[Federal Register Volume 63, Number 213 (Wednesday, November 4, 1998)]
[Notices]
[Pages 59584-59604]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: 98-29433]


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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--Tavg>350 deg.F'' 
and \3/4\.5.3, ``ECCS Subsystems--Tavg<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