to 5.0 weight percent Uranium-235 and to allow the use
of equivalent criticality control to that provided by the current TS
requirement of 2.35 milligrams of Boron-10 per linear inch loading in
the Integral Fuel Burnable Absorber pins.
Date of issuance: July 21, 1999.
Effective date: As of the date of issuance and shall be implemented
within 60 days.
[[Page 43785]]
Amendment Nos.: 223 and 204.
Facility Operating License Nos. DPR-70 and DPR-75. The amendments
revised the Technical Specifications.
Date of initial notice in Federal Register: March 10, 1999 (64 FR
11965).
The April 26, 1999, letter provided clarifying information that did
not change the initial proposed no significant hazards consideration
determination.
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated July 21, 1999.
No significant hazards consideration comments received: No.
Local Public Document Room location: Salem Free Public Library, 112
West Broadway, Salem, NJ 08079.
Dated at Rockville, Maryland, this 4th day of August 1999.
For the Nuclear Regulatory Commission.
Suzanne C. Black,
Deputy Director, Division of Licensing Project Management, Office of
Nuclear Reactor Regulation.
[FR Doc. 99-20545 Filed 8-10-99; 8:45 am]
BILLING CODE 7590-01-P