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Biweekly Notice; Applications and Amendments to Facility Operating Licenses Involving No Significant Hazards Considerations
Note: EPA no longer updates this information, but it may be useful as a reference or resource.
[Federal Register: November 4, 1998 (Volume 63, Number 213)]
[Notices]
[Page 59584-59604]
From the Federal Register Online via GPO Access [wais.access.gpo.gov]
[DOCID:fr04no98-102]
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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 fromOctober 9, 1998, through October 23, 1998.
The last biweekly notice was published on October 21, 1998 (63 FR
56238).
[[Page 59585]]
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
20555-0001, 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 December 4, 1998, 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
[[Page 59586]]
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.
Arizona Public Service Company, et al., Docket Nos. STN 50-528 and STN
50-529, Palo Verde Nuclear Generating Station, Units Nos. 1 and 2,
Maricopa County, Arizona
Date of application for amendment: October 6, 1998.
Description of amendment request: The proposed amendment would
clarify the power level threshold at which certain reactor protective
system (RPS) instrumentation trips must be enabled and may be bypassed,
and clarify that this level is a percentage of the neutron flux at
rated thermal power (RTP). The bypass power level, 1E-4% RTP, would be
specified as logarithmic power instead of thermal power. The intent of
(and the implementation of) the 1E-4% RTP RPS instrumentation bypass
threshold level in the technical specifications (TS) has always been
that this power level is neutron power, which would be indicated by
logarithmic power, and is not the heat transfer from the reactor core
to the coolant, including decay heat, which is the thermal power
definition in the TS.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. The proposed change does not involve a significant increase
in the probability or consequences of an accident previously
evaluated.
The proposed change would replace the words ``THERMAL POWER''
with ``logarithmic power'' for the 1E-4% rated thermal power (RTP)
level threshold in Table 3.3.1-1 footnotes (a) and (b), surveillance
requirement SR 3.3.1.7 Note 2, and Table 3.3.2-1 footnote (d) for
the reactor protective system (RPS) instrumentation. The purpose of
the 1E-4% RTP threshold is to (1) specify the power, below which,
the logarithmic power level trip is required to be operable and
surveilled, and (2) specify the power, above which, the local power
density (LPD) and departure from nucleate boiling ratio (DNBR) trips
are required to be operable. For these purposes, the appropriate
power threshold should be logarithmic power, which is the power
indicated on the logarithmic nuclear instrumentation, and not
thermal power. Thermal power is defined in TS section 1.1 as the
total reactor heat transfer rate to the reactor coolant, and would
include decay heat. Thermal power would therefore not drop to 1E-4%
RTP for a considerable period of time after shutdown, and would not
provide the plant protective function correlation required at 1E-4%
neutron RTP. However, logarithmic power, which is indicated by
neutron flux, does provide the plant protective function correlation
required at 1E-4% neutron RTP for the required reactor trips as
required by safety analyses. The logarithmic power level of 1E-4%
neutron RTP nominally correlates to the neutron flux measured by the
excore neutron instrumentation that is 1E-4% of the neutron flux at
100% RTP (3876 MWt) measured by the excore neutron instrumentation.
The proposed editorial amendment would also replace ``RTP'' with
``NRTP,'' in Table 3.3.1-1 footnotes (a) and (b), surveillance
requirement SR 3.3.1.7 Note 2, and Table 3.3.2-1 footnotes (c) and
(d). A definition would be added for NRTP (nuclear rated thermal
power) in section 1.1 as the indicated neutron flux at RTP. These
editorial clarifications will reflect the fact that the logarithmic
power level of 1E-4% is not a percentage of the ``total reactor core
heat transfer rate to the reactor coolant of 3876 MWt,'' as RTP is
defined in section TS 1.1, but is instead a percentage of the
indicated neutron flux at RTP.
An editorial change is also proposed to specify NRTP as the
``ALLOWABLE VALUE'' parameter for the high logarithmic power level
trip setpoint in Table 3.3.1-1 to correct the unintended omission of
the trip setpoint parameter during preparation of the Improved
Technical Specifications. This change will fill in the omitted
parameter with the correct parameter of NRTP that is also consistent
with the high logarithmic power trip setpoint parameter in Table
3.3.2-1.
These changes do not constitute a physical change to the Unit or
make changes in the RPS instrumentation setpoints, system logic or
manual actuation. In addition, these changes do not alter physical
plant equipment or the way in which plant equipment is operated.
This change is editorial in that it corrects the TS wording to match
the appropriate power parameter that was originally intended and
required by safety analyses, and that has been implemented since
original licensing of the PVNGS plants. Therefore, these 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.
The proposed change would replace the words ``THERMAL POWER''
with ``logarithmic power'' for the 1E-4% RTP level threshold in
Table 3.3.1-1 footnotes (a) and (b), surveillance requirement SR
3.3.1.7 Note 2, and Table 3.3.2-1 footnote (d) for the RPS
instrumentation. The purpose of the 1E-4% RTP threshold is to (1)
specify the power, below which, the logarithmic power level trip is
required to be operable and surveilled, and (2) specify the power,
above which, the LPD and DNBR trips are required to be operable. For
these purposes, the appropriate power threshold should be
logarithmic power, which is the power indicated on the logarithmic
nuclear instrumentation, and not thermal power. Thermal power is
defined in TS section 1.1 as the total reactor heat transfer rate to
the reactor coolant, and would include decay heat. Thermal power
would therefore not drop to 1E-4% RTP for a considerable period of
time after shutdown, and would not provide the plant protective
function correlation required at 1E-4% neutron RTP. However,
logarithmic power, which is indicated by neutron flux, does provide
the plant protective function correlation required at 1E-4% neutron
RTP for the required reactor trips as required by safety analyses.
The proposed editorial amendment would also replace ``RTP'' with
``NRTP,'' in Table 3.3.1-1 footnotes (a) and (b), surveillance
requirement SR 3.3.1.7 Note 2, and Table 3.3.2-1 footnotes (c) and
(d). A definition would be added for NRTP (nuclear rated thermal
power) in section 1.1 as the indicated neutron flux at RTP. These
editorial clarifications will reflect the fact that the logarithmic
power level of 1E-4% is not a percentage of the ``total reactor core
heat transfer rate to the reactor coolant of 3876 MWt,'' as RTP is
defined in section TS 1.1, but is instead a percentage of the
indicated neutron flux at RTP.
An editorial change is also proposed to specify NRTP as the
``ALLOWABLE VALUE'' parameter for the high logarithmic power level
trip setpoint in Table 3.3.1-1 to correct the unintended omission of
the trip setpoint parameter during preparation of the Improved
Technical Specifications. This change will fill in the omitted
parameter with the correct parameter of NRTP that is also consistent
with the high logarithmic power trip setpoint parameter in Table
3.3.2-1.
These changes do not constitute a physical change to the Unit or
make changes in the RPS instrumentation setpoints, system logic or
manual actuation. In addition, these changes do not alter physical
plant equipment or the way in which plant equipment is operated. The
proposed change does not introduce any new modes of plant operation
or new accident precursors. This change is editorial in that it
corrects the TS wording to match the appropriate power
[[Page 59587]]
parameter that was originally intended and required by safety
analyses, and that has been implemented since original licensing of
the PVNGS plants. Therefore, this change does not create the
possibility of a new or different kind of accident from any accident
previously evaluated.
3. The proposed change does not involve a significant reduction
in a margin of safety.
The proposed change would replace the words ``THERMAL POWER''
with ``logarithmic power'' for the 1E-4% RTP level threshold in
Table 3.3.1-1 footnotes (a) and (b), surveillance requirement SR
3.3.1.7 Note 2, and Table 3.3.2-1 footnote (d) for the RPS
instrumentation. The purpose of the 1E-4% RTP threshold is to (1)
specify the power, below which, the logarithmic power level trip is
required to be operable and surveilled, and (2) specify the power,
above which, the LPD and DNBR trips are required to be operable. For
these purposes, the appropriate power threshold should be
logarithmic power, which is the power indicated on the logarithmic
nuclear instrumentation, and not thermal power. Thermal power is
defined in TS section 1.1 as the total reactor heat transfer rate to
the reactor coolant, and would include decay heat. Thermal power
would therefore not drop to 1E-4% RTP for a considerable period of
time after shutdown, and would not provide the plant protective
function correlation required at 1E-4% neutron RTP. However,
logarithmic power, which is indicated by neutron flux, does provide
the plant protective function correlation required at 1E-4% neutron
RTP for the required reactor trips as required by safety analyses.
The proposed editorial amendment would also replace ``RTP'' with
``NRTP,'' in Table 3.3.1-1 footnotes (a) and (b), surveillance
requirement SR 3.3.1.7 Note 2, and Table 3.3.2-1 footnotes (c) and
(d). A definition would be added for NRTP (nuclear rated thermal
power) in section 1.1 as the indicated neutron flux at RTP. These
editorial clarifications will reflect the fact that the logarithmic
power level of 1E-4% is not a percentage of the ``total reactor core
heat transfer rate to the reactor coolant of 3876 MWt,'' as RTP is
defined in section TS 1.1, but is instead a percentage of the
indicated neutron flux at RTP.
An editorial change is also proposed to specify NRTP as the
``ALLOWABLE VALUE'' parameter for the high logarithmic power level
trip setpoint in Table 3.3.1-1 to correct the unintended omission of
the trip setpoint parameter during preparation of the Improved
Technical Specifications. This change will fill in the omitted
parameter with the correct parameter of NRTP that is also consistent
with the high logarithmic power trip setpoint parameter in Table
3.3.2-1.
These changes do not constitute a physical change to the Unit or
make changes in the RPS instrumentation setpoints, system logic or
manual actuation. In addition, these changes do not alter physical
plant equipment or the way in which plant equipment is operated.
This change is editorial in that it corrects the TS wording to match
the appropriate power parameter that was originally intended and
required by safety analyses, and that has been implemented since
original licensing of the PVNGS plants. Therefore, this 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: Phoenix Public Library, 1221
N. Central Avenue, Phoenix, Arizona 85004.
Attorney for licensee: Nancy C. Loftin, Esq., Corporate Secretary
and Counsel, Arizona Public Service Company, P.O. Box 53999, Mail
Station 9068, Phoenix, Arizona 85072-3999.
NRC Project Director: William H. Bateman.
Carolina Power & Light Company, Docket No. 50-261, H. B. Robinson Steam
Electric Plant, Unit No. 2, Darlington County, South Carolina
Date of amendment request: October 14, 1998.
Description of amendment request: The proposed change will revise
the H. B. Robinson, Unit 2, Technical Specification (TS) on Residual
Heat Removal Isolation Valve Interlock. The requested change modifies
the acceptance criterion for surveillance requirement (SR) 3.4.14.2
from setpoint value to the analytical limit for overpressurization of
the Residual Heat Removal System.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
The HBRSEP [H. B. Robinson Steam Electric Plant], Unit No. 2 TS
are proposed to be modified to increase the acceptance criterion for
Surveillance Requirement (SR) 3.4.14.2 from a RCS [reactor coolant
system] pressure of 465 psig to 474 psig. Carolina Power & Light
(CP&L) Company has evaluated the proposed Technical Specifications
(TS) change and has concluded that it does not involve a significant
hazards consideration. The conclusion is in accordance with the
criteria set forth in 10 CFR 50.92. The bases for the conclusion
that the proposed change does not involve a significant hazards
consideration is discussed below.
1. Does the change involve a significant increase in the
probability or consequences of an accident previously evaluated?
The proposed change increases the acceptance criterion for the
Residual Heat Removal (RHR) System interlock from 465 psig to 474
psig. The new value of 474 psig is the analytical limit for the RHR
System interlock setpoint that corresponds to the highest RCS
pressure that is allowable in the RHR System without
overpressurizing the RHR System above its design pressure. The RHR
System interlock prohibits remote manual operation of the RHR
Pressure Isolation Valves (PIVS) from the control room when Reactor
Coolant System (RCS) pressure is greater than the RHR System
interlock setpoint to avoid inadvertent overpressurization of the
RHR System due to operator action. Operating procedures prohibit
opening of the RHR PIVs when RCS pressure is greater than 375 psig.
Therefore, the probability of overpressurization of the RHR System
resulting in a Loss-of-Coolant Accident (LOCA) is not affected by
the change. The RHR System interlock provides no actuation function
to mitigate the consequences of a LOCA as a result of open RHR PIVs
with RCS pressure greater than the RHR System interlock setpoint.
Therefore, the consequences of overpressurization of the RHR System
is not affected by the change. Therefore, the proposed change does
not involve any increase in the probability or consequences of an
accident previously evaluated.
2. Does the change create the possibility of a new or different
kind of accident from any accident previously evaluated?
The proposed change does not involve any physical alteration of
plant systems, structures, or components. The proposed change
increases the acceptance criterion for the RHR System interlock SR
from 465 psig to the analytical limit of 474 psig. Performance of a
SR at the new acceptance criterion does not introduce any new
accident initiation scenarios since the SR is performed at
acceptable RCS pressure conditions. Therefore, the proposed change
does not create the possibility of a new or different kind of
accident from any accident previously evaluated.
3. Does this change involve a significant reduction in a margin
of safety?
The proposed change results in a new SR acceptance criterion
that corresponds to the analytical limit for the RHR System
interlock setpoint. The RHR System interlock is redundant to
administrative controls which prohibit opening the RHR System PIVs
under RCS pressure conditions which would overpressurize the RCS
System. Therefore, the proposed change does not result in 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: Hartsville Memorial Library,
147 West College Avenue, Hartsville, South Carolina 29550.
Attorney for licensee: William D. Johnson, Vice President and
Senior Counsel, Carolina Power & Light Company, Post Office Box 1551,
Raleigh, North Carolina 27602.
[[Page 59588]]
NRC Project Director: Frederick J. Hebdon.
Commonwealth Edison Company, Docket Nos. 50-237 and 50-249, Dresden
Nuclear Power Station, Units 2 and 3, Grundy County, Illinois
Docket Nos. 50-254 and 50-265, Quad Cities Nuclear Power Station, Units
1 and 2, Rock Island County, Illinois
Docket Nos. 50-373 and 50-374, LaSalle County Station, Units 1 and 2,
LaSalle County, Illinois
Date of application for amendment request: October 13, 1998.
Description of amendment request: The proposed amendments would
change the Dresden, Quad Cities, and LaSalle Technical Specifications
(TS) to reflect the use of Siemens Power Corporation (SPC) ATRIUM-9B
fuel. Specifically the proposed amendments incorporate the following
into the TS: (a) new methodologies that will enhance operational
flexibility and reduce the likelihood of future plant derates; (b)
administrative changes that eliminate the cycle-specific implementation
of ATRIUM-9B fuel and adopt Improved Standard Technical Specification
language where appropriate; and (c) changes to the Minimum Critical
Power Ratio (MCPR). This amendment request supplements the submittal of
August 14, 1998 (63 FR 48258). Changes in this supplement include only
a change in reference to a recently NRC-approved additive constant
uncertainty (ACU) generic methodology for ATRIUM-9B fuel (ANF-
1125(P)(A), Supplement 1, Appendix E) from Appendix D which provided an
interim value for ACU.
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 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.
a. Addition of SPC Revised Jet Pump Methodology (LaSalle Units 1 and 2)
The Reference 1 [ANF-91-048(P)(A), Supplement 1 and Supplement
2, ``BWR Jet Pump Model Revision for RELAX,'' October 1997 and NRC
SER, ``Review of Siemens Topical Report ANF-91-048(P), BWR Jet Pump
Revisison for RELAX (TAC No M995381), T.H. Essig to H.D. Curet,
September 19, 1997] methodology to be added to the Technical
Specifications is used as part of the LOCA [loss-of-coolant
accident] analysis and does not introduce physical changes to the
plant. The Reference 1 revised jet pump model changes the
calculational behavior of the jet pump under reversed drive flow
conditions. The revised jet pump model methodology makes the LOCA
model behave more realistically and calculates small break LOCA PCTs
[peak cladding temperature] that are comparable to the large break
LOCA results. Therefore, this change only affects the methodology
for analyzing the LOCA event and determining the protective APLHGR
[average planar linear heat generation rate] limits. The Technical
Specification requirements for monitoring APLHGR are not affected by
this change. The revised method will result in higher APLHGR limits,
thus the SPC fuel will be allowed to operate at higher nodal powers.
The approved methodology, however, still protects the fuel
performance limits specified by 10 CFR 50.46. Therefore, the
probability or consequences of an accident previously evaluated will
not change.
b. Addition of SPC Generic Methodology for Application of ANFB
[Advanced Nuclear Fuel for Boiling Water Reactors] Critical Power
Correlation to Non-SPC Fuel (Quad Cities Units 1 and 2 and LaSalle
Units 1 and 2)
The probability or consequences of a previously evaluated
accident are not increased by adding Reference 3 [EMF-1125(P)(A),
Supplement 1 Appendix C, ``ANFB Critical Power Correlation
Application for Coresident Fuel,'' August 1997, and NRC SER,
``Acceptance for Referencing of Licensing Topical Report EMF-
1125(P), Supplement 1 Appendix C, ``ANFB Critical Power Correlation
Application for Co-Resident Fuel,'' J.E. Lyons to R. A. Copeland,
May 9, 1997] to Section 6.9.A.6.b of the Quad Cities Technical
Specifications and Bases Section 2.1.2 and Section 6.6.A.6.b of the
LaSalle Technical Specifications. Reference 3 determines the
additive constants and the associated uncertainty for application of
the ANFB correlation to the coresident GE [General Electric Co.]
fuel. Therefore, it provides data that is used in the determination
of the MCPR Safety Limit. This approved methodology for applying the
ANFB critical power correlation to the GE fuel will protect the fuel
from boiling transition. Operational MCPR limits will also be
applied to ensure that the MCPR Safety Limit is protected during all
modes of operation and anticipated operational occurrences. Because
Reference 3 contains conservative methods and calculations and
because the operability of plant systems designed to mitigate any
consequences of accidents have not changed, the probability or
consequences of an accident previously evaluated will not increase.
c. Addition of SPC Topical for Revised ANFB Correlation Uncertainty
(Quad Cities Units 1 and 2, Dresden Units 2 and 3, and LaSalle Units 1
and 2)
The probability or consequences of a previously evaluated
accident are not increased by adding Reference 7 [ANF-1125(P),
Supplement 1, Appendix E, ``ANFB Critical Power Correlation
Determination of ATRIUM-9B Additive Constant Uncertainties,'' and
NRC SER, ``Acceptance for Referencing of Licensing Topical Report
ANF-1125(P), Supplement 1, Appendix E, ``ANFB Critical Power
Correlation Determination of ATRIUM-9B Additive Constant
Uncertainties'' (TAC No. MA2437), T.H. Essig to H.D. Curet,
September 23, 1998] to Section'' 6.9.A.6.b of the Quad Cities and
Dresden Technical Specifications and Bases Section 2.1.2 and Section
6.6.A.6.b of the LaSalle Technical Specifications. Reference 7
documents the additive constant uncertainty for the SPC ATRIUM-9B
fuel design with an internal water channel. This methodology is used
to determine an input to the MCPR Safety Limit calculations, which
ensures that at least 99.9 percent of the fuel rods avoid transition
boiling during normal operation as well as anticipated operational
occurrences. This change does not require any physical plant
modifications, physically affect any plant components, or entail
changes in plant operation. This methodology for determining the
ATRIUM-9B additive constant uncertainty for the MCPR Safety Limit
calculation will continue to support protecting the fuel from
boiling transition. Operational MCPR limits will be applied to
ensure the MCPR Safety Limit is not violated during all modes of
operation and anticipated operational occurrences. Therefore, no
individual precursors of an accident are affected and the
operability of plant systems designed to mitigate the probability or
the consequences of an accident previously evaluated is not affected
by these changes.
d. Change to Minimum Critical Power Ratio Safety Limit (Quad Cities
Units 1 and 2, Dresden Unit 3, and LaSalle Units 1 and 2)
Changing the MCPR Safety Limit at Quad Cities Units 1 and 2,
Dresden Unit 3, and LaSalle Units 1 and 2 will not increase the
probability or the consequences of an accident previously evaluated.
The MCPR Safety Limits for Quad Cities Units 1 and 2, Dresden Unit
3, and LaSalle Units 1 and 2 are anticipated to be conservative and
acceptable for future cycles. 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 Limits are being set at the CPR [critical power ratio] value
where less than 0.1 percent of the rods in the core are expected to
experience boiling transition. These Safety Limits are expected to
be applicable for future cycles of ATRIUM-9B. Therefore the
probability or consequences of an accident will not increase.
[[Page 59589]]
e. Removal of Footnotes Limiting Operation with ATRIUM-9B Fuel Reloads
(Quad Cities Unit 2 and Dresden Units 2 and 3)
The removal of footnotes from the Quad Cities and Dresden
Technical Specifications does not involve any significant increase
in the probability or consequences of an accident previously
evaluated. The footnotes were added to clarify that cycle specific
methods were used until the generic methodology was approved by the
NRC. Since the NRC has approved SPC's generic methodology for
application of the ANFB correlation to the coresident GE fuel
(Reference 3) and SPC has addressed the concerns regarding the
database used to calculate the ATRIUM-9B additive constant
uncertainties (Reference 7), the footnotes are no longer necessary.
The removal of the Unit 2 specific ``a'' pages, 2-1a and B2-3a, in
the Quad Cities Technical Specifications is justified by the removal
of the footnotes. Therefore, removing these footnotes and ``a''
pages does not require any physical plant modifications, nor does it
physically affect any plant components or entail changes in plant
operation. Therefore, the probability or consequences of an accident
previously evaluated are not expected to increase.
f. Revision to Thermal Limit Descriptions (Quad Cities Units 1 and 2,
Dresden Units 2 and 3, and LaSalle Units 1 and 2)
The revision to the Section 3 Technical Specification
description of the APLHGR limits has no implications on accident
analysis or plant operations. The purpose of the revision is to
allow flexibility for the MAPLHGR [maximum planar linear heat
generation rate] limits and their exposure basis to be specified in
the COLR [core operating limit report] and to establish consistency
with approved methodologies currently utilized by Siemens Power
Corporation, which calculate MAPLHGR limits based on bundle or
planar average exposures. This revision also provides for
consistency in the APLHGR limit Technical Specification wording
between the ComEd BWRs. The revision to the 3.11.D SLHGR [steady
state linear heat generation rate] Technical Specification for
Dresden also has no implications on accident analysis or plant
operations. The purpose of this revision is to allow flexibility for
the LHGR [linear heat generation rate] limits and their exposure
basis to be specified in the COLR. This revision makes the Dresden
LHGR definition consistent with NUREG 1433/1434, Revision 1 wording.
The definition of the Average Planar Exposure is deleted, because
the exposure basis of the APLHGR and LHGR is being removed.
Therefore, no plant equipment or processes are affected by this
change. Thus, there is no alteration 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:
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.
a. Addition of SPC Revised Jet Pump Methodology (LaSalle Units 1 and 2)
The revised jet pump model methodology will be used to analyze
the LOCA for LaSalle Units 1 and 2, and does not introduce any
physical changes to the plant or the processes used to operate the
plant. This change only affects the methods used to analyze the LOCA
event and determine the MAPLHGR limits. Therefore, the possibility
of a new or different kind of accident is not created.
b. Addition of SPC Generic Methodology for Application of ANFB Critical
Power Correlation to Non-SPC Fuel (Quad Cities Units 1 and 2 and
LaSalle Units 1 and 2)
Addition of the generic methodology for the application of the
ANFB critical power correlation to GE fuel in Section 6.9.A.6.b of
the Quad Cities Technical Specifications and Bases Section 2.1.2 and
Section 6.6.A.6.b of the LaSalle Technical Specifications does not
introduce any physical changes to the plant, the processes used to
operate the plant, or allowable modes of operation. This change only
involves adding an NRC approved methodology, which is used to
determine the additive constants and additive constant uncertainty
for GE fuel, to Section 6 of the Technical Specifications.
Therefore, no new precursors of an accident are created and no new
or different kinds of accidents are created.
c. Addition of SPC Topical for Revised ANFB Correlation Uncertainty
(Quad Cities Units 1 and 2, Dresden Units 2 and 3, and LaSalle Units 1
and 2)
Addition of the Reference 7 methodology to Section 6.9.A.6.b of
the Quad Cities and Dresden Technical Specifications and Bases
Section 2.1.2 and Section 6.6.A.6.b of the LaSalle Technical
Specifications will not create the possibility of a new or different
kind of accident from any accident previously evaluated. This
methodology describes the calculation of an input to the MCPR Safety
Limit--the ATRIUM-9B additive constant uncertainty. This change does
not introduce any physical changes to the plant, the processes used
to operate the plant, or allowable modes of operation. Therefore, no
new precursors of an accident are created and no new or different
kinds of accidents are created.
d. Change to Minimum Critical Power Ratio Safety Limit (Quad Cities
Units 1 and 2, Dresden Unit 3, and LaSalle Units 1 and 2)
Changing the MCPR Safety Limit will not create the possibility
of a new accident from an accident previously evaluated. This change
will not alter or add any new equipment or change modes of
operation. The MCPR Safety Limit is established to ensure that 99.9
percent of the rods avoid boiling transition.
The MCPR Safety Limit is changing for Quad Cities, Dresden Unit
3 and LaSalle due to the revised ATRIUM-9B additive constants and
the ATRIUM-9B additive constant uncertainty calculated in Reference
7. The new MCPR Safety Limit for Quad Cities Units 1 and 2, Dresden
Unit 3, and LaSalle Units 1 and 2 are greater than the current
values at Quad Cities Units 1 and 2, Dresden Unit 3, and LaSalle
Units 1 and 2 and are being increased now in anticipation of
bounding future reloads of ATRIUM-9B. This change does not introduce
any physical changes to the plant, the processes used to operate the
plant, or allowable modes of operation. Therefore, no new accidents
are created that are different from any accident previously
evaluated.
e. Removal of Footnotes Limiting Operation with ATRIUM-9B Fuel Reloads
(Quad Cities Unit 2 and Dresden Units 2 and 3)
The removal of the footnotes from the Quad Cities and Dresden
Technical Specifications does not create a new or different kind of
accident from any accident previously evaluated. The removal of the
footnotes does not affect plant systems or operation. The footnotes
were temporarily established to implement a conservative cycle
specific MCPR Safety Limit until the SPC generic methodology was
approved. With the approval of References 3 and 7, these footnotes
are no longer applicable. Removing these footnotes does not
introduce any physical changes to the plant, the processes used to
operate the plant, or allowable modes of operation. The removal of
the Unit 2 specific ``a'' pages, 2-1a and B2-3a, in the Quad Cities
Technical Specifications, which is justified by the removal of the
footnotes, also does not create a new or different kind of accident
from any accident previously evaluated.
f. Revision to Thermal Limit Descriptions (Quad Cities Units 1 and 2,
Dresden Units 2 and 3, and LaSalle 1 and 2)
The revision of the APLHGR and LHGR limit descriptions will not
create the possibility of a new or different kind of accident from
any accident previously evaluated. This revision will not alter any
plant systems, equipment, or physical conditions of the site. This
revision allows the flexibility of the APLHGR and the LHGR limits to
be specified in the COLR and to maintain consistency with the
calculated results of methodologies currently used to determine the
APLHGR. The definition of the Average Planar Exposure is deleted,
because it is being removed from LHGR and APLHGR Technical
Specifications. This change does not introduce any physical changes
to the plant, the processes used to operate the plant, or allowable
modes of operation. Therefore this change does not create the
possibility of a new or different kind of accident from any accident
previously evaluated.
3. Involve a significant reduction in the margin of safety for
the following reasons:
a. Addition of SPC Revised Jet Pump Methodology (LaSalle Units 1 and 2)
The revised jet pump model methodology, and the MAPLHGRs,
resulting from the revised jet pump methodology, will continue
[[Page 59590]]
to ensure fuel design criteria and 10 CFR 50.46 compliance. The
results of LOCA analyses performed with this methodology must
continue to comply with the requirements of 10 CFR 50.46. Therefore,
there is no significant reduction in the margin of safety.
b. Addition of SPC Generic Methodology for Application of ANFB Critical
Power Correlation to Non-SPC Fuel (Quad Cities Units 1 and 2 and
LaSalle Units 1 and 2)
The margin of safety is not decreased by adding Reference 3 to
Section 6.9.A.6.b of the Quad Cities Technical Specifications and
Bases Section 1.2 and Section 6.6.A.6.b of the LaSalle Technical
Specifications. Siemens Power Corporation methodology for
application of the ANFB Critical Power Correlation to coresident GE
fuel is approved by the NRC and is the same methodology used in the
cycle specific topicals for coresident fuel (References 4 [EMF-96-
021(P), Revision 1, Application of the ANFB Critical Power
Correlation to Coresident GE fuel for LaSalle Unit 2 Cycle 8,''
February 1996, and NRC SER, ``Safety Evaluation for Topical Report
EMF-96-021(P), Revision 1, `Application of the ANFB Critical Power
Correlation to Coresident GE Fuel for LaSalle Unit 2 Cycle 8' (TAC
NO. M94964),'' D.M. Skay to I. Johnson, September 26, 1996] and 5
[EMF-96-051(P), ``Application of the ANFB Critical Power Correlation
to Coresident GE Fuel for Quad Cities Unit 2 Cycle 15,'' May 1996,
and NRC SER, ``Approval of Topical Report EMF-96-051(P)--Quad
Cities, Unit 2 (TAC NO. M96213),'' R. Pulsifer to I. Johnson, May
16, 1997]). The MCPR Safety Limit will continue to ensure that
greater than 99.9 percent of the rods in the core avoid boiling
transition. Additionally, operating limits will be established to
ensure the MCPR Safety Limit is not violated during all modes of
operation.
c. Addition of SPC Topical for Revised ANFB Correlation Uncertainty
(Quad Cities Units 1 and 2, Dresden Units 2 and 3, and LaSalle Units 1
and 2)
The MCPR Safety Limit provides a margin of safety by ensuring
that less than 0.1 percent of the rods are expected to be in boiling
transition if the MCPR Safety Limit is not violated. This Technical
Specification amendment request proposes to insert the topical
report that describes SPC's calculation of the ATRIUM-9B additive
constant uncertainty. The new ATRIUM-9B additive constant
uncertainty calculation is conservative and is based on a larger
database than previous calculations. Because the criteria of
ensuring that 99.9 percent of the rods are expected to avoid boiling
transition has not been changed and a conservative method is used to
calculate the ATRIUM-9B additive constant uncertainty, a decrease in
the margin to safety will not occur due to adding this methodology
to the Technical Specifications. In addition, operational limits
will be established to ensure the MCPR Safety Limit is protected for
all modes of operation. This revised methodology will ensure that
the appropriate level of fuel protection is being employed.
d. Change to Minimum Critical Power Ratio Safety Limit (Quad Cities
Units 1 and 2, Dresden Unit 3, and LaSalle Units 1 and 2)
Changing the MCPR Safety Limit for Quad Cities Units 1 and 2,
Dresden Unit 3, and LaSalle Units 1 and 2 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 percent of the rods
are calculated to be in boiling transition if the MCPR Safety Limit
is not violated. The proposed Technical Specification amendment
request reflects the MCPR Safety Limit results from conservative
evaluations by SPC using the ANFB critical power correlation with
the ATRIUM-9B additive constant uncertainty calculated in Reference
7.
Because a conservative method is used to apply the ATRIUM-9B
additive constant uncertainty in the MCPR Safety Limit calculation,
a decrease in the margin to safety will not occur due to changing
the MCPR Safety Limit. The revised MCPR Safety Limit will ensure the
appropriate 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 occurrences. This
will ensure that the fuel design safety criterion of more than 99.9
percent of the fuel rods avoiding transition boiling during normal
operation as well as during an anticipated operational occurrence is
met.
e. Removal of Footnotes Limiting Operation with ATRIUM-9B Fuel Reloads
(Quad Cities Unit 2 and Dresden Units 2 and 3)
The removal of the cycle specific footnotes in Quad Cities and
Dresden Technical Specifications does not impose a change in the
margin of safety. These footnotes were added due to concerns
regarding the calculation of the additive constant uncertainty for
the ATRIUM-9B fuel and the cycle specific application of the ANFB
critical power correlation to coresident GE fuel in Quad Cities Unit
2 Cycle 15. Because the generic ANFB application to coresident GE
fuel MCPR methodology (Reference 3) has received NRC approval and
the topical report describing the increased database used to
calculate the additive constant uncertainties for ATRIUM-9B
(Reference 7) has also received NRC approval and both are proposed
to be added to the Technical Specifications in this amendment
request, there is no reason for the footnotes to remain. Removal of
the Unit 2 specific ``a'' pages, 2-1a and B2-3a, in the Quad Cities
Technical Specifications is justified by the removal of the
footnotes. Therefore, the removal of the ``a'' pages, 2-1a and B2-
3a, also does not impose a change in the margin of safety.
f. Revision to Thermal Limit Descriptions (Quad Cities Units 1 and 2,
Dresden Units 2 and 3, and LaSalle Units 1 and 2)
The revision to the APLHGR and LHGR limit descriptions will not
involve a reduction in the margin of safety. The methodology used to
calculate the APLHGR must comply with the guidelines of Appendix K
of 10 CFR Part 50, and the APLHGR and LHGR will still be required to
be maintained within the limits specified in the COLR. The
surveillance requirements for these two thermal limits remain
unchanged. Thus, there will be 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: For Dresden, Morris Area
Public Library District, 604 Liberty Street, Morris, Illinois 60450;
for Quad Cities, Dixon Public Library, 221 Hennepin Avenue, Dixon,
Illinois 61021; and for LaSalle, the Jacobs Memorial Library, 815 North
Orlando Smith Avenue, Illinois Valley Community College, Oglesby,
Illinois 61348-9692.
Attorney for licensee: Michael I. Miller, Esquire; Sidley and
Austin, One First National Plaza, Chicago, Illinois 60603. NRC Project
Director: Stuart A. Richards.
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 amendment request: September 30, 1998.
Description of amendment request: The proposed amendment would
increase the maximum fuel rod internal pressure in the spent fuel pool
from 1200 pounds per square inch gauge (psig) to 1300 psig by changing
the Updated Final Analysis Report (UFSAR) reference to the computer
code used to determine the fuel rod internal pressure (TACO3 computer
code would be added) in UFSAR Chapter 15. The proposed amendment would
also provide justification for not increasing the overall effective
decontamination factor for iodine as a consequence of a fuel handling
accident. In addition, the term ``fuel assembly gap gas pressure''
would be changed to ``fuel rod internal pressure'' to correct an UFSAR
error.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
The following discussion is a summary of the evaluation of the
changes contained in this proposed amendment against the 10 CFR
50.92 (c) requirements to demonstrate that all three standards for
no significant hazards consideration are satisfied. A no significant
hazards consideration is indicated if
[[Page 59591]]
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.
First Standard
Implementation of this amendment would not involve a significant
increase in the probability or consequences of an accident
previously evaluated. The increase in maximum rod internal pressure
in the spent fuel pool from 1200 psig to 1300 psig does not result
in a significant change in the calculated overall effective
decontamination factor for iodine (described in Attachment 1) [of
the licensee's submittal]. Therefore, the continued use of an
overall effective decontamination factor for iodine of 89 can be
justified. Therefore, there is no significant increase in the dose
consequences for a fuel handling accident at Oconee Nuclear Station.
Implementation of the BAW-10183P-A (Reference 4) methodology,
which allows fuel rod internal pressure to exceed system pressure,
also increases the fuel rod pressure at spent fuel pool conditions.
The fuel is currently licensed to rod internal pressure of system
pressure plus a proprietary amount above system pressure. This
criteria represents a separate limit from the maximum internal
pressure in the spent fuel pool criteria. Thus, an increase in the
maximum rod internal pressure in the spent fuel pool does not affect
the mechanical design limit specified in Reference 4. Therefore, an
increase in the maximum internal pressure in the spent fuel pool
does not constitute a significant increase in the probability of an
accident previously evaluated.
Second Standard
Implementation of this amendment will not create the possibility
of a new or different kind of accident from any previously
evaluated. The fuel handling accident is the bounding accident.
Implementation of this amendment will not impact any plant systems
that are accident initiators. No other modifications are being
proposed in the plant which would result in the creation of a new
accident mechanism. Also, no changes are being made to the way the
plant is operated; therefore, no new failure mechanisms will be
initiated.
Third Standard
Implementation of this amendment would not involve a significant
reduction in a margin of safety. As discussed in Attachment 1 [of
the licensee's submittal], the overall effective decontamination
factor (DF) of 522 was determined for a rod internal pressure of
1200 psig, and a DF of 443 for a rod internal pressure of 1300 psig
based on a spent fuel pool depth of 21.34 feet. Both of these
factors are well above the DF of 89 currently used in the fuel
handling accident analyses. The margin of safety is a factor of 5.
Based upon the preceding analysis, Duke proposes that ample
margin is retained to justify the continued use of a DF of 89 at a
maximum rod internal pressure of 1300 psig. Therefore, Duke has
concluded that the proposed amendment does not involve a significant
hazards consideration.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: Oconee County Library, 501
West South Broad Street, Walhalla, South Carolina.
Attorney for licensee: J. Michael McGarry III, Winston and Strawn,
1200 17th Street, NW., Washington, DC.
NRC Project Director: Herbert N. Berkow.
Duquesne Light Company, et al., Docket No. 50-412, Beaver Valley Power
Station, Unit 2, Shippingport, Pennsylvania.
Date of amendment request: September 24, 1998.
Description of amendment request: The proposed amendment would
revise technical specification (TS) 3.1.2.8 in two places to change the
term ``contained volume'' to ``usable volume.'' This change would
eliminate the potential for a non-conservative interpretation of the
specification values for the Refueling Water Storage Tank and Boric
Acid Storage System (BAT) and would eliminate the need for plant
administrative controls, which currently interpret these volumes as
usable volumes.
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 proposed Limiting Condition for Operation (LCO) change will
assure that the Refueling Water Storage Tank (RWST) minimum usable
volume is maintained consistent with that required by accident
analysis. The safety function of the RWST will not differ in any way
from its normal operational mode. The normal operation of plant
equipment is not a precursor to any accident. Therefore, operation
of equipment under this change will not increase the probability or
consequences of an accident previously evaluated.
2. Does the change create the possibility of a new or different
kind of accident from any accident previously evaluated?
The proposed amendment will not change the physical plant or the
modes of plant operation defined in the operating license. The
change does not involve the addition or modification of equipment
nor does it alter the design or operation of plant systems. The
proposed change will help to ensure that the analysis value of
minimum contained volume is available, so that the RWST can perform
its safety function.
Therefore, operation of the facility 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. Does the change involve a significant reduction in a margin
of safety?
RWST: The basis for TS 3.1.2.8.b is to ensure adequate water for
the Emergency Core Cooling System to respond to a Large Break Loss
Of Coolant Accident; supply the containment with cooling spray flow;
supply the containment sump with adequate water for Recirculation
Spray pump suction head concerns; and to provide adequate boron to
shut down the core. This change will ensure that the proper tank
volume is maintained to support the Design Basis Accident (DBA)
analysis.
BAT: These tanks are credited for ensuring adequate Shutdown
Margin in the event that the unit has to initiate an emergency
shutdown. Additional requirements are derived for the postulated
Anticipated Transient Without Scram event. This change will ensure
that the proper tank volume is maintained to support the DBA
analysis.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: B. F. Jones Memorial Library,
663 Franklin Avenue, Aliquippa, PA 1500l.
Attorney for licensee: Jay E. Silberg, Esquire, Shaw, Pittman,
Potts & Trowbridge, 2300 N Street, NW., Washington, DC 20037.
NRC Project Director: Robert A. Capra.
Duquesne Light Company, et al., Docket No. 50-412, Beaver Valley Power
Station, Unit 2, Shippingport, Pennsylvania
Date of amendment request: October 16, 1998.
Description of amendment request: The proposed amendment would
extend on a one time only basis, the surveillance interval for
technical specifications (TSs) 4.8.1.1.1.b and 4.8.1.2 from its current
due date of January 30, 1999, to the first entry into Mode 4 following
the seventh refueling outage (2R7), but not later than May 1, 1999, by
adding a new License Condition 2.C(12). The purpose of TSs 4.8.1.1.1.b
and 4.8.1.2 is to demonstrate the ability to transfer the unit power
[[Page 59592]]
supply from the unit circuit to the system circuit.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. Does the change involve a significant increase in the
probability or consequences of an accident previously evaluated?
The proposed change is temporary and allows a one time extension
of the automatic transfer function 18 month surveillance requirement
specified in Surveillance Requirement (SR) 4.8.1.1.1.b. This
surveillance requirement is also referenced in SR 4.8.1.2. The
proposed surveillance interval extension will not cause a
significant reduction in system reliability nor affect the ability
of a system to perform its design function. The proposed change does
not affect the UFSAR [Updated Final Safety Analysis Report] accident
analyses since a loss of offsite power is assumed during a design
basis accident. Therefore, this change does not involve a
significant increase in the probability or consequences of an
accident previously evaluated.
2. Does the change create the possibility of a new or different
kind of accident from any accident previously evaluated?
Extending the surveillance interval for the performance of
specific testing will not create the possibility of any new or
different kind of accidents. No change is required to any system
configurations, plant equipment or analyses. The UFSAR accident
analyses assume a loss of offsite power; therefore, loss of the
automatic bus transfer feature will not create the possibility of a
new or different kind of accident from any accident previously
evaluated.
3. Does the change involve a significant reduction in a margin
of safety?
Extending the surveillance interval for the automatic transfer
function will not impact any plant safety analyses since the UFSAR
accident analyses assume the loss of offsite power. The safety
limits assumed in the accident analyses and the design function of
the equipment required to mitigate the consequences of any
postulated accidents will not be changed since only the 18 month
surveillance test interval is being extended. Based on engineering
judgment, extending the surveillance test interval for the
performance of this specific test could slightly reduce the margin
of safety derived from the required surveillances. However, past
experience has shown that the system which automatically transfers
power from the unit to the system circuit supply is reliable. The
manual transfer requirement of SR 4.8.1.1.1.b demonstrates that the
breakers relied upon for the transfer of power are functional and
provides an opportunity to identify potential equipment degradation.
The manual transfer requirement of SR 4.8.1.1.1.b will continue to
be completed within the required surveillance interval. Therefore,
the plant will be maintained within the analyzed limits and the
proposed extension will not significantly 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 proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: B. F. Jones Memorial Library,
663 Franklin Avenue, Aliquippa, PA 1500l.
Attorney for licensee: Jay E. Silberg, Esquire, Shaw, Pittman,
Potts & Trowbridge, 2300 N Street, NW., Washington, DC 20037.
NRC Project Director: Robert A. Capra.
Entergy Gulf States, Inc., and Entergy Operations, Inc., Docket No. 50-
458, River Bend Station, Unit 1, West Feliciana Parish, Louisiana
Date of amendment request: September 22, 1998.
Description of amendment request: The proposed amendment would
delete license conditions associated with the River Bend Station (RBS)
Transamerica Delaval, Inc. (TDI) emergency diesel generators (EDGs),
which prescribe certain inspection requirements associated with various
overload conditions experienced by the EDGs. Current license
requirements were issued following publication of NUREG-1216, which
called for extensive periodic engine tear-downs as the major part of a
maintenance and surveillance program for TDI engines. The proposed
removal of license conditions appears to be consistent with the NRC's
approval of Generic Topical Report TDI-EDG-001-A ``Basis for
Modification to Inspection Requirements for Transamerica Delaval, Inc.,
Emergency Diesel Generators''. EOI currently inspects and maintains its
EDGs in accordance with Technical Requirements Manual (TRM)
surveillance requirement TSR 3.8.1.21. Periodicity of planned
inspections and maintenance are based upon the manufacturer's
recommendations for standby service.
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 the
consequences of an accident previously evaluated:
Diesel generators are not accident initiating equipment.
Elimination of the non-routine tear-downs and inspections will not
adversely affect the probability of an accident occurring. Regular
maintenance programs (which may include periodic tear-downs and
inspections) in lieu of this specific license condition would
decrease the consequences of an accident because of the availability
of the engines will increase as a result of the less frequent tear-
downs. (See Generic Topical Report TDI-EDG-001-A, ``Basis for
Modification to Inspection Requirements for Transamerica Delaval,
Inc., Emergency Diesel Generators'') Additionally, the high average
reliability of the TDI engines will not be negatively affected due
to this change. NRC research has shown there is a period of
decreased reliability immediately following intrusive tear-downs
(break-in period), followed by a long period of high reliability.
Continued monitoring and maintenance as implemented by Technical
Requirements Manual (TRM) surveillances will contribute to continued
high reliability of the EDGs.
2. Create the possibility of a new or different kind of accident
from any previously evaluated:
The proposed amendment does not affect the design or function of
any plant structure, system, or component, nor does it change the
way plant systems are operated. The proposed amendment will not
cause any physical change to the plant or the design or operation of
the diesel units. This change will only affect the frequency of
tear-down inspections of the EDGs, and not the physical activities
performed during such inspections. Therefore, the removal of the
existing condition from the operating license will not create the
possibility of a new or different kind of accident from any
previously evaluated.
3. Involve a significant decrease in the margin of safety.
The proposed amendment does not affect parameters which would
result in a significant reduction in margin of safety. Operating
experience and data have shown increased reliability can be achieved
by eliminating unnecessary tear-down inspections, such as those
prescribed by this license condition. Maintenance of the EDGs is
presently scheduled in accordance with the vendor's recommendations.
The RBS corrective action program provides a means to evaluate
future operational events and take the appropriate actions.
Therefore, the proposed amendment does not involve a significant
decrease in the margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: Government Documents
Department, Louisiana State University, Baton Rouge, LA 70803.
Attorney for licensee: Mark Wetterhahn, Esq., Winston & Strawn,
1400 L Street, N.W., Washington, D.C. 20005.
NRC Project Director: John N. Hannon.
[[Page 59593]]
Entergy Operations Inc., Docket No. 50-382, Waterford Steam Electric
Station, Unit 3, St. Charles Parish, Louisiana
Date of amendment request: May 28, 1998.
Description of amendment request: This amendment requests changes
to Technical Specification 3.7.1.2 and Surveillance Requirement 4.7.1.2
for the Emergency Feedwater System. The amendment will expand and
clarify the current specification. A change to Technical Specification
Bases 3/4.7.1.2 has been included to support the changes.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
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 proposed changes included in this amendment request are
being made to the Emergency Feedwater (EFW) System Technical
Specification. These changes include clarification of the LCO
[limiting conditions for operation], a 7 day allowed outage time for
an inoperable steam supply, additional ACTION requirements for
inoperable flow path(s), a requirement to test the pumps pursuant to
Specification 4.0.5, and rewording of numerous Surveillance
Requirements consistent with NUREG-1432, ``Standard Technical
Specifications Combustion Engineering Plants.''
The administrative and more restrictive changes will not affect
the assumptions, design parameters, or results of any accident
previously evaluated. The accident mitigation features of the plant
are not affected by these proposed changes. The proposed changes do
not add or modify any existing equipment. The administrative change
to test EFW pumps pursuant to the Inservice Test Program will ensure
the EFW pumps are tested against the more restrictive of the data
points required by either the safety analysis or the Inservice Test
Program. Therefore, the proposed administrative changes do not
involve a significant increase in the probability or consequences of
any accident previously evaluated.
The less restrictive changes (allowing 7 days for an inoperable
pump due to an inoperable steam supply, performing Surveillance
Requirements during other than shut down conditions, allowing the
use of actual actuation signals in addition to test signals, and
delaying the requirement to complete Surveillance Requirement ``d''
to just prior to Mode 2) will not affect the assumptions, design
parameters, or results of any accident previously evaluated. The
accident mitigation features of the plant are not affected by these
proposed changes. The proposed changes do not add or modify any
existing equipment. Therefore, the proposed less restrictive changes
do 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 type of
accident from any accident previously evaluated?
Response: No.
The proposed changes included in this amendment request are
being made to the EFW System Technical Specification. These changes
include clarification of the LCO, a 7 day allowed outage time for an
inoperable steam supply, additional ACTION requirements for
inoperable flow path(s), a requirement to test the pumps pursuant to
Specification 4.0.5, and rewording of numerous Surveillance
Requirements consistent with NUREG-1432. These changes do not alter
the design nor configuration of the plant. There has been no
physical change to plant systems, structures, or components. The
proposed changes will not reduce the ability of any of the safety-
related equipment required to mitigate Anticipated Operational
Occurrences or accidents. Therefore, the proposed changes will not
create the possibility of a new or different kind of accident from
any accident 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 proposed changes included in this amendment request are
being made to the EFW System Technical Specification. These changes
include clarification of the LCO, a 7 day allowed outage time for an
inoperable steam supply, additional ACTION requirements for
inoperable flow path(s), a requirement to test the pumps pursuant to
Specification 4.0.5, and rewording of numerous Surveillance
Requirements consistent with NUREG-1432.
The proposed change to the LCO requiring three pumps and two
flow paths be OPERABLE maintains the functionality of the EFW such
that it is capable of performing its design function as assumed in
the Updated Final Safety Analysis Report. If the functionality of
the system is not maintained, Technical Specifications require
ACTIONs be taken, within specified time limitations, to restore EFW
to OPERABLE status or shut down the reactor. This action is
consistent with the existing Technical Specification and NUREG-1432.
The allowed outage time for one inoperable steam supply has been
increased from 72 hours to 7 days in accordance with NUREG-1432.
This is acceptable due to the redundant OPERABLE steam supply, the
availability of redundant OPERABLE motor-driven EFW pumps, and the
low probability of an event requiring the inoperable steam supply.
This change is consistent (other than format) with NUREG-1432 and
has therefore been previously approved by the NRC.
The ACTION for one flow path inoperable (but capable of
delivering 100% flow) as proposed will allow a 72 hour completion
time for an inoperable flow path. This change is acceptable based on
the availability of at least two OPERABLE EFW pumps, a redundant
OPERABLE flow path capable of feeding the other steam generator and
the capability of the inoperable flow path to deliver 100% of the
required EFW flow to the affected steam generator.
The ACTION for one flow path inoperable (not capable of
delivering 100% flow) as proposed requires a unit shutdown be
initiated immediately. This change is appropriate due to the
seriousness of the condition and is acceptable due to the
availability of the remaining operable flow path to support the unit
shut down.
The ACTION for two flow paths not capable of delivering 100%
flow is the same as that for three pumps inoperable. With two flow
paths inoperable such that neither flow path is capable of
delivering 100% flow the unit is in a seriously degraded condition
just as it is with all three pumps inoperable. The ACTION as
proposed requires that immediate action be taken to restore one flow
path to OPERABLE status. This change is consistent with the intent
of the current EFW Technical Specification.
Testing pursuant to Specification 4.0.5 (Inservice Testing
Program) as proposed for Surveillance Requirement `b' will ensure
the EFW pumps are tested against the more restrictive of the data
points required by either the safety analysis or ASME Section XI.
The remaining changes to the EFW Technical Specification are
consistent (other than format) with NUREG-1432 and have therefore
been previously approved by the NRC.
Therefore, based on the above discussion, the proposed change
will not involve a significant reduction in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room Location: University of New Orleans
Library, Louisiana Collection, Lakefront, New Orleans, LA 70122.
Attorney for licensee: N.S. Reynolds, Esq., Winston & Strawn 1400 L
Street N.W., Washington, D.C. 20005-3502.
NRC Project Director: John N. Hannon.
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: September 28, 1998.
Description of amendment request: In 1997 Northeast Nuclear Energy
Company (the licensee) changed the Final Safety Analysis Report (FSAR)
Section 8.7.3.1 electrical separation requirements from 12 inches to 6
[[Page 59594]]
inches. At that time, the licensee concluded that the FSAR changes did
not involve an unreviewed safety question. Therefore, the licensee did
not request a license amendment to implement the FSAR change. The
licensee has since determined that, although the changes were safe, an
unreviewed safety question was involved. Therefore, the licensee is now
requesting NRC's review and approval, through an amendment to Operating
License No. DPR-65 pursuant to 10 CFR 50.90, regarding the separation
requirement of 6 inches in Millstone Unit No. 2 FSAR (which is applied
to redundant vital cables, internal wiring of redundant vital circuits,
and associated devices).
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:
In accordance with 10CFR50.92, NNECO [Northeast Nuclear Energy
Company] has reviewed the proposed changes and has concluded that
they do not involve a Significant Hazards Consideration (SHC). The
basis for this conclusion is that the three criteria of
10CFR50.92(c) are not compromised. The proposed changes do not
involve an SHC because the changes would not:
1. Involve a significant increase in the probability or
consequences of an accident previously evaluated.
The FSAR changes reduce the minimum allowable separation between
redundant vital wires/devices of different channels from twelve
inches to six inches. Reducing the physical separation between
wires/devices does not in itself increase the probability of any
credible event that would challenge circuit operability since the
wire/device characteristics have not changed and there is no change
in the circuit the wires/devices are in. The probability that an
accident could occur due to the change in separation is not
increased since the remaining separation will still prevent adverse
channel interactions (i.e. short circuit, etc.). The six inch
standard is acceptable in accordance with IEEE standard 384-1981
[IEEE standard 384-1981, ``Standard Criteria for Independence of
Class 1E Equipment and Circuits''], sections 6.6.2 and 6.6.5, and
IEEE standard 420-1982, [IEEE standard 420-1982, ``Design Standards
and Qualification of class 1E Control Boards, panels, and Racks Used
in Nuclear Power Generating Stations''], sections 4.3.1, 4.3.2, and
4.3.3 which have been endorsed by the NRC in Regulatory Guide 1.75
[Regulatory Guide 1.75, ``Physical Independence of Electrical
Systems'']. Therefore, these changes will not significantly increase
the probability or consequences of an accident previously evaluated.
2. Create the possibility of a new or different kind of accident
from any accident previously evaluated.
The FSAR changes reduce the minimum allowable separation between
redundant vital wires/devices of different channels from twelve
inches to six inches. The new minimum allowable separation will not
introduce any new or unanalyzed failure modes of equipment or
systems, and does not change the configuration of the plant. These
changes will not require any new or unusual operator actions, alter
the way any structure, system, or component functions and do not
alter the manner in which the plant is operated. Therefore, there
are no new or different types of failures of systems or equipment
important to safety which could cause a new or different type of
accident from any accident previously evaluated.
3. Involve a significant reduction in a margin of safety.
The FSAR changes reduce the minimum allowable separation between
redundant vital wires/devices of different channels from twelve
inches to six inches. The probability that a single wire/device
failure could cause the failure of redundant vital channels may be
increased. However, the new minimum allowed separation has been
found acceptable by IEEE standard 384-1981, sections 6.6.2 and
6.6.5, and IEEE standard 420-1982, sections 4.3.1, 4.3.2, and 4.3.3
which have been endorsed by the NRC in Regulatory Guide 1.75. The
new minimum allowed separation does not change any plant equipment
configuration, does not change the functionality of any equipment,
and does not change any operating setpoints. This change does not
alter the acceptance limits of the safety parameters of the accident
analyses stated in the FSAR. No new analysis assumptions are
required based on this change (e.g. common-cause failures).
Therefore, there is no impact on the margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: Learning Resources Center,
Three Rivers Community-Technical College, 574 New London Turnpike,
Norwich, 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 Project Director: William M. Dean.
Public Service Electric & Gas Company, Docket No. 50-311, Salem Nuclear
Generating Station, Unit No. 2, Salem County, New Jersey
Date of amendment request: October 12, 1998.
Description of amendment request: The proposed amendment would
allow a one-time extension of the Technical Specification (TS)
surveillance interval to the end of fuel cycle 10 for certain TS
surveillance requirements (SRs). Specifically, SR 4.3.2.1.3 requires
the instrumentation response time testing of each engineered safety
features actuation system function at least once per 18 months and SRs
4.8.2.3.2.f and 4.8.2.5.2.d require that the 125 volt DC and the 28
volt DC distribution system batteries, respectively, be capacity
service tested at least once per 18 months, during shutdown.
Additionally, SR 4.8.2.5.2.c.2 requires that the 125 volt DC battery
connections be verified clean, tight, and coated with anti-corrosion
material at least once per 18 months. Because of the length of the last
outage and delays in restart, the SRs will be overdue prior to reaching
the next refueling outage (2R10). The SRs are to be completed during
the 2R10 outage, prior to returning the unit to Mode 4 (hot shutdown)
upon outage completion.
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:
4.3.2.1.3 (Instrumentation, Engineered Safety Feature Actuation
System Instrumentation)
1. The proposed change does not involve a significant increase
in the probability or consequences of an accident previously
evaluated.
The deferral of the surveillance requirement does not involve
any physical changes to the plant nor does it change the way the
plant is operated. Thus, the proposal does not increase the
probability of an accident previously evaluated.
The SEC [safeguard equipment control] automatic self-test
feature, the monthly functional surveillance testing and the
positive surveillance testing history provide sufficient assurance
of the operability of the system. These features also provide
assurance that a degraded condition, if it did occur, would be
detected.
Thus, it is reasonable to conclude that this proposal represents
no significant increase in the consequences of an accident
previously analyzed.
2. The proposed change does not create the possibility of a new
or different kind of accident from any accident previously
evaluated.
Deferral of the surveillance requirement does not involve any
physical changes to the plant nor does it change the way the plant
is operated.
Thus, it can be concluded that deferring the surveillance
requirement to the refueling outage cannot create the possibility of
a different kind of accident from any accident previously evaluated.
[[Page 59595]]
3. The proposed change does not involve a significant reduction
in a margin of safety.
Deferral of the surveillance requirement does not involve any
physical changes to the plant nor does it change the way the plant
is operated. The self-test feature and the monthly functional
testing will provide reasonable assurance that the SECs will remain
operable during the few weeks of deferral to the refueling outage.
Also the ability to detect a degraded condition in the SEC will not
be affected during the deferral period.
Therefore, the plant's response to accident conditions during
the period of deferral will not be affected.
Thus, it can be reasonably concluded that this proposal to amend
the Salem Unit 2 Technical Specifications, on a one-time basis, to
defer surveillance requirement 4.3.2.1.3 does not involve a
significant reduction in any margin of safety.
4.8.2.3.2.f, (Electrical Power Systems, 125 Volt D.C.
Distribution), and 4.8.2.5.2.c.2 and 4.8.2.5.2.d (Electrical Power
Systems, 28 Volt D.C. Distribution)
1. The proposed change does not involve a significant increase
in the probability or consequences of an accident previously
evaluated.
The deferral of the battery service tests to the refueling
outage does not involve any physical changes to the power plant or
to the manner in which the power plant is operated. Therefore, the
probability of an accident previously evaluated is not increased.
Weekly and quarterly testing and performance monitoring by the
system manager along with the current condition of the batteries
(past test results demonstrating above 100% capacity) provide
assurance that battery condition and performance will not
deteriorate during the deferral period. Other positive industry
experience for similar batteries on 24 month cycles also support
this assurance. Therefore, the consequences of a loss of power
accident will not be increased due to the deferral of the
surveillance requirements.
2. The proposed change does not create the possibility of a new
or different kind of accident from any accident previously
evaluated.
The deferral of the battery service tests to the refueling
outage does not involve any physical changes to the power plant or
to the manner in which the power plant is operated. No new failure
mechanisms will be introduced by the surveillance deferral.
Therefore, the proposed change does not create the possibility of a
new or different kind of accident from any accident previously
evaluated.
3. The proposed change does not involve a significant reduction
in a margin of safety.
The deferral of the battery service tests to the refueling
outage does not involve any physical changes to the power plant or
to the manner in which the power plant is operated. Continuing
weekly and quarterly testing and performance monitoring along with
the current condition of the batteries provides assurance that
battery condition and performance will not deteriorate to an
unacceptable level during the deferral period and that any
degradation that may occur will be detected. Therefore, the plant's
response to accident conditions during the period of deferral will
not be affected.
Thus, it can be reasonably concluded that this proposal to amend
the Salem Unit 2 Technical Specifications, on a one-time basis, to
defer surveillance requirements 4.8.2.3.2.f and 4.8.2.5.2.d does not
involve a significant reduction in any margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
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 Project Director: Robert A. Capra.
Southern California Edison Company, et al., Docket Nos. 50-361 and 50-
362, San Onofre Nuclear Generating Station, Unit Nos. 2 and 3, San
Diego County, California
Date of amendment request: May 7, 1998.
Description of amendment request: This change would revise the
reference for obtaining the thyroid dose conversion factors used in the
definition of Dose Equivalent Iodine 131 (I-131) in Technical
Specification (TS) Section 1.1, ``Definitions'' for each plant.
Specifically, the reference to ``Table E-7 of Regulatory Guide 1.109,
Rev. 1, NRC 1977'' is to be replaced with a reference to the
International Commission on Radiological Protection Publication 30
(ICRP-30), Supplement to Part 1, Pages 192-212, Tables titled,
``Committed Dose Equivalent in Target Organs or Tissues per Intake of
Unit Activity.''
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. The proposed change does not involve a significant increase
in the probability or consequences of an accident previously
evaluated.
The proposed change, which utilizes International Committee on
Radiological Protection (ICRP)-30 methodology for determining dose
equivalent Iodine-131, and therefore for evaluating thyroid dose
consequences, does not involve any change to the method of operation
of any plant equipment, nor does it modify any plant equipment. In
addition, utilization of the ICRP-30 Dose Conversion Factors (DCFs)
will effectively reduce calculated thyroid dose consequences of
design basis accidents, thereby decreasing the calculated thyroid
dose consequences of previously evaluated accidents.
Therefore, the proposed changes will not increase the
probability or consequences of an accident previously evaluated.
2. The proposed change does not create the possibility of a new
or different kind of accident from any accident previously
evaluated.
The proposed change does not modify the configuration of the
units, involve any change to plant equipment or change the method of
plant operation. The utilization of the ICRP methodology for
determining DCFs uses more recent data which only affects
calculations for determining thyroid dose consequences.
Therefore, this change does not create the possibility of a new
or different kind of accident from any previously evaluated
accident.
3. The proposed change does not involve a significant reduction
in a margin of safety.
The change to utilize the ICRP methodology for determining DCFs
allows the use of more recent data which only affects calculations
for determining thyroid dose consequences. ICRP-30 is recognized in
Revision 1 of NUREG-1432, ``Standard Technical Specifications,
Combustion Engineering Plants,'' as an acceptable source document
for DCFs. The new methodology will result in more accurate DCFs that
will be used in the determination of dose consequences. Utilization
of the ICRP-30 DCFs will effectively reduce calculated thyroid dose
consequences of design basis accidents, thereby providing additional
design margin.
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 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, P.O. Box 800, Rosemead, California 91770.
NRC Project Director: William H. Bateman.
STP Nuclear Operating Company, Docket Nos. 50-498 and 50-499, South
Texas Project, Units 1 and 2, Matagorda County, Texas
Date of amendment request: September 30, 1998.
Description of amendment request: Revises Units 1 and 2 Technical
[[Page 59596]]
Specification (TS) Section 3/4.4.5, ``Steam Generator'' Surveillance
Requirements. The installation of the new Delta 94 steam generators at
the South Texas Project Units 1 and 2 necessitates changes to the steam
generator tube sample selection and inspection requirements; inservice
inspection frequencies; acceptance criteria; and inspection reporting
requirements.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. The proposed change does not involve a significant increase
in the probability or consequences of an accident previously
evaluated.
Eliminating provisions in the Technical Specifications for
applications of the voltage-based repair criteria, the F* alternate
repair criteria, and laser-welded sleeves for the Delta 94 steam
generators is an administrative adjustment, since the voltage-based
repair criteria, the F* alternate repair criteria, and laser-welded
sleeves are not applicable to the Delta 94 steam generators.
The Delta 94 steam generator tubing is designed and evaluated
consistent with the margins of safety specified in ASME Code Section
III.
The program for periodic inservice inspection of steam
generators monitors the integrity of the steam generator tubing to
ensure that there is sufficient time to take proper and timely
corrective action if tube degradation is present.
The ASME Section XI basis for the 40% through-wall plugging
limit is applicable to the Delta 94 steam generators just as it was
applicable to the Model E steam generators prior to the
implementation of voltage-based repair criteria, F* alternate repair
criteria, and laser-welded sleeves. In addition, analysis per
Regulatory Guide 1.121 (WCAP-15095/WCAP-15096) has confirmed the
applicability of the 40% plugging limit for the Delta 94 steam
generators.
The changes also clarify that inservice inspection is required
following steam generator replacement, and that inservice inspection
is not required during the steam generator replacement outage. This
is an administrative change in that it only provides clarification
of requirements written without steam generator replacement
considerations, and therefore, reduces the possibility for confusion
in the application of the subject technical specification
provisions. Therefore, 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.
Eliminating provisions in the Technical Specifications for
application of the voltage-based repair criteria, the F* alternate
repair criteria, and laser-welded sleeves to the Delta 94 steam
generators is an administrative adjustment, since the voltage-based
repair criteria, the F* alternate repair criteria, and laser-welded
sleeves are not applicable to the Delta 94 steam generators.
The changes also clarify that inservice inspection is required
following steam generator replacement, and that inservice inspection
is not required during the steam generator replacement outage. These
are administrative changes in that they only provide clarification
of requirements written without steam generator replacement
considerations, and therefore, reduce the possibility for confusion
in the application of the subject technical specification
provisions. Therefore, these proposed changes do 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 the margin of safety.
Eliminating provisions in the Technical Specifications for
applications of the voltage-based repair criteria, the F* alternate
repair criteria, and laser-welded sleeves for the Delta 94 steam
generators is an administrative adjustment, since the voltage-based
repair criteria, the F* alternate repair criteria, and laser-welded
sleeves are not applicable to the Delta 94 steam generators.
The Delta 94 steam generator tubing is designed and evaluated
consistent with the margins of safety specified in ASME Code Section
III. The program for periodic inservice inspection of steam
generators monitors the integrity of the steam generator tubing to
ensure that there is sufficient time to take proper and timely
corrective action if tube degradation is present.
The ASME Section XI basis for the 40% through-wall plugging
limit is applicable to the Delta 94 steam generators just as it was
applicable to the Model E steam generators prior to the
implementation of voltage-based repair criteria, F* alternate repair
criteria, and laser-welded sleeves. In addition, analysis per
Regulatory Guide 1.121 (WCAP-15095/WCAP-15096) has confirmed the
applicability of the 40% plugging limit for the Delta 94 steam
generators.
The changes also clarify that inservice inspection is required
following steam generator replacement, and that inservice inspection
is not required during the steam generator replacement outage. These
are administrative changes in that they only provide clarification
of requirements written without steam generator replacement
considerations, and therefore, reduce the possibility for confusion
in the application of the subject technical specification
provisions. Therefore, these 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 standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
request for amendments involves no significant hazards consideration.
Local Public Document Room location: Wharton County Junior College,
J. M. Hodges Learning Center, 911 Boling Highway, Wharton, TX 77488.
Attorney for licensee: Jack R. Newman, Esq., Morgan, Lewis &
Bockius, 1800 M Street, N.W., Washington, DC 20036-5869.
NRC Project Director: John N. Hannon.
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: September 20, 1996 (TS 96-09).
Brief description of amendments: The amendments would change the
Sequoyah Nuclear Plant (SQN) Technical Specifications by clarifying the
types of work shifts that are acceptable when considering the
requirements to ensure heavy use of overtime is not used routinely by
unit staff. The current ``8-hour day'' criteria in Section 6.2.2.g will
be expanded to include 10-hour and 12-hour allowances. In addition, the
``40-hour week'' criteria will be changed to a ``nominal 40-hour week''
to provide the necessary flexibility associated with the use of the
proposed shift durations.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the Tennessee Valley
Authority (TVA), the licensee, has provided its analysis of the issue
of no significant hazards consideration, which is presented below:
TVA has evaluated the proposed technical specification (TS)
change and has determined that it does not represent a significant
hazards consideration based on criteria established in 10 CFR
50.92(c). Operation of Sequoyah Nuclear Plant (SQN) in accordance
with the proposed amendment will not:
1. Involve a significant increase in the probability or
consequences of an accident previously evaluated.
This change affects the requirements that ensure unit staff
personnel do not routinely incur heavy use of overtime. These
requirements are not changed by the proposed revision, but are
clarified to accommodate the various shift durations used at SQN.
The overtime usage by unit staff is not considered to be the
initiator for any postulated accident; therefore, the clarification
of associated requirements will not increase the probability of an
accident. Limiting the use of overtime by staff personnel enhances
the operation and maintenance of critical plant equipment that are
necessary to mitigate accidents. The proposed revision clarifies
these provisions, but does not reduce their adequacy. Therefore, the
proposed revision will not increase the consequences of an accident
previously evaluated.
[[Page 59597]]
2. Create the possibility of a new or different kind of accident
from any previously analyzed.
This change only affects the clarification of shift durations
use by unit staff and is not associated with the initiators of
accidents. Therefore, the possibility of a new or different kind of
accident from any previously analyzed is not created by the proposed
clarifications.
3. Involve a significant reduction in a margin of safety.
The proposed changes do not affect plant equipment setpoints or
operating policies at SQN. The overtime provisions that ensure the
unit staff are capable to operate and maintain the plant in an
acceptable manner to provide safe operation and mitigation of
accidents is maintained by this change. Therefore, the margin of
safety is not reduced by the proposed changes.
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
amendment 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 Project Director: Frederick J. Hebdon.
TU Electric Company, Docket Nos. 50-445 and 50-446, Comanche Peak Steam
Electric Station, Units 1 and 2, Somervell County, Texas
Date of amendment request: October 2, 1998.
Brief description of amendments: The proposed change would revise
Technical Specification (TS) 4.0.6, ``Steam Generator Surveillance
Requirements,'' to add definitions required for the F* alternate steam
generator tube plugging criterion and identify the portion of the tube
subject to the criteria.
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?
The supporting technical evaluation of the subject criterion
[Westinghouse WCAP-15004, listed as Reference 1 (Proprietary)],
demonstrates that the presence of the tubesheet enhances the tube
integrity in the region of the hardroll by precluding tube
deformation beyond its initial expanded outside diameter. The result
of hardrolling of the tube into the tubesheet is an interference fit
between the tube and the tubesheet. A tube rupture cannot occur
because the contact between the tube and tubesheet does not permit
sufficient movement of tube material. In a similar manner, the
tubesheet does not permit sufficient movement of tube material to
permit buckling collapse of the tube during postulated LOCA
loadings. Analysis and testing have been done to determine the
resistive strength of roll expanded tubes within the tubesheet. This
evaluation provides the basis for the acceptance criterion for tube
degradation subject to the F* criterion. The F* distance of roll
expansion is sufficient to preclude tube axial translation or
pullout from tube degradation located below the F* distance,
regardless of the extent of the tube degradation. The necessary
engagement length applicable to the Comanche Peak Unit 1 steam
generators is determined to be 1.13 inches, plus an allowance for
eddy current measurement uncertainty, based on preload analyses.
Verification that this value is significantly conservative was
demonstrated by both pullout and hydraulic proof testing.
Application of the F* criterion provides a level of protection for
tube degradation in the tubesheet region commensurate with that
afforded by RG 1.121. Leakage testing of roll expanded tubes
indicates that for roll lengths approximately equal to the F*
distance, any postulated faulted condition primary to secondary
leakage from F* tubes would be insignificant. No leakage occurred
from any of the hydraulic proof test specimens for pressures up to
and exceeding faulted condition events. The existing Technical
Specification leakage rate requirements and accident analysis
assumptions remain unchanged.
Based on the above, it is concluded that the proposed F*
criterion does not adversely impact any other previously evaluated
design basis accidents and operation of Comanche Peak Unit 1 in
accordance with the proposed license amendment does not involve a
significant increase in the probability or consequences of an
accident previously evaluated.
2. Do the proposed changes create the possibility of a new or
different kind of accident from any accident previously evaluated?
Implementation of the proposed F* criterion does not introduce
any significant changes to the plant design basis. Use of the F*
criterion does not provide a mechanism to result in an accident
initiated outside of the region of the tubesheet expansion. Even if
it is postulated that a circumferential separation of a F* tube were
to occur below the F* distance, tube structural and leakage
integrity will be maintained consistent with the assumptions of the
design basis accidents during all plant conditions. Verification of
the F* distance of non-degraded tube roll expansion prevents a
postulated separated tube from lifting out of the tubesheet during
all plant conditions. The F* criterion does not create a possibility
for simultaneous failures of multiple tubes. Any other hypothetical
accident as a result of any degradation in the expanded portion of
the tube would be bounded by the existing steam generator tube
rupture accident analysis.
Therefore, it is concluded that the proposed license amendment
does not create the possibility of a new or different kind of
accident from any accident previously evaluated.
(3) Do the proposed changes involve a significant reduction in a
margin of safety?
The use of the F* criterion has been demonstrated to maintain
the integrity of the tube bundle commensurate with the requirements
of RG 1.121 (intended for indications in the free span of tubes) and
the primary to secondary pressure boundary under normal and
postulated accident conditions. Acceptable tube degradation for the
F* criterion is any degradation indication in the tubesheet region,
more than the F* distance below the bottom of the transition between
the roll expansion and the unexpanded tube or the bottom of the
tubesheet (whichever is lower). The safety factors used in the
verification of the strength of the degraded tube are consistent
with the safety factors in the ASME Boiler and Pressure Vessel Code
used in steam generator design. The F* distance has been verified by
pullout and hydraulic proof testing of tubes in tubesheet simulating
collars to be greater than the length of roll expansion required to
preclude both tube pullout and significant leakage during normal and
postulated accident conditions. Resistance to tube pullout is based
upon the primary to secondary pressure differential as it acts on
the surface area of the tube, which includes the tube wall cross-
section, in addition to the inner diameter based area of the tube.
The leak testing acceptance criteria are based on the primary to
secondary leakage limit in the Technical Specifications and the
leakage assumptions used in the FSAR accident analyses.
Implementation of the proposed F* criterion will decrease the
number of tubes which must be taken out of service with tube plugs.
Plugged tubes reduce the RCS flow margin, thus implementation of the
F* alternate plugging criterion will maintain the margin of flow
that would otherwise be reduced in the event of increased plugging.
Therefore, it is concluded that the proposed change does not
result in a significant reduction in margin to plant safety as
defined in the Final Safety Analysis Report or the bases of the
Technical Specifications.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: University of Texas at
Arlington Library, Government Publications/Maps, 702 College, P.O. Box
19497, Arlington, TX 76019.
Attorney for licensee: George L. Edgar, Esq., Morgan, Lewis and
Bockius, 1800 M Street, N.W., Washington, DC 20036.
[[Page 59598]]
NRC Project Director: John N. Hannon.
Virginia Electric and Power Company, Docket Nos. 50-280 and 50-281,
Surry Power Station, Unit Nos. 1 and 2, Surry County, Virginia
Date of amendment request: September 12, 1996, as supplemented
April 24, 1997, and September 24, 1998.
Description of amendment request: The staff had previously
published a Notice of Consideration of Amendments and Proposed No
Significant Hazards Consideration Determination for the licensee's
September 12, 1996, application in the Federal Register on April 23,
1997 (62 FR 19835). As a result of the staff's requests for additional
information, the licensee supplemented its original proposal to
relocate the fire protection requirements from the Technical
Specifications (TS) to the Updated Final Safety Analysis Report (UFSAR)
by letters dated April 24, 1997, and September 24, 1998. The April 24,
1997, letter corrected two minor administrative oversights and does not
affect the No Significant Hazards Consideration Determination (NSHCD).
However, the September 24, 1998, letter revised the original
application to require the Station Nuclear Safety and Operating
Committee to submit recommended changes to the offsite review group. In
addition, a requirement was added for the establishment,
implementation, and maintenance of the Fire Protection Program and
implementing procedures. The NSHCD for these changes, as provided in
the September 24, 1998, letter, is addressed below.
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:
Since these two changes only deal with administrative
requirements, neither of these two specific changes would result in
a significant hazards consideration. Therefore, the operation of
Surry Power Station with the proposed amendment will not:
1. Involve a significant increase in the probability or
consequences of an accident previously evaluated.
The probability of an accident is not increased as a result of
this Technical Specifications change request. This is an
administrative change and merely incorporates two additional
requirements for ensuring that the Fire Protection Program and
implementing procedures are appropriately established, implemented
and maintained, and that changes to the Program and implementing
procedures receive the appropriate offsite review. The consequences
of an accident previously evaluated are not increased since the
station will not be operated differently, and no physical
modifications are being made to plant systems or components.
2. Create the possibility of a new or different kind of accident
from any accident previously evaluated.
A new or different type of accident is not being created since
this TS change request is administrative. As noted above, the
station will not be operated differently, and no physical
modifications are being made to plant systems or components.
Administrative revisions regarding the establishment, implementation
and maintenance of a TS requirement for a Fire Protection Program
and implementing procedures and the imposition of an offsite review
for changes thereto [do] not create a new or different type of
accident from any accident previously evaluated.
3. Involve a significant reduction in a margin of safety.
The margin of safety as defined in the Technical Specifications
is not reduced since system/component performance as assumed in the
existing safety analyses is not being affected by the proposed TS
change. The TS change is administrative in nature and, as such, has
no effect on station operation. The Fire Protection Program is being
retained and maintained in the UFSAR and station procedures.
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: Swem Library, College of
William and Mary, Williamsburg, Virginia 23185.
Attorney for licensee: Michael W. Maupin, Esq., Hunton and
Williams, Riverfront Plaza, East Tower, 951 E. Byrd Street, Richmond,
Virginia.
NRC Project Director: Herbert N. Berkow.
Previously Published Notices of Consideration of Issuance of Amendments
to Facility Operating Licenses, Proposed no Significant Hazards
Consideration Determination and Opportunity for a Hearing
The following notices were previously published as separate
individual notices. The notice content was the same as above. They were
published as individual notices either because time did not allow the
Commission to wait for this biweekly notice or because the action
involved exigent circumstances. They are repeated here because the
biweekly notice lists all amendments issued or proposed to be issued
involving no significant hazards consideration.
For details, see the individual notice in the Federal Register on
the day and page cited. This notice does not extend the notice period
of the original notice.
Illinois Power Company, Docket, No. 50-461, Clinton Power Station,
DeWitt County, Illinois
Date of application for amendment: October 5, 1998.
Brief description of amendment request: The proposed amendment
requests deferral of the next scheduled local leak rate test for valve
1MC-042 until the seventh refueling outage.
Date of publication of individual notice in Federal Register:
October 23, 1998 (63 FR 56949).
Expiration date of individual notice: November 23, 1998.
Local Public Document Room location: Vespasian Warner Public
Library, 310 N. Quincy Street, Clinton, IL 61727.
Northeast Nuclear Energy Company, Docket No. 50-423, Millstone Nuclear
Power Station, Unit 3, New London County, Connecticut
Date of amendment request: August 6, 1998, as supplemented by
letters dated September 3 and 21, 1998.
Description of amendment request: The proposed amendment allows a
one-time extension to the steam generator tube inspection surveillance
interval until the next refueling outage or July 1, 1999, whichever
date is earlier.
Date of publication of individual notice in Federal Register:
August 17, 1998 (63 FR 43964).
Expiration date of individual notice: September 16, 1998.
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.
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.
[[Page 59599]]
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: August 27, 1998, as supplemented
by letter dated October 1, 1998.
Brief description of amendment: This amendment revises Technical
Specifications (TS) 3.0.4 and 4.0.4 in accordance with the guidance
provided in Generic Letter 87-09. The revision to TS 3.0.4 removes the
need to explicitly reference its applicability for certain TS. As a
result, several other TS were also amended by deleting references to TS
3.0.4.
Date of issuance: October 20, 1998.
Effective date: October 20, 1998.
Amendment No: 84.
Facility Operating License No. NPF-63. Amendment revises the
Technical Specifications.
Date of initial notice in Federal Register: September 8, 1998 (63
FR 47529).
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated October 20, 1998.
No significant hazards consideration comments received: No.
Local Public Document Room location: Cameron Village Regional
Library, 1930 Clark Avenue, Raleigh, North Carolina 27605.
Commonwealth Edison Company, Docket Nos. 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 application for amendments: August 23, 1996.
Brief description of amendments: The amendments revise the
Technical Specifications related to the Non-Accessible Area Exhaust
Filter Plenum Ventilation System to reflect the design lineup and to
make provisions for the performance of maintenance and testing.
Date of issuance: October 15, 1998.
Effective date: Immediately, to be implemented within 30 days.
Amendment Nos.: 105; 105 & 97; 97.
Facility Operating License Nos. NPF-37, NPF-66, NPF-72 and NPF-77:
The amendments revised the Technical Specifications.
Date of initial notice in Federal Register: March 12, 1997 (62 FR
11488).
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated October 15, 1998.
No significant hazards consideration comments received: No.
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.
Entergy Operations, Inc., Docket No. 50-382, Waterford Steam Electric
Station, Unit 3, St. Charles Parish, Louisiana
Date of amendment request: October 16, 1996, as supplemented by
letters dated December 22, 1997, and May 27, 1998.
Brief description of amendment: The amendment changes the Appendix
A Technical Specifications by relocating certain administrative
controls to Quality Assurance Program Manual as described in
Administrative Letter 95-06, ``Relocation of Technical Administrative
Controls related to Quality Assurance;'' changing shift coverage from
8-hour day, 40-hour weeks to an option of 8 or 12 hour days and nominal
40-hour weeks; and making editorial changes to the titles of certain
organizational positions.
Date of issuance: October 19, 1998.
Effective date: October 19, 1998, to be implemented within 60 days.
Amendment No.: 146.
Facility Operating License No. NPF-38: Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: April 9, 1997 (62 FR
17233).
The December 22, 1997, and May 27, 1998 letters, provided
clarifying information that did not change the initial proposed no
significant hazards consideration determination.
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated October 19, 1998.
No significant hazards consideration comments received: No.
Local Public Document Room location: University of New Orleans
Library, Louisiana Collection, Lakefront, New Orleans, LA 70122.
Florida Power and Light Company, et al., Docket Nos. 50-335 and 50-389,
St. Lucie Plant, Unit Nos. 1 and 2, St. Lucie County, Florida
Date of application for amendment: June 21, 1995.
Brief description of amendment: The amendments revise the Technical
Specification action statements and certain surveillances of TS 3/
4.5.1, Safety Injection Tanks (SITs). These revisions include a two-
tiered extension of the action completion/allowed outage time for the
SITs. The revisions are also consistent with the guidance provided in
Generic Letter 93-05, ``Line-Item Technical Specifications Improvements
to Reduce surveillance requirements for Testing During Power
Operation.''
Date of Issuance: October 16, 1998.
Effective Date: To be implemented within 30 days from date of
receipt.
Amendment Nos.: 157 and 96.
Facility Operating License No. NPF-16: Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: September 27, 1995 (60
FR 49936).
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated October 16, 1998.
No significant hazards consideration comments received: No.
Local Public Document Room location: Indian River Community College
Library, 3209 Virginia Avenue, Fort Pierce, Florida 34981-5596.
Florida Power and Light Company, et al., Docket Nos. 50-335 and 50-389,
St. Lucie Plant, Unit Nos. 1 and 2, St. Lucie County, Florida
Date of amendment request: October 31, 1996, supplemented October
31, 1997, May 27, 1998, and September 25, 1998.
[[Page 59600]]
Description of amendment request: The amendments revise the
administrative control specifications to reduce the administrative
burden carried by the Facility Review Group and the Plant General
Manager by making more efficient use of site personnel possessing the
requisite experience and qualifications in the review and approval
process for plant procedures.
Date of Issuance: October 16, 1998.
Effective Date: October 16, 1998.
Amendment Nos.: 158 and 97.
Facility Operating License Nos. DPR-67 and NPF-16: Amendments
revised the Technical Specifications.
Date of Initial Notice in Federal Register: December 18, 1996 (61
FR 66707) The October 31, 1997, May 27, 1998, and September 25, 1998,
submittals provided clarifying information that did not change the
original no significant hazards consideration determination.
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated October 16, 1998.
No significant hazards consideration comments received: No.
Local Public Document Room location: Indian River Community College
Library, 3209 Virginia Avenue, Fort Pierce, Florida 34981-5596.
GPU Nuclear, Inc. et al., Docket No. 50-219, Oyster Creek Nuclear
Generating Station, Ocean County, New Jersey
Date of application for amendment: August 21, 1998.
Brief description of amendment: The amendment removes the
requirement for the Automatic Depressurization System function of the
Electromatic Relief Valves to be operable during Reactor Vessel
Pressure Testing. Additionally, it clarifies Note h of Technical
Specification Table 3.1.1.
Date of Issuance: October 14, 1998.
Effective date: October 14, 1998, to be implemented within 30 days.
Amendment No.: 199.
Facility Operating License No. DPR-16: Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: September 10, 1998 (63
FR 48527).
The Commission's related evaluation of this amendment is contained
in a Safety Evaluation dated October 14, 1998.
No significant hazards consideration comments received: No.
Local Public Document Room location: Ocean County Library,
Reference Department, 101 Washington Street, Toms River, NJ 08753.
GPU Nuclear, Inc. et al., Docket No. 50-219, Oyster Creek Nuclear
Generating Station, Ocean County, New Jersey
Date of application for amendment: May 28,1998.
Brief description of amendment: The amendment revises Technical
Specification 4.5.A.1 such that the first Type A test required by the
primary containment leakage rate testing program be performed during
refueling outage 18 rather than refueling outage 17.
Date of Issuance: October 15, 1998.
Effective date: October 15, 1998, to be implemented within 30 days.
Amendment No.: 200.
Facility Operating License No. DPR-16. Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: July 15, 1998 (63 FR
38201).
The Commission's related evaluation of this amendment is contained
in a Safety Evaluation dated October 15, 1998.
No significant hazards consideration comments received: No.
Local Public Document Room location: Ocean County Library,
Reference Department, 101 Washington Street, Toms River, NJ 08753.
Illinois Power Company, Docket No. 50-461, Clinton Power Station, Unit
1, DeWitt County, Illinois.
Date of application for amendment: May 4, 1998, as supplemented
September 23, 1998.
Brief description of amendment: The amendment incorporates
Technical Specification requirements for the protection systems for the
new static VAR compensators.
Date of issuance: October 9, 1998.
Effective date: October 9, 1998.
Amendment No.: 117.
Facility Operating License No. NPF-62: The amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: June 3, 1998 (63 FR
30264).
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated October 9, 1998.
No significant hazards consideration comments received: No.
Local Public Document Room location: The Vespasian Warner Public
Library, 120 West Johnson Street, Clinton, IL 61727.
Northeast Nuclear Energy Company, et al., Docket No. 50-423, Millstone
Nuclear Power Station, Unit No. 3, New London County, Connecticut.
Date of application for amendment: May 9, 1997, as supplemented
August 4, 1998.
Brief description of amendment: The amendment revises the shutdown
margin requirements and adds Technical Specification 3/4.3.5 to provide
the limiting condition for operation and surveillance requirements for
the shutdown margin monitors. The amendment also makes administrative
changes and revises the associated Bases section.
Date of issuance: October 21, 1998.
Effective date: As of the date of issuance, to be implemented
within 60 days from the date of issuance.
Amendment No.: 164.
Facility Operating License No. NPF-49: Amendment revised the
Facility Operating License and the Technical Specifications.
Date of initial notice in Federal Register: June 18, 1997 (62 FR
33129).
The August 4, 1998, letter provided clarifying information that did
not change the scope of the May 9, 1997, application, and the initial
proposed no significant hazards consideration determination.
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated October 21, 1998.
No significant hazards consideration comments received: No.
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.
Omaha Public Power District, Docket No. 50-285, Fort Calhoun Station,
Unit No. 1, Washington County, Nebraska
Date of amendment request: July 11, 1995.
Brief description of amendment: The amendment revises Technical
Specifications (TS) 2.3(2)f and 2.3(2)g to increase allowed outage
times for the safety injection tanks (SIT).
Date of issuance: October 19, 1998.
Effective date: October 19, 1998.
Amendment No.: 186.
Facility Operating License No. DPR-40. Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: August 2, 1995 (60 FR
39447). The Commission's related evaluation of the amendment is
contained in a Safety Evaluation dated October 19, 1998.
No significant hazards consideration comments received: No.
Local Public Document Room location: W. Dale Clark Library, 215
South 15th Street, Omaha, Nebraska 68102.
[[Page 59601]]
Omaha Public Power District, Docket No. 50-285, Fort Calhoun Station,
Unit No. 1, Washington County, Nebraska
Date of amendment request: October 3, 1997, as supplemented by
letter dated May 18, 1998.
Brief description of amendment: The amendment revises Technical
Specifications (TS) 3.9 to clarify required flow paths for testing the
auxiliary feedwater system (AFW) and to delete specific AFW pump
discharge pressure.
Date of issuance: October 19, 1998.
Effective date: October 19, 1998, to be implemented 30 days from
the date of issuance.
Amendment No.: 187.
Facility Operating License No. DPR-40: Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: December 3, 1997 (62 FR
63982).
The May 18, 1998, supplemental letter provided additional
clarifying information that did not change the staff's original no
significant hazards consideration determination. The Commission's
related evaluation of the amendment is contained in a Safety Evaluation
dated October 19, 1998.
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, Public Service Electric and Gas Company, Delmarva
Power and Light Company, and Atlantic City Electric Company, Docket No.
50-277, Peach Bottom Atomic Power Station, Unit No. 2, York County,
Pennsylvania
Date of application for amendment: July 10, 1998, as supplemented
by two letters dated September 11, 1998. The supplemental letters
provided clarifying information but did not change the initial no
significant hazards consideration determination.
Brief description of amendment: This amendment revises the
Technical Specifications for safety limit Minimum Critical Power Ratio
from its current value of 1.11 to 1.10 for two recirculation loop
operation, and from 1.13 to 1.12 for single recirculation loop
operation.
Date of issuance: October 26, 1998.
Effective date: As of date of issuance, to be implemented prior to
startup for Cycle 13 operations, scheduled for October 1998.
Amendment No.: 226.
Facility Operating License No. DPR-44: Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: September 9, 1998 (63
FR 48261).
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated October 26, 1998.
No significant hazards consideration comments received: No.
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.
Pennsylvania Power and Light Company, Docket Nos. 50-387 and 50-388
Susquehanna Steam Electric Station, Units 1 and 2, Luzerne County,
Pennsylvania
Date of application for amendments: March 16, 1998, as supplemented
by letters dated May 22, August 10, and September 17, 1998, and also by
letter dated February 9, 1998.
Brief description of amendments: The amendment authorized changes
to the Final Safety Analysis Report to incorporate the increases in the
main steam line radiation monitor setpoint and allowable values and the
change to the design basis of the offgas system to a detonation
resistant design.
Date of issuance: October 13, 1998.
Effective date: October 13, 1998.
Amendment Nos.: 179 and 152.
Facility Operating License Nos. NPF-14 and NPF-22: The amendments
revised the Final Safety Analysis Report.
Date of initial notice in Federal Register: May 20, 1998 (63 FR
27764).
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated October 13, 1998.
No significant hazards consideration comments received: No.
Local Public Document Room location: Osterhout Free Library,
Reference Department, 71 South Franklin Street, Wilkes-Barre, PA 18701.
Pennsylvania Power and Light Company, Docket Nos. 50-387 and 50-388,
Susquehanna Steam Electric Station, Units 1 and 2, Luzerne County,
Pennsylvania
Date of application for amendments: April 23, 1998.
Brief description of amendments: These amendments change the name
``Pennsylvania Power & Light Company'' to ``PP&L, Inc.'' in the
operating licenses and appendices to reflect the licensee's corporate
name change.
Date of issuance: October 19, 1998.
Effective date: Both units, as of the date of issuance to be
implemented within 30 days.
Amendment Nos.: 180 and 153.
Facility Operating License Nos. NPF-14 and NPF-22. The amendments
revised the operating licenses and Appendix B to each licensee and
Attachment 1 to the Unit 1 license.
Date of initial notice in Federal Register: July 1, 1998 (63 FR
35993).
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated October 19, 1998.
No significant hazards consideration comments received: No.
Local Public Document Room location: Osterhout Free Library,
Reference Department, 71 South Franklin Street, Wilkes-Barre, PA 18701.
Philadelphia Electric Company, Docket Nos. 50-352 and 50-353, Limerick
Generating Station, Units 1 and 2, Montgomery County, Pennsylvania
Date of application for amendments: February 25, 1997, as
supplemented September 8 and November 18, 1997 and January 8 and July
2, 1998. The supplemental letters provided clarifying information and
did not change the initial proposed no significant hazards
consideration determination.
Brief description of amendments: These amendments revise the
Facility Operating Licenses, Technical Specifications, and
Environmental Protection Plans to reflect a corporate name change,
remove obsolete information, and correct typographical errors.
Date of issuance: October 23, 1998.
Effective date: Both units, as of date of issuance and shall be
implemented within 30 days.
Amendment Nos.: 131 and 92.
Facility Operating License Nos. NPF-39 and NPF-85. The amendments
revised the Technical Specifications and Licenses.
Date of initial notice in Federal Register: June 4, 1997 (62 FR
30642).
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated October 23, 1998.
No significant hazards consideration comments received: No.
Local Public Document Room location: Pottstown Public Library, 500
High Street, Pottstown, PA 19464.
[[Page 59602]]
Power Authority of The State of New York, Docket No. 50-286, Indian
Point Nuclear Generating Unit No. 3, Westchester County, New York
Date of application for amendment: June 25, 1997, as supplemented
August 3, 1998.
Brief description of amendment: The amendment allows the use of
zirconium or stainless steel filler rods in fuel assemblies to replace
failed or damaged fuel rods.
Date of issuance: October 8, 1998.
Effective date: As of the date of issuance to be implemented within
30 days.
Amendment No.: 183.
Facility Operating License No. DPR-64: Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: June 17, 1998 (63 FR
33107).
The August 3, 1998, submittal fell within the scope of, and did not
change, the initial proposed finding of no significant hazards
consideration.
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated October 8, 1998.
No significant hazards consideration comments received: No.
Local Public Document Room location: White Plains Public Library,
100 Martine Avenue, White Plains, New York 10610.
Southern Nuclear Operating Company, Inc., Georgia Power Company,
Oglethorpe Power Corporation, Municipal Electric Authority of Georgia,
City of Dalton, Georgia, Docket Nos. 50-321 and 50-366, Edwin I. Hatch
Nuclear Plant, Units 1 and 2, Appling County, Georgia
Date of application for amendments: August 8, 1997, as supplemented
by letters dated March 9, May 6, July 6, July 31, September 4, and
September 11, 1998.
Brief description of amendments: The amendments revise the
Technical Specifications to accommodate an increase in the maximum
licensed thermal power level from 2558 megawatts thermal (MWt) to 2763
MWt.
Date of issuance: October 22, 1998.
Effective date: As of the date of issuance to be implemented on
Unit 1 prior to startup from the next refueling outage and on Unit 2
prior to startup from the current refueling outage.
Amendment Nos.: Unit 1-214; Unit 2-155.
Facility Operating License Nos. DPR-57 and NPF-5: The amendments
revised the Technical Specifications and Operating Licenses.
Public comments requested as to proposed no significant hazards
consideration: Yes. (63 FR 53730 dated October 6, 1998.) The notice
provided an opportunity to submit comments on the Commission's proposed
no significant hazards consideration determination. No comments have
been received. The notice also provided for an opportunity to request a
hearing by November 5, 1998, but indicated that if the Commission makes
a final no significant hazards consideration determination, any such
hearing would take place after issuance of the amendments.
The Commission's related evaluation of the amendments, finding of
exigent circumstances, and a final no significant hazards consideration
determination are contained in a Safety Evaluation dated October 22,
1998.
Local Public Document Room location: Appling County Public Library,
301 City Hall Drive, Baxley, Georgia.
Southern Nuclear Operating Company, Inc., Docket Nos. 50-348 and 50-
364, Joseph M. Farley Nuclear Plant, Units 1 and 2, Houston County,
Alabama
Date of amendments request: May 27, 1997.
Brief Description of amendments: The amendments revise the
Technical Specifications (TSs) to change the Applicable Modes for
Source Range (SR) Nuclear Instrumentation (NI) (TS \3/4\.3.1, ``Reactor
Trip System Instrumentation''), provide allowances for an exception to
the requirements for the state of the power supplies for residual heat
removal discharge to charging pump suction valves following Mode
changes (TS \3/4\.5.2, ``ECCS Subsystems--T<INF>avg</INF>>350 deg.F''
and \3/4\.5.3, ``ECCS Subsystems--T<INF>avg</INF><350 deg.F''), and
delete cycle-specific guidance concerning manual engineered safety
feature functional input checks.
Date of issuance: October 15, 1998.
Effective date: As of the date of issuance to be implemented within
30 days from the date of issuance.
Amendment Nos.: Unit 1-138; Unit 2-130.
Facility Operating License Nos. NPF-2 and NPF-8: Amendments revise
the Technical Specifications.
Date of initial notice in Federal Register: June 18, 1997 (62 FR
33134).
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated October 15, 1998.
No significant hazards consideration comments received: No.
Local Public Document Room location: Houston-Love Memorial Library,
212 W. Burdeshaw Street, Post Office Box 1369, Dothan, Alabama.
Tennessee Valley Authority, Docket No. 50-390 Watts Bar Nuclear Plant,
Unit 1, Rhea County, Tennessee
Date of application for amendment: June 5, 1997, as supplemented
April 21 and August 12, 1998.
Brief description of amendment: The requested changes would revise
the Technical Specifications (TS) to allow testing of diesel
generators, pursuant to Surveillance Requirement (SR) 3.8.1.14, during
operational modes 1 or 2. The requested changes would also revise the
TS to allow testing of the diesel generator batteries and associated
battery chargers, pursuant to SRs 3.8.4.12, 3.8.4.13 and 3.8.4.14
during operational modes 1, 2, 3 or 4.
Date of issuance: October 19, 1998.
Effective date: October 19, 1998.
Amendment No.: 12.
Facility Operating License No. NPF-90: Amendment revises the TS.
Date of initial notice in Federal Register: July 29, 1998 (63 FR
40561).
The supplemental letter dated August 12, 1998, contained clarifying
information and did not change the original no significant hazards
consideration determination.
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated October 19, 1998.
No significant hazards consideration comments received: None.
Local Public Document Room location: Chattanooga-Hamilton County
Library, 1001 Broad Street, Chattanooga, TN 37402.
Wisconsin Public Service Corporation, Docket No. 50-305, Kewaunee
Nuclear Power Plant, Kewaunee County, Wisconsin
Date of application for amendment: April 8, 1998, as revised by
letter dated August 27, 1998.
Brief description of amendment: The amendment reduces the allowable
reactor coolant system specific activity from 1.0 microcurie/gram to
0.20 microcurie/gram dose equivalent I-131, a means described by
Generic Letter 95-05 to support the reduction of reactor coolant system
specific activity limits.
Date of issuance: October 27, 1998.
Effective date: October 27, 1998.
Amendment No.: 140.
Facility Operating License No. DPR-43: Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: September 14, 1998 (63
FR 49137).
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated October 27, 1998.
[[Page 59603]]
No significant hazards consideration comments received: No.
Local Public Document Room location: University of Wisconsin,
Cofrin Library, 2420 Nicolet Drive, Green Bay, WI 54311-7001.
Notice of Issuance of Amendment to Facility Operating License and
Final No Significant Hazards Consideration Determination
During the period since publication of the last biweekly notice,
individual notices of issuance of amendments have been issued for the
facilities as listed below. These notices were previously published as
separate individual notices. They are repeated here because this
biweekly notice lists all amendments that have been issued for which
the Commission has made a final determination that an amendment
involves no significant hazards consideration.
In this case, a prior Notice of Consideration of Issuance of
Amendment, Proposed No Significant Hazards Consideration Determination,
and Opportunity for a Hearing was issued, a hearing was requested, and
the amendment was issued before any hearing because the Commission made
a final determination that the amendment involves no significant
hazards consideration.
Details are contained in the individual notice as cited.
Notice of Issuance of Amendments to Facility Operating Licenses and
Final Determination of No Significant Hazards Consideration and
Opportunity for a Hearing (Exigent Public Announcement or Emergency
Circumstances)
During the period since publication of the last biweekly notice,
the Commission has issued the following amendments. The Commission has
determined for each of these amendments that the application for the
amendment complies with the standards and requirements of the Atomic
Energy Act of 1954, as amended (the Act), and the Commission's rules
and regulations. The Commission has made appropriate findings as
required by the Act and the Commission's rules and regulations in 10
CFR Chapter I, which are set forth in the license amendment.
Because of exigent or emergency circumstances associated with the
date the amendment was needed, there was not time for the Commission to
publish, for public comment before issuance, its usual 30-day Notice of
Consideration of Issuance of Amendment, Proposed No Significant Hazards
Consideration Determination, and Opportunity for a Hearing.
For exigent circumstances, the Commission has either issued a
Federal Register notice providing opportunity for public comment or has
used local media to provide notice to the public in the area
surrounding a licensee's facility of the licensee's application and of
the Commission's proposed determination of no significant hazards
consideration. The Commission has provided a reasonable opportunity for
the public to comment, using its best efforts to make available to the
public means of communication for the public to respond quickly, and in
the case of telephone comments, the comments have been recorded or
transcribed as appropriate and the licensee has been informed of the
public comments.
In circumstances where failure to act in a timely way would have
resulted, for example, in derating or shutdown of a nuclear power plant
or in prevention of either resumption of operation or of increase in
power output up to the plant's licensed power level, the Commission may
not have had an opportunity to provide for public comment on its no
significant hazards consideration determination. In such case, the
license amendment has been issued without opportunity for comment. If
there has been some time for public comment but less than 30 days, the
Commission may provide an opportunity for public comment. If comments
have been requested, it is so stated. In either event, the State has
been consulted by telephone whenever possible.
Under its regulations, the Commission may issue and make an
amendment immediately effective, notwithstanding the pendency before it
of a request for a hearing from any person, in advance of the holding
and completion of any required hearing, where it has determined that no
significant hazards consideration is involved.
The Commission has applied the standards of 10 CFR 50.92 and has
made a final determination that the amendment involves no significant
hazards consideration. The basis for this determination is contained in
the documents related to this action. Accordingly, the amendments have
been issued and made effective as indicated.
Unless otherwise indicated, the Commission has determined that
these amendments satisfy the criteria for categorical exclusion in
accordance with 10 CFR 51.22. Therefore, pursuant to 10 CFR 51.22(b),
no environmental impact statement or environmental assessment need be
prepared for these amendments. If the Commission has prepared an
environmental assessment under the special circumstances provision in
10 CFR 51.12(b) and has made a determination based on that assessment,
it is so indicated.
For further details with respect to the action see (1) the
application for amendment, (2) the amendment to Facility Operating
License, and (3) the Commission's related letter, Safety Evaluation
and/or Environmental Assessment, as indicated. All of these items are
available for public inspection at the Commission's Public Document
Room, the Gelman Building, 2120 L Street, NW., Washington, DC, and at
the local public document room for the particular facility involved.
The Commission is also offering an opportunity for a hearing with
respect to the issuance of the amendment. By December 4, 1998, 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
[[Page 59604]]
the proceeding; and (3) the possible effect of any order which may be
entered in the proceeding on the petitioner's interest. The petition
should also identify the specific aspect(s) of the subject matter of
the proceeding as to which petitioner wishes to intervene. Any person
who has filed a petition for leave to intervene or who has been
admitted as a party may amend the petition without requesting leave of
the Board up to 15 days prior to the first prehearing conference
scheduled in the proceeding, but such an amended petition must satisfy
the specificity requirements described above.
Not later than 15 days prior to the first prehearing conference
scheduled in the proceeding, a petitioner shall file a supplement to
the petition to intervene which must include a list of the contentions
which are sought to be litigated in the matter. Each contention must
consist of a specific statement of the issue of law or fact to be
raised or controverted. In addition, the petitioner shall provide a
brief explanation of the bases of the contention and a concise
statement of the alleged facts or expert opinion which support the
contention and on which the petitioner intends to rely in proving the
contention at the hearing. The petitioner must also provide references
to those specific sources and documents of which the petitioner is
aware and on which the petitioner intends to rely to establish those
facts or expert opinion. Petitioner must provide sufficient information
to show that a genuine dispute exists with the applicant on a material
issue of law or fact. Contentions shall be limited to matters within
the scope of the amendment under consideration. The contention must be
one which, if proven, would entitle the petitioner to relief. A
petitioner who fails to file such a supplement which satisfies these
requirements with respect to at least one contention will not be
permitted to participate as a party.
Those permitted to intervene become parties to the proceeding,
subject to any limitations in the order granting leave to intervene,
and have the opportunity to participate fully in the conduct of the
hearing, including the opportunity to present evidence and cross-
examine witnesses. Since the Commission has made a final determination
that the amendment involves no significant hazards consideration, if a
hearing is requested, it will not stay the effectiveness of the
amendment. Any hearing held would take place while the amendment is in
effect.
A request for a hearing or a petition for leave to intervene must
be filed with the Secretary of the Commission, U.S. Nuclear Regulatory
Commission, Washington, DC 20555-0001, 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 the
factors specified in 10 CFR 2.714(a)(1)(i)-(v) and 2.714(d).
Arizona Public Service Company, et al., Docket No. STN 50-530, Palo
Verde Nuclear Generating Station, Unit No. 3, Maricopa County, Arizona
Date of application for amendment: October 6, 1998
Brief description of amendment: The amendment revises TS 3.3.1,
``Reactor Protective System (RPS) Instrumentation--Operation,'' and TS
3.3.2, ``Reactor Protective System (RPS) Instrumentation--Shutdown.''
The proposed amendment would clarify the power level threshold at which
certain RPS instrumentation trips must be enabled and may be bypassed,
and would clarify that this level is a percentage of the neutron flux
at rated thermal power (RTP). The bypass power level, 1E-4% RTP, would
be specified as logarithmic power instead of thermal power.
Date of issuance: October 19, 1998.
Effective date: October 19, 1998.
Amendment No.: 119.
Facility Operating License No. NPF-74: The amendment revised the
Technical Specifications.
Press release issued requesting comments as to proposed no
significant hazards consideration: Yes. October 13, 1998. Arizona
Republic Newspaper (Arizona).
Comments received: No. The Commission's related evaluation of the
amendment, finding of exigent circumstances, consultation with the
State of Arizona and final determination of no significant hazards
consideration are contained in a Safety Evaluation dated October 19,
1998.
Local Public Document Room location: Phoenix Public Library, 1221
N. Central Avenue, Phoenix, Arizona 85004.
Attorney for licensee: Nancy C. Loftin, Esq., Corporate Secretary
and Counsel, Arizona Public Service Company, P.O. Box 53999, Mail
Station 9068, Phoenix, Arizona 85072-3999.
NRC Project Director: William H. Bateman.
Dated at Rockville, Maryland, this 28th day of October 1998.
For the Nuclear Regulatory Commission
Elinor G. Adensam,
Acting Director, Division of Reactor Projects--III/IV, Office of
Nuclear Reactor Regulation.
[FR Doc. 98-29433 Filed 11-3-98; 8:45 am]
BILLING CODE 7590-01-P