[Federal Register Volume 63, Number 174 (Wednesday, September 9, 1998)]
[Notices]
[Pages 48256-48278]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: 98-24130]


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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

[[Page 48257]]

Commission that such amendment involves no significant hazards 
consideration, notwithstanding the pendency before the Commission of a 
request for a hearing from any person.
    This biweekly notice includes all notices of amendments issued, or 
proposed to be issued from August 17, 1998, through August 28, 1998. 
The last biweekly notice was published on August 26, 1998 (63 FR 
45521).
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 October 9, 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

[[Page 48258]]

significant hazards consideration, any hearing held would take place 
before the issuance of any amendment.
    A request for a hearing or a petition for leave to intervene must 
be filed with the Secretary of the Commission, U.S. Nuclear Regulatory 
Commission, Washington, DC 20555-0001, Attention: Rulemakings and 
Adjudications Staff, or may be delivered to the Commission's Public 
Document Room, the Gelman Building, 2120 L Street, NW., Washington DC, 
by the above date. A copy of the petition should also be sent to the 
Office of the General Counsel, U.S. Nuclear Regulatory Commission, 
Washington, DC 20555-0001, and to the attorney for the licensee.
    Nontimely filings of petitions for leave to intervene, amended 
petitions, supplemental petitions and/or requests for a hearing will 
not be entertained absent a determination by the Commission, the 
presiding officer or the Atomic Safety and Licensing Board that the 
petition and/or request should be granted based upon a balancing of 
factors specified in 10 CFR 2.714(a)(1)(i)-(v) and 2.714(d).
    For further details with respect to this action, see the 
application for amendment which is available for public inspection at 
the Commission's Public Document Room, the Gelman Building, 2120 L 
Street, NW., Washington, DC, and at the local public document room for 
the particular facility involved.

Carolina Power & Light Company, et al., Docket Nos. 50-325 and 50-324, 
Brunswick Steam Electric Plant, Units 1 and 2, Brunswick County, North 
Carolina

    Date of amendment request: August 17, 1998.
    Description of amendment request: The Carolina Power & Light 
Company, licensee for the Brunswick Steam Electric Plant (BSEP), Unit 
Nos. 1 and 2, proposed amendments to the Technical Specifications (TS) 
to revise the requirement that the operations manager hold or has held 
a senior reactor operator (SRO) license. The proposed revision would 
require that either the operations manager or assistant operations 
manager hold an SRO license.
    The licensee has concluded that the proposed license amendments do 
not involve a Significant Hazards Consideration. In support of this 
determination, an evaluation of each of the three standards set forth 
in 10 CFR 50.92 is provided 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:

    1. Would operation of the facility in accordance with the 
proposed amendment involve a significant increase in the probability 
or consequences of an accident previously evaluated?
    The proposed license amendments do not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated. The change to Technical Specification 5.2.2.f 
to require the operations manager or assistant operations manager to 
hold an SRO license is administrative in nature and does not 
directly affect plant operations. The change does not physically 
alter the facility in any manner and, as such, does not affect the 
means in which any safety-related system performs its intended 
safety function.
    2. Would operation of the facility in accordance with the 
proposed amendment create the possibility of a new or different kind 
of accident from any accident previously evaluated?
    The proposed license amendments will not create the possibility 
of a new or different kind of accident from any accident previously 
evaluated. As stated above, the proposed change is administrative in 
nature. It does not involve physical alterations of the plant 
configuration or changes in setpoints or operating parameters. 
Therefore, there is no possibility of creating a new or different 
kind of accident.
    3. Would operation of the facility in accordance with the 
proposed amendment involve a significant reduction in a margin of 
safety?
    The proposed license amendments do not involve a significant 
reduction in a margin of safety. The proposed change to Technical 
Specification 5.2.2.f, requiring the operations manager or assistant 
operations manager to hold an SRO license is consistent with (1) 10 
CFR 50.54(l), which requires individuals responsible for directing 
the licensed activities of licensed operators to hold an SRO 
license, (2) the previously approved wording of Revision 1 of NUREG-
1433, ``Standard Technical Specifications General Electric Plants, 
BWR/4,'' and Technical Specification Traveler Form (TSTF) 65, 
Revision 1, and (3) the intent of ANSI-N18.1.-1971, ``Selection and 
Training of Nuclear Power Plant Personnel.'' Therefore, the proposed 
change does not represent a reduction in the margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: University of North Carolina 
at Wilmington, William Madison Randall Library, 601 S. College Road, 
Wilmington, North Carolina 28403-3297.
    Attorney for licensee: William D. Johnson, Vice President and 
Senior Counsel, Carolina Power & Light Company, Post Office Box 1551, 
Raleigh, North Carolina 27602.
    NRC Project Director: Pao-Tsin Kuo (Acting).

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: August 14, 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 supersedes in its entirety a 
letter from J. Hosmer (ComEd) to U.S. NRC, ``Technical Specification 
Changes for Transition to Siemens Power Corporation ATRIUM-9B Fuel,'' 
dated August 29, 1997 (63 FR 2274).
    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.

[[Page 48259]]

a. Addition of SPC Revised Jet Pump Methodology (LaSalle Units 1 and 2)

    The Reference 1 methodology to be added to the Technical 
Specifications is used as part of the LOCA 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 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 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 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 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 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 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. Approval of Reference 7 (Reference 20) 
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% 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. 
This change implements the MCPR Safety Limits resulting from the SPC 
ANFB critical power correlation methodology using the ATRIUM-9B 
additive constant uncertainty resulting from approval of Reference 7 
(Reference 20). 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 value where less than 0.1% 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.

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 limits and their exposure basis to 
be specified in the COLR 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 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 
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.

[[Page 48260]]

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% 
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 resulting from approval 
of Reference 7 (Reference 20). 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 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 2.1.2 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 (Reference 4 and 5). The 
MCPR Safety Limit will continue to ensure that greater than 99.9% 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% 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% 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% 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 resulting from approval of Reference 7 
(Reference 20).
    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% 
of the fuel rods avoiding transition boiling during normal operation 
as well as during an anticipated operational occurrence is met.

[[Page 48261]]

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 (Reference 20) 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, IL 60450; for Quad 
Cities, Dixon Public Library, 221 Hennepin Avenue, Dixon, IL 61021; and 
for LaSalle, the Jacobs Memorial Library, 815 North Orlando Smith 
Avenue, Illinois Valley Community College, Oglesby, IL 61348-9692.
    Attorney for licensee: Michael I. Miller, Esquire; Sidley and 
Austin, One First National Plaza, Chicago, IL 60603.
    NRC Project Director: Stuart A. Richards.

Entergy Operations Inc., Docket No. 50-382, Waterford Steam Electric 
Station, Unit 3, St. Charles Parish, Louisiana

    Date of amendment request: July 17, 1996.
    Description of amendment request: The proposed change extends the 
surveillance interval for the Reactor Trip Breakers (RTBs) from monthly 
to quarterly and increases the allowed outage time for operation with 
an inoperable RTB from one hour to two hours.
    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 any accident previously evaluated?
    Response: No.
    The proposed change to increase RTB surveillance interval will 
have no significant effect on the probability or consequences of any 
accident previously evaluated. As previously stated, all of the 
transient and accident analyses that call for a reactor trip assume 
that the reactor trip breakers (RTBs) operate and interrupt power to 
the control element drive mechanism (CEDMs). Extensive testing 
results, indicate that the RTBs are available and capable of 
performing their safety-related function. Currently RTBs are 
verified operable every 4 weeks. Under the proposed change RTBs 
would be verified operable at least every 6 weeks. This reduced 
testing frequency is intended to increase component reliability. The 
increase in the testing interval cannot increase component failure 
rate or the potential for component failure.
    The proposed change to increase the allowed outage time for RTBs 
from 1 hour to 2 hours will have no significant impact on 
probability or consequences of any accident previously evaluated. 
When an RTB is inoperable, Functional Testing and other breaker 
operations becomes more difficult. The current technical 
specification allows an inoperable breaker to be closed for 1 hour 
to perform testing of other RTBs. This provision is infrequently 
required, but when it is required, the allowed outage time is very 
short and rushing to complete a test may lead to an inadvertent 
reactor trip. Increasing this allowed outage time is an improvement 
item identified in NUREG 1366 and consistent with philosophy 
provided in Generic Letter 89-07.
    Therefore, the proposed change will 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.
    This proposed change does not involve any changes in equipment 
and will not alter the manner in which the plant will be operated.
    Therefore, the proposed change 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 change will not adversely affect the performance of 
the safety function of the RTBs. In fact, it is expected that the 
performance of the RTBs will improve as a result of this change 
based on less wear and tear on the equipment. The proposed change 
will have no adverse impact on the protective boundaries, safety 
limits or margin of safety.
    Therefore, the proposed change will not involve a significant 
reduction in a margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room Location: University of New Orleans 
Library, Louisiana Collection, Lakefront, New Orleans, 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.

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.
    Description of amendment request: The proposed amendment would 
revise the Technical Specifications (TSs) to incorporate revised Safety 
Limit Minimum Critical Power Ratios (SLMCPRs) for the use of cycle-
specific analysis performed for Peach Bottom Atomic Power Station 
(PBAPS), Unit 2, Cycle 13.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:
    1. The proposed TS changes do not involve a significant increase 
in the probability or consequences of an accident previously 
evaluated.

[[Page 48262]]

    The derivation of the cycle-specific SLMCPRs for incorporation 
into the TS, and its use to determine cycle-specific thermal limits, 
have been performed using the methodology discussed in ``General 
Electric Standard Application for Reactor Fuel,'' NEDE-24011-P-A-13, 
and U.S. Supplement, NEDE-24011-P-A-13-US, August, 1996, and the 
``Proposed Amendment 25 to GE Licensing Topical Report NEDE-24011-P-
A (GESTAR II) on Cycle Specific Safety Limit MCPR.'' Amendment 25 
was submitted by [General Electric Nuclear Energy] GENE to the U.S. 
Nuclear Regulatory Commission (USNRC) on December 13, 1996. This 
change in SLMCPRs cannot increase the probability or severity of an 
accident.
    The basis of the SLMCPR calculation is to ensure that greater 
than 99.9% of all fuel rods in the core avoid transition boiling if 
the limit is not violated. The new SLMCPRs preserve the existing 
margin to transition boiling and fuel damage in the event of a 
postulated accident. The fuel licensing acceptance criteria for the 
SLMCPR calculation apply to PBAPS, Unit 2, Cycle 13 in the same 
manner as they have applied previously. The probability of fuel 
damage is not increased. Therefore, the proposed TS changes do not 
involve an increase in the probability or consequences of an 
accident previously evaluated.
    In addition to the change to the SLMCPR, the footnote to TS 
2.1.1.2 is being revised, and a footnote is being added to TS 
5.6.5.b.1. The revision to the footnote associated with TS 2.1.1.2 
will ensure that the SLMCPR value is reconfirmed for the cycle 
subsequent to PBAPS, Unit 2, Cycle 13, and the footnote to TS 
5.6.5.b.1 is being added due to the use of the proposed Amendment 25 
and the use of a proposed R-factor calculation methodology (``R-
Factor Calculation Method for GE11, GE12, and GE13 Fuel,'' NEDC-
32505P, Revision 1, June 1997), which has not yet been approved for 
generic use by the USNRC. The revision to the footnote associated 
with TS 2.1.1.2 and the addition of the footnote to TS 5.6.5.b.1 are 
administrative changes that do not involve an increase in the 
probability or consequences of an accident previously evaluated.
    2. The proposed TS changes do not create the possibility of a 
new or different kind of accident from any accident previously 
evaluated.
    The SLMCPR is a TS numerical value, designed to ensure that 
transition boiling does not occur in 99.9% of all fuel rods in the 
core during the limiting postulated accident. It cannot create the 
possibility of any new type of accident. The new SLMCPRs are 
calculated using methodology discussed in ``Generic Electric 
Standard Application for Reactor Fuel,'' NEDE-24011-P-A-13, and U.S. 
Supplement, NEDE-24011-P-A-13-US, August, 1996, and the ``Proposed 
Amendment 25 to GE Licensing Topical Report NEDE-24011-P-A (GESTAR 
II) on Cycle Specific Safety Limit MCPR.'' Amendment 25 was 
submitted by GENE to the USNRC on December 13, 1996. Therefore, the 
revision to the SLMCPR will not create the possibility of a new or 
different kind of accident from any accident previously evaluated.
    Additionally, this proposed change will revise the footnote to 
TS 2.1.1.2, and add a footnote to TS 5.6.5.b.1. The revision to the 
footnote associated with TS 2.1.1.2, and the addition of the 
footnote to TS 5.6.5.b.1, are administrative changes that do not 
create the possibility of a new or different kind of accident from 
any previously evaluated.
    3. The proposed TS changes do not involve a significant 
reduction in a margin of safety.
    There is no significant reduction in the margin of safety 
previously approved by the USNRC as a result of the proposed change 
to the SLMCPR, and the proposed change that will revise the footnote 
to TS 2.1.1.2, and add a footnote to TS 5.6.5.b.1. The new SLMCPRs 
are calculated using methodology discussed in ``General Electric 
Standard Application for Reactor Fuel,'' NEDE-24011-P-A-13, and U.S. 
Supplement, NEDE-24011-P-A-13-US, August, 1996, and the ``Proposed 
Amendment 25 to GE Licensing Topical Report NEDE-24011-P-A (GESTAR 
II) on Cycle Specific Safety Limit MCPR.'' Amendment 25 was 
submitted by GENE to the USNRC on December 13, 1996. The fuel 
licensing acceptance criteria for the calculation of the SLMCPR 
apply to PBAPS, Unit 2 Cycle 13 in the same manner as they have 
applied previously. The SLMCPRs ensure that greater than 99.9% of 
all fuel rods in the core will avoid transition boiling if the limit 
is not violated, thereby preserving the fuel cladding integrity. 
Therefore, the proposed TS changes will not significantly reduce the 
margin of safety previously approved by the USNRC.
    Additionally, the proposed change that will revise the footnote 
to TS 2.1.1.2, and add a footnote to TS 5.6.5.b.1 is an 
administrative change that will not significantly reduce the margin 
of safety previously approved by the USNRC.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: Government Publications 
Section, State Library of Pennsylvania, (Regional Depository) Education 
Building, Walnut Street and Commonwealth Avenue, Box 1601, Harrisburg, 
PA 17105.
    Attorney for Licensee: J. W. Durham, Sr., Esquire, Sr. V.P. and 
General Counsel, PECO Energy Company, 2301 Market Street, Philadelphia, 
PA 19101.
    NRC Project Director: Robert A. Capra.

Pennsylvania Power and Light Company, Docket No. 50-388 Susquehanna 
Steam Electric Station, Unit 2, Luzerne County, Pennsylvania.

    Date of amendment request: August 4, 1998
    Description of amendment request: The amendment would modify the 
Susquehanna Steam Electric Station, Unit 2, Technical Specifications to 
replace figures 2.1.1.2-1 and 2.1.1.2-2, and associated footnotes, with 
single value minimum critical power ratio (MCPR) Safety Limits of 
Section 2.1.1.2; remove references from Section 5.6.5 which do not 
directly support the generation of Core Operating Limits; remove 
references from Section 5.6.5 which were previously included to address 
the application of the ANFB-10 correlation to ATRIUM-10 fuel; include 
Siemens Power Corporation ANFB-10 topical report in Section 5.6.5; and 
to change the Bases to reflect inclusion of the ANFB-10 critical power 
correlation.
    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 applicable sections of the FSAR [Final Safety Analysis 
Report] are Chapters 4.4 and 15. FSAR Chapter 4.4 describes the MCPR 
Safety Limit, and Chapter 15 describes the transient and accident 
analyses. The reference to be added to Section 5.6.5 of the Unit 2 
Technical Specifications describes an NRC approved critical power 
correlation for ATRIUMTM-10 fuel appropriate for use in 
conservative methodologies for generating MCPR Safety Limits and 
MCPR Operating Limits to assure safe operation of Unit 2 with 
ATRIUMTM-10 fuel. A discussion of the impact of the 
proposed Technical Specification change is provided below.
    The proposed change in critical power correlation does not 
physically affect the plant or its systems. Thus, it does not 
increase the probability of an accident previously evaluated.
    A Unit 2 Cycle 10 MCPR Safety Limit analysis was performed for 
PP&L by SPC. This analysis used NRC approved methods described in 
ANF-524(P)(A), Revision 2 and Supplement 1 Revision 2. These methods 
will be used each cycle to calculate the Unit 2 Safety Limits. For 
Unit 2 Cycle 10, the critical power performance of the 9x9-2 and 
ATRIUMTM-10 fuel was determined using the NRC approved 
ANFB and ANFB-10 correlations, respectively. The SAFETY LIMIT MCPR 
calculations statistically combine uncertainties on feedwater flow, 
feedwater temperature, core flow, core pressure, core power 
distribution, and uncertainties in the Critical Power Correlations. 
The SPC analysis used cycle specific power distributions and 
calculated MCPR values such that at least 99.91% of the fuel rods 
are expected to avoid boiling transition during normal operation or 
anticipated operational occurrences. The resulting two-loop and 
single-loop MCPR Safety Limits are included in the proposed 
Technical Specification change. Thus, the cladding integrity and its 
ability to contain fission products are not adversely affected.

[[Page 48263]]

    Analyses of the Single Loop Pump Seizure accident with the NRC 
approved ANFB-10 correlation for the ATRIUMTM-10 fuel 
(Reference 1) [Reference 1 refers to the reference listed in the 
application dated August 4, 1998] will be performed to demonstrate 
that the NRC acceptance criterion (i.e., small fraction of 10 CFR 
100 dose limits) is met. Analyses will also be performed to validate 
the conclusion that single-loop transients are less severe than the 
those events analyzed for two-loop operation.
    Changes to Section 2.1.1.2 reflect the change from a flow 
dependent MCPR Safety Limit to a single value MCPR Safety Limit for 
two-loop operation and single-loop operation.
    Changes to Reference 5.6.5 delete the methodology used for 
critical power analyses for ATRIUMTM-10 fuel and add the 
NRC approved ANFB-10 methodology to the list of approved 
methodologies. Other changes in Reference 5.6.5 are administrative 
in nature because they delete references that are not directly 
related to the generation of Core Operating Limits. No new analysis 
approaches are used due to the removal of these references.
    Changes to BASES Sections 2.1.1 and 3.2.2 reflect the inclusion 
of the ANFB-10 critical power correlation. The range of the 
applicability of the ANFB-10 is valid for pressures > 571 psia and 
bundle mass fluxes > 0.115  x  10 \6\ lb/hr-ft \2\. These values 
assure that a valid CPR calculation will result at or above 25% of 
rated core thermal power, that is, reactor steam dome pressure 
785 psig and core flow 10 Mlbm/hr.
    The consequences of transients and accidents will remain within 
the criteria approved by the NRC. The methodology used to perform 
the analyses have been previously approved by the NRC. Thus, 
analysis results using the new methodology will continue to provide 
assurance that the reactor will perform its design safety function 
during normal operation and design basis events. Therefore, the 
proposed action does not involve an 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 changes to the Unit 2 Technical Specifications 
(MCPR Safety Limits, removal of methodology references not directly 
supporting the generation of Core Operating Limits, removal of the 
two references describing previously approved methodology for 
applying ANFB to ATRIUMTM-10 fuel, and inclusion of the 
ANFB-10 correlation reference) do not require any physical plant 
modifications, physically affect any plant components, or entail 
changes in plant operation. Removal of the Unit 2 Cycle 9 footnote 
allows Unit 2 Cycle 10 and future cycle operation with thermal 
limits generated using NRC approved methodology. Thus, the proposed 
change does not create the possibility of a previously unevaluated 
operator error or a new single failure. The consequences of 
transients and accidents will remain within the criteria approved by 
the NRC. Therefore, the proposed change does not create the 
possibility of a new or different kind of accident from any 
previously evaluated.
    3. The proposed change does not involve a significant reduction 
in a margin of safety.
    The applicable Technical Specification Sections include 2.1.1.2 
and 5.6.5.
    The changes to the Unit 2 Technical Specifications discussed in 
item 1 above do not require any physical plant modifications, 
physically affect any plant components, or entail changes in plant 
operation. Therefore, the proposed change will not jeopardize or 
degrade the function or operation of any plant system or component 
governed by Technical Specifications. The consequences of transients 
and accidents will remain within the criteria approved by the NRC. 
The proposed MCPR Safety Limits and use of the NRC approved ANFB-10 
critical power correlation described in the reference added to 
Section 5.6.5 do not involve a significant reduction in the margin 
of safety as currently defined in the BASES of the applicable 
Technical Specification sections.
    Therefore, the proposed change does not involve a significant 
reduction in the margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: Osterhout Free Library, 
Reference Department, 71 South Franklin Street, Wilkes-Barre, PA 18701.
    Attorney for licensee: Jay Silberg, Esquire, Shaw, Pittman, Potts 
and Trowbridge, 2300 N Street NW., Washington, DC 20037.
    NRC Project Director: Robert A. Capra.

Pennsylvania Power and Light Company (PP&L), Docket No. 50-388, 
Susquehanna Steam Electric Station, Unit 2, Luzerne County, 
Pennsylvania

    Date of amendment request: August 5, 1998.
    Description of amendment request: The amendment would modify the 
Susquehanna Steam Electric Station, Unit 2, Technical Specifications 
Table 3.3.5.1-1 ``Emergency Core Cooling System Instrumentation.'' The 
change updates the allowable values for both the Core Spray (CS) and 
Low Pressure Coolant Injection System (LPCI) ``Reactor Steam Dome 
Pressure--Low'' functions for initiation and injection permissive. 
Specifically, the allowable values are changed from a specified minimum 
pressure to a specified allowable pressure band. This more restrictive 
allowable value range will prevent CS and LPCI system 
overpressurization while still permitting injection to prevent fuel 
clad temperature limits from being exceeded.
    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.
    This proposal does not involve an increase in the probability or 
consequences of an accident previously evaluated. The proposed 
amendment changes the ``Reactor Steam Dome Pressure--Low'' Allowable 
Values so to provide further assurance that the Core Spray and 
[Residual Heat Removal] RHR systems will perform their [loss-of-
coolant accident] LOCA design basis function.
    The functional design basis of the Core Spray and LPCI is to 
inject water into the reactor vessel to cool the core during a LOCA 
by opening the Core Spray and LPCI injection valves when reactor 
pressure drops below the reactor vessel low pressure permissive. The 
upper analytical limit for the permissive is the Core Spray and LPCI 
systems' maximum design pressure, and the lower analytical limit is 
the lowest pressure which allows injection to prevent exceeding the 
fuel cladding temperature limit. The new allowable values were 
selected to lie within the upper and lower limits to ensure there 
will be no change in the required logic or functions of the Core 
Spray and LPCI systems. These new values do not affect the LOCA nor 
its ``limiting fault'' frequency of occurrence and do not introduce 
any new accidents or malfunctions of equipment important to safety. 
Since they do not affect the LOCA, they do not change the 
probability of occurrence of the LOCA. The new allowable values do 
not change the logic or function of the reactor vessel low pressure 
permissive. These new values simply provide the basis for which the 
associated pressure instruments are to be set to ensure proper 
operation of Core Spray and LPCI within the design pressures as 
described above. Therefore, the change in allowable values does not 
increase the probability of occurrence or the consequences of an 
accident or malfunction of equipment important to safety.
    Based upon the analysis presented above, PP&L concludes that the 
proposed action does not involve an 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.
    This proposal does not create the probability of a new or 
different type of accident from any accident previously evaluated. 
The new allowable values do not change any plant systems, 
structures, or components, nor do they change any existing or create 
any new Core Spray and LPCI logic or functions. The new allowable 
values were selected to ensure the required operation of the Core 
Spray and LPCI systems within the maximum design pressures.

[[Page 48264]]

    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 the margin of safety.
    The change does not involve a reduction in the margin of safety. 
Technical Specification Bases Section B3.3.5.1 [9 sic] (ECCS 
Instrumentation) identifies that the low reactor steam dome pressure 
signals are used as permissives for operation of the low pressure 
ECCS subsystems. The new allowable values were selected so as to not 
impact the logic, redundancy, operability or surveillance 
requirements for these subsystems. The new allowable values maintain 
the margin requirements of the Core Spray and LPCI system pressures 
such that they do not exceed their system maximum design pressures 
and that system pressures are high enough to ensure that the ECCS 
injection prevents the fuel peak cladding temperature from exceeding 
the limits of 10 CFR50.46.
    Therefore, the margin of safety is enhanced by the proposed 
changes.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: Osterhout Free Library, 
Reference Department, 71 South Franklin Street, Wilkes-Barre, PA 18701.
    Attorney for licensee: Jay Silberg, Esquire, Shaw, Pittman, Potts 
and Trowbridge, 2300 N Street NW., Washington, DC 20037.
    NRC Project Director: Robert A. Capra.

Power Authority of the State of New York, Docket No. 50-333, James A. 
FitzPatrick Nuclear Power Plant, Oswego County, New York

    Date of amendment request: August 3, 1998.
    Description of amendment request: The proposed changes provide for 
applicability of the safety limit minimum critical power ratio (SLMCPR) 
to fuel cycle 14.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:
    Operation of the FitzPatrick plant in accordance with the proposed 
amendment would not involve a significant hazards consideration as 
defined in 10 CFR 50.92, since it would not:

    1. involve a significant increase in the probability or 
consequences of an accident previously evaluated.
    A change to a note stating that the SLMCPR remains applicable 
through Cycle 14 does not affect the initiation of any accident. 
Operation in accordance with the current SLMCPR ensures the 
consequences of previously analyzed accidents are not changed. 
Therefore, this proposed change does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    2. create the possibility of a new or different kind of accident 
from any accident previously evaluated.
    The SLMCPR establishes a performance limit for the fuel. This 
limit remains unchanged. Changing a note to reflect this is an 
administrative change and will not initiate any accident. Therefore, 
this proposed change does not create the possibility of a new or 
different kind of accident from any accident previously evaluated.
    3. involve a significant reduction in a margin of safety.
    GE [General Electric] has performed an evaluation of the SLMCPR 
for Cycle 14 and found that the cycle specific value, based on 
current reload plans, is bounded by the generic value calculated for 
GE 12 fuel. The existing SLMCPR remains unchanged for Cycle 14 and 
the margin of safety for the prevention of onset of transition 
boiling is unchanged. Therefore, this proposed change does not 
involve a significant reduction in a margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: Reference and Documents 
Department, Penfield Library, State University of New York, Oswego, New 
York 13126.
    Attorney for licensee: Mr. David E. Blabey, 1633 Broadway, New 
York, New York 10019.
    NRC Project Director: S. Singh Bajwa, Director.

Public Service Electric & Gas Company, Docket Nos. 50-272 and 50-311, 
Salem Nuclear Generating Station, Unit Nos. 1 and 2, Salem County, New 
Jersey

    Date of amendment request: July 30, 1998.
    Description of amendment request: The proposed amendments would 
revise Technical Specification (TS) 3/4.7.6, ``Control Room Emergency 
Air Conditioning System.'' Specifically, the acceptance criteria for 
the control room envelope would be revised to maintain a \1/8\-inch 
positive pressure with respect to all areas directly accessible from 
the control room and a positive pressure with respect to all other 
areas adjacent to the control room.
    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.
    CREACS [Control Room Emergency Air Conditioning System] ensures 
adequate protection after an accident and is not an accident 
initiator. The change to the acceptance criteria for CREACS does not 
affect the probability of an accident.
    Revising the acceptance criteria for the CREACS from a `1/8-inch 
W.G. [water gauge] positive pressure in the control room with 
respect to the adjacent area' to `a 1/8-inch W.G. positive pressure 
in the control room with respect to all areas directly accessible 
(Work Control Center and Control Room Equipment Rooms) from the 
control room and a positive pressure to all other areas adjacent to 
the control room' does not alter the assumptions in the radiological 
dose assessment provided to the NRC and approved under Amendments 
190 (Unit 1) and 173 (Unit 2). Therefore the conclusions of the 
radiological dose assessment reviewed and approved by the NRC under 
the above Amendments remain unchanged. The radiological dose 
assessment provided under Amendments 190 and 173 demonstrates that 
operation of the CREAS in the pressurized mode at the initiation of 
an accident will ensure that the requirements of General Design 
Criterion (GDC) 19 will be met.
    Therefore, the proposed TS change does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    2. The proposed change does not create the possibility of a new 
or different kind of accident from any accident previously 
evaluated.
    Since the CREACS is an accident mitigation system that does not 
communicate with the Reactor Coolant Pressure boundary or interface 
with Emergency Core Cooling Systems (ECCS), the proposed change to 
the acceptance criteria for CREACS pressurization cannot result in 
new accident scenarios. The function of the CREACS system is to 
maintain the habitability of the CRE [control room envelope] 
following an accident.
    Therefore, the proposed TS 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 CREACS ensures that (1) the ambient air temperature does not 
exceed the allowable temperature for continuous duty rating for 
equipment and instrumentation cooled by the CREACS and (2) the 
Control Room will remain habitable for operations personnel during 
and following all credible radiological accident conditions. 
Revising the

[[Page 48265]]

acceptance criteria to maintaining the control room at a \1/8\-inch 
W.G. positive pressure in the control room with respect to all areas 
directly accessible (Work Control Center and Control Room Equipment 
Rooms) from the control room and a positive pressure to all other 
areas adjacent to the control room does not alter the assumptions 
used in the radiological dose assessment nor revise the conclusions 
of the dose assessment which was reviewed under Amendments 190 and 
173. Since the assumptions and conclusions of the dose assessment 
remain unchanged, the CREACS continues to ensure that the 
requirements of GDC 19 continue to be met, and there is no reduction 
in the safety provided to the control room operators.
    Therefore, the proposed change to the TS does not involve a 
reduction in the margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: 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.

Public Service Electric & Gas Company, Docket Nos. 50-272 and 50-311, 
Salem Nuclear Generating Station, Unit Nos. 1 and 2, Salem County, New 
Jersey

    Date of amendment request: August 12, 1998.
    Description of amendment request: The proposed amendments would 
revise Technical Specification (TS) 3/4.6.1.3, ``Containment Air 
Locks,'' to change the action statements for an inoperable airlock. The 
proposed amendments would also correct an editorial error in TS Bases 
3/4.6.1.2, ``Containment Leakage.''
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Will not involve a significant increase in the probability or 
consequences of an accident previously evaluated.
    The reactor containment serves to mitigate the consequences of a 
Design Basis Accident (DBA). That is, the containment is designed to 
provide a barrier to ensure that in the event of a DBA, a release of 
radioactive material will not result in the radiation dose to the 
general public exceeding the limits of 10 CFR 100. Each unit's 
containment has been provided with two air locks. These air locks 
permit personnel to access components and systems within the 
containment boundary without compromising the containment's ability 
to carry out its design function. In this capacity, the air locks 
serve as part of the containment boundary and as such are not 
considered as a contributor to the probability of an accident.
    To carry out their design function, the air locks are designed 
and tested to certify their ability to withstand a pressure in 
excess of the maximum expected following a DBA. Each door is 
individually tested to verify that leakage will remain below design 
values with the containment at design pressure. An interlock is 
provided to ensure that containment integrity is maintained during 
personnel passage by allowing only one air lock door to be open at a 
time. This interlock is also periodically tested to verify its 
functionality.
    The proposed changes will allow continued operation with one air 
lock door inoperable or with the air lock door interlock mechanism 
disabled but will specify the actions necessary under those 
conditions to assure that containment integrity is not compromised. 
This will ensure that the consequences of an accident previously 
evaluated are not significantly increased. Additionally, the 
proposed changes specify that in the event that an air lock is 
inoperable for a reason other than an inoperable air lock door, or 
air lock interlock mechanism, the unit must be placed in a condition 
in which the analyzed accident could not occur.
    Based upon the above, the proposed changes do not involve a 
significant increase in the probability or consequences of an 
accident previously evaluated.
    2. Create the possibility of a new or different kind of 
accident.
    The proposed changes to the Containment Air Lock Technical 
Specifications do not affect the ability of the containment to carry 
out its design function. The changes also do not introduce any new 
equipment; nor do they result in the operation of the plant in a 
manner contrary to the safety analysis. Therefore, the proposed 
changes will not increase the probability of a new or different kind 
of accident from any accident previously identified.
    3. Will not involve a significant reduction in a margin of 
safety.
    The proposed changes do not affect any design or functional 
requirements of the Containment or the Containment Air Locks. 
Additionally, the proposed changes do not affect any of the 
conditions or assumptions of the applicable safety analyses. 
Containment Air Lock leakage rates are determined based upon 
containment leakage at design pressure. The proposed changes will 
not affect containment design pressure nor will they affect the peak 
containment pressures expected for analyzed accidents.
    Based upon the above, the proposed change will not involve a 
reduction in a margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: 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 requests: May 11, 1998 (Supersedes the May 30, 
1996, amendment request). This Notice supersedes the staff's proposed 
no significant hazards consideration determination for the requested 
changes that was published on September 11, 1996 (61 FR 47981).
    Description of amendment requests: The proposed amendments would 
revise the Technical Specifications (TS) to allow use of performance-
based criteria to establish containment leak rate test intervals and 
add a new ``Containment Leakage Rate Testing Program'' to the 
administrative section of TS to codify the program used to determine 
the testing program. The proposed program implements 10 CFR Part 50, 
Appendix J, Option B, by referring to Regulatory Guide 1.163, 
``Performance-Based Containment Leak Test Program,'' dated September 
1995.
    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.
    Since the interval between containment leakage rate tests is not 
related in any way to conditions which cause accidents, and plant 
structures, systems, and components will not be operated in a 
different manner as a result of the proposed Technical Specification 
(TS) change, the proposed changes will not increase the probability 
of an accident previously evaluated.
    Containment leakage may result from accidents which are 
evaluated in the Updated Final Safety Analysis Report. The proposed 
TS changes may result in an acceptably small increase in post-
accident containment leakage. Using a statistical approach, NUREG-
1493 determined that the increase in hypothetical dose to the public 
resulting from extending the testing interval is extremely small. 
NUREG-1493 concluded that such small hypothetical dose increases

[[Page 48266]]

to the public are justifiable due to the real reduction in 
occupational exposure resulting from interval extension. Therefore, 
the proposed change does not significantly increase the 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 only incorporates the performance based 
approach for containment leak rate testing authorized in the new 
Option B to Appendix J of 10 CFR Part 50. The interval extensions 
allowed, through this approach, do not have the potential for 
creating the possibility of new or different kinds of accidents from 
those previously evaluated because plant structures, systems, and 
components will not be operated in a different manner as a result of 
the TS change and, therefore, will not introduce any new or 
different failure modes or initiators. 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 proposed Technical Specification does not alter the 
allowable containment leakage rate. The proposed change replaces the 
current, prescriptive testing requirements with a new performance 
based approach for establishing the testing intervals. 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.

Southern California Edison Company, et al., Docket Nos. 50-361 and 50-
362, San Onofre Nuclear Generating Station, Unit Nos. 2 and 3, San 
Diego County, California

    Date of amendment requests: June 19, 1998.
    Description of amendment requests: The proposed amendments would 
revise Technical Specification (TS) 3.4.1, ``RCS DNB (Pressure, 
Temperature and Flow) Limits.'' Specifically, the proposed changes 
would include (1) a reduction in the minimum primary reactor coolant 
system (RCS) cold leg temperature (Tcold) from 554 F to 535 
F between the 70 percent and 100 percent rated thermal power levels, 
(2) a conversion of the specified RCS minimum flow rate from a ``Mass'' 
(i.e., lb/hr) to a ``Volumetric'' (gpm) flow basis, and (3) elimination 
of the maximum RCS flow rate limit.
    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 to Technical Specification (TS) 3.4.1 does 
not adversely impact structure, system, or component design or 
operation in a manner which would result in a change in the 
frequency of occurrence of accident initiation. Nor are the affected 
parameters themselves accident initiators. As such, the proposed TS 
change will not significantly increase the probability of accidents 
previously evaluated. Likewise, the proposed TS change does not 
significantly increase the consequences of an accident previously 
evaluated. The safety analysis assessments confirm that the existing 
Analyses of Record (AORs) for San Onofre Units 2 and 3 remain valid 
or have been re-analyzed to demonstrate continued compliance with 
applicable Acceptance Criteria.
    The change in Reactor Coolant System (RCS) ``Mass'' flow to 
``Volumetric'' flow is a change in measuring units to be consistent 
with the measure used in the performance of the safety analysis. 
Therefore, there is no impact on any evaluated accidents.
    The elimination of the upper RCS flow limit has no effect on 
Departure from Nucleate Boiling which is a concern at lower flows, 
and the maximum flow that is physically possible is less than the 
current upper limit.
    Therefore, this amendment request does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    2. The proposed change does not create the possibility of a new 
or different kind of accident from any accident previously 
evaluated.
    Tcold is an input parameter used in event analysis, 
it is not an event initiator. No new or different accidents have 
been identified which could result from operating at the proposed 
Tcold. The safety analysis assessments performed confirm 
that the existing safety system settings for San Onofre Units 2 and 
3 remain valid, thereby assuring continued conformance to the 
Acceptance Criteria for all events.
    A change in RCS flow measuring units can not initiate an 
accident, nor can the elimination of an upper RCS flow limit which 
can not be attained.
    Therefore, this amendment 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.
    Updated Final Safety Analysis Report (UFSAR) safety analyses 
have been assessed and remain valid or have been re-analyzed to 
demonstrate continued compliance with applicable Acceptance Criteria 
for operation at the reduced Tcold. All other safety 
limits and safety system settings remain unchanged.
    A change in measuring units for RCS flow does not reduce the 
margin of safety.
    Elimination of an RCS flow limit that can not physically be 
reached does not reduce the margin of safety. The shiftly 
surveillance requirement for maximum flow has no practical basis or 
safety benefit. Additionally, the margin to departure from nuclear 
boiling increases as the flow rate increases.
    Therefore, this amendment 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: July 6, 1998.
    Description of amendment request: The proposed amendment would 
relocate the description of the reactor coolant system design features 
from Technical Specification 5.4 to the Updated Final Safety Analysis 
Report.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. The proposed change does not involve a significant increase 
in the probability or consequences of an accident previously 
evaluated.
    The proposed change relocates the description of the Reactor 
Coolant System design features to the Updated Final Safety Analysis 
Report (UFSAR), a licensee-controlled document. The description of 
the Reactor Coolant System design features, currently a part of the 
UFSAR, is maintained in accordance with 10 CFR 50.59 and 50.71. 
Existing South Texas Project procedures ensure that changes to the 
facility as described in the UFSAR, such as the replacement of the 
steam generators, are reviewed to determine if an unreviewed

[[Page 48267]]

safety question exists. The proposed amendment does not result in 
any hardware or operating procedure changes. The initiators of any 
accident previously evaluated are not affected by the relocation of 
the Reactor Coolant System design features. Therefore, the proposed 
change does not involve a significant increase in the probability or 
consequences of an accident previously evaluated.
    2. The proposed change does not create the possibility of a new 
or different kind of accident from any accident previously 
evaluated.
    The change does not alter the plant configuration or make 
changes in the methods governing plant operation. The proposed 
change does not impose different requirements, and adequate control 
of information will be maintained in accordance with existing 
procedures. The change does not alter assumptions made in the safety 
analysis and licensing basis. Therefore, the proposed change does 
not create the possibility of a new or different kind of accident.
    3. The proposed change does not involve a significant reduction 
in a margin of safety.
    The relocation of a description of Reactor Coolant System design 
features has no impact on any safety analysis assumptions. There are 
no changes to the plant configuration or operating procedures. 
Future changes to the relocated information are governed by existing 
procedures in accordance with 10 CFR 50.59 and 50.71. Consequently, 
there is no significant reduction in a margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the 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.

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: July 6, 1998.
    Description of amendment request: Relocates the Technical 
Specification 3/4.3.3.3 requirements for the Seismic Instrumentation to 
the Technical Requirements Manual.
    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 relocates requirements and surveillances for 
the Seismic Monitoring System that do not meet the criteria for 
inclusion in Technical Specifications as identified in 10 CFR 
50.36(c)(2)(ii). The affected systems and components are not assumed 
to be initiators of analyzed events and are not assumed to mitigate 
accident or transient events. The requirements and surveillances for 
these affected systems and components will be relocated from the 
Technical Specifications to the Technical Requirements Manual, which 
is incorporated in the STP UFSAR and will be maintained pursuant to 
10 CFR 50.59. In addition, the Seismic Monitoring System components 
are addressed in existing surveillance procedures which are also 
controlled by 10 CFR 50.59 and subject to the change control 
provisions imposed by plant administrative procedures, which endorse 
applicable regulations and standards. The associated changes to the 
Index are administrative. Therefore, the change does not involve a 
significant increase in the probability or consequences of an 
accident previously evaluated.
    2. Does the change create the possibility of a new or different 
kind of accident from any accident previously evaluated?
    The proposed change relocates requirements and surveillances for 
the Seismic Monitoring System that do not meet the criteria for 
inclusion in Technical Specifications as identified in 10 CFR 
50.36(c)(2)(ii). The change does not involve a physical alteration 
of the plant (no new or different type of equipment will be 
installed) or make changes in the methods governing normal plant 
operation. The change will not impose different requirements, and 
adequate control of information will be maintained. This change will 
not alter assumptions made in the safety analysis and licensing 
basis. The associated changes to the Index are administrative. 
Therefore, the 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 relocates requirements and surveillances for 
the Seismic Monitoring System, which does not meet the 10 CFR 50.36 
criteria for inclusion in Technical Specifications. The change will 
not reduce a margin of safety because the change has no impact on 
any safety analysis assumptions. In addition, the relocated 
requirements and surveillances for the affected structures, systems, 
components, or variables remain the same as the existing Technical 
Specifications. Since any future changes to these requirements or 
the surveillance procedures will be evaluated per the requirements 
of 10 CFR 50.59, there will be no reduction in a margin of safety. 
The associated changes to the Index are administrative and have no 
potential effect on the margin of safety.
    Therefore, the change does not involve a significant reduction 
in the margin of safety.


    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the 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.

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: July 6, 1998.
    Description of amendment request: Relocates the Technical 
Specification 3/4.7.13 requirements for the Area Temperature Monitoring 
System to the Technical Requirements Manual.
    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 relocates requirements and surveillances for 
Technical Specification 3/4.7.13, which does not meet the criteria 
for inclusion in Technical Specifications as identified in 10 CFR 
50.36(c)(2)(ii). The affected systems and components are not assumed 
to be initiators of analyzed events and are not assumed to mitigate 
accident or transient events. The requirements and surveillances for 
these affected systems and components will be relocated from the 
Technical Specifications to the Technical Requirements Manual, which 
is incorporated in the STP UFSAR and will be maintained pursuant to 
10 CFR 50.59. In addition, the Area Temperature Monitoring System 
components are addressed in existing surveillance procedures which 
are also controlled by 10 CFR 50.59 and subject to the change 
control provisions imposed by plant administrative procedures, which 
endorse applicable regulations and standards. The associated changes 
to the Index are administrative. Therefore, the 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?

[[Page 48268]]

    The proposed change relocates requirements and surveillances for 
the Area Temperature Monitoring System, which does not meet the 
criteria for inclusion in Technical Specifications as identified in 
10 CFR 50.36(c)(2)(ii). The change does not involve a physical 
alteration of the plant (no new or different type of equipment will 
be installed) or make changes in the methods governing normal plant 
operation. The change will not impose different requirements, and 
adequate control of information will be maintained. This change will 
not alter assumptions made in the safety analysis and licensing 
basis. The associated changes to the Index are administrative. 
Therefore, the 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 relocates requirements and surveillances for 
the Area Temperature Monitoring System, which does not meet the 10 
CFR 50.36 criteria for inclusion in Technical Specifications. The 
change will not reduce a margin of safety since it has no impact on 
any safety analysis assumptions. In addition, the relocated 
requirements and surveillances for the affected structure, system, 
component, or variable remain the same as the existing Technical 
Specifications. Since any future changes to these requirements or 
the surveillance procedures will be evaluated per the requirements 
of 10 CFR 50.59, there will be no reduction in a margin of safety. 
The associated changes to the Index are administrative and have no 
potential effect on 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.

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: July 22, 1998.
    Description of amendment request: The proposed amendment would 
revise the Technical Specifications to reflect the steam generator 
water level low-low trip setpoint differences between the existing 
Model E and the replacement Model Delta-94 steam generators for the 
Reactor Trip System and the Engineered Safety Features Actuation System 
instrumentation.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. The proposed change does not involve a significant increase 
in the probability or consequences of an accident previously 
evaluated.
    This proposed change includes changing the low-low steam 
generator water level trip setpoint. The setpoint is being changed 
to enhance the operational flexibility associated with the RSGs 
[replacement steam generators].
    The minimum setpoint change proposed in this request establishes 
controls to ensure that an adequate heat sink is maintained by 
providing an adequate secondary liquid mass to remove primary system 
sensible heat and core decay heat shortly after reactor trip and 
initiating auxiliary feedwater flow for long-term cooling. The 
accidents analyzed for this requirement are the Loss of Non-
Emergency AC Power to the Plant Auxiliaries, Loss of Normal 
Feedwater and Feedwater Line Break transients. These accidents were 
analyzed utilizing the Westinghouse RETRAN model. All acceptance 
criteria were shown to be met for both these events. Therefore, the 
proposed steam generator water level low-low trip setpoint change is 
demonstrated not to result in an increase in the consequences for 
these accidents.
    The steam generator water level low-low trip setpoint is not 
considered a precursor to any of the analyzed accidents, and 
therefore, these proposed changes do not result in an increase in 
the probability or consequences of any 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.
    The proposed setpoint change does not create any new operating 
conditions or modes. The proposed change only revises the actuation 
setpoints for the Reactor Trip System and Engineered Safety Features 
Actuation System. The actions of these systems continue to be 
performed in accordance with existing requirements, which are 
sufficient to ensure plant safety is maintained.
    3. The proposed change does not involve a significant reduction 
in a margin of safety.
    The events potentially affected by the setpoint change in the 
steam generator water level low-low reactor trip (Table 2.2-1, 
Function 13) and ESFAS Auxiliary Feedwater System actuation (Table 
3.3-4, Function 6.d) are the Loss of Normal Feedwater and Feedwater 
System Pipe Break. These events were analyzed and it was 
demonstrated that all acceptance criteria were met for both of these 
events.

    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.

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: July 28, 1998.
    Description of amendment request: The proposed amendment addresses 
the operator action to reduce the steam generator power-operated relief 
valve setpoint consistent with the revised small-break loss-of-coolant 
accident (SBLOCA) analysis for the replacement Delta-94 steam 
generators. The operator action and the associated revised SBLOCA 
analysis are reflected in a proposed revision to the South Texas 
Project Updated Final Safety Analysis Report.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. The proposed change does not involve a significant increase 
in the probability or consequences of an accident previously 
evaluated.
    The proposed operator action associated with the re-analysis of 
the Delta-94 SGs [steam generators] will not result in a significant 
increase in the probability of an accident previously evaluated. The 
initiators of any design basis accident are not affected by this 
operator action. The operator action would facilitate the automatic 
mitigation capability of the SG PORVs [power-operated relief 
valves], and would not initiate the mitigating safety function. The 
operator action will be incorporated into the EOPs [Emergency 
Operating Procedures] and would not be performed until after the 
initiation of an accident. The automatic actuation of the SG PORVs 
is not a new design feature. The effects of inadvertent opening of a 
single steam dump, relief or safety valve are currently analyzed as 
described in Section 15.1.4 of the UFSAR [Updated Final Safety 
Analysis Report]. Consequently, there is no significant impact on 
any previously evaluated accident probabilities.

[[Page 48269]]

    The proposed operator action associated with the re-analysis of 
the Delta-94 SGs does not result in a significant increase in the 
consequences of any accidents previously evaluated. The operator 
action will not adversely affect the integrated ability of the plant 
systems to perform their intended safety functions to mitigate the 
consequences of a small break LOCA [loss-of-coolant accident], or 
any other accident previously evaluated. In fact, the re-analysis 
has demonstrated that the use of the operator action reduces the 
consequences of a small break LOCA in that the Peak Cladding 
Temperature for the most limiting small break LOCA transient is 
reduced and continues to be substantially below the acceptance limit 
of 10 CFR 50.46.
    The operator action does not affect the integrity of any fission 
product barrier such that their function in the control of 
radiological consequences is not affected. The radiological 
consequences for the small break LOCA presented in the UFSAR remain 
unchanged as a result of the proposed operator action.
    2. The proposed change does not create the possibility of a new 
or different kind of accident from any accident previously 
evaluated.
    The proposed license amendment does not create the possibility 
of a new or different kind of accident from any accident previously 
evaluated. The proposed amendment is not the result of any physical 
changes to the existing facility. The operator action does not 
represent a different initiator for any design basis accident and 
does not create new design basis scenarios. Small break LOCA 
mitigation, utilizing a combination of automatic and manual actions, 
is already part of the STP [South Texas Project, Units 1 and 2] 
licensing basis. Written procedures address those operator actions 
required for small break LOCA mitigation. The current STP EOPs have 
an operator action for a steam generator tube rupture (SGTR) similar 
to the operator action for the small break LOCA addressed by this 
proposed license amendment. The operator action for the SGTR is to 
raise the safety-grade SG PORV setpoints. The operator action 
credited in the small break LOCA analysis for the Delta-94 SGs is to 
lower the safety-grade SG PORV setpoints. The purpose of the action 
is to provide a more rapid cooldown of the primary side by 
depressurizing the secondary side during a small break LOCA using 
the steam dumps first, then the SG PORVs, if steam dumps are 
unavailable. The inadvertent operation of a single steam dump, 
relief or safety valve is currently addressed in UFSAR Section 
15.1.4.
    The proposed amendment does not alter any original design 
specification, such as seismic requirements, electrical separation 
requirements and environmental qualification, and is not the result 
of any physical changes to the facility. In addition, the proposed 
amendment does not result in exposure of additional equipment used 
in accident mitigation to an adverse environment beyond that 
currently identified in the UFSAR.
    3. The proposed change does not involve a significant reduction 
in a margin of safety.
    The proposed operator action does not involve a significant 
reduction in the margin of safety. The plant systems required for 
the mitigation of any design basis accidents will continue to be 
able to perform their safety function. In fact, the re-analysis has 
demonstrated that the use of the operator action reduces the 
consequences of a small break LOCA in that the Peak Cladding 
Temperature for the most limiting small break LOCA transient is 
reduced and continues to be substantially below the acceptance 
criteria of 10 CFR 50.46.

    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 No. 50-259, 50-260 and 50-296, 
Browns Ferry Nuclear Plant Units 1, 2, 3, Limestone County, Alabama

    Date of amendment request: June 12, and August 14, 1998.
    Description of amendment request: The proposed amendment would 
revise the technical specifications (TS) for the Browns Ferry Nuclear 
Plant (BFN) Units 1, 2 and 3. The proposed changes would revise 
surveillance frequency of ``once-per-cycle'' surveillance requirements 
(SR) from 18 to 24 months to accommodate a 24-month fuel cycle. The 
licensee also proposed changes to the associated TS Bases (TS-390).
    Basis for proposed no significant hazards consideration 
determination: Tennesee Valley Authority addressed the affected SRs 
into two groups: (1) non-instrument calibration related, and (b) those 
involving instrument calibrations. 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:

Group 1: Non-instrument Calibration Related SRs

    (1) The proposed amendment does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    The proposed amendment changes the surveillance frequency from 
18 months to 24 months for SRs in the Units 2 and 3 TS that are 
normally a function of the refueling interval. In addition, the 
proposed amendment changes the surveillance frequency from 18 months 
to 24 months for those SRs in the Unit 1 TS that control the test 
interval for components and systems that are common to Units 1, 2, 
and 3. Under certain circumstances SR 3.0.2 would allow a maximum 
surveillance interval of 30 months for these SRs. The evaluations in 
Section III [Licensee's June 12, 1998 application, Section III, 
Safety Analysis] have shown that the reliability of protective 
instrumentation and equipment will be preserved for the maximum 
allowable surveillance interval. The proposed changes do not involve 
any change to the design or functional requirements of plant 
systems, and the surveillance test methods will be unchanged. The 
proposed changes will not give rise to any increase in operating 
power level, fuel operating limits, or effluents. In addition, the 
proposed changes will not significantly increase any radiation 
levels. Based on the foregoing considerations and the evaluations 
completed in accordance with the guidance of Generic Letter 91-04, 
it is concluded that the proposed amendment does not involve a 
significant increase in the probability or consequences of an 
accident previously evaluated.
    (2) The proposed amendment does not create the possibility of a 
new or different kind of accident from any accident previously 
evaluated.
    The proposed amendment requires no change to the plant design or 
the mode of operation, for any item of equipment. No new equipment 
is either added or substituted for any existing equipment. Based on 
the Section III [Licensee's June 12, 1998 application, Section III, 
Safety Analysis] evaluations, the extension of surveillance 
intervals is shown to have no significant impact on equipment 
performance. The proposed changes do not create the possibility of 
any new failure mechanisms. Therefore, the proposed amendment does 
not create the possibility of a new or different kind of accident 
from any accident previously evaluated.
    (3) The proposed amendment does not involve a significant 
reduction in a margin of safety.
    The proposed amendment seeks to change surveillance intervals 
from 18 to 24 months. Although the proposed TS changes will result 
in an increase in the interval between surveillance tests, the 
impact on system availability is small based on other, more frequent 
testing or redundant systems or equipment. There is no evidence of 
any failures that would impact the availability of the systems. This 
change does not alter the existing setpoints, TS allowable values or 
analytical limits. The assumptions in the current safety analyses 
are not impacted and the proposed amendment does not reduce a margin 
of safety.
    Therefore, it is concluded that the proposed amendment does not 
involve a significant reduction in a margin of safety.

Group 2: SRs that Involve Instrument Calibrations

    (1) The proposed amendment does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.

[[Page 48270]]

    The proposed amendment changes the surveillance Frequency from 
18 months to 24 months for SRs in the Units 2 and 3 TS that are 
normally a function of the refueling interval. In addition, the 
proposed amendment changes the surveillance Frequency from 18 months 
to 24 months for those SRs in the Unit 1 TS that control the test 
interval for components and systems that are common to Units 1, 2, 
and 3. Under certain circumstances SR 3.0.2 would allow a maximum 
surveillance interval of 30 months for these SRs. The evaluations in 
Section III [Licensee's August 14, 1998 application, Section III, 
Safety Analysis] have shown that the reliability of protective 
instrumentation will be preserved for the maximum allowable 
surveillance interval. The proposed changes do not involve any 
change to the design or functional requirements of plant systems, 
and the surveillance test methods will be unchanged. The proposed 
changes will not give rise to any increase in operating power level, 
fuel operating limits, or effluents. In addition, the proposed 
changes will not significantly increase any radiation levels. Based 
on the foregoing considerations and the evaluations completed in 
accordance with the guidance of Generic Letter 91-04, it is 
concluded that the proposed amendment does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    (2) The proposed amendment does not create the possibility of a 
new or different kind of accident from any accident previously 
evaluated.
    The proposed amendment requires no change to the plant design or 
the mode of operation, for any item of equipment. The proposed 
changes do not create the possibility of any new failure mechanisms. 
Therefore, the proposed amendment does not create the possibility of 
a new or different kind of accident from any accident previously 
evaluated.
    (3) The proposed amendment does not involve a significant 
reduction in a margin of safety.
    The proposed amendment seeks to change instrument calibration 
surveillance intervals from 18 to 24 months. The primary 
consideration relative to safety margin is that of exceeding 
analytical limits for the current safety analyses as a result of 
increased instrument drift over the extended surveillance interval. 
The drift studies discussed in Section III.A have shown that the 
existing setpoints and TS allowable values can be retained without 
challenging the current analytical limits; thereby preserving the 
assumptions in the current safety analyses and ensuring that safety 
limits will not be exceeded.
    To confirm that the drift errors remain within projected values, 
instruments subjected to the longer interval between calibrations 
will continue to be monitored as required by current plant 
procedures. This practice will assure that no significant reduction 
in safety margin is incurred by adoption of the proposed amendment.
    Therefore, it is concluded that the proposed amendment does not 
involve a significant reduction in a margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
its review, it appears that the three standards of 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: Athens Public Library, 405 E. 
South Street, Athens, Alabama 35611
    Attorney for licensee: General Counsel, Tennessee Valley Authority, 
400 West Summit Hill Drive, ET l0H, Knoxville, Tennessee 37902.
    NRC Project Director: Frederick J. Hebdon.

Tennessee Valley Authority, Docket Nos. 50-327 and 50-328, Sequoyah 
Nuclear Plant, Units 1 and 2, Hamilton County, Tennessee

    Date of application for amendments: August 22,1996 (TS 97-04), as 
supplemented on August 27, 1998.
    Brief description of amendments: The amendments would change the 
Sequoyah (SQN) Technical Specifications (TS) by extending the emergency 
diesel generator allowed outage time from 72 hours to 7 days. This 
amendment request was previously noticed on October 9, 1996 (61 FR 
52969). The scope of the amendment request was changed by the August 
27, 1998 submittal.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), Tennessee Valley 
Authority (TVA), the licensee, has provided its analysis of the issue 
of no significant hazards consideration, which is presented below:

    TVA has concluded that operation of SQN Units 1 and 2, in 
accordance with the proposed change to the TSs [Technical 
Specifications] and operating licenses, does not involve a 
significant hazards consideration. TVA's conclusion is based on its 
evaluation, in accordance with 10 CFR 50.91(a)(1), of the three 
standards set forth in 10 CFR 50.92(c).
    A. The proposed amendment does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    The EDGs [emergency diesel generators] supply backup power to 
the essential safety systems in the event of a loss-of-offsite 
(normal) power. The EDGs are not postulated to be an initiator of a 
design basis accident. The requested change to provide a 7-day AOT 
[allowed outage time] for the EDGs and the deletion of the 
additional 72-hour extension for this AOT will not impact the plant 
design, components or operational practices. The increased out-of-
service time does not invalidate assumptions used in evaluating the 
radiological consequences of an accident and does not provide a new 
or altered release path. In addition, the administrative changes to 
delete EDG reporting requirements and an obsolete License Condition 
will not impact plant equipment or operating practices. Therefore, 
this change does not involve an increase in the probability of any 
accident previously evaluated.
    An increase in the AOT for the EDGs would not change the 
conditions, operating configuration, or minimum amount of operable 
equipment assumed in the plant Final Safety Analysis Report for 
accident mitigation. The longer AOT would provide a longer time 
window for maintenance, but would lesson the overall EDG 
unavailability, therefore, it would reduce plant risk. The CDF [core 
damage frequency] associated with a 7-day AOT increases from the 
base case in the SQN [Sequoyah Nuclear Plant] IPE [individual plant 
examination] but is not risk-significant. This CDF increase is based 
on sensitivity studies performed in accordance with the guidance in 
Draft Regulatory Guide DG-1065, dated June 1997. These studies 
assume additional unavailability of the EDGs for an increase in AOT 
even though plant practices are not expected to change. The EDG 
availability improvements and CDF reductions during 12- and 6-year 
maintenance activities compensates for this potential increase to 
provide an overall safety benefit.
    The deletion of the footnote for extending the AOT for fuel tank 
cleaning removes inappropriate extensions of EDG out-of-service 
time. SQN's implementation of the Maintenance Rule, 10 CFR 50.65, 
also supports the proper scheduling and performance of maintenance 
activities to ensure EDG unavailability is adequately controlled. 
Based on no change in plant risk during routine maintenance, because 
work activity durations are unchanged, and the decrease in overall 
plant risk during the 12-and 6-year maintenance activities, as a 
result of the 7-day EDG action time, this change will not result in 
a significant increase in the consequences of an accident. In 
addition, the administrative deletions of reporting requirements 
that are not necessary based on Maintenance Rule implementation and 
obsolete License Condition deletion will not increase the 
consequences of an accident.
    B. The proposed amendment does not create the possibility of a 
new or different kind of accident from any accident previously 
evaluated.
    The proposed change to extend the AOT for the EDGs and delete 
unnecessary TS and operating license provisions does not alter the 
physical design or configuration of the plant. The EDG operation 
remains unchanged, therefore, this change does not create the 
possibility of a new or different kind of accident from any 
previously analyzed.
    C. The proposed amendment does not involve a significant 
reduction in a margin of safety.
    The proposed extension of the EDG action time for inoperable 
units to 7 days will not alter plant equipment, setpoints or 
operating practices that provide the necessary margin of safety. The 
extension will reduce EDG unavailability and plant risk such that 
the

[[Page 48271]]

EDG's ability to react to accident situations is increased. Overall 
CDF, as a result of a 7-day AOT, indicates a slight increase but it 
is not significant. The AOT extension deletion for fuel tank 
cleaning is a conservative change to maintain appropriate EDG out-
of-service times. The deletions of administrative requirements for 
reporting EDG reliability and obsolete License Conditions do not 
impact functions that maintain the margins of safety and have been 
or are continuing to be satisfied by other regulatory requirements. 
Therefore, the proposed change does not involve a significant 
reduction in the margin of safety.

    The NRC has reviewed the licensee's analysis and, based on this 
review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
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.

Union Electric Company, Docket No. 50-483, Callaway Plant, Unit 1, 
Callaway County, Missouri

    Date of application request: June 29, 1998.
    Description of amendment request: The amendment would revise 
technical specification 3.7.1.7 to (1) address operability of all four 
atmospheric steam dump (ASD) lines, (2) retain an action statement for 
excessive ASD seat leakage, and (3) incorporate action statements for 
multiple inoperable ASD lines.
    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.
    Revising the LCO to refer to the ASD lines rather than the ASD 
valves; requiring four ASD lines to be operable rather than three; 
limiting the LCO 3.0.4 exception to one ASD line inoperable; and 
adding a surveillance for the manual isolation valves constitutes a 
more restrictive change from the current Specification. The proposed 
changes impose more stringent requirements to ensure that ASD 
operability is maintained consistent with the safety analysis and 
licensing basis, and also to address all potential single failure 
scenarios.
    Therefore these changes do not involve a significant increase in 
the probability or consequences of an accident previously evaluated.
    When two ASD lines are inoperable due to causes other than 
excessive ASD seat leakage, the proposed change increases the 
allowed outage time for restoration of all but one required ASD line 
from 24 hours to 72 hours. The increase in time is not significant 
when balanced against the availability of the condenser steam dump 
system and/or the main steam safety valves, and the low probability 
of an event occurring during the restoration period that would 
require the ASD lines. Therefore the increase in allowed outage time 
for restoration of all but one ASD line does not involve a 
significant increase in the probability or consequences of an 
accident previously evaluated.
    The proposed change revising the required completion time from 
hot standby to hot shutdown from six hours to twelve hours is 
consistent with NUREG-1431, Rev. 1, where the required completion 
time to shut the plant down is revised to achieving hot standby in 
six hours and hot shutdown within the following twelve hours. The 
proposed change does not alter the plant configuration or operation 
or the function of any safety system. Consequently, the change does 
not increase the probability of an accident as defined in the 
accident analysis. The proposed change permits a longer time to 
cooldown to RHR entry conditions; however, this would not affect the 
consequences of any postulated accidents and is appropriate due to 
the need to avoid any transients while cooling down. Therefore the 
proposed change would not involve a significant increase in the 
probability or consequences of an accident.
    Therefore, it is concluded that all of the above-proposed 
changes do not significantly increase the probability or 
consequences of any accident previously evaluated.
    2. The proposed change does not create the possibility of a new 
or different kind of accident from any accident previously 
evaluated.
    Revising the LCO to refer to the ASD lines rather than the ASD 
valves; requiring four ASD lines to be operable rather than three; 
limiting the LCO 3.0.4 exception to one ASD line inoperable; and 
adding a surveillance for the manual isolation valves does not 
involve a physical alteration of the plant (no new or different type 
of equipment will be installed) or changes in controlling 
parameters. The proposed change does impose different requirements. 
However, these changes are consistent with assumptions made in the 
safety analysis and licensing basis. Thus, this change does not 
create the possibility of a new or different kind of accident from 
any accident previously evaluated.
    When two ASD lines are inoperable due to causes other than 
excessive ASD seat leakage, the proposed change increases the 
allowed outage time for restoration of all but one required ASD line 
from 24 hours to 72 hours. The increase in time is not significant 
when balanced against the availability of the condenser steam dump 
system and/or the main steam safety valves, and the low probability 
of an event occurring during the restoration period that would 
require the ASD lines. The increase in the allowed outage time does 
not result in a condition not previously considered or analyzed, and 
therefore does not create the possibility of a new or different kind 
of accident.
    The proposed change revising the required completion time from 
hot standby to hot shutdown from six hours to twelve hours is 
consistent with NUREG-1431, Rev. 1, where the required completion 
time to shut the plant down is revised to achieving hot standby in 
six hours and hot shutdown within the following twelve hours. The 
proposed change does not require physical alteration to any plant 
system or change the method by which any safety-related system 
performs its function. The change does allow additional time to 
complete the transfer from the steam generator method for heat 
removal to the RHR system, but does not alter the basic methodology. 
Therefore, the proposed change would not create the possibility of a 
new or different kind of accident.
    All of the proposed changes discussed above do not create the 
potential for a new or previously unanalyzed accident.
    3. The proposed change does not involve a significant reduction 
in a margin of safety.
    Revising the LCO to refer to the ASD lines rather than the ASD 
valves; requiring four ASD lines to be operable rather than three; 
limiting the LCO 3.0.4 exception to one ASD line inoperable; and 
adding a surveillance for the manual isolation valves imposes more 
stringent requirements. These requirements either have no impact on 
or increase the margin of safety by increasing the scope of the 
specification to include additional plant equipment; by adding 
additional requirements; and by imposing a new surveillance. The 
change is consistent with the safety analysis and licensing basis, 
and does not involve a reduction in a margin of safety.
    When two ASD lines are inoperable due to causes other than 
excessive seat leakage, the proposed change increases the allowed 
outage time for restoration from 24 hours to 72 hours. The increase 
in time is not significant when balanced against the availability of 
the condenser steam dump system and/or the main steam safety valves, 
and the low probability of an event occurring during the restoration 
period that would require the ASD lines. The increase in the allowed 
outage time does not result in a condition not previously considered 
and does not involve a significant reduction in a margin of safety.
    The proposed change revising the required completion time from 
hot standby to hot shutdown from six hours to twelve hours is 
consistent with NUREG-1431, Rev. 1, where the required completion 
time to shut the plant down is revised to achieving hot standby in 
six hours and hot shutdown within the following twelve hours. The 
change does not alter the basic regulatory requirements or change 
any accident analysis assumptions, initial conditions or results. 
Therefore, the proposed change would have no significant adverse 
effect on margins of safety.
    None of the proposed changes have any significant adverse effect 
on margins of safety.


[[Page 48272]]


    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: Elmer Ellis Library, 
University of Missouri, Columbia Missouri 65201.
    Attorney for licensee: John O'Neill, Esq., Shaw, Pittman, Potts & 
Trowbridge, 2300 N Street, N.W., Washington, D.C. 20037.
    NRC Project Director: William H. Bateman.

Virginia Electric and Power Company, Docket Nos. 50-338 and 50-339, 
North Anna Power Station, Units No. 1 and No. 2, Louisa County, 
Virginia

    Date of amendment request: July 28, 1998.
    Description of amendment request: The North Anna Power Station 
(NAPS), Unit 1 and 2, Technical Specifications (TS) Surveillance 
Requirement (SR) 4.6.2.2.1.b requires verification, during 
recirculation flow, that each outside recirculation spray (ORS) pump 
develops a discharge pressure of greater than or equal to 115 pounds 
per square inch (psig) and that each Casing Cooling pump develops a 
discharge pressure of greater than or equal to 58 psig for Unit 1 and 
46 psig for Unit 2 when tested. The proposed changes will revise the 
testing acceptance criteria being verified from discharge pressure to 
the required developed head. The frequency of testing shall be in 
accordance with the Inservice Testing Program.
    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:

    Virginia Electric and Power Company has reviewed the 
requirements of 10 CFR 50.92 as they relate to the proposed changes 
for the North Anna Units 1 and 2 and determined that the changes do 
not pose a significant hazards consideration * * * Specifically, 
operation of the North Anna Power Station in accordance with the 
proposed Technical Specification changes will not:
    (a) Involve a significant increase in the probability or 
consequences of an accident as previously evaluated
    The applicable UFSAR [Updated Final Safety Analysis Report] 
accidents previously evaluated are the LOCA [loss-of-coolant 
accident] and MSLB [main steamline break]. The proposed changes 
ensure that the Casing Cooling and ORS pumps will perform properly 
with no unacceptable degradation by using the correct pump test 
acceptance criteria as controlled by the PT program. This does not 
increase the probability of a LOCA or MSLB.
    (b) Create the possibility of a new or different type from any 
accident previously evaluated
    The proposed changes to the Technical Specifications will ensure 
that the Casing Cooling and ORS pumps are tested at the frequency 
established by the lnservice Testing Program to confirm their 
ability to provide design basis flow during a LOCA/MSLB. This will 
not result in any physical alteration to any plant system, nor would 
there be a change in the method by which any safety related system 
performs its function. The design and operation of the Casing 
Cooling and ORS systems are not being changed. Also, the proposed 
changes do not affect the design, operation or failure modes of the 
Casing Cooling and ORS pumps and other components within the Casing 
Cooling and ORS systems. Therefore, the proposed changes do not 
create the possibility of a new or different kind of accident from 
any accident previously evaluated.
    (c) Involve a significant reduction in a margin of safety
    Implementation of the proposed changes ensures that the Casing 
Cooling and ORS pumps do not operate with unacceptable degraded 
flows during a LOCA/MSLB that are less than their containment 
analysis design basis flow. Therefore, the proposed changes would 
not reduce the margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: The Alderman Library, Special 
Collections Department, University of Virginia, Charlottesville, 
Virginia 22903-2498.
    Attorney for licensee: Michael W. Maupin, Esq., Hunton and 
Williams, Riverfront Plaza, East Tower, 951 E. Byrd Street, Richmond, 
Virginia 23219.
    NRC Project Director: Pao-Tsin Kuo, Acting.

Virginia Electric and Power Company, Docket Nos. 50-338 and 50-339, 
North Anna Power Station, Units No. 1 and No. 2, Louisa County, 
Virginia

    Date of amendment request: July 28, 1998.
    Description of amendment request: The North Anna Power Station 
(NAPS), Unit 1 and 2, Technical Specifications (TS) Surveillance 
Requirements (SR) 4.8.1.1.2.a.4, 4.8.1.1.2.c, 4.8.1.1.2.d.2, 
4.8.1.1.2.d.4.b, 4.8.1.1.2.d.5, 4.8.1.1.2.d.6.b, 4.8.1.1.2.d.11.b, and 
4.8.1.1.2.e currently require each Emergency Diesel Generator (EDG) to 
be demonstrated OPERABLE by the performance of specific Surveillance 
Requirements. One significant part of demonstrating operability of the 
EDG requires verification that the frequency is within a specified 
range, which is currently 60 plus or minus 1.2 Hz. The proposed changes 
would change the frequency limit from 60 plus or minus 1.2 Hz to 60 
plus or minus 0.5 Hz and separate the requirement of the EDG start from 
the steady state voltage and frequency limits.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    Virginia Electric and Power Company has reviewed the proposed 
Technical Specification changes against the requirements of 10 CFR 
50.92 and has determined that the proposed changes would not pose a 
significant hazards consideration. Specifically, operation of the 
North Anna Power Station in accordance with the proposed Technical 
Specifications changes will not:
    1. Involve a significant increase in the probability or 
consequences of an accident previously evaluated.
    The proposed change provides a more stringent requirement for 
the EDG frequency limit at steady state operation of 60 [plus or 
minus] 0.5 Hz from the current 60 [plus or minus] 1.2 Hz. The change 
additionally provides a separation of the start requirements from 
the steady state limits for voltage and frequency. The change to the 
EDG frequency limit does not result in operation that will increase 
the probability of initiating an analyzed event and does not alter 
assumptions relative to mitigation of an accident or transient 
event. The change to the frequency limit is acceptable because the 
safety analyses assumptions for emergency power limits the frequency 
variations to 60 [plus or minus] 0.5 Hz and assumes that the EDG 
supplies the emergency bus with electrical power within 10 seconds 
of receiving an emergency start signal. The EDG output breaker will 
close with no electrical power applied to the emergency bus when the 
EDG output reaches 95% of rated voltage. The minimum frequency 
requirement of 59.5 Hz is based on the steady state limit for the 
EDG. The EDG supplies the electrical power for the required 
equipment to mitigate the consequences of design basis events. The 
minimum voltage and frequency (3740 volts and 59.5 Hz) limits ensure 
that the ESF [engineered safety feature] equipment is maintained 
with the required electrical power to mitigate the consequences of 
an accident previously evaluated. Therefore, this change does not 
involve a significant increase in the probability or consequences of 
an accident previously evaluated.
    2. Create the possibility of a new or different type from any 
accident previously evaluated.

[[Page 48273]]

    The proposed change provides a more stringent requirement for 
the EDG frequency at steady state operation of 60 [plus or minus] 
0.5 Hz from the current 60 [plus or minus] 1.2 Hz. The change 
additionally provides a separation of the start requirements from 
the steady state limits for voltage and frequency. The change does 
not introduce a new mode of plant operation and does not involve 
physical modification to the plant. The proposed change does impose 
different requirements. However, these changes are consistent with 
the assumptions in the safety analyses. Thus, this change does not 
create the possibility of a new or different kind of accident from 
any accident previously evaluated.
    3. Involve a significant reduction in a margin of safety.
    The proposed change provides a more stringent requirement for 
the EDG frequency at steady state operation of 60 [plus or minus] 
0.5 Hz from the current 60 [plus or minus] 1.2 Hz. The change 
additionally provides a separation of the start requirements from 
the steady state limits for voltage and frequency. The change to the 
frequency limit is acceptable because the safety analyses 
assumptions for emergency power limits the frequency variations to 
60 [plus or minus] 0.5 Hz and assumes that the EDG supplies the 
emergency bus with electrical power within 10 seconds of receiving 
an emergency start signal. The EDG output breaker will close with no 
electrical power applied to the emergency bus when the EDG output 
reaches 95% of rated voltage. The minimum frequency requirement of 
59.5 Hz is based on the steady state limit for the EDG.
    The EDG supplies the electrical power for the required equipment 
to mitigate the consequences of design basis events. The minimum 
voltage and frequency (3740 volts and 59.5 Hz) limits ensure that 
the ESF equipment will be supplied with the required electrical 
power to mitigate previously evaluated accidents. The margin of 
safety is established through the design of the plant structures, 
systems and components, the parameters within which the plant is 
operated, and the establishment of the setpoints for the actuation 
of equipment relied upon to respond to an event. The change allowing 
the separation of the start requirements from the steady state 
voltage and frequency limits, due to the short time period allowed 
in this condition, does not significantly impact the performance of 
structures; systems or components relied upon for accident 
mitigation or any safety analysis assumptions. Therefore, the 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 request involves no significant hazards consideration.
    Local Public Document Room location: The Alderman Library, Special 
Collections Department, University of Virginia, Charlottesville, 
Virginia 22903-2498.
    Attorney for licensee: Michael W. Maupin, Esq., Hunton and 
Williams, Riverfront Plaza, East Tower, 951 E. Byrd Street, Richmond, 
Virginia 23219.
    NRC Project Director: Pao-Tsin Kuo, Acting.

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.

Duke Energy Corporation, Docket Nos. 50-413 and 50-414, Catawba Nuclear 
Station, Units 1 and 2, York County, South Carolina

    Date of amendment request: August 6, 1998.
    Description of amendment request: The proposed amendments would 
revise Technical Specification (TS) Surveillance Requirement 
4.8.1.1.2.i.2. This requirement is in conflict with a relief granted by 
the NRC staff in February 1995. The deletion of TS Surveillance 
Requirement 4.8.1.1.2.i.2 would remove such a conflict.
    Date of publication of individual notice in Federal Register: 
August 17, 1998 (63 FR 43962).
    Expiration date of individual notice: September 16, 1998.
    Local Public Document Room location: York County Library, 138 East 
Black Street, Rock Hill, South Carolina.

Niagara Mohawk Power Corporation, Docket No. 50-220, Nine Mile Point 
Nuclear Station Unit No. 1, Oswego County, New York

    Date of application for amendment: July 16, 1998. This notice 
supersedes a previous notice (62 FR 40851, published July 30, 1997) 
that was based upon an amendment request dated July 2, 1997. The 
request dated July 2, 1997, was superseded in its entirety by the 
amendment request dated July 16, 1998.
    Brief description of amendment: The amendment would change 
Technical Specification 3/4.2.3 regarding reactor coolant chemistry in 
accordance with a report by Electrical Power Research Institute, Inc. 
TR-103515-R1, ``BWR Water Chemistry Guidelines, 1996 Revision,'' also 
known as Boiling Water Reactor Vessel and Internals Project-29.
    Date of publication of individual notice in Federal Register: 
August 13, 1998 (63 FR 43432).
    Expiration date of individual notice: September 14, 1998.
    Local Public Document Room location: Reference and Documents 
Department, Penfield Library, State University of New York, Oswego, New 
York 13126.

Northern States Power Company, Docket Nos. 50-282 and 50-306, Prairie 
Island Nuclear Generating Plant, Units 1 and 2, Goodhue County, 
Minnesota

    Date of amendment requests: February 27, 1998, as supplemented July 
14, 1998.
    Brief description of amendment requests: The proposed amendments 
would allow a design modification to the existing Anticipated Transient 
Without Scram (ATWS) Mitigation System Actuation Circuitry (AMSAC). The 
design modification would install a Diverse Scram System (DSS) designed 
to meet the requirements of a DSS described by 10 CFR 50.62 (ATWS Rule) 
for non-Westinghouse designed plants and make major modifications to 
the existing AMSAC.
    Date of publication of individual notice in Federal Register: 
August 17, 1998 (63 FR 4365).
    Expiration date of individual notice: September 16, 1998.
    Local Public Document Room location: Minneapolis Public Library, 
Technology and Science Department, 300 Nicollet Mall, Minneapolis, 
Minnesota 55401.

Power Authority of the State of New York, Docket No. 50-333, James A. 
FitzPatrick Nuclear Power Plant, Oswego County, New York

    Date of application for amendment: October 14, 1997, as 
supplemented July 23, 1998.
    Brief description of amendment: The proposed amendment would change 
the James A. FitzPatrick Technical Specifications to provide for 
installation of additional racks to increase spent fuel pool capacity, 
and to correct the maximum exposure dependent, infinite lattice 
multiplication factor for fuel bundles.

[[Page 48274]]

    Date of initial notice in Federal Register: August 24, 1998 (63 FR 
45096).
    Expiration date of individual notice: September 23, 1998.
    Local Public Document Room location: Reference and Documents 
Department, Penfield Library, State University of New York, Oswego, New 
York 13126.

Notice of Issuance of Amendments to Facility Operating Licenses

    During the period since publication of the last biweekly notice, 
the Commission has issued the following amendments. The Commission has 
determined for each of these amendments that the application complies 
with the standards and requirements of the Atomic Energy Act of 1954, 
as amended (the Act), and the Commission's rules and regulations. The 
Commission has made appropriate findings as required by the Act and the 
Commission's rules and regulations in 10 CFR Chapter I, which are set 
forth in the license amendment.
    Notice of Consideration of Issuance of Amendment to Facility 
Operating License, Proposed No Significant Hazards Consideration 
Determination, and Opportunity for A Hearing in connection with these 
actions was published in the Federal Register as indicated.
    Unless otherwise indicated, the Commission has determined that 
these amendments satisfy the criteria for categorical exclusion in 
accordance with 10 CFR 51.22. Therefore, pursuant to 10 CFR 51.22(b), 
no environmental impact statement or environmental assessment need be 
prepared for these amendments. If the Commission has prepared an 
environmental assessment under the special circumstances provision in 
10 CFR 51.12(b) and has made a determination based on that assessment, 
it is so indicated.
    For further details with respect to the action see (1) the 
applications for amendment, (2) the amendment, and (3) the Commission's 
related letter, Safety Evaluation and/or Environmental Assessment as 
indicated. All of these items are available for public inspection at 
the Commission's Public Document Room, the Gelman Building, 2120 L 
Street, NW., Washington, DC, and at the local public document rooms for 
the particular facilities involved.

Commonwealth Edison Company, Docket Nos. STN 50-456 and STN 50-457, 
Braidwood Station, Unit Nos. 1 and 2, Will County, Illinois

    Date of application for amendments: January 14, 1998.
    Brief description of amendments: The amendments revise the 
Technical Specifications to support replacement of the 125 volt direct 
current (Vdc) AT&T batteries with new Charter Power Systems, Inc. (C&D) 
batteries. In addition, the crosstie loading limitation is revised to 
reflect the larger capacity of the C&D batteries.
    Date of issuance: August 18, 1998.
    Effective date: Immediately, to be implemented within 30 days.
    Amendment Nos.: 94 and 94.
    Facility Operating License Nos. NPF-72 and NPF-77: The amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: May 20, 1998 (63 FR 
27758). The Commission's related evaluation of the amendments is 
contained in a Safety Evaluation dated August 18, 1998.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: Wilmington Public Library, 201 
S. Kankakee Street, Wilmington, Illinois 60481.

Detroit Edison Company, Docket No. 50-341, Fermi 2, Monroe County, 
Michigan

    Date of application for amendment: April 2, 1998 (NRC-98-0057).
    Brief description of amendment: The amendment revises Technical 
Specification 3.3.7.5 to permit entering Operational Conditions 1 and 2 
prior to completion of Surveillance Requirements for the primary 
containment hydrogen and oxygen monitors in order to establish the 
conditions necessary (inerted containment) to properly perform the 
calibrations. The amendment also allows an increase in the frequency of 
the calibration for the oxygen monitors from once every 18 months to 
quarterly and corrects the nomenclature for the hydrogen and oxygen 
monitors in tables 3.3.7.5-1 and 4.3.7.5-1.
    Date of issuance: August 20, 1998.
    Effective date: August 20, 1998, with full implementation within 90 
days.
    Amendment No.: 125.
    Facility Operating License No. NPF-43. Amendment revises the 
Technical Specifications.
    Date of initial notice in Federal Register: April 22, 1998 (63 FR 
19968).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated August 20, 1998.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: Monroe County Library System, 
Ellis Reference and Information Center, 3700 South Custer Road, Monroe, 
Michigan 48161.

Detroit Edison Company, Docket No. 50-341, Fermi 2, Monroe County, 
Michigan

    Date of application for amendment: March 27, 1998 (NRC-98-0034), as 
supplemented May 28 and July 31, 1998.
    Brief description of amendment: The amendment revises footnotes 
associated with the emergency core cooling system (ECCS) in Technical 
Specifications 3.5.1, ``ECCS--Operating,'' and 3.5.2, ``ECCS--
Shutdown,'' to indicate that a low pressure coolant injection system 
loop may be considered operable during alignment and operation for 
decay heat removal if it is capable of being manually realigned and is 
not otherwise inoperable. The associated Bases are also revised.
    Date of issuance: August 25, 1998.
    Effective date: August 25, 1998, with full implementation within 90 
days.
    Amendment No.: 126.
    Facility Operating License No. NPF-43: Amendment revises the 
Technical Specifications.
    Date of initial notice in Federal Register: April 22, 1998 (63 FR 
19968). The May 28 and July 31, 1998, letters provided clarifying 
information that was within the scope of the original Federal Register 
notice and did not change the staff's initial proposed no significant 
hazards considerations determination.
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated August 25, 1998.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: Monroe County Library System, 
Ellis Reference and Information Center, 3700 South Custer Road, Monroe, 
Michigan 48161.

Detroit Edison Company, Docket No. 50-341, Fermi 2, Monroe County, 
Michigan

    Date of application for amendment: September 25, 1996 (NRC-96-
0085), as supplemented by letters dated November 26, 1997, and March 10 
and June 17, 1998.
    Brief description of amendment: The amendment revises Surveillance 
Requirement 4.8.4.3 to clarify the situational testing requirement for 
thermal overload devices to indicate that this portion of the 
requirement must be completed upon initial installation of a thermal 
overload device and following any maintenance that could affect its 
performance.
    NRC has also granted the request of Detroit Edison Company to 
withdraw a portion of its September 25, 1996,

[[Page 48275]]

application. The proposed change would have deleted the requirement for 
periodically testing motor-operated valve thermal overload protective 
devices. However, by letter dated June 17, 1998, the licensee withdrew 
this portion of the amendment request. For further details with respect 
to these actions, see the application for amendment dated September 25, 
1996, as supplemented above, and the licensee's letter dated June 17, 
1998, which withdrew this portion of the application for license 
amendment, and the staff's safety evaluation enclosed with the 
amendment. The above documents 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 listed 
below.
    Date of issuance: August 25, 1998.
    Effective date: August 25, 1998, with full implementation within 90 
days.
    Amendment No.: 127.
    Facility Operating License No. NPF-43: Amendment revises the 
Technical Specifications.
    Date of initial notice in Federal Register: October 23, 1996 (61 FR 
55030).
    The November 26, 1997, and March 10 and June 17, 1998, submittals 
provided additional clarifying information within the scope of the 
original Federal Register notice and did not change the staff's initial 
proposed no significant hazards consideration determination.
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated August 25, 1998.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: Monroe County Library System, 
Ellis Reference and Information Center, 3700 South Custer Road, Monroe, 
Michigan 48161.

Duke Energy Corporation, et al., Docket Nos. 50-413 and 50-414, Catawba 
Nuclear Station, Units 1 and 2, York County, South Carolina

    Date of application for amendments: April 8, 1998.
    Brief description of amendments: The amendments revise Technical 
Specification Section 3/4.6.5.1, regarding the ice condenser, to reduce 
the total ice weight from 2,475,252 to 2,330,856 pounds, and to reduce 
individual ice basket ice weight from 1273 to 1199 pounds. The 
associated Bases section is also revised to reflect the changed 
requirements.
    Date of issuance: August 25, 1998.
    Effective date: As of the date of issuance to be implemented within 
30 days from the date of issuance.
    Amendment Nos.: Unit 1--168; Unit 2--160.
    Facility Operating License Nos. NPF-35 and NPF-52: Amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: May 6, 1998 (63 FR 
25107).
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated August 25, 1998.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: York County Library, 138 East 
Black Street, Rock Hill, South Carolina.

Duke Energy Corporation, et al., Docket Nos. 50-413 and 50-414, Catawba 
Nuclear Station, Units 1 and 2, York County, South Carolina

    Date of application for amendments: December 11, 1997.
    Brief description of amendments: The amendments revise Technical 
Specification Table 3.3-4, Engineered Safety Feature Actuation System 
Instrumentation Trip Setpoints, to require that suction of the Nuclear 
Service Water System be swapped from Lake Wylie to the Standby Nuclear 
Service Water Pond at a higher minimum water level of Lake Wylie. 
Specifically, the amendments change the swap setpoint from greater than 
or equal to 554.4 feet to greater than or equal to 557.5 feet, and the 
allowable value from greater than or equal to 552.9 feet to greater 
than or equal to 555.4 feet.
    Date of issuance: August 25, 1998.
    Effective date: As of the date of issuance to be implemented within 
30 days from the date of issuance.
    Amendment Nos.: Unit 1--169; Unit 2--161.
    Facility Operating License Nos. NPF-35 and NPF-52: Amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: February 11, 1998 (63 
FR 6983).
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated August 25, 1998.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: York County Library, 138 East 
Black Street, Rock Hill, South Carolina.

Duke Energy Corporation, et al., Docket Nos. 50-413 and 50-414, Catawba 
Nuclear Station, Units 1 and 2, York County, South Carolina

    Date of application for amendments: September 15, 1997, as 
supplemented by letters dated March 5, April 27, June 15, July 22, and 
August 10, 1998.
    Brief description of amendments: The amendments revise Technical 
Specification Figures 3.4-2 and 3.4-3 (pressure-temperature limits 
curves), Table 4.4-5 (reactor vessel surveillance capsule withdrawal 
schedule), and Sections 3/4.4.9.3 and 3.5.3 (requirements concerning 
overpressure protection). The associated Bases are also revised.
    Date of issuance: August 28, 1998.
    Effective date: As of the date of issuance to be implemented within 
60 days.
    Amendment Nos.: Unit 1--170; Unit 2--162.
    Facility Operating License Nos. NPF-35 and NPF-52: Amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: October 8, 1997 (62 FR 
52580); and July 29, 1998 (63 FR 40553).
    The March 5, April 27, July 22, and August 10, 1998, letters 
provided additional information that did not change the scope of the 
September 15, 1997, application and the initial proposed no significant 
hazards consideration determination.
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated August 28, 1998.
    No significant hazards consideration commets received: No.
    Local Public Document Room location: York County Library, 138 East 
Black Street, Rock Hill, South Carolina.

Duke Energy Corporation, Docket No. 50-287, Oconee Nuclear Station, 
Unit 3, Oconee County, South Carolina

    Date of application of amendment: July 20, 1998.
    Brief description of amendment: The amendment extends, on a one-
time basis, Technical Specification Surveillance 4.18.3 for hydraulic 
and mechanical snubber testing. The tests are required to be performed 
at a frequency of 18 months, with a maximum allowed frequency of 22 
months, 15 days. The amendment extends this to a maximum of 25 months.
    Date of Issuance: August 26, 1998.
    Effective date: As of the date of issuance to be implemented within 
30 days from the date of issuance.
    Amendment No.: 229.

[[Page 48276]]

    Facility Operating License No. DPR-55: The amendment revises the 
Technical Specifications.
    Date of initial notice in Federal Register: July 27, 1998 (63 FR 
40137).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated August 26, 1998.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: Oconee County Library, 501 
West South Broad Street, Walhalla, South Carolina.

Duke Energy Corporation, Docket No. 50-287, Oconee Nuclear Station, 
Unit 3, Oconee County, South Carolina

    Date of application of amendment: July 16, 1998.
    Brief description of amendment: The amendment extends, on a one-
time basis, during Operating Cycle 17, certain specified Technical 
Specification surveillances that are required to be performed at a 
frequency of 18 months from the maximum allowed frequency of 22 months, 
15 days, to a maximum of 24 months.
    Date of Issuance: August 28, 1998.
    Effective date: As of the date of issuance to be implemented within 
30 days from the date of issuance.
    Amendment No.: 230.
    Facility Operating License No. DPR-55: The amendment revises the 
Technical Specifications.
    Date of initial notice in Federal Register: July 29, 1998 (63 FR 
40555)
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated August 28, 1998.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: Oconee County Library, 501 
West South Broad Street, Walhalla, South Carolina.

Duquesne Light Company, et al., Docket Nos. 50-334 and 50-412, Beaver 
Valley Power Station, Unit Nos. 1 and 2 (BVPS-1 and BVPS-2), 
Shippingport, Pennsylvania

    Date of application for amendments: December 19, 1997, as 
supplemented June 16, July 9, and July 15, 1998.
    Brief description of amendments: These amendments revise the 
requirements for the source range neutron flux channels in Modes 2 
(Below P-6), 3, 4, and 5 to incorporate the guidance provided in NUREG-
1431, the NRC's improved Standard Technical Specifications with some 
modifications to address plant-specific design features. This change 
allows (1) the use of alternate detectors provided the required 
functions are provided, and (2) plant cooldown with inoperable 
detectors provided the shutdown margin accounts for the temperature 
change. This change also modifies the BVPS-2 Technical Specification 
(TS) Table 3.3-1 Channels To Trip and Minimum Channels Operable 
requirements to 0 and 1, respectively. This portion of the amendment 
makes these BVPS-2 requirements consistent with the current BVPS-1 
requirements. For both BVPS-1 and BVPS-2, TS Table 4.3-1 is modified to 
include a notation exempting the alternate source range detectors from 
surveillance testing until they are required for operability.
    Date of issuance: August 26, 1998.
    Effective date: Both units, effective immediately, to be 
implemented within 60 days.
    Amendment Nos.: 217 and 94.
    Facility Operating License Nos. DPR-66 and NPF-73: Amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: March 11, 1998 (63 FR 
11918).
    The June 16, July 9, and July 15, 1998, letters provided clarifying 
information that did not change the initial no significant hazards 
consideration determination or expand the amendment request beyond the 
scope of the March 11, 1998, Federal Register notice.
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated August 26, 1998.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: B.F. Jones Memorial Library, 
663 Franklin Avenue, Aliquippa, PA 15001.

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 amendments: June 3, 1998.
    Brief description of amendments: Revise the surveillance 
requirements of TS Section 4.11.2.5.1, Explosive Gas Mixture, to add a 
reference the St. Lucie Units 1 and 2 Updated Final Safety Analysis 
Reports for clarification of an alternative monitoring method to be 
used in the event that continuous monitoring of explosive gas mixtures 
in the waste decay tanks becomes inoperable.
    Date of Issuance: August 10, 1998.
    Effective Date: August 10, 1998, and shall be implemented within 30 
days of receipt.
    Amendment Nos.: 156 and 94.
    Facility Operating License Nos. DPR-67 and NPF-16: Amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: July 1, 1998 (63 FR 
35990).
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated August 10, 1998.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: Indian River Junior College 
Library, 3209 Virginia Avenue, Fort Pierce, Florida 34954-9003.

Florida Power and Light Company, et al., Docket No. 50-389, St. Lucie 
Plant, Unit No. 2, St. Lucie County, Florida

    Date of application for amendment: March 3, 1998.
    Brief description of amendment: This amendment revises the TS in 
three areas. First, the amendment revises TS 3.4.7, Reactor Coolant 
System-Chemistry, to eliminate the need for sampling of reactor coolant 
system chemistry in the defueled condition. Second, the amendment 
revises TS 5.6.1.a.1, Design Features-Fuel Storage-Criticality, to 
reflect the total uncertainty associated with the unborated criticality 
analysis previously approved by NRC. And third, the amendment revises 
TS 6.5.2.9.d, Technical Review Responsibilities, to be consistent with 
the quality assurance process previously approved by NRC.
    Date of Issuance: August 18, 1998.
    Effective Date: As of date of issuance, and shall be implemented 
within 30 days of receipt.
    Amendment No.: 95.
    Facility Operating License No. NPF-16: Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: April 8, 1998 (63 FR 
17224).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated August 18, 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.

Northeast Nuclear Energy Company, et al., Docket No. 50-336, Millstone 
Nuclear Power Station, Unit No. 2, New London County, Connecticut

    Date of application for amendment: April 6, 1998.
    Brief description of amendment: The amendment changes the Technical 
Specifications (TSs) by (1) adding a surveillance requirement to verify

[[Page 48277]]

pressurizer heater capacity to TS 3.4.4, ``Reactor Coolant System--
Pressurizer,'' (2) moving the identification of the location of the 
containment air temperature detectors from the surveillance 
requirements portion of TS 3.6.1.5, ``Containment Systems--Air 
Temperature,'' to the TS Bases for Containment Systems, Section 3/
4.4.6.1.5, ``Air Temperature,'' and (3) modifying the action statements 
and surveillance requirements of TS 3.7.1.5, ``Plant Systems--Main 
Steam Isolation Valves.'' The TS Bases are updated to include the 
location of containment air temperature detectors and reflect the 
changes.
    Date of issuance: August 21, 1998.
    Effective date: As of the date of issuance to be implemented within 
60 days.
    Amendment No.: 219.
    Facility Operating License No. DPR-65: Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: May 6, 1998 (63 FR 
25113).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated August 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.

Northern States Power Company, Docket No. 50-263, Monticello Nuclear 
Generating Plant, Wright County, Minnesota

    Date of application for amendment: June 19, 1998, as supplemented 
on July 1, 1998. The June 19, 1998, submittal superseded in its 
entirety Northern States Power (NSP) Company's previous letters dated 
July 26, 1996, and April 11, 1997. NSP letter dated May 5, 1997, was 
also considered in the staff's review of the amendment request.
    Brief description of amendment: The amendment revises Section 
3.6.C, Coolant Chemistry, and 3/4.17.B, Control Room Emergency 
Filtration System, of the Technical Specifications (TS) to establish TS 
requirements that are consistent with modified analysis inputs used for 
the evaluation of the radiological consequences of a postulated main 
steam line break accident, and of a postulated line break in the 
reactor water cleanup system.
    This amendment request was originally noticed in the Federal 
Register on May 6, 1998 (63 FR 25115).
    Date of issuance: August 28, 1998.
    Effective date: August 28, 1998. Implementation of the license 
conditions shall be as specified in Appendix C to DPR-22.
    Amendment No.: 101.
    Facility Operating License No. DPR-22: Amendment revised the 
License and the Technical Specifications.
    Date of publication of individual notice in Federal Register: July 
28, 1998 (63 FR 40321).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated August 28, 1998.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: Minneapolis Public Library, 
Technology and Science Department, 300 Nicollet Mall, Minneapolis, 
Minnesota 55401.

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: January 17, 1995, as 
supplemented by letters dated March 30, 1995; July 2, 1996; February 28 
and September 22, 1997; and January 23, July 9 and July 29, 1998.
    Brief description of amendment: These amendments revise the 
technical specifications to support the replacement of the Source Range 
and Intermediate Range Monitors with the Wide Range Neutron Monitoring 
System.
    Date of issuance: August 24, 1998.
    Effective date: As of the date of issuance and is to be implemented 
upon completion of Modification P00271.
    Amendment No.: 222.
    Operating License No. DPR-44: Amendment revised the Technical 
Specifications.
    Date of initial notice in Federal Register: June 6, 1995 (60 FR 
29885)
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated August 24, 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

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: December 18, 1997, as 
supplemented July 14, 1998.
    Brief description of amendments: The amendments revise the Unit 1 
and Unit 2 Facility Operating Licenses by modifying or deleting 
obsolete conditions.
    Date of issuance: August 18, 1998.
    Effective date: As of the date of issuance to be implemented within 
30 days.
    Amendment Nos.: Unit 1-212; Unit 2-153.
    Facility Operating License Nos. DPR-57 and NPF-5: Amendments 
revised the Facility Operating Licenses.
    Date of initial notice in Federal Register: January 28, 1998 (63 FR 
4324).
    The July 14, 1998, letter provided clarifying information that did 
not change the scope of the December 18, 1997, application and the 
initial proposed no significant hazards consideration determination.
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated August 18, 1998.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: Appling County Public Library, 
301 City Hall Drive, Baxley, Georgia.

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: October 29, 1996, as 
supplemented February 19, June 20, and October 21, 1997.
    Brief description of amendments: The amendments revise the 
Technical Specifications (TSs) associated with the oscillation power 
range monitor portion of the digital Power Range Neutron Monitoring 
system. The TSs associated with the average power range monitor portion 
of the system were issued on March 21, 1997.
    Date of issuance: August 20, 1998.
    Effective date: As of the date of issuance to be implemented on 
each

[[Page 48278]]

unit prior to the next refueling outage of that unit.
    Amendment Nos.: Unit 1-213; Unit 2-154.
    Facility Operating License Nos. DPR-57 and NPF-5: Amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: January 2, 1997 (62 FR 
130).
    The letters dated February 19, June 20, and October 21, 1997, 
provided clarifying information that did not change the scope of the 
October 29, 1996, application and the initial proposed no significant 
hazards consideration determination.
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated August 20, 1998.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: Appling County Public Library, 
301 City Hall Drive, Baxley, Georgia.

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 application for amendments: September 16, 1997, as 
supplemented by letter dated February 23, 1998.
    Brief description of amendments: The amendments would allow 
sleeving of steam generator tubes with sleeves designed by the vendor, 
ASEA Brown Boveri/Combustion Engineering (ABB/CE). Additionally, the 
proposed TS amendment would require that sleeves be removed from 
service upon detection of service-induced degradation, require post 
weld heat treatment (PWHT) of sleeve welds, and reduce the allowable 
primary-to-secondary leakage through any one steam generator to 150 
gallons per day (gpd).
    Date of issuance: August 26, 1998.
    Effective date: August 26, 1998, to be implemented 30 days from the 
date of issuance.
    Amendment Nos.: Unit 2-140; Unit 3-132
    Facility Operating License Nos. NPF-10 and NPF-15: The amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: January 28, 1998 (63 FR 
4323).
    The February 23, 1998, supplemental letter provided additional 
clarifying information and 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 
August 26, 1998.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: Main Library, University of 
California, P. O. Box 19557, Irvine, California 92713.

Union Electric Company, Docket No. 50-483, Callaway Plant, Unit 1, 
Callaway County, Missouri

    Date of application for amendment: March 9, 1998, as supplemented 
by letter dated July 8, 1998.
    Brief description of amendment: The amendment revises Technical 
Specification 4.5.2b.1 and its associated Bases to eliminate the 
requirement to vent the centrifugal charging pump casings.
    Date of issuance: August 17, 1998.
    Effective date: August 17, 1998, to be implemented within 30 days 
from the date of issuance.
    Amendment No.: 127.
    Facility Operating License No. NPF-30: The amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: May 6, 1998 (63 FR 
25118)
    The July 8, 1998, supplemental letter provided additional 
clarifying information and 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 August 17, 1998.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: University of Missouri-
Columbia, Elmer Ellis Library, Columbia, Missouri 65201-5149.

Virginia Electric and Power Company, et al., Docket Nos. 50-338 and 50-
339, North Anna Power Station, Units No. 1 and No. 2, Louisa County, 
Virginia

    Date of application for amendments: September 1, 1995, as 
supplemented April 8, 1996; April 22, 1996; April 23, 1996; November 
18, 1997; February 9, 1998; March 25, 1998; May 5, 1998; June 25, 1998; 
and June 29, 1998.
    Brief description of amendments: The proposed action would revise 
the Technical Specifications (TS) changing the Emergency Diesel 
Generator (EDG) outage time from 72 hours to 14 days.
    Date of issuance: August 26, 1998.
    Effective date: August 26, 1998.
    Amendment Nos.: 214 and 195.
    Facility Operating License Nos. NPF-4 and NPF-7: Amendments revised 
the Licenses and the Technical Specifications.
    Date of initial notice in Federal Register: June 17, 1998 (63 FR 
33110), which superseded the notice of September 27, 1995 (60 FR 
49949).
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated August 26, 1998.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: The Alderman Library, Special 
Collections Department, University of Virginia, Charlottesville, 
Virginia 22903-2498.

Virginia Electric and Power Company, et al., Docket Nos. 50-280 and 50-
281, Surry Power Station, Units 1 and 2, Surry County, Virginia

    Date of application for amendments: June 19, 1998, as supplemented 
July 14, 1998.
    Brief Description of amendments: These amendments revise the 
Licenses and Technical Specifications (TS) to allow the use of a 
temporary jumper line for providing service water to component cooling 
water heat exchangers while maintenance is performed on existing 
service water supply piping. In addition, editorial changes have been 
made to TS Table 3.7-2, item 3, and to TS Bases Section 3.14.
    Date of issuance: August 26, 1998.
    Effective date: August 26, 1998.
    Amendment Nos.: 216 and 216.
    Facility Operating License Nos. DPR-32 and DPR-37: Amendments 
change the License and Technical Specifications.
    Date of initial notice in Federal Register: July 14, 1998 (63 FR 
38206). The Commission's related evaluation of the amendments is 
contained in a Safety Evaluation dated August 26, 1998.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: Swem Library, College of 
William and Mary, Williamsburg, Virginia 23185.

    Dated at Rockville, Maryland, this 2nd day of September 1998.

    For the Nuclear Regulatory Commission.

Elinor G. Adensam,
Acting Director, Division of Reactor Projects--III/IV, Office of 
Nuclear Reactor Regulation.
[FR Doc. 98-24130 Filed 9-8-98; 8:45 am]
BILLING CODE 7590-01-P