[Federal Register Volume 63, Number 97 (Wednesday, May 20, 1998)]
[Notices]
[Pages 27757-27772]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: 98-13223]



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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 Pub. L. 97-415, the U.S. Nuclear Regulatory Commission 
(the Commission or NRC staff) is publishing this regular biweekly 
notice. Pub. L. 97-415 revised section 189 of the Atomic Energy Act of 
1954, as amended (the Act), to require the Commission to publish notice 
of any amendments issued, or proposed to be issued, under a new 
provision of section 189 of the Act. This provision grants the 
Commission the authority to issue and make immediately effective any 
amendment to an operating license upon a determination by the 
Commission that such amendment involves no significant hazards 
consideration, notwithstanding the pendency before the Commission of a 
request for a hearing from any person.
    This biweekly notice includes all notices of amendments issued, or 
proposed to be issued from April 25, 1998, through May 8, 1998. The 
last biweekly notice was published on May 6, 1998 (63 FR 25101).

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 June 19, 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.

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    Those permitted to intervene become parties to the proceeding, 
subject to any limitations in the order granting leave to intervene, 
and have the opportunity to participate fully in the conduct of the 
hearing, including the opportunity to present evidence and cross-
examine witnesses.
    If a hearing is requested, the Commission will make a final 
determination on the issue of no significant hazards consideration. The 
final determination will serve to decide when the hearing is held.
    If the final determination is that the amendment request involves 
no significant hazards consideration, the Commission may issue the 
amendment and make it immediately effective, notwithstanding the 
request for a hearing. Any hearing held would take place after issuance 
of the amendment.
    If the final determination is that the amendment request involves a 
significant hazards consideration, any hearing held would take place 
before the issuance of any amendment.
    A request for a hearing or a petition for leave to intervene must 
be filed with the Secretary of the Commission, U.S. Nuclear Regulatory 
Commission, Washington, DC 20555-0001, Attention: Rulemakings and 
Adjudications Staff, or may be delivered to the Commission's Public 
Document Room, the Gelman Building, 2120 L Street, NW., Washington DC, 
by the above date. A copy of the petition should also be sent to the 
Office of the General Counsel, U.S. Nuclear Regulatory Commission, 
Washington, DC 20555-0001, and to the attorney for the licensee.
    Nontimely filings of petitions for leave to intervene, amended 
petitions, supplemental petitions and/or requests for a hearing will 
not be entertained absent a determination by the Commission, the 
presiding officer or the Atomic Safety and Licensing Board that the 
petition and/or request should be granted based upon a balancing of 
factors specified in 10 CFR 2.714(a)(1)(i)-(v) and 2.714(d).
    For further details with respect to this action, see the 
application for amendment which is available for public inspection at 
the Commission's Public Document Room, the Gelman Building, 2120 L 
Street, NW., Washington, DC, and at the local public document room for 
the particular facility involved.

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

    Date of amendment request: January 14, 1998.
    Description of amendment request: The proposed amendments would 
change the Technical Specifications to allow replacement of the 125 
volt direct current (DC) AT&T batteries with new Charter Power Systems, 
Inc. (C&D) batteries, and revise the crosstie loading limitation.
    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:

    A. The proposed change does not involve a significant increase 
in the probability of consequences of an accident previously 
evaluated.
    The replacement C&D battery has been selected to meet or exceed 
the design, functional, and operational requirements of those of the 
present AT&T battery, including crosstie load limitations. The C&D 
batteries are similar in design to the previously installed Gould 
batteries (e.g. electrolyte specific gravity and construction of the 
plates) except for capacity. The replacement C&D batteries have a 
significantly larger capacity than either the previously installed 
Gould, or the currently installed AT&T, batteries. This increased 
capacity can provide additional margin for future use. Also, the C&D 
batteries are qualified for a 20 year life and meet the latest 
applicable standards. The short circuit current provided by the C&D 
batteries is well within the interrupting capability of the existing 
DC system [c]ircuit breakers.
    Additionally, the crosstie limit is increased to take advantage 
of the larger C&D battery capacity. The C&D batteries were sized 
based on having sufficient capacity to energize the design basis DC 
loads of an operating unit with the [Institute of Electrical and 
Electronics Engineers] IEEE-485 design margin while maintaining the 
desired limited DC load of 200 amps for a shutdown unit. This 
proposed change allows use of the C&D batteries' larger capacity.
    Also, although adherence to the performance testing intervals 
stated in IEEE Std 450 could result in a planned shutdown and 
possible subsequent increase in the probability of occurrence of an 
accident (e.g. Turbine Trip), it would be part of a controlled and 
planned shutdown, therefore the increases would not be considered 
significant.
    The overall design, function, and operation of the DC system and 
equipment has not been altered by these changes. The proposed 
changes do not affect any accident initiators of precursors and do 
not alter the design assumptions for the systems or components used 
to mitigate the consequences of an accident as analyzed in UFSAR 
[Updated Final Safety Analysis Report] Chapter 15. Therefore, there 
is no increase in the probability or consequences of an accident 
previously evaluated.
    B. The proposed change does not create the possibility of a new 
or different kind of accident from any accident previously 
evaluated.
    The replacement C&D batteries will provide the same function as 
those of the installed AT&T batteries and will be operated with the 
same types of operational controls. These limits include battery 
float terminal voltage, individual cell voltage and electrolyte 
specific gravity, and crosstie loading. Crosstie conditions are 
allowed under the present Technical Specifications. The crosstie 
limit is increased to take advantage of the larger C&D battery 
capacity. The remaining changes are administrative in nature or 
provide clarification to maintain consistency with other Technical 
Specifications.
    The DC system and its equipment will continue to perform the 
same function and be operated in the same fashion. The proposed 
changes do not create any new or common failure modes. The proposed 
changes do not introduce any new accident initiators or precursors, 
or any new design assumptions for the systems or components used to 
mitigate the consequences of an accident. Therefore, the possibility 
of a new or different kind of accident from any accident previously 
evaluated has not been created.
    C. The proposed change does not involve a significant reduction 
in a margin of safety.
    The replacement C&D batteries will meet or exceed the design, 
functional, and qualifications of the installed AT&T batteries. The 
proposed Technical Specification limitations for the C&D batteries 
are derived from the same methodology as the AT&T batteries with 
applied margins in accordance with IEEE 485. Increasing the crosstie 
loading limits takes advantage of the larger C&D battery capacity 
with its increased design margin. The proposed change to the 
crosstie loading limit will continue to conservatively envelope the 
postulated design requirements. The remaining changes are 
administrative in nature or provide clarification to maintain 
consistency with other Technical Specifications.
    The inherent design conservatism of the DC system and its 
equipment has not been altered. The DC system and its equipment will 
continue to be operated with the same degree of conservatism. 
Therefore, there is no reduction in the margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
requested amendments involve no significant hazards consideration.
    Local Public Document Room location: Wilmington Public Library, 201 
S. Kankakee Street, Wilmington, Illinois 60481.
    Attorney for licensee: Michael I. Miller, Esquire; Sidley and 
Austin, One First National Plaza, Chicago, Illinois 60603.
    NRC Project Director: Stuart A. Richards.

[[Page 27759]]

Commonwealth Edison Company, Docket Nos. 50-373 and 50-374, LaSalle 
County Station, Units 1 and 2, LaSalle County, Illinois

    Date of amendment request: May 27, 1997, as supplemented on August 
1, 1997, and March 24, 1998.
    Description of amendment request: The proposed amendments would 
revise Technical Specification Section 6, ``Administrative Controls,'' 
to incorporate revised organizational titles and would delete the Unit 
1 License Condition 2.C.(30)(a) related to the function of the Shift 
Technical Advisor. In addition, the proposed amendments would change 
the submittal frequency of the Radiological Effluent Release Report 
from semiannually to annually. The proposed amendments will also make 
several administrative and editorial changes. The staff's proposed no 
significant hazards consideration determination for the requested 
change was published on July 30, 1997 (62 FR 40848).
    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:

    A. The proposed changes do not involve a significant increase in 
the probability or consequences of an accident previously evaluated.
    The proposed changes do not affect any accident initiators or 
precursors and do not change or alter the design assumptions for 
systems or components used to mitigate the consequences of an 
accident. The proposed changes do not affect the design or operation 
of any system, structure, or component in the plant. There are no 
changes to parameters governing plant operation, and, no new or 
different type of equipment will be installed.
    The proposed changes provide clarification, consistency with 
station procedures, programs, the Code of Federal Regulations (10 
CFR), other Technical Specifications, and Improved Technical 
Specifications. These changes do not impact any accident previously 
evaluated in the UFSAR [Updated Final Safety Analysis Report]. There 
is no relaxation of applicable administrative controls. Those 
administrative requirements which have no effect on safe operation 
of the plant are eliminated.
    B. The proposed changes do not create the possibility of a new 
or different kind of accident from any accident previously 
evaluated.
    The proposed changes do not affect the design or operation of 
any plant system, structure, or component. There are no changes to 
parameters governing plant operation, and, no new or different type 
of equipment will be installed.
    C. The proposed changes do not involve a significant reduction 
in a margin of safety.
    The proposed changes do not affect the margin of safety for any 
Technical Specification. The initial conditions and methodologies 
used in the accident analyses remain unchanged; therefore, accident 
analyses results are not impacted. Plant safety parameters or 
setpoints are not affected. All responsibilities described in the 
Technical Specifications for administrative controls will continue 
to be performed by individuals possessing the requisite 
qualifications. Clarifications, relocations, and nomenclature 
changes neither result in a reduction of personnel responsibilities, 
nor do they cause a relaxation of programmatic controls. There are 
no resulting effects on plant safety parameters or setpoints.
    Guidance has been provided in ``Final Procedures and Standards 
on No Significant Hazards Considerations,'' Final Rule, 51 FR 7744, 
for the application of standards to license change requests for 
determination of the existence of significant hazards 
considerations. This document provides examples of amendments which 
are and are not considered likely to involve significant hazards 
considerations. These proposed amendments most closely fit the 
example of a purely administrative change to the Technical 
Specifications to achieve consistency throughout the Technical 
Specifications, correction of an error, or a change in nomenclature.
    The proposed amendment does not involve a significant relaxation 
of the criteria used to establish safety limits, a significant 
relaxation of the bases for the limiting safety system settings, or 
a significant relaxation of the bases for the limiting conditions 
for operations. The proposed change does not reduce the margin of 
safety as defined in the basis for any Technical Specification.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
requested amendments involve no significant hazards consideration.
    Local Public Document Room location: Jacobs Memorial Library, 
Illinois Valley Community College, Oglesby, Illinois 61348.
    Attorney for licensee: Michael I. Miller, Esquire; Sidley and 
Austin, One First National Plaza, Chicago, Illinois 60603.
    NRC Project Director: Stuart A. Richards.

Commonwealth Edison Company, Docket Nos. 50-254 and 50-265, Quad Cities 
Nuclear Power Station, Units 1 and 2, Rock Island County, Illinois.

    Date of amendment request: April 13, 1998.
    Description of amendment request: Unreviewed Safety Question 
involving additional manual actions incorporated in new fire protection 
procedures as a result of a revised Appendix R Safe Shutdown Safety 
Analysis.
    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) No significant increase in the probability or consequences 
of an accident previously evaluated is involved because of the 
following:
    Two types of previously evaluated accidents are relevant to this 
criterion: (1) A fire; (2) other accidents evaluated in the Updated 
Final Safety Analysis Report. For these previously evaluated 
accidents, the change would not result in an increase in either 
their probabilities of occurrence or the consequences of their 
occurrence, for the following reasons:
    The additional operator manual actions do not significantly 
change the probability or consequences of a fire. The likelihood of 
a fire is unchanged. Additional operations do not significantly 
change the fire loading nor introduce significant new ignition 
sources. The quantities and arrangement of combustible materials are 
not changed through additional manual actions.
    The consequences of a fire are unchanged because operator manual 
actions serve to support the station's ability to achieve and 
maintain shutdown in the event of a fire.
    Additional manual operations are for purposes of safe shutdown 
in the event of a fire in areas requiring alternate shutdown 
capability and do not impact other accident scenarios. Also, there 
is no increase in the predicted frequency of other accidents as a 
result of this change. Accordingly there is no significant change in 
the probability or consequences of other accidents previously 
evaluated because they are independent of this change in procedures 
for fire scenarios.
    (2) The possibility of a new or different kind of accident from 
any accident previously evaluated is not created because:
    The proposed change does not create the possibility of a new or 
different kind of accident from that previously evaluated for the 
Quad Cities Station. Although the number of manual actions increased 
and there may be some compression in the time for taking necessary 
actions relative to the current safe shutdown analysis and 
procedures, there is no significant change in the operation of plant 
equipment following the postulated fire event. The existing safe 
shutdown analysis already relies on operator manual actions which 
perform the same type of actions.
    The overall approach and methodology to performing these 
operator actions are not significantly different from the prior 
approach and methodology. This proposed change does not involve an 
accident initiator or failure not previously considered. The results 
or effects of equipment malfunctions

[[Page 27760]]

previously evaluated are unchanged as the result of potential 
operator errors. No new failures would occur, and no new modes of 
operation are introduced by the proposed changes.
    Additional manual actions and the timing thereof provide a 
somewhat different demand on the plant equipment operators, but 
still provide an effective method for achieving and maintaining 
post-fire safe shutdown for areas requiring alternate shutdown 
capability. As such, the proposed changes do not create the 
possibility of a new or different kind of accident.
    (3) No significant reduction in the margin of safety is involved 
because:
    A change in the fire protection program does not result in a 
significant reduction in the margin of safety if the change does not 
result in a significant adverse impact on the plant's ability to 
achieve and maintain safe shutdown in the event of a fire. The 
proposed operator manual actions to achieve and maintain safe 
shutdown in a fire scenario do not significantly affect the 
capability or reliability of the equipment assumed to operate in the 
safety analysis.
    The types of manual actions to be performed in support of 
Appendix R safe shutdown functions are not significantly different 
from those previously considered. The complexity of actions is not 
significantly changed. Indeed many of the additional actions are 
designed to provide additional protection from spurious operations 
which could result from a fire.
    Any reduction in margin associated with changes in the time 
before which certain manual actions must occur is largely a result 
of re-analyses which incorporate conservatisms not previously 
considered. In total, the proposed changes do not adversely impact 
the capability to meet the requirements of Appendix R. Any reduction 
in margin associated with additional manual actions to achieve and 
maintain post fire safe shutdown in areas requiring alternate 
capabilities does not involve a significant reduction in margin.

    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 are 
satisfied. Therefore, the NRC staff proposes to determine that the 
requested amendments involve no significant hazards consideration.
    Local Public Document Room location: Dixon Public Library, 221 
Hennepin Avenue, Dixon, Illinois 61021.
    Attorney for licensee: Michael I. Miller, Esquire; Sidley and 
Austin, One First National Plaza, Chicago, Illinois 60603.
    NRC Project Director: Stuart A. Richards.

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

    Date of amendment request: May 27, 1997, as supplemented by letters 
dated March 9, March 20, and April 20, 1998.
    Description of amendment request: The proposed amendments would 
revise the current Technical Specifications (TS) of each unit to 
conform with NUREG-1431, Revision 1, ``Standard Technical 
Specifications--Westinghouse Plants.'' The Commission had previously 
issued a Notice of Consideration of Issuance of Amendments published in 
the Federal Register on July 14, 1997 (62 FR 37628) covering all the 
proposed changes that were indeed within the scope of NUREG-1431. In 
DEC's March 9, March 20, and April 20, 1998, supplements, there are 
proposed changes that are beyond the scope of NUREG-1431, which were, 
thus, not covered by the staff's July 14, 1997, notice. The following 
descriptions and proposed no significant hazard analyses cover only 
those beyond-scope changes. Associated with each change are 
administrative/editorial changes such that the new or revised 
requirements would fit into the format of NUREG-1431.
    1. Table 3.3-3 of the current TS contains an entry regarding the 
Containment Pressure Control System, allowing an inoperable channel be 
placed in trip in 1 hour. DEC proposed to tighten this requirement such 
that the system supported by the inoperable channel be declared 
inoperable immediately. No changes to the design of the Containment 
Pressure Control System or other systems were proposed by DEC.
    2. Table 4.3-1 of the Unit 1 current TS has a footnote (No. 13) 
that specifies a filter time constant of 1.5 seconds in the steam 
generator low-low level reactor trip circuitry. DEC proposed to delete 
this time constant since it was never used. No design changes to the 
instrumentation and control systems are involved.
    3. Section 4.5.1.1.c of the current TS requires that power be 
removed from the accumulator isolation valve when the reactor coolant 
system pressure is greater than 2000 pounds per square inch gauge 
(psig). DEC proposed to make this requirement more restrictive, 
lowering this threshold to 1000 psig on the recommendation of the 
nuclear vendor, Westinghouse. No design changes to the accumulator 
system are involved.
    4. Section 4.6.5.1.b.1 of the current TS requires that the boron 
concentration of the ice in the ice condenser be verified once every 9 
months to be at least 1800 ppm. DEC proposed to relax the frequency 
from 9 months to 18 months on the basis that boron, in the form of 
sodium tetraborate, does not decrease in quantity even though the ice 
sublimates. No design changes to the ice condenser are involved.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), DEC has provided its 
analyses of the issue of no significant hazards consideration for each 
of the above proposed changes. The NRC staff has reviewed DEC's 
analyses against the standards of 10 CFR 50.92(c). The NRC staff's 
analysis is presented below.
    1. Will the changes involve a significant increase in the 
probability or consequences of an accident previously evaluated?
    For all the changes the answer is ``no.'' The proposed changes will 
not affect the safety function of the subject systems. There will be no 
direct effect on the design or operation of any plant structures, 
systems, or components. No previously analyzed accidents were initiated 
by the functions of these systems, and the systems were not factors in 
the consequences of previously analyzed accidents. Therefore, the 
proposed changes will have no impact on the consequences or 
probabilities of any previously evaluated accidents.
    2. Will the changes create the possibility of a new or difference 
kind of accident from any accident previously evaluated? For all the 
changes the answer is ``no.'' The proposed changes would not lead to 
any hardware or operating procedure change. Hence, no new equipment 
failure modes or accidents from those previously evaluated will be 
created.
    3. Will the changes involve a significant reduction in a margin of 
safety?
    For all the changes the answer is ``no.'' Margin of safety is 
associated with confidence in the design and operation of the plant. 
The proposed changes to the TS do not involve any change to plant 
design, operation, or analysis. Thus, the margin of safety previously 
analyzed and evaluated is maintained.
    Based on this analysis, it appears that the three standards of 10 
CFR 50.92(c) are satisfied for each of the proposed changes. Therefore, 
the NRC staff proposes to determine that the amendment request involves 
no significant hazards consideration.
    Local Public Document Room location: York County Library, 138 East 
Black Street, Rock Hill, South Carolina.

[[Page 27761]]

    Attorney for licensee: Mr. Paul R. Newton, Legal Department 
(PB05E), Duke Energy Corporation, 422 South Church Street, Charlotte, 
North Carolina.
    NRC Project Director: Herbert N. Berkow

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

    Date of amendment request: April 20, 1998.
    Description of amendment request: The Control Room Area Ventilation 
System (CRAVS) can be actuated by a number of ways, including by the 
engineered safety features actuation signal (ESFAS) when safety 
injection is also initiated. The only relationship between automatic 
actuation of the CRAVS and the ESFAS is through safety injection 
initiation, applicable in Modes 1, 2, 3, and 4. However, in Tables 3.3-
3 and 4.3-2 of the units' Technical Specifications, regarding 
operability and surveillance requirements, the CRAVS automatic 
actuation has been erroneously specified for all modes (Modes 1, 2, 3, 
4, 5, and 6). The licensee proposed to correct this error by the 
proposed amendment.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

First Standard

    Implementation of this amendment would not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated. The Control Room Area Ventilation System and 
ESFAS are not accident initiating systems; they are accident 
mitigating systems. Therefore, changing the mode requirements for 
the subject ESFAS functional unit cannot impact accident initiating 
probabilities. The technical justification associated with this 
proposed amendment shows that the current Technical Specification 
mode requirements for the subject functional unit are incorrect as 
written. The Control Room Area Ventilation System and ESFAS will 
remain fully capable of performing their design accident mitigation 
functions for the modes in which they are required. The Control Room 
Area Ventilation System operability requirement of Technical 
Specification 3/4.7.6 will continue to be met. Therefore, no 
accident consequences will be impacted.

Second Standard

    Implementation of this amendment would not create the 
possibility of a new or different kind of accident from any accident 
previously evaluated. As noted previously, the Control Room Area 
Ventilation System and ESFAS are not accident initiating systems. 
Correcting the mode requirements as specified will not impact any 
plant systems that are accident initiators. No other modifications 
are being proposed to the plant which would result in the creation 
of new accident mechanisms. Also, no changes are being made to the 
way in which the plant is operated, so no new failure mechanisms 
will be initiated.

Third Standard

    Implementation of this amendment would not involve a significant 
reduction in a margin of safety. Margin of safety is related to the 
confidence in the ability of the fission product barriers to perform 
their design functions during and following an accident situation. 
These barriers include the fuel cladding, the reactor coolant 
system, and the containment system. The performance of the fission 
product barriers will not be impacted by implementation of this 
proposed amendment. Both the Control Room Area Ventilation System 
and the ESFAS will remain fully capable of performing their design 
functions for the modes in which they are required. Therefore, no 
safety margin will be significantly impacted.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: York County Library, 138 East 
Black Street, Rock Hill, South Carolina. Attorney for licensee: Mr. 
Paul R. Newton, Legal Department (PB05E), Duke Energy Corporation, 422 
South Church Street, Charlotte, North Carolina.
    NRC Project Director: Herbert N. Berkow.

Duke Energy Corporation (DEC), Docket Nos. 50-369 and 50-370, McGuire 
Nuclear Station, Units 1 and 2, Mecklenburg County, North Carolina

    Date of amendment request: May 27, 1997, as supplemented by letter 
dated March 9, 1998.
    Description of amendment request: The three proposed changes are 
associated with DEC's application to convert to the Improved Technical 
Specifications. The first change would increase the surveillance 
interval for the boron concentration of the ice bed from once per 9 
months, to every 18 months. This change is supported by operating 
experience data, establishes surveillance intervals that coincide with 
refueling outages, and minimizes containment entries during power 
operation. The second change would decrease the Reactor Coolant System 
pressure level at which power is removed from the accumulator isolation 
valve from 2000 pounds per square inch gauge (psig) to 1000 psig. This 
change is considered a more restrictive change, and is based on 
recommendations by Westinghouse Nuclear Safety Advisory Letter 97-003. 
The third change would revise the Turbine Trip and Feedwater Isolation 
function to include an initiation signal from the average-low 
temperature. This change is considered a more restrictive change, and 
is consistent with the plant design and safety analysis.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration for each of the above proposed changes. The NRC staff has 
reviewed the licensee's analyses against the standards of 10 CFR 
50.92(c). The NRC staff's analysis is presented below:
    1. Will the changes involve a significant increase in the 
probability or consequences of an accident previously evaluated?
    The proposed changes will not affect the safety function of the 
subject systems. There will be no direct effect on the design or 
operation of any plant structures, systems, or components. No 
previously analyzed accidents were initiated by functions of these 
systems, and the systems were not factors in the consequences of 
previously analyzed accidents. Therefore, the proposed changes will 
have no impact on the consequences or probabilities of any previously 
evaluated accidents.
    2. Will the changes create the possibility of a new or different 
kind of accident from any accident previously evaluated?
    The proposed changes would not lead to any hardware or operating 
procedure change. Hence, no new equipment failure modes or accidents 
from those previously evaluated will be created.
    3. Will the changes involve a significant reduction in a margin of 
safety?
    Margin of safety is associated with confidence in the design and 
operation of the plant. The proposed changes do not involve any change 
to the plant design, operation, or analysis. Thus, the margin of safety 
previously analyzed and evaluated is maintained.
    Based on this analysis, it appears that the three standards of 10 
CFR 50.92(c) are satisfied for each of the proposed changes. Therefore, 
the NRC staff proposes to determine that the amendment request involves 
no significant hazards consideration.
    Local Public Document Room location: J. Murrey Atkins Library, 
University of North Carolina at

[[Page 27762]]

Charlotte, 9201 University City Boulevard, North Carolina.
    Attorney for licensee: Mr. Albert Carr, Duke Energy Corporation, 
422 South Church Street, Charlotte, North Carolina.
    NRC Project Director: Herbert N. Berkow.

Entergy Gulf States, Inc., and Entergy Operations, Inc., Docket No. 50-
458, River Bend Station, Unit 1, West Feliciana Parish, Louisiana

    Date of amendment request: April 9, 1998.
    Description of amendment request: The proposed amendment would 
revise license condition 2.C(13) to allow Final Feedwater Temperature 
Reduction (FFWTR) at the River Bend Station, Unit No.1(RBS). FFWTR is 
to be used at the end of each fuel cycle to allow approximately 
fourteen additional effective full power days of operation.
    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 changes do not significantly increase the 
probability or consequences of an accident previously evaluated.
    The abnormal operational occurrences or accidents analyzed in 
the SAR [Safety Analysis Report] have been examined for impact 
caused by partial feedwater heating during cycle extension or at 
coastdown condition. The limiting abnormal operation transients, 
including the Load Rejection with no Bypass (LRNBP) event and the 
Feedwater Controller Failure (FWCF) maximum demand event, Turbine 
Trip with No Bypass (TTNBP) and Pressure Regulator Failure Downscale 
(PRFD) have been analyzed based upon the core nuclear characteristic 
at end-of-cycle (EOC) conditions including the effects of increased 
core flow and the proposed reduction in feedwater temperature with 
an all-rods-out condition.
    The LOCA (Loss of Coolant Accident), fuel loading error, rod 
drop accident, rod withdrawal error, overpressure protections and 
ATWS (anticipated transient without scram) analyses have been 
evaluated for the effects of reduced feedwater temperature operation 
and found acceptable. In addition, the case of the analyzed 
operational events the current fuel OLMCPR (operating limit minimum 
critical power ratio) and MAPLHGR (maximum average planar linear 
heat generation rate) limits bound those necessary for operation and 
therefore, are not affected by operation with FFWTR therefore, these 
events are bounded by the current RBS analysis. Because the accident 
results are acceptable and the current operating fuel limits are 
unaffected, the consequence of an event previously evaluated remains 
unaffected.
    The probability of an accident is not affected by the proposed 
changes since no systems or equipment which could initiate an 
accident are affected. Therefore, the proposed changes do not 
significantly increase the probability or consequences of any 
previously evaluated accident.
    2. The request does not create the possibility of occurrence of 
a new or different kind of accident from any accident previously 
evaluated.
    The FFWTR mode of operation is functionally similar to operation 
with Feedwater Heaters Out of Service (USAR (Updated Safety Analysis 
Report) Section 15.1.7). All abnormal operational transients or 
accidents have been evaluated and the most limiting cases have been 
analyzed for applicability for the FFWTR operation. Limits on 
MAPLHGR and OLMCPR (including the power and flow dependent MCPR) 
which are included in the Core Operating Limits Report (COLR) as 
part of the normal reload licensing process will continue to assure 
that operations are within the assumptions, initial conditions and 
assumed power distribution and therefore will not create a new type 
of accident. The proposed changes do not involve new setpoints, new 
system interactions, or physical modifications to the plant. 
Therefore, the proposed changes do not create the possibility of a 
new or different kind of accident from any previous analyzed.
    3. The request does not involve a significant reduction in a 
margin of safety.
    The proposed changes do not involve any setpoint changes and 
would allow steady state power operation at off-rated feedwater 
temperature conditions as defined in current plant procedures. The 
transient and accidents described in the SAR are evaluated for 
effects caused by the reduced feedwater temperature of 100 (degrees) 
F. As described in Attachment 4 (to the April 9, 1998, amendment 
request), * * * the FWCF is the most limiting transient under such 
condition and the required OLMCPR for this event is bounded by the 
EOC OLMCPR limits set forth in the RBS COLR. The thermal limits MCPR 
and LHGR curves, and the MAPLHGR limits establish limits on power 
operation and thereby ensure that the core is operated within the 
assumptions and initial conditions of the transient or accident 
analyses.
    Operation within these limits set forth by the MCPR limits, the 
LHGR limits and the MAPLHGR criteria will ensure that the margin of 
safety will be maintained to the same level described in the 
Technical Specifications Bases and the SAR. As a result the 
consequences of postulated transients or accidents are not 
increased.
    The MCPR safety limit, mechanical performance limits and 
overpressure limits are not exceeded during any transient or 
postulated accident at normal feedwater temperature or at reduced 
feedwater temperature condition. Therefore, the proposed changes to 
allow partial feedwater heating for cycle extension do not involve a 
significant reduction in margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: Government Documents 
Department, Louisiana State University, Baton Rouge, LA 70803.
    Attorney for licensee: Mark Wetterhahn, Esq., Winston & Strawn, 
1400 L Street, N.W., Washington, DC 20005.
    NRC Project Director: John N. Hannon.

Illinois Power Company, Docket No. 50-461, Clinton Power Station, Unit 
1, DeWitt County, Illinois

    Date of amendment request: April 27, 1998.
    Description of amendment request: The proposed amendment would 
change the title of ``shift supervisor'' to ``shift manager'' in the 
Technical Specifications (TS).
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration which is presented below:

    (1) The proposed change replaces the title of ``shift 
supervisor'' with the title of ``shift manager'' as it pertains to 
the responsibilities of the position described in TS Section 5.1.2. 
The proposed change does not involve a change to the plant design or 
to the operation of the plant by qualified operators and senior 
operators. Although this change involves changes to the Operations 
department, individuals in those positions comprising the operating 
crews will continue to have to meet the same licensing, experience, 
training, and education requirements, notwithstanding the proposed 
change in the title of the individual with ultimate command 
authority in the main control room, from ``shift supervisor'' to 
``shift manager.'' Therefore, the operation of CPS is not affected 
by this change. Further, as also noted, the proposed change does not 
affect plant design. It therefore would not affect systems, 
structures, or components important to safety, particularly those 
associated with the plant accident analyses. As a result, the 
proposed change does not affect any parameters or conditions that 
may contribute to the initiation of any accidents previously 
evaluated, nor does it affect the operation or response of systems, 
structures, or components assumed to mitigate postulated accidents 
that have been evaluated/analyzed. On this basis, IP has concluded 
that the proposed change will not result in a significant increase 
in the probability or consequences of any accident previously 
evaluated.
    (2) As noted above, the proposed change does not involve a 
change to design or operation of the plant. As a result, the 
proposed change, which is only administrative in nature, cannot 
introduce

[[Page 27763]]

any new failure modes or precursors, parameters, or conditions that 
could cause or contribute to the initiation of any new accidents not 
previously evaluated. On this basis, IP has concluded that the 
proposed change will not create the possibility of a new or 
different kind of accident not previously evaluated.
    (3) As noted above, the proposed change is an administrative 
change that involves no changes to plant design or operation, 
including the design or operation of systems, components, or 
structures important to safety. On this basis there are no margins 
of safety affected by the proposed change. As a result, IP has 
concluded that the proposed change will not result in 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: Vespasian Warner Public 
Library, 120 West Johnson Street, Clinton, IL 61727.
    Attorney for licensee: Leah Manning Stetzner, Vice President, 
General Counsel, and Corporate Secretary, 500 South 27th Street, 
Decatur, IL 62525.
    NRC Project Director: Ronald R. Bellamy, Acting.

Maine Yankee Atomic Power Company, Docket No. 50-309, Maine Yankee 
Atomic Power Station, Lincoln County, Maine

    Date of amendment request: April 13, 1998.
    Description of amendment request: The proposed amendment would 
amend the Technical Specifications to base the Limiting Condition for 
Operation for the fuel storage pool water level on a revised analysis 
of the fuel handling accident and on a new analysis for radiological 
shielding during movement of irradiated fuel.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    The proposed change does not:
    1. Involve a significant increase in the probability or 
consequence of an accident previously evaluated.
    The proposed restrictions on the water level in the spent fuel 
pool has no impact on the probability or consequences of the 
remaining applicable design basis accidents. These restrictions are 
fulfilled by normal operating conditions, preserve initial 
conditions assumed in the analyses of postulated DBAs and ensure 
that the conditions of such DBAs are consistent with the analyses. 
Revised analysis was performed assuming a fuel handling accident 
occurs after the spent fuel fission products have decayed at least 
1-year. The initial conditions assumed a minimum of 19 feet of water 
for iodine absorption. No credit was taken for control room or spent 
fuel pool ventilation filtration. The results of the revised 
analysis demonstrate that the projected doses resulting from a 
postulated fuel handling accident are insignificant in comparison to 
10 CFR part 100 limits. Therefore, the proposed changes to the 
Technical Specifications do not involve any increase in the 
probability or consequences of any accident previously evaluated.
    2. Create the possibility of a new or different kind of accident 
from any accident previously evaluated.
    The proposed restrictions on the water level in the spent fuel 
pool are fulfilled by normal operating conditions and preserve 
initial conditions assumed in the analysis of postulated DBAs. These 
additional restrictions do not involve changes to any structure or 
equipment affecting the safe storage of irradiated fuel. The results 
of the revised analysis of a fuel handling accident demonstrate that 
the projected doses are insignificant in comparison to 10 CFR part 
100 limits with a minimum of 19 feet of water for iodine absorption. 
In addition, maintaining this minimum water level will also provide 
sufficient shielding for personnel radiation protection during fuel 
movement. Therefore, the proposed changes to the Technical 
Specifications would not create the possibility of a new or 
different accident from any accident previously evaluated.
    3. Involve a significant reduction in a margin of safety.
    The proposed restrictions on the water level in the spent fuel 
pool preserve initial conditions assumed in the analyses of 
postulated DBAs and ensure that margins of safety contained in the 
analyses are maintained. The margin of safety for the fuel handling 
accident relates to the acceptance limit which the NRC approved 
during its review of the license. The fuel handling accident 
acceptance limit defined in the basis for the Maine Yankee Technical 
Specification (formerly specified as TS 3.13.D.10) is 10% of 10 CFR 
part 100 limits. A reduction in margin of safety occurs when the 
acceptance limit would no longer be met as a result of a proposed 
change. Since the acceptance limit is met, there is no reduction in 
margin of safety. The projected dose rates at the specified Fuel 
Storage Pool water level during fuel movement with a fuel assembly 
raised to its highest allowable height would result in personnel 
exposures within that previously assumed. There is no reduction in a 
margin of safety. The NRC acceptance limit which is that combination 
of occupancy time and dose rate that maintains personnel doses 
within 10 CFR 20.1201 limits is not exceeded. Therefore, the 
proposed changes to the MYTS would not involve a significant 
reduction in any margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: Wiscasset Public Library, High 
Street, PO Box 367, Wiscasset, ME 04578.
    Attorney for licensee: Mary Ann Lynch, Esquire, Maine Yankee Atomic 
Power Company, PO Box 408, Wiscasset, ME 04578.
    NRC Project Director: Seymour H. Weiss.

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: March 2, 1998.
    Description of amendment requests: The proposed amendments would 
remove the spent fuel pool special ventilation system operability-based 
restriction on crane operations in the spent fuel pool enclosure, while 
maintaining that restriction during spent fuel handling operations.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. The proposed amendment[s] will not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    The proposed change does not affect any system that is a 
contributor to initiating events for previously evaluated 
anticipated operational occurrences and design basis accidents. 
Therefore, the proposed change will not increase the probability of 
any previously evaluated accident.
    The proposed change does not impact the required availability of 
the spent fuel pool special ventilation system during spent fuel 
handling operations to mitigate the consequences of a fuel handling 
accident.
    The proposed change does impact the required availability of the 
spent fuel pool special ventilation system during heavy load 
handling operations. However, this system is not required to 
mitigate the consequences of a heavy load dropping onto a spent fuel 
assembly. Such a requirement is not applicable at Prairie Island, 
because the heavy loads in the spent fuel pool enclosure are either 
handled with single-failure-proof cranes, rigging and plant 
procedures implementing Prairie Island commitments to NUREG-0612, or 
handled with spent fuel pool protective covers in place as described 
in the Prairie Island USAR (updated safety analysis report). The use 
of a single-failure-proof crane with rigging and procedures that 
implement the requirements of NUREG-0612 assures that the potential 
for a heavy load

[[Page 27764]]

drop is extremely small and therefore consideration of the effects 
of heavy load drops is not required. Spent fuel pool covers prevent 
dropped loads* (*The covers do have a limit on the weight load they 
are analyzed to withstand.) from falling into the spent fuel pool 
and therefore consideration of the effects of heavy load drops is 
also not required. These actions taken to reduce the accident 
initiator probabilities to insignificant magnitudes negate any 
theoretically small increase in the consequence of a postulated 
heavy load drop accident resulting from the removal of a requirement 
to have one train of the spent fuel pool special ventilation system 
operable during crane operations. It is concluded in summary that 
the proposed change does not involve a significant increase in the 
consequences of any accident previously evaluated.
    2. The proposed amendment[s] will not create the possibility of 
a new or different kind of accident from any accident previously 
analyzed.
    The proposed change does impact the required availability of the 
spent fuel pool special ventilation system during heavy load 
handling operations. Load drop events over spent fuel are well 
understood and have been thoroughly evaluated. The proposed change 
will not create any new accident scenarios or create the possibility 
of a new or different kind of accident from any accident previously 
analyzed.
    3. The proposed amendment[s] will not involve a significant 
reduction in the margin of safety.
    The proposed change does not impact the required availability of 
the spent fuel pool special ventilation system during spent fuel 
handling operations to mitigate the consequences of a fuel handling 
accident as described in the USAR. As a result the safety margin 
inherent in the 10 CFR part 100 dose limits is not reduced.
    The proposed change does impact the required availability of the 
spent fuel pool special ventilation system during heavy load 
handling operations. However, this system is not required to 
mitigate the consequences of a heavy load dropping onto a spent fuel 
assembly because the potential for a load drop is extremely small. 
Provision of single-failure-proof equipment and compliance with the 
other requirements of NUREG-0612 (provide) a defense-in-depth 
approach to assure the safe handling of heavy loads which would 
otherwise be demonstrated to be safe by the deterministic analysis 
of the radiological effects of dropped heavy loads.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment requests involve no significant hazards consideration.
    Local Public Document Room location: Minneapolis Public Library, 
Technology and Science Department, 300 Nicollet Mall, Minneapolis, 
Minnesota 55401.
    Attorney for licensee: Jay Silberg, Esq., Shaw, Pittman, Potts, and 
Trowbridge, 2300 N Street, NW, Washington, DC 20037.
    NRC Project Director: Cynthia A. Carpenter.

Pennsylvania Power and Light Company, Docket Nos. 50-387 and 50-388 
Susquehanna Steam Electric Station, Units 1 and 2, Luzerne County, 
Pennsylvania

    Date of amendment request: November 26, 1997.
    Description of amendment request. The amendments to the Units 1 and 
2 Technical Specifications Surveillance Requirement Section 4.7.1.3.a 
involve lowering the Ultimate Heat Sink (UHS) surveillance requirement 
maximum acceptable spray pond average temperature from 88  deg.F to 85 
deg.F. This temperature is specified to assure that the post design 
basic accident (DBA) loss-of-coolant (LOCA) accident/loss of offsite 
power maximum UHS temperature will be maintained less than the UHS 
design temperature.
    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 
change lowers the UHS temperature surveillance requirement so that 
the maximum post DBA UHS temperature is maintained less than that 
reported previously.
    The UHS provides cooling to equipment and systems required for 
the safe shutdown of the plant following an accident with 
radiological consequence potential, such as a LOCA. The change in 
UHS initial temperature limit to 85  deg.F assures that the peak 
temperature will remain less than that reported previously. 
Therefore, the components cooled by the UHS will not be impacted and 
will be capable of performing their function as designed.
    Based upon the analysis presented above, PP&L (Pennsylvania 
Power and Light Company) 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 proposed change lowers the UHS surveillance requirement 
temperature so that the maximum post DBA UHS temperature is 
maintained less than that reported previously. Therefore the 
operation of the components cooled by the UHS will not be impacted 
and will be capable of performing their design function.
    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. 
The proposed change lowers the UHS surveillance temperature so that 
the maximum post DBA UHS temperature is maintained less than that 
reported previously. The margin of safety is unaffected since the 
maximum post DBA UHS temperature is not affected. Performance of 
equipment cooled by the UHS is unaffected.

    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, Docket Nos. 50-387 and 50-388, 
Susquehanna Steam Electric Station, Units 1 and 2, Luzerne County, 
Pennsylvania

    Date of amendment request: March 16, 1998.
    Description of amendment request: The proposed amendment would 
change the design basis and Technical Specifications to support the 
implementation of Hydrogen Water Chemistry.
    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.
    No Design Basis Event requiring functioning of the Main Steam 
Line Radiation monitors is defined in the FSAR. FSAR Section 
7.2.1.1.4.2.(i) describing Main Steam Line Radiation monitoring 
states that for accidents resulting in gross fission

[[Page 27765]]

product release ``the primary variables for trip initiation would be 
reactor vessel low level, reactor vessel high pressure, or high 
neutron flux''. Because the Main Steam Line Radiation Monitors 
[MSLRM] trip function is not used in any accident analyses this 
proposed setpoint change does not involve an increase in the 
probability or consequences of an accident previously evaluated.
    In conformance to SRP 15.4.9, the analysis of the design basis 
CRDA assumed release of activity by leakage from an isolated 
condenser. As described in the FSAR, the main steam line radiation 
monitors will shut down the mechanical vacuum pump if operating and 
close its suction valves, thus isolating the condenser in the event 
of a Main Steam Line-High Radiation trip. Operation of the 
mechanical vacuum pump following burst failures of fuel rods 
insufficient to cause a main steam line radiation monitor trip was 
evaluated to better understand the potential impacts of raising the 
setpoint. Doses calculated under conservative conditions were small 
compared to the acceptance criteria for offsite dose of 25% of 10 
CFR part 100 limits for offsite dose for the CRDA and 10 CFR 50 
limits for control room dose.
    Relocation of the Main Condenser Offgas Treatment System 
Explosive Gas Monitoring System requirements to the FSAR Section 
16.3 (Technical Requirements Manual (TRM)) and procedures involves 
the use of an alternate regulatory process for controlling the 
instrumentation requirements. The change does not introduce any new 
modes of plant operation, make any physical changes, alter any 
operational setpoints, or change the surveillance requirements. Any 
change in the Main Condenser Offgas Treatment System Explosive Gas 
Monitoring System requirements would be evaluated pursuant to the 
requirements of 10 CFR 50.59.
    The Technical Specifications, the Explosive Gas Mixture 
description contained in LCO/Surveillance 3.11.2.6/4.11.2.6 and 
associated bases will be moved and retained in TS Section 6.0 
``Administrative Controls''. The LCO specific limit and program 
details will be relocated to the FSAR Section 16.3 (TRM) and 
procedures and any changes controlled by the 10 CFR 50.59 process. 
Therefore, this change does not involve an increase in the 
probability or consequences of an accident previously evaluated.
    These proposed changes to Technical Specifications do not 
require physical changes to instrument channels other than the Main 
Steam Radiation Monitor setpoint, or to any systems or component 
that interfaces with the instrumentation channels, therefore there 
is no change in the probability or consequences of any accident 
analyzed in the FSAR.
    Finally, revising the TS index is an administrative change.
    2. Create the possibility of a new or different kind of accident 
from any accident previously evaluated.
    The proposed Main Steam Line Radiation setpoint change does not 
result in any design or physical configuration changes to the 
instrumentation channels. Operation incorporating the proposed 
change will not impair the instrumentation channels from performing 
as provided in the design basis.
    Relocation of the Main Condenser Offgas Treatment System 
Explosive Gas Monitoring System requirements to the FSAR Section 
16.3 (TRM) and procedures involves the use of an alternate 
regulatory process for controlling the instrumentation requirements. 
Therefore, the above change does not introduce any accident 
initiators as it does not involve any new modes of plant operation, 
make any physical changes, alter any operational setpoints, or 
change the surveillance requirements.
    The Technical Specifications, the Explosive Gas Mixture 
description contained in LCO/Surveillance 3.11.2.6/4.11.2.6 and 
associated bases will be moved and retained in TS Section 6.0 
``Administrative Controls''. The LCO specific limit and program 
details involves the use of an alternate regulatory process for 
controlling the requirements.
    Since the proposed changes to the Technical Specifications do 
not adversely impact the reliability of the safety required systems, 
no new or different kind of accident is created.
    3. Involve a significant reduction in a margin of safety.
    Raising the trip setpoint does not significantly reduce the 
sensitivity of the MSLRM's to alarm and initiate actions in response 
to gross fuel failures during power operation or to the design basis 
control rod drop accident. The source term assumed for the design 
basis CRDA greatly exceeds that required to initiate the main steam 
line high radiation trip. Raising the setpoint does not induce a 
delay in reaching the setpoint that would result in an increase in 
offsite dose from the design basis control rod drop accident. The 
delay time from fuel failure to monitor response is determined by 
the transport time for steam flow from the reactor vessel to the 
monitor location, which is not changed by either hydrogen water 
chemistry or by the monitor setpoint. Consequently, raising the trip 
setpoint will not result in an incremental increase in activity 
release, control room dose or offsite dose. Therefore, there is no 
reduction in the margin of safety for the design basis event.
    The radiological consequences of small fuel rupture events, that 
would produce main steam line radiation levels below the proposed 
trip setpoint, are not significant. These postulated events were 
evaluated to better understand the potential impacts of raising the 
setpoint. The potential offsite doses from such an event, in the 
absence of a trip, would be small compared to the limits of 10 CFR 
part 50 for control room dose and to the acceptance criteria of 25% 
of 10 CFR part 100 limits for offsite dose from the design basis 
CRDA.
    Relocation of the Main Condenser Offgas Treatment System 
Explosive Gas Monitoring System requirements to FSAR Section 16.3 
(TRM) involves the use of an alternate regulatory process for 
controlling the instrumentation requirements. Any change in the Main 
Condenser Offgas Treatment System Explosive Gas Monitoring System 
requirements would be evaluated pursuant to the requirements of 10 
CFR 50.59. Also, revising the TS index is an administrative change.
    The Explosive Gas Mixture description contained in LCO/
Surveillance 3.11.2.6/4.11.2.6 and associated bases will be moved 
and retained in TS Section 6.0 ``Administrative Controls''. The LCO 
specific limit and program details will be relocated to the FSAR 
Section 16.3 (TRM) and procedures and any changes controlled by the 
10 CFR 50.59 process.
    Therefore, the proposed change does not involve a significant 
reduction in a margin of safety.

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

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: February 16, 1998, as supplemented by 
letter dated April 2, 1998.
    Description of amendment request: The proposed amendment request 
would revise Technical Specification     3/4.4.5, ``Steam Generators,'' 
and its Bases to allow the implementation of 1-volt voltage-based 
repair criteria for the steam generator tube support plate-to-tube 
intersections for Unit 2 in accordance with Generic Letter 95-05, and 
make related Unit 1 administrative changes for consistency of wording 
(the NRC had previously approved a similar 1-volt voltage-based repair 
criteria application for Unit 1). In addition, the proposed amendment 
would make an administrative change to Bases     3/4.4.6.2, 
``Operational Leakage,'' to clarify that the allowable steam generator 
leakage specification applies to both Unit 1 and Unit 2.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Operation of the facility in accordance with the proposed 
amendment would not involve a significant increase in the 
probability or consequences of an accident previously evaluated.

[[Page 27766]]

Structural Considerations

    Industry testing of model boiler and operating plant tube 
specimens for free span tubing at room temperature conditions shows 
typical burst pressures in excess of 5000 psi for indications of 
ODSCC (outer diameter stress corrosion cracking) with voltage 
measurements at or below the current structural limit of 5.45 volts. 
One model boiler specimen with a voltage amplitude of 19 volts also 
exhibited a burst pressure greater than 5000 psi. Burst testing 
performed on one intersection pulled from STP (South Texas Project) 
Unit 1 in 1993 with a 0.51 volt indication yielded a measured burst 
pressure of 8900 psi at room temperature. Burst testing performed on 
another intersection pulled from STP Unit 1 in 1995 with a 0.48 volt 
indication yielded a measured burst pressure of 9950 psi at room 
temperature.
    The next projected end-of-cycle (EOC) voltage compares favorably 
with the current structural limit considering the voltage growth 
rate for indications at STP. Using the methodology of Generic Letter 
95-05, the structural limit is reduced by allowances for uncertainty 
and growth to develop a beginning-of-cycle (BOC) repair limit which 
should preclude EOC indications from growing in excess of the 
structural limit. The non-destructive examination (NDE) uncertainty 
to be applied per Generic Letter 95-05 is approximately 20%. The 
growth allowance will be 30%/EFPY [effective full power year] or a 
STP Unit 2-specific growth rate, to be calculated in accordance with 
Generic Letter 95-05, whichever is greater. Where the generator-
specific growth rate exceeds both the Unit 2-specific average growth 
rate and 30%/EFPY, that generator-specific growth rate will be used 
for that generator. Each succeeding cycle upper voltage repair limit 
will also be conservatively established based on Generic Letter 95-
05 methodology. By adding NDE uncertainty allowances and a growth 
allowance to the repair limit, the structural limit can be 
validated.
    The upper voltage repair limit could be applied to bobbin coil 
voltages between the lower and upper repair limits to leave such 
indications in service independent of RPC [rotating pancake coil] 
confirmation. However, RPC-confirmed indications will be 
conservatively removed from service consistent with Generic Letter 
95-05.

Leakage Considerations

    As part of the implementation of voltage-based repair criteria, 
the distribution of EOC degradation indications at the TSP (tube 
support plate) intersections has been used to calculate the primary-
to-secondary leakage which is bounded by the maximum leakage 
required to remain within the applicable dose limits of 10 CFR 100 
(10 CFR part 100) and GDC (General Design Criterion) 19. This limit 
was calculated using the Technical Specification Reactor Coolant 
System (RCS) Iodine-131 transient spiking values consistent with 
NUREG-0800. Application of the voltage-based repair criteria 
requires the projection of postulated Main Steam Line Break (MSLB) 
leakage based on the projected EOC voltage distribution from the 
beginning of cycle voltage distribution. Projected EOC voltage 
distribution is developed using the most recent EOC eddy current 
results and a voltage measurement uncertainty. Draft NUREG-1477 and 
Generic Letter 95-05 require that all indications to which voltage-
based repair criteria are applied must be included in the leakage 
projection.
    The projected MSLB leakage rate calculation methodology 
prescribed in Generic Letter 95-05 will be used to calculate the EOC 
leakage. A Monte Carlo approach will be used to determine the EOC 
leakage, accounting for all of the bobbin coil eddy current test 
uncertainties, voltage growth, and an assumed probability of 
detection of 0.6. The fitted log-logistic probability of leakage 
correlation will be used to establish the MSLB leak rate for each 
cycle. This leak rate will be used for comparison with a bounding 
allowable leak rate in the faulted loop which would result in 
radiological consequences which are within the dose limits of 10 CFR 
part 100 for offsite doses and GDC 19 for control room doses. Due to 
the relatively low voltage levels of indications at STP to date and 
low voltage growth rates, it is expected that the actual calculated 
leakage values will be far less than this limit for each successive 
cycle.

Other Considerations

    Those changes associated with grammatical corrections, deleting 
tube diameter information not applicable to South Texas, and 
applying the additional reporting requirements to Unit 2, are 
administrative and do not involve a change to, or the operation of, 
any safety-related system.
    Therefore, implementation of voltage-based repair criteria does 
not adversely affect steam generator tube integrity and the 
radiological consequences will remain below the limits of 10 CFR 
part 100 and GDC 19. Operation of the facility in accordance with 
the proposed amendment would not result in any increase in the 
probability or consequences of an accident previously evaluated.
    2. Operation of the facility in accordance with the proposed 
amendment would not create the possibility of a new or different 
kind of accident from any accident previously evaluated.
    Implementation of the proposed steam generator tube voltage-
based repair criteria for ODSCC at the TSP intersections does not 
introduce any significant changes to the plant design basis. Use of 
the criteria does not provide a mechanism which could result in an 
accident outside of the region of the TSP elevations because the 
criteria do not apply outside the thickness of the TSPs. It is 
therefore expected that for all plant conditions, neither a single 
nor multiple tube rupture event would likely occur in a steam 
generator where voltage-based repair criteria has been applied.
    Specifically, STP has implemented a maximum leakage rate of 150 
gpd [gallons-per-day] per steam generator to help preclude the 
potential for excessive leakage during all plant conditions. The 
draft Reg Guide 1.121 criterion for establishing operational leakage 
rate limits governing plant shutdown is based upon leak-before-break 
(LBB) considerations to detect a free span crack before potential 
tube rupture as a result of faulted plant conditions. The 150 gpd 
limit is intended to provide for leakage detection and plant 
shutdown in the event of unexpected crack propagation outside the 
tube support plate thickness resulting in excessive leakage. Draft 
Reg Guide 1.121 acceptance criteria for establishing operating 
leakage limits are based on LBB considerations such that plant 
shutdown is initiated if permissible degradation is exceeded.
    Thus, the 150 gpd limit provides for plant shutdown prior to 
reaching critical degradation lengths. Additionally, the leak-
before-break evaluation assumes that the entire crevice area is 
uncovered during the secondary side blowdown of a MSLB. Typically, 
it is expected for the vast majority of intersections, that only 
partial uncovery will occur. Therefore, the proximity of the TSP 
will enhance the burst capacity of the tube.
    Steam generator tube integrity is continually maintained through 
inservice inspection and primary-to-secondary leakage monitoring. 
Any tubes falling outside the voltage-based repair criteria limits 
are removed from service.
    Those changes associated with grammatical corrections, deleting 
tube diameter information not applicable to South Texas, and 
applying the additional reporting requirements to Unit 2, are 
administrative and do not involve a change to, or the operation of, 
any safety-related system.
    Therefore, operating the facility in accordance with the 
proposed amendment would not create the possibility of a new or 
different kind of accident from any accident previously evaluated.
    3. Operation of the facility in accordance with the proposed 
amendment would not involve a significant reduction in a margin of 
safety.
    The use of the voltage-based bobbin probe for dispositioning 
ODSCC degraded tubes within TSP intersections is demonstrated to 
maintain steam generator tube integrity in accordance with the 
requirements of draft Reg Guide 1.121. Draft Reg Guide 1.121 
describes a method acceptable to the NRC staff for meeting GDCs 14, 
15, 31, and 32 by reducing the probability or the consequences of 
steam generator tube rupture. This is accomplished by determining 
the limiting conditions of degradation of steam generator tubing, as 
established by inservice inspection, for which tubes with 
unacceptable degradation are removed from service. Upon 
implementation of the criteria, even under the worst case 
conditions, the occurrence of ODSCC at the TSP elevation is not 
expected to lead to a steam generator tube rupture event during 
normal or faulted plant conditions. The EOC distribution of 
indications at the TSP elevations for each successive cycle will be 
confirmed to result in acceptable primary-to-secondary leakage 
during all plant conditions.
    In addressing the combined effects of loss of coolant accident 
(LOCA) and safe shutdown earthquake (SSE) on the steam generators, 
as required by GDC 2, it has been determined that tube collapse may 
occur in the steam generators at some plants. This is not the case 
at STP Unit 2 as the TSPs do not become sufficiently deformed as a 
result

[[Page 27767]]

of lateral loads at the wedge supports at the periphery of the plate 
due to the combined effects of the leak-before-break-limited LOCA 
rarefaction wave and SSE loadings to affect tube integrity.
    Because the leak-before-break methodology is applicable to the 
STP reactor coolant loop piping, the probability of breaks in the 
primary loop piping is sufficiently low that they need not be 
considered in the structural design of the plant. Implementation 
practices using the bobbin probe voltage based tube plugging 
criteria bounds Reg Guide 1.83, Rev. 1, considerations by:
    (1) Using enhanced eddy current inspection guidelines consistent 
with those used by EPRI in developing the correlations. This 
provides consistency in voltage normalization.
    (2) Performing a 100% bobbin coil inspection for all hot leg 
tube support plate intersections and all cold leg intersections down 
to the lowest cold leg tube support plate with known ODSCC 
indications at each cycle. The determination of the tube support 
plate intersections having ODSCC indications shall be based on the 
performance of at least a 20% random sampling of tubes inspected 
over their full length, and
    (3) Incorporating rotating pancake coil inspection for all tubes 
with bobbin voltages greater than 1.0 volt. This further establishes 
the principal degradation morphology as ODSCC.
    Implementation of voltage-based repair criteria at TSP 
intersections will decrease the number of tubes which must be 
repaired at each subsequent inspection. Since the installation of 
tube plugs to remove ODSCC degraded tubes from service reduces the 
RCS flow margin, voltage-based repair criteria implementation will 
help preserve the margin of flow.
    For each cycle the projected EOC primary-to-secondary leak rate 
allowed is bounded by a leak rate which limits the radiological 
consequences of a EOC MSLB to within the dose limits of 10CFR100 for 
offsite doses and 10CFR50, Appendix A, General Design Criteria (GDC) 
19 for control room doses. Therefore, this change does not involve a 
significant reduction in the margin to safety.
    The assessment of radiological consequences of an assumed steam 
line break applicable to STP Unit 1 was provided in Attachment 2 to 
ST-HL-AE-5359 on May 2, 1996. The submittal was made in response to 
questions from the Emergency Preparedness and Radiation Protection 
Branch and is applicable to Unit 2 as well. The staff concluded that 
the thyroid doses for the Exclusion Area Boundary (EAB), Low 
Population Zone (LPZ), and control room are within the acceptance 
criteria.
    Those changes associated with grammatical corrections, deleting 
tube diameter information not applicable to South Texas, and 
applying the additional reporting requirements to Unit 2, are 
administrative and do not involve a change to, or the operation of, 
any safety-related system.

    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, NW., Washington, DC 20036-5869.
    NRC Project Director: John N. Hannon.

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.

Carolina Power & Light Company, et al., Docket No. 50-400, Shearon 
Harris Nuclear Power Plant, Unit 1, Wake and Chatham Counties, North 
Carolina

    Date of amendment request: April 24, 1998.
    Brief description of amendment: The proposed amendment would revise 
Technical Specification 3/4.3.2, ``Engineered Safety Features Actuation 
System Instrumentation'' to allow a 2-hour surveillance interval to 
facilitate testing of the 6.9 kV Emergency Bus Undervoltage relays.
    Date of publication of individual notice in the Federal Register: 
May 4, 1998 (63 FR 24574).
    Expiration date of individual notice: May 18, 1998 for comments; 
June 3, 1998 for hearings.
    Local Public Document Room location: Cameron Village Regional 
Library, 1930 Clark Avenue, Raleigh, North Carolina 27605.

Northeast Nuclear Energy Company, Docket No. 50-423, Millstone Nuclear 
Power Station, Unit No. 3, New London, County, Connecticut

    Date of amendment request: April 7, 1998.
    Description of amendment request: The proposed amendment would 
replace the pressurizer maximizer water inventory requirement with a 
pressurizer maximizer indicated level requirement.
    Date of publication of individual notice in Federal Register: April 
23, 1998 (63 FR 20219)
    Expiration date of individual notice: May 26, 1998.
    Local Public Document Room location: Learning Resources Center, 
Three Rivers Community-Technical College, 574 New London Turnpike, 
Norwich, Connecticut, and the Waterford Library, ATTN: Vince Juliano, 
49 Rope Ferry Road, Waterford, Connecticut.

Northeast Nuclear Energy Company, Docket No. 50-423, Millstone Nuclear 
Power Station, Unit 3, New London, County, Connecticut

    Date of amendment request: April 14, 1998.
    Description of amendment request: The proposed amendment addresses 
an earlier identified condition relating to the plant operators' 
ability to meet the operator response time of 10 minutes assumed in 
Chapter 15 of the Final Safety Analysis Report for termination of an 
Inadvertent Safety Injection event.
    Date of publication of individual notice in Federal Register: April 
20, 1998 (63 FR 19532).
    Expiration date of individual notice: May 20, 1998.
    Local Public Document Room location: Learning Resources Center, 
Three Rivers Community-Technical College, 574 New London Turnpike, 
Norwich, Connecticut, and the Waterford Library, ATTN: Vince Juliano, 
49 Rope Ferry Road, Waterford, Connecticut.

Pennsylvania Power and Light Company, Docket Nos. 50-387 and 50-388 
Susquehanna Steam Electric Station, Units 1 and 2, Luzerne County, 
Pennsylvania

    Date of amendment request: August 1, 1996, as supplemented on March 
2, 1998.
    Brief description of amendment request: The proposed amendments 
would revise the Technical Specifications as follows: (1.n.) Change the 
surveillance requirement frequency for verification that the average 
planar

[[Page 27768]]

heat generation rate, minimum critical power ratio, linear heat 
generation rate, and average power range monitor gain and setpoint are 
within specified limits. Specifically, the frequency would be changed 
from within 12 hours after completion of a thermal power increase of at 
least 15 percent of rated thermal power (RTP) to once within 24 hours 
after greater than or equal to 25 percent RTP, 24 hours thereafter, and 
prior to exceeding 50 percent RTP; (2.o.) Change the surveillance 
requirement for the verification of the average power range monitor 
flow biased simulated thermal power-high time constant from 6 seconds 
plus or minus 1 second to less than 7 seconds. The lower limit of 5 
seconds will be relocated to plant procedures since it is not a 
condition for operability of this reactor protection system function; 
(3.p.) Change the frequency of surveillance requirement for rod worth 
minimizer channel functional test; and (4.q.) Relocate the main steam 
line radiation monitor reactor protection system and isolation trips 
from the Technical Specifications to the plant-controlled Technical 
Requirements Manual.
    Date of publication of individual notice in Federal Register: 1.n. 
April 27, 1998 (63 FR 20664); 2.o. April 27, 1998 (63 FR 20669); 3.p. 
April 27, 1998 (63 FR 20665); 4.q. April 27, 1998 (63 FR 20667).
    Expiration date of individual notices: May 27, 1998 (all 4 
notices).
    Local Public Document Room location: Osterhout Free Library, 
Reference Department, 71 South Franklin Street, Wilkes-Barre, PA 18701.

Power Authority of The State of New York, Docket No. 50-286, Indian 
Point Nuclear Generating Unit No. 3, Westchester County, New York

    Date of application for amendment: February 27, 1998.
    Brief Description of amendment: The amendment revised the Technical 
Specifications by revising the pressure-temperature and overpressure 
limits.
    Date of publication of individual notice in Federal Register: March 
9, 1998 (63 FR 11456).
    Expiration date of individual notice: April 8, 1998.
    Local Public Document Room location: White Plains Public Library, 
100 Martine Avenue, White Plains, New York 10601.

Tennessee Valley Authority, Docket No. 50-390 Watts Bar Nuclear Plant, 
Unit 1, Rhea County, Tennessee

    Date of application for amendment: April 29, 1998.
    Brief description of amendments: To amend the Watts Bar Nuclear 
Plant, Unit 1, Technical Specifications (TS) for the Hydrogen 
Mitigation System igniters. The amendment revises the TS limiting 
condition for operation, LCO 3.6.8, to provide temporary requirements 
for hydrogen ignitors to address the two Train A ignitors which are 
currently out of service.
    Date of publication of individual notice in the Federal Register: 
May 7, 1998 (63 FR 25243).
    Expiration date of individual notice: June 8, 1998.
    Local Public Document Room location: Chattanooga-Hamilton County 
Library, 1001 Broad Street, Chattanooga, TN 37402.

Notice of Issuance of Amendments to Facility Operating Licenses

    During the period since publication of the last biweekly notice, 
the Commission has issued the following amendments. The Commission has 
determined for each of these amendments that the application complies 
with the standards and requirements of the Atomic Energy Act of 1954, 
as amended (the Act), and the Commission's rules and regulations. The 
Commission has made appropriate findings as required by the Act and the 
Commission's rules and regulations in 10 CFR Ch. 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.

Baltimore Gas and Electric Company, Docket Nos. 50-317 and 50-318, 
Calvert Cliffs Nuclear Power Plant, Unit Nos. 1 and 2, Calvert County, 
Maryland

    Date of application for amendments: December 4, 1996, as 
supplemented March 27, June 9, June 18, July 21, August 14, August 19, 
September 10, October 6, October 20, October 23, November 5, 1997, and 
January 12, January 28 and March 16, 1998.
    Brief description of amendments: The amendments include the 
following:
    1. The amendments added a new surveillance requirement (SR) 3.4.9.2 
to the Improved Technical Specifications (ITS) which requires 
verification that the capacity of each required bank of pressurizer 
heaters is equal to or greater than 150 kW every 24 months.
    2. The amendments changed the current TS applicability for the 
pressurizer safety valves for Mode 3 to specify that two safety valves 
shall be operable with all reactor coolant system (RCS) cold leg 
temperature 365  deg.F for Unit 1 and >301  deg.F for Unit 
2. This is a less restrictive change.
    3. As part of the conversion to the ITS, the amemdment changed a 
requirement that the power-operated relief valves be demonstrated 
operable by performing a channel functional test once per 31 days to 
once per 92 days.
    4. The ITS LCO 3.4.1.3 eliminated the limit of 1 gpm total primary-
to-secondary leakage through all steam generators and thus will only 
require a limit of 100 gallons per day through any one steam generator. 
This is an administrative change.
    5. The amendment retains the requirement of SR 4.5.2.f.2 and 
specifies a frequency of 24 months. The amendment also adds a new SR 
3.5.2.7 which requires verification that each LPSI pump stops on an 
actual or simulated actuation signal.
    6. The amendment regarding the control room emergency ventilation 
system (CREVS) changes the surveillance interval from 18 months to 24 
months (each refueling cycle) for SR 4.7.6.1.e.2 requires that each 
train of CREVS is demonstrated operable at least once every 18 months 
by verifying that on a control room high radiation test signal, the 
system automatically switches into a recirculation mode of operation 
with flow through the HEPA filter and charcoal adsorber banks and that 
both of the isolation valves in each duct and common exhaust duct, and 
isolation valve in the toilet exhaust area duct, close. The above 
change is less restrictive.

[[Page 27769]]

    7. The amendment changes the surveillance interval regarding the 
control room emergency temperature system (CRETS) from 62 days on a 
staggered basis (one train every 31 days) to 24 months (each refueling 
interval) for SR 4.7.6.1.a.
    8. The amendment changes the surveillance interval regarding the 
spent fuel pool exhaust ventilation system (SFPEVS) from 18 months to 
24 months (each refueling interval) for SR 4.9.12.d. This is a less 
restrictive change.
    9. The amendment changes the surveillance interval regarding the 
penetration room exhaust ventilation system (PREVS) from 18 months to 
24 months (each refueling interval) for SR 4.6.6.1.d.2.
    Date of issuance: May 4, 1998.
    Effective date: As of the date of issuance to be implemented by 
August 31, 1998.
    Amendment Nos.: 227 and 201.
    Facility Operating License Nos. DPR-53 and DPR-69: Amendments 
revised the Technical Specifications in its entirety.
    Date of initial notice in Federal Register: March 6, 1998 (63 FR 
11312) The Commission's related evaluation of these amendments is 
contained in a Safety Evaluation dated May 4, 1998.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: Calvert County Library, Prince 
Frederick, Maryland 20678.

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

    Date of application for amendments: October 29, 1997, as 
supplemented January 28 and April 20, 1998.
    Brief Description of amendments: The amendments update the 
Technical Specification description of Control Rod Assemblies to allow 
for boron carbide or hafnium absorber materials, as approved by the NRC 
staff.
    Date of issuance: April 27, 1998.
    Effective date: April 27, 1998.
    Amendment Nos.: 193 and 224.
    Facility Operating License Nos. DPR-71 and DPR-62: Amendments 
change the Technical Specifications.
    Date of initial notice in Federal Register: December 17, 1997 (62 
FR 66137) The Commission's related evaluation of the amendments is 
contained in a Safety Evaluation dated April 27, 1998.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: University of North Carolina 
at Wilmington, William Madison Randall Library, 601 S. College Road, 
Wilmington, North Carolina 28403-3297.

Commonwealth Edison Company, Docket Nos. STN 50-454 and STN 50-455, 
Byron Station, Unit Nos. 1 and 2, Ogle County, Illinois

    Date of application for amendments: November 7, 1997, as 
supplemented on March 24, 1998, and April 9, 1998.
    Brief description of amendments: The amendments defer the next 
scheduled Type A containment integrated leak rate test for Byron, Unit 
2, until the next refueling outage in 1999.
    Date of issuance: May 8, 1998.
    Effective date: Immediately, to be implemented within 30 days.
    Amendment Nos.: 102 and 102.
    Facility Operating License Nos. NPF-37 and NPF-66: The amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: April 7, 1998 (63 FR 
17036) The April 9, 1998, supplement provided clarifying information 
which did not change the staff's initial proposed no significant 
hazards consideration.
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated May 8, 1998.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: Byron Public Library District, 
109 N. Franklin, P.O. Box 434, Byron, Illinois 61010.

Commonwealth Edison Company, Docket Nos. 50-373 and 50-374, LaSalle 
County Station, Units 1 and 2, LaSalle County, Illinois

    Date of application for amendments: September 26, 1997, as 
supplemented on April 7, 1998.
    Brief description of amendments: The amendments revise Technical 
Specification 3.6.1.8 to prohibit the simultaneous opening of the 
drywell and suppression chamber purge system isolation valves and 
revise the surveillance requirements of TS 3/4.6.5.3, ``Standby Gas 
Treatment System'' to upgrade the filter testing methods to more 
current industry standards. This amendment approves only a portion of 
the request dated September 26, 1997. The remainder of the request will 
be addressed in separate correspondence.
    Date of issuance: April 27, 1998.
    Effective date: Immediately, to be implemented prior to startup of 
LaSalle, Unit 1, from the current outage and prior to restart of 
LaSalle, Unit 2, from the current outage.
    Amendment Nos.: 125 and 110.
    Facility Operating License Nos. NPF-11 and NPF-18: The amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: November 19, 1997 (62 
FR 61840) The April 7, 1998, submittal provided additional clarifying 
information that did not change the initial proposed no significant 
hazards consideration determination. The Commission's related 
evaluation of the amendments is contained in a Safety Evaluation dated 
April 27, 1998.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: Jacobs Memorial Library, 
Illinois Valley Community College, Oglesby, Illinois 61348.

Consumers Energy Company, Docket No. 50-255, Palisades Plant, Van Buren 
County, Michigan

    Date of application for amendment: December 11, 1995, as 
supplemented January 18, September 3, October 2, October 18, and 
October 25, 1996, and March 28, 1997.
    Brief description of amendment: The amendment revises 
administrative controls technical specifications (TS) and related 
surveillance requirements. Amendment 174, issued on October 31, 1996, 
provided a partial response to the licensee's request. This amendment 
completes action on the request.
    NRC has also granted the request of Consumers Energy to withdraw a 
portion of its December 11, 1996, application. The proposed change 
would have deleted the requirements of current TS 4.5.4, ``Surveillance 
for Prestressing System,'' TS 4.5.5, ``End Anchorage Concrete 
Surveillance,'' and TS 4.5.8, ``Dome Delamination Surveillance,'' and 
replaced the requirements with proposed TS 6.5.5, ``Containment 
Structural Integrity Surveillance Program.'' However, by letter dated 
March 28, 1997, the licensee withdrew the proposed change. In addition, 
the staff has denied a portion of the amendment request regarding 
limitations on the dose rates resulting from radioactive material 
released in gaseous effluents to areas beyond the site boundary. A 
separate Notice of Partial Denial of Amendment to Facility Operating 
License and Opportunity for Hearing has been published in the Federal 
Register. For further details with respect to these actions, see the 
application for amendment dated December 11, 1996, as supplemented 
above, the licensee's letter dated March 28, 1997, which withdrew this 
portion of the application for license amendment, and the staff's 
Safety Evaluation enclosed with the

[[Page 27770]]

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 room 
listed below.
    Date of issuance: May 7, 1998.
    Effective date: May 7, 1998, to be implemented within 60 days from 
date of issuance.
    Amendment No.: 181.
    Facility Operating License No. DPR-20: Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: September 20, 1996 (61 
FR 49493) The October 2, October 18, and October 25, 1996, and March 
28, 1997, letters provided clarifying information and updated TS pages 
that were 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 May 7, 1998.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: Van Wylen Library, Hope 
College, Holland Michigan 49423.

Consumers Energy Company, Docket No. 50-255, Palisades Plant, Van Buren 
County, Michigan

    Date of application for amendment: December 27, 1995, as 
supplemented September 4, October 18, and November 26, 1996, June 27 
and November 21, 1997, and January 29, and April 10, 1998.
    Brief description of amendment: The amendment revises specification 
requirements and associated bases regarding the electrical power 
systems to closely emulate the Standard Technical Specifications for 
Combustion Engineer Plants, NUREG-1432, Revision 1.
    Date of issuance: April 29, 1998.
    Effective date: The license amendment is effective as of the date 
of issuance with full implementation within 60 days after Cold Shutdown 
following completion of the 1998 refueling outage, but no later than 
October 2, 1998.
    Amendment No.: 180.
    Facility Operating License No. DPR-20: Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: April 9, 1997 (62 FR 
17229) The June 27 and November 21, 1997, and January 29 and April 10, 
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 April 29, 1998.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: Van Wylen Library, Hope 
College, Holland, Michigan 49423.

Entergy Operations, Inc., System Energy Resources, Inc., South 
Mississippi Electric Power Association, and Entergy Mississippi, Inc., 
Docket No. 50-416, Grand Gulf Nuclear Station, Unit 1, Claiborne 
County, Mississippi

    Date of application for amendment: September 18, 1997, as 
supplemented by letter dated February 24, 1998.
    Brief description of amendment: The amendment decreases the safety 
limit for the minimum critical power ratio (MCPR) from 1.12 to 1.11 for 
two recirculation loop operation and from 1.14 to 1.12 for single 
recirculation loop operation in Technical Specification (TS) 2.1.1.2. 
Because the proposed amendment is for Cycle 10 operation, the amendment 
would also revise the footnotes to TSs 2.1.1.2 and 5.6.5 to state that 
the MCPR values and the items 19 and 20, two topical reports being 
added to the core operating limits report in TS 5.6.5, are ``applicable 
only for Cycle 10 operation.'' Cycle 10 operation begins at the plant 
restart from the current refueling outage No. 9.
    Date of issuance: May 8, 1998.
    Effective date: May 8, 1998.
    Amendment No: 136.
    Facility Operating License No. NPF-29: Amendment revises the 
Technical Specifications.
    Date of initial notice in Federal Register: October 22, 1998 (62 FR 
54872) The Commission's related evaluation of the amendment is 
contained in a Safety Evaluation dated May 8, 1998.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: Judge George W. Armstrong 
Library, 220 S. Commerce Street, Natchez, MS 39120.

Florida Power Corporation, et al., Docket No. 50-302, Crystal River 
Unit No. 3 Nuclear Generating Plant, Citrus County, Florida

    Date of application for amendment: October 4, 1997.
    Brief description of amendment: To revise the Final Safety Analysis 
Report (FSAR) and the Improved Technical Specification (TS) Bases to 
reflect the modified reactor building fan control logic for fan AHF-1C.
    Date of issuance: April 29, 1998.
    Effective date: April 29, 1998.
    Amendment No.: 166.
    Facility Operating License No. DPR-72: Amendment revised the 
updated FSAR and TS Bases.
    Date of initial notice in Federal Register: November 13, 1997 (62 
FR 60921) The Commission's related evaluation of the amendment is 
contained in a Safety Evaluation dated April 29, 1998.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: Coastal Region Library, 8619 
W. Crystal Street, Crystal River, Florida 32629.

Florida Power and Light Company, et al., Docket No. 50-335, St. Lucie 
Plant, Unit No. 1, St. Lucie County, Florida

    Date of application for amendment: July 22, 1997.
    Brief description of amendment: The amendment will incorporate a 
recent evaluation of a postulated inadvertent opening of a main steam 
safety valve into the current licensing basis for St. Lucie Unit 1.
    Date of Issuance: April 30, 1998.
    Effective Date: April 30, 1998.
    Amendment No.: 154.
    Facility Operating License No. NPF-16: Amendment revised the 
Updated Final Safety Evaluation Report.
    Date of initial notice in Federal Register: August 27, 1997 (62 FR 
45457) The Commission's related evaluation of the amendment is 
contained in a Safety Evaluation dated April 30, 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.

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 application for amendments: March 6, 1998, as supplemented 
March 30, March 31, and April 13, 1998.
    Brief description of amendments: The amendments update the 
Technical Specification heatup and cooldown rate curves and extend 
their reactor vessel fluence limit from the current 20 effective full 
power years (EFPYs) to a new value of 35 EFPYs, incorporate into 
Technical Specifications the use of a Pressure and Temperature Limits 
Report, and change the power-operated relief valves temperature 
requirement for operability.
    Date of issuance: May 4, 1998.
    Effective date: May 4, 1998, with full implementation within 30 
days.

[[Page 27771]]

    Amendment Nos.: 135, 127.
    Facility Operating License Nos. DPR-42 and DPR-60. Amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: March 27, 1998 (63 FR 
14972) The March 30, March 31, and April 13, 1998, letters provided 
clarifying information and updated Technical Specification pages 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 amendments is contained 
in a Safety Evaluation dated May 4, 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.

Philadelphia Electric Company, Docket No. 50-352, Limerick Generating 
Station, Unit 1, Montgomery County, Pennsylvania.

    Date of application for amendment: February 9, 1998, as 
supplemented April 8 and 24, 1998.
    Brief description of amendment: The amendment revises the minimal 
critical power ratio safety limits for operation Cycle 8.
    Date of issuance: May 4, 1998.
    Effective date: As of date of issuance, and shall be implemented 
within 30 days.
    Amendment No.: 127.
    Facility Operating License No. NPF-39: This amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: February 25, 1998 (63 
FR 9613) The April 8 and 24, 1998, letters provided clarifying 
information that did not change the initial proposed no significant 
hazards consideration determination.
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated May 4, 1998.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: Pottstown Public Library, 500 
High Street, Pottstown, PA 19464.

Southern Nuclear Operating Company, Inc., Docket Nos. 50-348 and 50-
364, Joseph M. Farley Nuclear Plant, Units 1 and 2, Houston County, 
Alabama.

    Date of amendments request: February 14, 1997, as supplemented by 
letters dated June 20, August 5, September 22, November 19, December 9, 
December 17, and December 31, 1997, January 23, February 12, February 
26, March 3, March 6, March 16, April 3, April 13, and two letters on 
April 17, 1998.
    Brief Description of amendments: The amendments change the maximum 
reactor core power level for facility operation from 2652 megawatts-
thermal (MWt) to 2775 MWt for the Joseph M. Farley Nuclear Plant, Units 
1 and 2. The amendments also approve changes to the Technical 
Specifications to implement uprated power operation.
    Date of issuance: April 29, 1998.
    Effective date: As of the date of issuance to be implemented prior 
to entering Mode 4 for Cycle 16 (fall 1998) for Unit 1 and prior to 
entering Mode 4 for Cycle 13 (spring 1998) for Unit 2.
    Amendment Nos.: Unit 1--137; Unit 2--129.
    Facility Operating License Nos. NPF-2 and NPF-8: Amendments revise 
the Technical Specifications, Operating Licenses, and adds a new 
Appendix C to the Operating Licenses.
    Date of initial notice in Federal Register: October 8, 1997 (62 FR 
52588) The November 19, December 9, December 17, and December 31, 1997, 
January 23, February 12, February 26, March 3, March 6, March 16, April 
3, April 13, and two letters on April 17, 1998, provided additional and 
clarifying information that did not change the scope of the February 
14, 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 April 29, 1998, and an Environmental 
Statement was prepared and dated April 17, 1998.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: Houston-Love Memorial Library, 
212 W. Berdache Street, Post Office Box 1369, Dothan, Alabama.

TU Electric Company, Docket Nos. 50-445 and 50-446, Comanche Peak Steam 
Electric Station, Unit Nos. 1 and 2, Somervell County, Texas.

    Date of amendment request: April 9, 1998 (TXX-98107).
    Brief description of amendments: The proposed amendment would allow 
on a one time basis, the verification of the proper operation of the 
Unit 2 load shed seal-in contacts and the diesel generator trip bypass 
contacts at power and crediting performance of Surveillance 
Requirements (SR) 4.8.1.1.2f.4) and 4.8.1.1.2f.6), at power as opposed 
to ``during shutdown'' as currently required by those SR. The proposed 
amendment would also allow on a one time basis the verification of the 
proper operation of the Unit 2 lockout relays and contacts to be 
deferred until the startup from the Unit 2 fourth refueling outage 
(2RFO4) or earlier outage to at least MODE 3.
    Date of issuance: May 8, 1998.
    Effective date: May 8, 1998.
    Amendment Nos.: Unit 1--Amendment No. 59; Unit 2--Amendment No. 45.
    Facility Operating License Nos. NPF-87 and NPF-89: The amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: April 20, 1998, (63 FR 
19534). The Commission's related evaluation of the amendments is 
contained in a Safety Evaluation dated May 8, 1998.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: University of Texas at 
Arlington Library, Government Publications/Maps, 702 College, P.O. Box 
19497, Arlington, TX 76019.

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

    Date of application for amendment: August 8, 1997, as supplemented 
by letter dated November 10, 1997.
    Brief description of amendment: The amendment revises the feedwater 
isolation engineered safety feature actuation system (ESFAS) functions 
in Technical Specification Tables 3.3-3, 3.3-4, and 4.3-2.
    Date of issuance: April 23, 1998.
    Effective date: April 23, 1998, to be implemented within 30 days 
from the date of issuance.
    Amendment No.: 126.
    Facility Operating License No. NPF-30: The amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: December 17, 1997 (62 
FR 66144) The November 10, 1997, supplemental letter provided 
additional clarifying information that did not change the staff's 
original no significant hazards consideration determination. The 
Commission's related evaluation of the amendment is contained in a 
Safety Evaluation dated April 23, 1998.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: University of Missouri--
Columbia, Elmer Ellis Library, Columbia, Missouri 65201-5149.

Wolf Creek Nuclear Operating Corporation, Docket No. 50-482, Wolf Creek 
Generating Station, Coffey County, Kansas

    Date of amendment request: February 4, 1998.

[[Page 27772]]

    Brief description of amendment: The amendment would revise 
Technical Specification 3.2.4, ``Quadrant Power Tilt Ratio,'' (QPTR) 
and its associated Bases to reflect (1) a change in the action for 
determining QPTR when QPTR is above 1.02, (2) a change in the 
completion time for resetting the power range neutron flux-high trip 
setpoints after QPTR is determined to be above 1.02, and (3) deletion 
of actions requiring QPTR to be restored within 24 hours, QPTR to be 
verified during a return to power operation, resetting the power range 
neutron flux-high trip setpoint to less than 55 percent following a 
power reduction to 50 percent reactor thermal power or below, and 
actions for QPTR in excess of 1.09.
    Date of issuance: April 27, 1998.
    Effective date: April 27, 1998, to be implemented within 60 days 
from the date of issuance.
    Amendment No.: 116.
    Facility Operating License No. NPF-42: The amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: March 25, 1998 (63 FR 
14489) The Commission's related evaluation of the amendment is 
contained in a Safety Evaluation dated April 27, 1998.
    No significant hazards consideration comments received: No.
    Local Public Document Room locations: Emporia State University, 
William Allen White Library, 1200 Commercial Street, Emporia, Kansas 
66801 and Washburn University School of Law Library, Topeka, Kansas 
66621.

    Dated at Rockville, MD., this 13th day of May 1998.

    For the Nuclear Regulatory Commission.
Elinor G. Adensam,
Acting Director, Division of Reactor Projects--III/IV, Office of 
Nuclear Reactor Regulation.
[FR Doc. 98-13223 Filed 5-19-98; 8:45 am]
BILLING CODE 7590-01-P