[Federal Register Volume 62, Number 127 (Wednesday, July 2, 1997)]
[Notices]
[Pages 35846-35858]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: 97-17140]


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NUCLEAR REGULATORY COMMISSION


Biweekly Notice; Applications and Amendments to Facility 
Operating Licenses Involving No Significant Hazards Considerations

I. Background

    Pursuant to Public Law 97-415, the U.S. Nuclear Regulatory 
Commission (the Commission or NRC staff) is publishing this regular 
biweekly notice. Public Law 97-415 revised section 189 of the Atomic 
Energy Act of 1954, as amended (the Act), to require the Commission to 
publish notice of any amendments issued, or proposed to be issued, 
under a new provision of section 189 of the Act. This provision grants 
the Commission the authority to issue and make immediately effective 
any amendment to an operating license upon a determination by the 
Commission that such amendment involves no significant hazards 
consideration, notwithstanding the pendency before the Commission of a 
request for a hearing from any person.
    This biweekly notice includes all notices of amendments issued, or 
proposed to be issued from June 9, 1997, through June 20, 1997. The 
last biweekly notice was published on June 18, 1997 (62 FR 33117).

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 
Review and Directives Branch, Division of Freedom of Information and 
Publications 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 August 1, 1997, 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

[[Page 35847]]

which is available at the Commission's Public Document Room, the Gelman 
Building, 2120 L Street, NW., Washington, DC and at the local public 
document room for the particular facility involved. If a request for a 
hearing or petition for leave to intervene is filed by the above date, 
the Commission or an Atomic Safety and Licensing Board, designated by 
the Commission or by the Chairman of the Atomic Safety and Licensing 
Board Panel, will rule on the request and/or petition; and the 
Secretary or the designated Atomic Safety and Licensing Board will 
issue a notice of a hearing or an appropriate order.
    As required by 10 CFR 2.714, a petition for leave to intervene 
shall set forth with particularity the interest of the petitioner in 
the proceeding, and how that interest may be affected by the results of 
the proceeding. The petition should specifically explain the reasons 
why intervention should be permitted with particular reference to the 
following factors: (1) The nature of the petitioner's right under the 
Act to be made a party to the proceeding; (2) the nature and extent of 
the petitioner's property, financial, or other interest in the 
proceeding; and (3) the possible effect of any order which may be 
entered in the proceeding on the petitioner's interest. The petition 
should also identify the specific aspect(s) of the subject matter of 
the proceeding as to which petitioner wishes to intervene. Any person 
who has filed a petition for leave to intervene or who has been 
admitted as a party may amend the petition without requesting leave of 
the Board up to 15 days prior to the first prehearing conference 
scheduled in the proceeding, but such an amended petition must satisfy 
the specificity requirements described above.
    Not later than 15 days prior to the first prehearing conference 
scheduled in the proceeding, a petitioner shall file a supplement to 
the petition to intervene which must include a list of the contentions 
which are sought to be litigated in the matter. Each contention must 
consist of a specific statement of the issue of law or fact to be 
raised or controverted. In addition, the petitioner shall provide a 
brief explanation of the bases of the contention and a concise 
statement of the alleged facts or expert opinion which support the 
contention and on which the petitioner intends to rely in proving the 
contention at the hearing. The petitioner must also provide references 
to those specific sources and documents of which the petitioner is 
aware and on which the petitioner intends to rely to establish those 
facts or expert opinion. Petitioner must provide sufficient information 
to show that a genuine dispute exists with the applicant on a material 
issue of law or fact. Contentions shall be limited to matters within 
the scope of the amendment under consideration. The contention must be 
one which, if proven, would entitle the petitioner to relief. A 
petitioner who fails to file such a supplement which satisfies these 
requirements with respect to at least one contention will not be 
permitted to participate as a party.
    Those permitted to intervene become parties to the proceeding, 
subject to any limitations in the order granting leave to intervene, 
and have the opportunity to participate fully in the conduct of the 
hearing, including the opportunity to present evidence and cross-
examine witnesses.
    If a hearing is requested, the Commission will make a final 
determination on the issue of no significant hazards consideration. The 
final determination will serve to decide when the hearing is held.
    If the final determination is that the amendment request involves 
no significant hazards consideration, the Commission may issue the 
amendment and make it immediately effective, notwithstanding the 
request for a hearing. Any hearing held would take place after issuance 
of the amendment.
    If the final determination is that the amendment request involves a 
significant hazards consideration, any hearing held would take place 
before the issuance of any amendment.
    A request for a hearing or a petition for leave to intervene must 
be filed with the Secretary of the Commission, U.S. Nuclear Regulatory 
Commission, Washington, DC 20555-0001, Attention: Docketing and 
Services Branch, 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 50-454 and STN 50-455, 
Byron Station, Unit Nos. 1 and 2, Ogle County, Illinois

    Date of amendment request: May 6, 1997.
    Description of amendment request: The proposed amendment would 
revise Technical Specification (TS) 3/4.7.5, ``Ultimate Heat Sink,'' 
and the associated bases to support steam generator replacement and to 
incorporate recent Ultimate Heat Sink (UHS) design evaluations. The 
replacement steam generators have a larger primary side volume which 
results in a larger mass/energy release to the containment in the event 
of a loss-of-coolant accident (LOCA), and a corresponding increase in 
the heat load to the UHS.
    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.
    TS 3/4.7.5 establishes the operating requirements for the UHS. 
Operation of the UHS within its design basis ensures the following: 
(1) Sufficient cooling capacity is available for continued operation 
of safety related equipment during normal and accident conditions 
and (2) adequate inventory is available to provide a 30-day cooling 
water supply to safety related equipment. Design analyses supporting 
the proposed TS changes provide full qualification of the UHS.
    A loss of off site power (LOOP) coincident with a loss of 
coolant accident (LOCA), designated a LOOP/LOCA, on one unit, in 
conjunction with the non-accident unit proceeding to an orderly 
shutdown and cooldown from maximum power using normal operating 
procedures, remains the limiting design basis event for the UHS 
basin temperature.
    The proposed changes to the UHS Limiting Condition for Operation 
for basin temperature and the number of fans running do not, in 
themselves, factor into any initiating event for Updated Final 
Safety Analysis Report (UFSAR) Chapter 15 accidents and, 
consequently, do not increase the probability of occurrence for 
these previously evaluated accidents.
    The UHS plays a vital role in mitigating the consequences of any 
accident or transient. The proposed changes will ensure that the

[[Page 35848]]

minimum conditions necessary for the UHS to perform its design 
functions will always be met. Engineering calculations demonstrate 
that the SX [essential service water] pump discharge design 
temperature limit of 100 deg.F, which was assumed as an initial 
input for the accident analyses, is preserved. Consequently, the 
proposed changes to the number of cooling tower fans required to be 
running in high speed relative to the SX pump discharge temperature 
do not increase the consequences of any accident previously 
evaluated.
    The two unit plant trip from full power with the loss of normal 
auxiliary feedwater (AF) supply source has been shown to be more 
limiting than the LOOP/LOCA scenario for UHS makeup and volume 
considerations.
    The proposed changes to the UHS LCO for minimum basin water 
level do not, in themselves, factor into any initiating event for 
the UFSAR Chapter 15 accidents and, consequently, do not increase 
the probability of occurrence for these previously evaluated 
accidents.
    The proposed changes to increase the minimum basin water levels 
ensure there is a sufficient volume of water in the UHS basin at all 
times. With these proposed changes, the UHS will perform its design 
function for the required 30 days, and the consequences of any 
accident previously evaluated are not increased.
    2. The proposed change does not create the possibility of a new 
or different kind of accident from any accident previously 
evaluated.
    The supporting analyses for the revised TS 3/4.7.5 do not 
involve a new or different kind of accident from any accident 
previously evaluated. The proposed limits on SX basin minimum water 
level, maximum basin temperature, and the number of fans operating 
are within the design capabilities of the UHS, and ensure that the 
UHS will always be in a condition to perform its design function in 
the event of an accident or transient. New and revised analyses 
which support the requested TS changes ensure the full qualification 
of the UHS. The UHS will not be operated in a different manner such 
that the possibility of a new or different kind of accident would be 
created. Consequently, these changes do not create the possibility 
of a new or different kind of accident from those previously 
evaluated.
    3. The proposed change does not involve a significant reduction 
in a margin of safety.
    The proposed limits on SX basin minimum water level and maximum 
temperature are based on the results of new and revised design 
analyses which ensure that the margin of safety is not reduced. 
Required operator actions with appropriate times are incorporated 
into the analyses. The new limits on temperature and volume will 
ensure that, under the most limiting accident or transient scenario, 
cooling water from the basin will meet the accident analyses SX 
design temperature limit of 100 degrees Fahrenheit and will ensure 
that adequate inventory is available to provide a 30-day cooling 
water supply to safety related equipment. Therefore, the proposed 
changes do not involve a significant reduction in a margin of 
safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 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: Byron Public Library District, 
109 N. Franklin, P.O. Box 434, Byron, Illinois 61010.
    Attorney for licensee: Michael I. Miller, Esquire; Sidley and 
Austin, One First National Plaza, Chicago, Illinois 60603.
    NRC Project Director: Robert A. Capra.

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

    Date of amendment request: June 12, 1997.
    Description of amendment request: The proposed license amendment 
request would change the licensee's name from ``Duke Power Company'' to 
``Duke Energy Corporation'' in the facility operating licenses for the 
Catawba, McGuire, and Oconee nuclear stations as a result of a 
corporate merger of Duke Power Company with PanEnergy Corporation.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    (1) Will the change involve a significant increase in the 
probability or consequences of an accident previously evaluated?
    No. These LARs (license amendment requests) involve an 
administrative change only. The Oconee, McGuire, and Catawba FOLs 
(Facility Operating Licenses) are being changed to reference the new 
corporate name of the licensee. No actual plant equipment or 
accident analyses will be affected by the proposed changes. 
Therefore, these LARs will have no impact on the possibility of any 
type of accident: new, different, or previously evaluated.
    (2) Will the change create the possibility of a new or different 
kind of accident from any accident previously evaluated?
    No. These LARs involve an administrative change only. The 
Oconee, McGuire, and Catawba FOLs are being changed to reference the 
new corporate name of the licensee. No actual plant equipment or 
accident analyses will be affected by the proposed changes and no 
failure modes not bounded by previously evaluated accidents will be 
created. Therefore, these LARs will have no impact on the 
possibility of any type of accident: new, different, or previously 
evaluated.
    (3) Will the change involve a significant reduction in a margin 
of safety?
    No. Margin of safety is associated with confidence in the 
ability of the fission product barriers (i.e., fuel and fuel 
cladding, Reactor Coolant System pressure boundary, and containment 
structure) to limit the level of radiation dose to the public. These 
LARs involve an administrative change only. The Oconee, McGuire, and 
Catawba FOLs are being changed to reference the new corporate name 
of the licensee.

    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 29730.
    Attorney for licensee: Mr. Paul R. Newton, Legal Department 
(PB05E), Duke Power Company, 422 South Church Street, Charlotte, North 
Carolina 28242.
    NRC Project Director: Herbert N. Berkow.

Duke Power Company, Docket Nos. 50-369 and 50-370, McGuire Nuclear 
Station, Units 1 and 2, Mecklenburg County, North Carolina

    Date of amendment request: June 12, 1997
    Description of amendment request: The proposed license amendment 
request would change the licensee's name from ``Duke Power Company'' to 
``Duke Energy Corporation'' in the facility operating licenses for the 
Catawba, McGuire, and Oconee nuclear stations as a result of a 
corporate merger of Duke Power Company with PanEnergy Corporation.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    (1) Will the change involve a significant increase in the 
probability or consequences of an accident previously evaluated?
    No. These LARs (license amendment requests) involve an 
administrative change only. The Oconee, McGuire, and Catawba FOLs 
(Facility Operating Licenses) are being changed to reference the new 
corporate name of the licensee. No actual plant equipment or 
accident analyses will be affected by the proposed changes. 
Therefore, these LARs will have no impact on the possibility of any 
type of accident: new, different, or previously evaluated.
    (2) Will the change create the possibility of a new or different 
kind of accident from any accident previously evaluated?

[[Page 35849]]

    No. These LARs involve an administrative change only. The 
Oconee, McGuire, and Catawba FOLs are being changed to reference the 
new corporate name of the licensee. No actual plant equipment or 
accident analyses will be affected by the proposed changes and no 
failure modes not bounded by previously evaluated accidents will be 
created. Therefore, these LARs will have no impact on the 
possibility of any type of accident: new, different, or previously 
evaluated.
    (3) Will the change involve a significant reduction in a margin 
of safety?
    No. Margin of safety is associated with confidence in the 
ability of the fission product barriers (i.e., fuel and fuel 
cladding, Reactor Coolant System pressure boundary, and containment 
structure) to limit the level of radiation dose to the public. These 
LARs involve an administrative change only.
    The Oconee, McGuire, and Catawba FOLs are being changed to 
reference the new corporate name of the licensee.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: J. Murrey Atkins Library, 
University of North Carolina at Charlotte, 9201 University City 
Boulevard, North Carolina 28223-0001.
    Attorney for licensee: Mr. Albert Carr, Duke Power Company, 422 
South Church Street, Charlotte, North Carolina 28242.
    NRC Project Director: Herbert N. Berkow.

Duke Power Company, Docket Nos. 50-269, 50-270 and 50-287, Oconee 
Nuclear Station, Units 1, 2, and 3, Oconee County, South Carolina

    Date of amendment request: June 12, 1997.
    Description of amendment request: The proposed license amendment 
request would change the licensee's name from ``Duke Power Company'' to 
``Duke Energy Corporation'' in the facility operating licenses for the 
Catawba, McGuire, and Oconee nuclear stations as a result of a 
corporate merger of Duke Power Company with PanEnergy Corporation.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    (1) Will the change involve a significant increase in the 
probability or consequences of an accident previously evaluated?
    No. These LARs (license amendment requests) involve an 
administrative change only. The Oconee, McGuire, and Catawba FOLs 
(Facility Operating Licenses) are being changed to reference the new 
corporate name of the licensee. No actual plant equipment or 
accident analyses will be affected by the proposed changes. 
Therefore, these LARs will have no impact on the possibility of any 
type of accident: new, different, or previously evaluated.
    (2) Will the change create the possibility of a new or different 
kind of accident from any accident previously evaluated?
    No. These LARs involve an administrative change only. The 
Oconee, McGuire, and Catawba FOLs are being changed to reference the 
new corporate name of the licensee. No actual plant equipment or 
accident analyses will be affected by the proposed changes and no 
failure modes not bounded by previously evaluated accidents will be 
created. Therefore, these LARs will have no impact on the 
possibility of any type of accident: new, different, or previously 
evaluated.
    (3) Will the change involve a significant reduction in a margin 
of safety?
    No. Margin of safety is associated with confidence in the 
ability of the fission product barriers (i.e., fuel and fuel 
cladding, Reactor Coolant System pressure boundary, and containment 
structure) to limit the level of radiation dose to the public. These 
LARs involve an administrative change only. The Oconee, McGuire, and 
Catawba FOLs are being changed to reference the new corporate name 
of the licensee.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: Oconee County Library, 501 
West South Broad Street, Walhalla, South Carolina 29691.
    Attorney for licensee: J. Michael McGarry III, Winston and Strawn, 
1200 17th Street, NW., Washington, DC 20036.
    NRC Project Director: Herbert N. Berkow.

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

    Date of amendment request: May 30, 1997.
    Description of amendment request: Technical Specification (TS) 
Surveillances 4.5.2.f and 4.6.2.2.b require the periodic flow testing 
of the recirculation spray system pumps. The proposed amendment would 
change the surveillances by replacing the pump differential acceptance 
criteria with a pump acceptance curve.
    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:

    NNECO has reviewed the proposed revision in accordance with 
10CFR50.92 and has concluded that the revision does not involve a 
significant hazards consideration (SHC). The basis for this 
conclusion is that the three criteria of 10CFR50.92(c) are not 
satisfied. The proposed revision does not involve [an] SHC because 
the revision would not:
    1. Involve a significant increase in the probability or 
consequence of an accident previously evaluated.
    The proposed changes to Technical Specification Surveillances 
4.5.2.f and 4.6.2.2.b will modify the surveillance acceptance 
criteria to require that each Recirculation Spray System (RSS) pump 
develop a differential pressure greater than or equal to the pump 
performance curve contained on Figure 3.5-1 when tested according to 
the requirements of Specification 4.0.5. Because it is undesirable 
to test the pumps on recirculation flow to the RWST [reactor water 
storage tank], pump testing will now be performed at lower flows 
than previously performed. Consistent with Specification 4.0.5, one 
point on Figure 3.5-1 will be used to meet the proposed surveillance 
acceptance criteria. Periodically comparing the reference 
differential pressure developed at this reduced flow detects trends 
that might be indicative of pump degradation. The proposed changes 
are consistent with RSS pump design criteria and performing 
surveillance testing does not significantly increase the probability 
of an accident previously evaluated.
    The proposed changes to modify the surveillance acceptance 
criteria to require that each RSS pump develop a differential 
pressure greater than or equal to the pump performance curve 
provides the necessary assurance that the pumps will function as 
required in previous evaluations and does not significantly increase 
the consequence of an accident previously evaluated.
    Therefore, the proposed changes do not involve a significant 
increase in the probability or consequence of an accident previously 
evaluated.
    2. Create the possibility of a new or different kind of accident 
from any accident previously evaluated.
    The proposed changes to the surveillance acceptance criteria of 
the RSS pumps does not change the operation of the Recirculation 
Spray System or any of its components during normal or accident 
evaluations.
    Therefore, the proposed revision does not create the possibility 
of a new or different kind of accident from any accident previously 
evaluated.
    3. Involve a significant reduction in a margin of safety.
    The proposed changes will change the surveillance requirements 
needed to demonstrate operability for each of the RSS pumps. 
Technical Specification Surveillances 4.5.2.f and 4.6.2.2.b will now 
require that each pump meet its acceptance criteria in accordance 
with Figure 3.5-1

[[Page 35850]]

when tested according to the requirements of Specification 4.0.5. 
Figure 3.5-1 will be inserted into the Technical Specifications.
    The new acceptance criteria for the RSS Technical Specification 
surveillance is above the accident analysis curve and is more 
restrictive than the current inservice inspection curve in the 
accident analysis region. The proposed TS curve has been degraded in 
accordance with the recommendations of ASME XI (American Society of 
Mechanical Engineers Boiler and Pressure Vessel Code, Section XI) 
for the full range of flow and will be used to meet the TS 
requirements.
    Therefore, the proposed revision does not involve a significant 
reduction in a margin of safety.
    In conclusion, based on the information provided, it is 
determined that the proposed revision does not involve an SHC.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: Learning Resources Center, 
Three Rivers Community-Technical College, 574 New London Turnpike, 
Norwich, Connecticut, and the Waterford Library, ATTN: Vince Juliano, 
49 Rope Ferry Road, Waterford, Connecticut.
    Attorney for licensee: Lillian M. Cuoco, Esq., Senior Nuclear 
Counsel, Northeast Utilities Service Company, P.O. Box 270, Hartford, 
CT 06141-0270.
    NRC Deputy Director: Phillip F. McKee.

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

    Date of amendment request: June 13, 1997.
    Description of amendment request: The proposed amendment would 
modify Technical Specification (TS) Surveillance Requirement 4.4.1.3.3 
to be consistent with the requirements of TS 3.4.1.3. Specifically, the 
change would bring TS Surveillance 4.4.1.3.3 into agreement with TS 
3.4.1.3 that would require at least two reactor coolant system loops to 
be operable and in operation when the reactor trip system breakers are 
closed during Mode 4.
    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:

    NNECO has reviewed the proposed revision in accordance with 
10CFR50.92 and has concluded that the revision does not involve a 
significant hazards consideration (SHC). The basis for this 
conclusion is that the three criteria of 10CFR50.92(c) are not 
satisfied. The proposed revision does not involve (an) SHC because 
the revision would not:
    1. Involve a significant increase in the probability or 
consequence of an accident previously evaluated.
    The proposed change to Technical Specification Surveillance 
4.4.1.3.3 is being made to bring Technical Specification 
Surveillance 4.4.1.3.3 into agreement with Technical Specification 
3.4.1.3 that requires at least two reactor coolant system loops to 
be operable and in operation when the reactor trip system breakers 
are closed during Mode 4. This requirement was incorporated into 
Technical Specification 3.4.1.3 in Amendment 7. This change to the 
surveillance does not alter the design, operation, maintenance or 
testing of the associated systems as previously analyzed.
    Therefore, the proposed revision does not involve a significant 
increase in the probability or consequence of an accident previously 
evaluated.
    2. Create the possibility of a new or different kind of accident 
from any accident previously evaluated.
    This proposed change does not introduce any new failure modes or 
malfunctions, since the changes only bring Surveillance 4.4.1.3.3 in 
agreement with Technical Specification 3.4.1.3. Additionally, the 
proposed change does not alter the operation of the reactor coolant 
system during normal or accident conditions.
    Therefore, the proposed revision does not create the possibility 
of a new or different kind of accident from any accident previously 
evaluated.
    3. Involve a significant reduction in a margin of safety.
    The proposed change to Technical Specification Surveillance 
4.4.1.3.3 will reword the surveillance to ensure compliance with 
Technical Specification 3.4.1.3. Technical Specification 3.4.1.3 was 
changed in Amendment No. 7 to address the closure of the Reactor 
Trip System breakers in Mode 4. As written, Technical Specification 
Surveillance 4.4.1.3.3 does not adequately ensure compliance with 
Technical Specification 3.4.1.3. This proposed change is necessary 
to bring Surveillance 4.4.1.3.3 in agreement with Technical 
Specification 3.4.1.3 as it was amended.
    Therefore, the proposed revision does not involve a significant 
reduction in a margin of safety.
    In conclusion, based on the information provided, it is 
determined that the proposed revision does not involve an SHC.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: Learning Resources Center, 
Three Rivers Community-Technical College, 574 New London Turnpike, 
Norwich, Connecticut, and the Waterford Library, ATTN: Vince Juliano, 
49 Rope Ferry Road, Waterford, Connecticut.
    Attorney for licensee: Lillian M. Cuoco, Esq., Senior Nuclear 
Counsel, Northeast Utilities Service Company, P.O. Box 270, Hartford, 
CT 06141-0270.
    NRC Deputy Director: Phillip F. McKee.

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

    Date of amendment requests: May 7, 1997, as supplemented May 30, 
1997.
    Description of amendment requests: The proposed amendments would 
remove from the Technical Specifications certain limitations on crane 
operations in the spent fuel pool enclosure relating to spent fuel pool 
special ventilation system operability.
    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.
    Operation of the Prairie Island plant in accordance with the 
proposed changes does not involve a significant increase in the 
probability or consequences of an accident previously evaluated. The 
proposed changes do not involve a physical modification to the 
plant.
    The spent fuel pool special ventilation system is provided to 
mitigate the consequences of a design basis fuel handling accident 
which involves dropping a spent fuel assembly directly onto a stored 
spent fuel assembly. Spent fuel pool special ventilation system 
performance and environmental consequences were based on the 
conservative assumption that all fuel rods in one fuel assembly 
fail. However, evaluation of the mechanical performance of spent 
fuel stored in the spent fuel racks demonstrated that no fuel rods 
fail.
    The proposed changes will continue to require the spent fuel 
pool special ventilation system to be operable to mitigate the 
consequences of a fuel handling accident in accordance with its 
original design intent. Spent fuel pool special ventilation system 
operability is not required in conjunction with crane operations. 
Heavy loads in the spent fuel pool enclosure are handled (1) by 
single-failure-proof cranes with rigging and plant procedures which 
implement Prairie Island commitments to NUREG-0612 [``Control of 
Heavy Loads at Nuclear Power Plants''] or (2) over spent fuel pool 
protective

[[Page 35851]]

covers as described in the Prairie Island USAR [updated safety 
analysis report]. In accordance with the requirements of NUREG-0612, 
use of a single-failure-proof crane with rigging and procedures 
which implement the requirements of NUREG-0612 assures that the 
potential for a load drop is extremely small and the effects of 
heavy load drops are not considered. Spent fuel pool covers prevent 
dropped loads from falling into the spent fuel pool. Thus, there are 
no radiological releases resulting from handling heavy loads in the 
spent fuel pool enclosure for which spent fuel pool special 
ventilation system operability would be required. Therefore, these 
changes do not involve a significant increase in the probability or 
consequences of the fuel handling 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 Technical Specification changes continue to require 
the spent fuel pool special ventilation system to be operable during 
handling of irradiated fuel as originally designed. Heavy loads in 
the spent fuel pool enclosure are handled by means which assure that 
the potential for a dropped load is extremely small (through use of 
single-failure-proof cranes with rigging and plant procedures which 
implement Prairie Island commitments to NUREG-0612) or prevent 
dropped loads from falling into the spent fuel pool (through use of 
spent fuel pool protective covers as described in the USAR). Thus, 
the proposed changes do not create the possibility of a new or 
different kind of accident from any accident previously evaluated 
because the proposed changes, in themselves, do not introduce a new 
mode of plant operation, surveillance requirement or involve a 
physical modification to the plant.
    The proposed changes do not alter the design, function, or 
operation of any plant components and therefore, no new accident 
scenarios are created. Therefore, the possibility of a new or 
different kind of accident from any accident previously evaluated 
would not be created by these amendments.
    3. The proposed amendment(s) will not involve a significant 
reduction in the margin of safety.
    The proposed amendment(s) will continue to require the spent 
fuel pool special ventilation system to operate following a fuel 
handling accident as originally designed. Heavy load crane 
operations in the spent fuel pool enclosure are handled (1) by 
single-failure-proof cranes with rigging and plant procedures which 
implement Prairie Island commitments to NUREG-0612; or (2) over 
spent fuel pool protective covers as described in the Prairie Island 
USAR. Provision of single-failure-proof equipment and compliance 
with the other requirements of NUREG-0612 provides an equivalent 
margin of safety to that which would be demonstrated by analysis of 
the radiological effects of dropped loads. Use of protective covers 
has been previously reviewed and approved by the NRC. Therefore, 
th[ese] proposed amendment(s) (do) not involve a significant 
reduction in the margin of safety.

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

PECO Energy Company, Public Service Electric and Gas Company, Delmarva 
Power and Light Company, and Atlantic City Electric Company, Dockets 
Nos. 50-277 and 50-278, Peach Bottom Atomic Power Station, Units Nos. 2 
and 3, York County, Pennsylvania

    Date of application for amendments: May 9, 1997.
    Description of amendment request: The proposed change revises the 
Peach Bottom Atomic Power Station, Units 2 and 3 technical 
specifications to extend the interval for replacing the primary 
containment purge and exhaust valve inflatable seals.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. The proposed TS (technical specification) changes do not 
involve a significant increase in the probability or consequences of 
an accident previously evaluated.
    Revising SR [surveillance requirement] 3.6.1.3.16 to replace the 
inflatable seals for the Primary Containment purge and exhaust 
valves from every 48 months to every 96 months will not involve a 
significant increase in the probability or consequences of an 
accident previously evaluated. The valves will continue to be leak 
tight throughout the lifetime of the plant. This change will not 
result in increased onsite or offsite radiological dose. This change 
will result in reduced occupational dose exposure.
    This submittal does not propose any change to the existing 
requirements contained in the PBAPS [Peach Bottom Atomic Power 
Station] Technical Specifications for leak testing of the Primary 
Containment purge and exhaust valves per 10 CFR 50, Appendix J, 
``Primary Reactor Containment Leakage Testing For Water-Cooled Power 
Reactors.'' This continued testing will assure the leak tightness of 
the purge and exhaust valves.
    The T-ring materials (Ethylene Propylene) has been found to 
withstand normal and accident thermal exposures for the design life 
of the plant based on thermal aging analysis. The elastomer seat 
material will provide acceptable seat tightness when exposed to a 
total integrated radiation dose of 10E7 rads based on information 
provided by EPRI [Electric Power Research Institute] in technical 
report NP-2129, entitled ``Radiation Effects on Organic Material in 
Nuclear Plants.'' The radiation dose of 10E7 rads bounds the design 
basis accident dose to which these valves would be exposed. The 
radiation dose these valves are exposed to during normal operation 
is insignificant as compared to the accident dose. Based on this, 
radiation effects from the additional exposure resulting from the 
extended replacement frequency will not adversely impact the T-ring 
seat material.
    2. The proposed TS changes do not create the possibility of a 
new or different kind of accident from any accident previously 
evaluated.
    Revising SR 3.6.1.3.16 to replace the inflatable seals for the 
Primary Containment purge and exhaust valves from every 48 months to 
every 96 months does not create the possibility of a new or 
different kind of accident from any accident previously evaluated. 
This change does not involve any physical changes to a plant 
structure, system, or component (SSC) which could act as an accident 
initiator. The design, function, and reliability of the Primary 
Containment purge and exhaust valves are also not impacted by this 
change. This activity does not adversely influence any equipment, 
which is required to be maintained operable for the prevention or 
mitigation of accidents or transients. Furthermore, implementation 
of the proposed changes will not adversely affect the manner in 
which plant SSC are operated.
    3. The proposed TS changes do not involve a significant 
reduction in a margin of safety.
    No margins of safety are reduced as a result of the proposed TS 
changes. The proposed changes do not alter the intended operation of 
plant structures, systems, or components utilized in the mitigation 
of accidents or transients. The operating experience of these valves 
and the testing performed in accordance with 10 CFR 50, Appendix J 
provides a high level of confidence in the ability of these valves 
to perform their intended safety function with respect to valve leak 
tightness.

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

[[Page 35852]]

Counsel, PECO Energy Company, 2301 Market Street, Philadelphia, PA 
19101.
    NRC Project Director: John F. Stolz.

PECO Energy Company, Public Service Electric and Gas Company, Delmarva 
Power and Light Company, and Atlantic City Electric Company, Dockets 
Nos. 50-277 and 50-278, Peach Bottom Atomic Power Station, Units Nos. 2 
and 3, York County, Pennsylvania

    Date of application for amendments: May 23, 1997.
    Description of amendment request: The proposed change revises the 
Peach Bottom Atomic Power Station (PBAPS), Units 2 and 3 Technical 
Specifications (TS) to exclude the measured Main Steam Isolation Valves 
(MSIVs) leakage from the total Type B and C local leak rate test (LLRT) 
results.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. The proposed TS changes do not involve a significant increase 
in the probability or consequences of an accident previously 
evaluated.
    Excluding the MSIV leakage from the total Type B and C LLRT 
results does not involve any change in the safety function or method 
of operation of any plant component, system, or structure. No new 
accident initiators or failure modes are created as a result of this 
change. Therefore, this change will not result in an increase in the 
probability of an accident previously evaluated.
    The MSIV leakage release pathway is of significance only for the 
evaluation of the design basis LOCA (loss-of-coolant accident) as 
described in the PBAPS, Units 2 and 3 UFSAR (updated final safety 
analysis report). The doses effectively reflected in the PBAPS, 
Units 2 and 3 UFSAR reflect the impact of a 0.635% Primary 
Containment volume per day Primary to Secondary Containment leakage, 
plus a 0.145% Secondary Containment bypass leakage to the condenser. 
Since accident consequences already reflect both leakage release 
pathways, the consequences of the design basis LOCA are not 
increased.
    2. The proposed TS changes do not create the possibility of a 
new or different kind of accident from any accident previously 
evaluated.
    The MSIV's provide the means for mitigating the radiological 
consequences of an accident. Revising Section 5.5.12 of the PBAPS, 
Units 2 and 3 TS to exclude the measured MSIVs leakage from the 
total Type B and C LLRT results has no effect on accident initiators 
which lead to a new or different kind of accident. This change will 
not involve any changes to plant systems, structures, or components 
which could act as new accident initiators. The design, function, 
and reliability of the MSIVs are also not impacted by this change. 
Therefore, this change will not create the possibility of a new or 
different kind of accident from any previously evaluated.
    3. The proposed TS changes do not involve a significant 
reduction in a margin of safety.
    No margins of safety are reduced as a result of this change to 
the TS. No safety limits will be changed as a result of this TS 
change. The MSIVs will continue to perform their intended safety 
function. The combined dose rates from the two release paths (i.e., 
Primary to Secondary Containment leakage and Secondary Containment 
bypass leakage) are unchanged as a result of this change, and are 
within the limits of 10 CFR 100, and in conformance with NUREG-0737 
post-accident access requirements.

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

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

    Date of amendment request: March 27, May 28, and June 4, 1997.
    Description of amendment request: The proposed amendment would 
revise the technical specifications (TSs) as follows:
Part 1--Boron Concentration Changes
    The Cycle 2 core design for Watts Bar (WBN) will include a longer 
fuel cycle and more highly enriched fuel (from 3.1 percent to 3.7 
percent). To accommodate this design, the refueling water storage tank 
(RWST) and accumulator boron concentrations will be increased to 
provide enough boron in the sump to meet the large break loss-of-
coolant accident (LBLOCA) requirement for sump boron concentration. 
This requirement is that during a LBLOCA, the core will remain 
subcritical from boron provided by the emergency core cooling system 
(ECCS), which takes suction from the RWST and containment sump.
    The increase in RWST (TS 3.5.4) and accumulator (TS 3.5.1) boron 
concentrations will be from a range of 2000-2100 ppm to 2500-2700 ppm 
and from 1900-2100 to a range of 2400-2700 ppm, respectively. 
Associated changes are proposed for TS Bases B 3.5.4.
Part 2--Safety Limits, Instrumentation, and Reactor Coolant System
    Watts Bar has experienced hot leg temperature fluctuations, 
including random spikes, which decrease the operating margin to both 
the overtemperature delta temperature (OTDT) and overpower delta 
temperature (OPDT) reactor trip setpoints. These fluctuations have 
caused, in some cases, the plant to experience OT alarms during steady-
state operation since the temperature fluctuations reduced the 
operating margin. To mitigate the temperature fluctuations and 
associated alarms, the OTDT and OPDT setpoints have been enhanced to 
increase the operating margin associated with these trip functions.
    In addition, Watts Bar has decided to reduce the plant thermal 
design flow from 97,500 gpm per loop to 93,100 gpm per loop (total of 
390,000 gpm) to accommodate 10 percent steam generator tube plugging 
and a 2 percent reduction in thermal design flow (RTDF).
    Also, Watts Bar has decided to implement a tolerance of 0.6 deg.F 
for the TS Surveillance for indicated differential temperature and 1 
deg.F tolerance for the surveillance of TAVG (identified as 
T prime and T double prime in the TSs). The use of this tolerance will 
help to determine whether the indicated DT and TAVG should 
be left as is, or rescaled during the surveillance. These tolerances 
have been incorporated as biases into the uncertainty analysis for the 
affected protection system functions. These functions include the OTDT, 
OPDT and vessel DT equivalent to power (used in the steam generator 
low-low water level trip functions). As a result of implementing these 
biases into the protection system functions (and the changes to the 
OTDT/OPDT setpoints and reduced TDF), the Allowable Value in the TSs 
for the OTDT, OPDT and vessel DT equivalent to power functions have 
been modified.
    The licensee's safety evaluation has been prepared to allow for 
plant operation during Cycle 2 with the revised OTDT and OPDT 
setpoints, the thermal design flow of 93,100 gpm and the tolerances for 
indicated differential temperature, T prime and T double prime. To 
obtain sufficient departure from nucleate boiling (DNB) margin for the 
OTDT/OPDT setpoint, reduced TDF and Cycle 2 design features, it was 
necessary to implement the RTDP. The

[[Page 35853]]

RTDP program changes the uncertainty treatment for core power, 
TAVG, pressurizer pressure, and RCS flow. These 
uncertainties have been incorporated, where applicable, into the safety 
analyses addressed in the Safety Evaluation.
    The following TSs will be changed to incorporate the OTDT/OPDT 
margin enhancement, thermal design flow of 93,100 gpm and tolerances 
for indicated differential temperature, T prime and T double prime.
    The Reactor Core Safety Limits (TS Figure 2.1.1-1 of the licensee's 
application) have been modified to improve DNB margin. The Allowable 
Values for the Vessel DT Equivalent to Power input to Steam Generator 
Water Level Low-Low in the Reactor Trip System Instrumentation (Table 
3.3.1-1, page 4) and Engineered Safety Feature Actuation System (ESFAS) 
Instrumentation (Table 3.3.2-1, page 4), have been changed to reflect 
the addition of a 0.6+F tolerance to the measurement of 
indicated differential temperature.
    The revised reactor core safety limits lines allow for changes in 
the OTDT/OPDT reactor trip setpoints to improve operating margin. The 
allowable values for these functions in the Reactor Trip System 
Instrumentation (TS Table 3.3.1-1) have changed as a result of 
including tolerances for indicated differential temperature, T prime 
and T double prime in the uncertainty analysis. Several setpoint gains 
and time constants have been modified to enhance plant operation.
    Regarding the RCS Pressure, Temperature and Flow DNB Limits 
(Section 3.4.1), the RCS average temperature limit has been revised to 
account for the change in uncertainty from implementing RTDP. The total 
RCS flow has been modified to account for the reduced thermal design 
flow from 97,500 gpm to 93,100 gpm. The total flow value in the 
Technical Specification includes an allowance for instrument 
uncertainty.
    Associated changes have been made to the following TS Bases 
sections: Reactor Core Safety Limits (Section B 2.1.1); Nuclear 
Enthalpy Rise Hot Channel Factor (Section B 3.2.2); Reactor Trip System 
Functions OTDT, OPDT and Steam Generator Water Level Low-low (Vessel 
Delta T Equivalent to Power) (Section B 3.3.1); Reactor Trip System 
Functions--Reactor Coolant Flow--Low (Single Loop and Two Loops) 
(Section B 3.3.1); ESFAS Instrumentation (Section B 3.3.2); RCS 
Pressure, Temperature, and Flow DNB (Section B 3.4.1).
Part 3--Addition To Core Operating Limit Report Methodologies
    The amendment would revise the Core Operating Limits Report (COLR) 
methodologies listed in TS 5.9.5.b to add the reference to the 
Westinghouse report WCAP-12610-P-A, ``Vantage + Fuel Assembly Reference 
Core Report.'' The report reflects use of fuel assemblies in Cycle 2 
using ZIRLO fuel rod cladding.
    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:

Part 1--Boron Concentration Changes
    The Nuclear Regulatory Commission has provided standards for 
determining whether a significant hazards consideration exists (10 CFR 
50.92 (c)). A proposed amendment to an operating license for a facility 
involves no significant hazards consideration if 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;
    The RWST and accumulator boron concentrations do not affect any 
initiating event for accidents currently evaluated in the FSAR 
[final safety analysis report]. The increased concentrations will 
not adversely affect the performance of any system or component 
which is placed in contact with the RWST or accumulator water. The 
integrity and operability of the stainless steel surfaces in the 
RWST, accumulator and affected NSSS [nuclear steam supply system] 
components/systems will be maintained. The decrease in solution pH 
is small and will not degrade the stainless steel. Also, the 
integrity of the Class 1E instrumentation and control equipment will 
be maintained since the lower sump pH, resulting from the increased 
boron concentrations, is still within the applicable equipment 
qualification [EQ] limits. These limits are set to preclude the 
possibility of chloride induced stress corrosion cracking and assure 
that there is no significant degradation of polymer materials. The 
design, material and construction standards of all components which 
are placed in contact with the RWST and accumulator water remain 
unaffected.
    For the evaluations, the consequences of an accident previously 
evaluated in the FSAR will not be increased. There is no increase in 
the LOCA accident consequences. The changes in the concentrations 
increase the amount of boron in the sump during a LOCA. The 
increased boron in the sump is sufficient to maintain the core in a 
subcritical condition during a LOCA. Also, a revised hot leg 
switchover time has been calculated and will be implemented in the 
plant EOPs (emergency operating procedures). Thus, there will be no 
boron precipitation in the core during a LOCA.
    Furthermore, there is no increase in consequences of the non-
LOCA events. The concentration changes are a benefit to the SLB 
(steam line break) at full power analysis due to the reduction in 
power during the accident. The loss of normal feedwater event is not 
sensitive to changes in the RWST and accumulator boron 
concentrations. The concentration changes do not affect the 
inadvertent operation of ECCS analysis since the minimum DNBR 
(departure from nucleate boiling ratio) occurs at the event 
initiation, and the concentration changes do not affect the analysis 
trend.
    Finally, the concentration changes are a benefit for the SLB M&E 
(mass and energy) release and SGTR (steam generator tube rupture) 
events since the increased boron increases the available shutdown 
margin for these events. In addition, the increase in RWST and 
accumulator boron concentrations and subsequent slight decrease in 
containment sump and a spray pH does not impact the LOCA dose 
evaluation since pH is not a function of radionuclide concentration. 
Therefore, the present analysis remains bounding. Also, the slight 
decrease in sump, core and spray fluid pH has been evaluated to not 
impact the corrosion rate (and subsequent generation of Hydrogen) of 
Aluminum and Zinc inside containment significantly that the present 
analysis does not remain bounding. Further, the decreased sump, core 
and spray fluid pH has been evaluated to not affect the amount of 
hydrogen generated from the radiolytic decomposition of the sump and 
core solution. In view of the preceding, it is concluded that the 
proposed change will not increase the consequences of an accident 
previously evaluated in the FSAR.
    (2) or create the possibility of a new or different kind of 
accident from any accident previously evaluated;
    The changes to the RWST and accumulator concentrations do not 
cause the initiation of any accident nor create any new credible 
limiting single failure. The changes do not result in a condition 
where the design, material, and construction standards of the RWST 
and accumulators and other potentially affected NSSS components, 
that were applicable prior to the changes, are altered. * * * *
    The changes do not invalidate any of the accident analyses 
results or conclusions. All of the safety analysis acceptance 
criteria continue to be met. The changes in the concentrations 
increase the amount of boron in the sump during a LOCA. The 
increased boron in the sump is sufficient to maintain the core in a 
subcritical condition during a LOCA. Also, a revised hot leg 
switchover time has been calculated and will be implemented in the 
plant EOPs. Thus, there will be no boron precipitation in the core 
during a LOCA.
    Furthermore, there is no possibility of a different kind of non-
LOCA event. The concentration changes are a benefit to the SLB at 
full power analysis due to the reduction in power increase during 
the accident. The loss of normal feedwater event is not sensitive to 
changes in the RWST and

[[Page 35854]]

accumulator boron concentrations. The concentration changes do not 
affect the inadvertent operation at ECCS analysis since the minimum 
DNBR occurs at the event initiation, and the concentration changes 
do not affect the analysis trend.
    Finally, the concentration changes are a benefit for the SLB M&E 
release and SGTR events since the increased boron increases the 
available shutdown margin for these events.
    (3) or involve a significant reduction in a margin of safety.
    The changes do not invalidate any of the non-LOCA safety 
analysis results or conclusions, and all of the non-LOCA safety 
analysis acceptance criteria continue to be met. The margin of 
safety associated with the licensing basis LBLOCA and SBLOCA (small-
break loss-of-coolant accident) analyses is not reduced as a result 
of the proposed changes. Since adequate margin to the PCT (peak 
cladding temperature) limit of 2200+F has been 
maintained, no degradation in the margin of safety to the design 
failure point (fuel melt) has been calculated. The licensing basis 
containment and steam line break mass and energy releases remain 
bounding, and the SGTR event acceptance criteria continue to be met. 
Furthermore, the changes do not affect the safety related 
performance of the RWST, accumulator or related NSSS components.

Part 2--Safety Limits, Instrumentation, and Reactor Coolant System.

    The Nuclear Regulatory Commission has provided standards for 
determining whether a significant hazards consideration exists (10 
CFR 50.92 (c)). A proposed amendment to an operating license for a 
facility involves no significant hazards consideration if 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;
    The proposed changes do not result in a condition where the 
design, material, and construction standards, which were applicable 
prior to the changes, are altered. The revised OTDT and OPDT 
setpoints do not require any hardware changes and are used for 
accident mitigation. Thus, the setpoint changes do not increase the 
probability of the accident.
    All of the affected NSSS systems and components have been 
evaluated with the TDF (thermal design flow) of 93,100 gpm. The 
primary loop components (reactor vessel, reactor internals, CRDMs 
(control rod drive mechanism), loop piping and supports, reactor 
coolant pump, steam generator, and pressurizer) meet the applicable 
structural limits with the revised TDF of 93,100 gpm and will 
continue to perform their design functions. The RCCA (rod cluster 
control assembly) drop time remains unaffected and the current 
design core bypass flow remains valid. No additional steam generator 
tubes need to be plugged to mitigate the potential for U-Bend 
fatigue. Also, all of the NSSS systems will still perform their 
intended design functions. The pressurizer spray flow remains above 
the design value and the pressurizer relief system remains 
unaffected since the TDF is lower than the current design flow and 
the required pressure drop is lower. The design of the auxiliary 
system components remains bounding for the revised TDF and the 
corresponding changes to the NSSS thermal hydraulic parameters. In 
addition, all of the NSSS/BOP (nuclear steam supply system/balance 
of plant) interface systems will perform their intended design 
functions. The steam generator safety valves will provide adequate 
relief capacity to maintain the steam generator within applicable 
design limits. The ADVs [atmospheric dump valves] will still relieve 
20 percent of the maximum full load steam flow. The steam dump 
system will still relieve 40 percent of the maximum full load steam 
flow.
    All of the applicable acceptance criteria for the accidents 
described in the FSAR continue to be met. The LBLOCA analysis 
currently uses a TDF of 93,100 gpm. Thus, no adjustments are 
required for the LBLOCA input parameters to accommodate the TDF of 
93,100 gpm. The SBLOCA has been performed with the TDF of 93,100 
gpm, and the corresponding PCT is well below the 2200+F 
limit. The post LOCA boron concentration and the hot leg switchover 
time are unaffected. The revised thermal design procedure has been 
implemented to obtain sufficient DNB margin to account for the TDF 
of 93,100 gpm, the new OTDT/OPDT setpoints and the Cycle 2 design 
features. All of the non-LOCA analyses have been re-analyzed or re-
evaluated and all of the applicable acceptance criteria continue to 
be met.
    The SLB radiological doses are unaffected and are still within 
the existing licensing basis limits. The margin to overfill during 
the SGTR event has been improved and the offsite doses during an 
SGTR have been re-calculated and shown to be well within the 
10CFR100 guidelines. The plant control systems will still provide 
adequate response for the Condition 1 transients without causing a 
reactor trip on OTDT and OPDT.
    Finally, the changes in the tolerances for indicated 
differential temperature, T prime and T double prime do not require 
any hardware modifications and only require changes to the Technical 
Specification Allowable Values for the OPDT and OTDT setpoints and 
for the vessel DT equivalent to power functions. Thus, there is no 
increase in the probability of an accident since the appropriate 
Allowable Values have been modified to determine channel operability 
for these functions.
    (2) or create the possibility of a new or different kind of 
accident from any accident previously evaluated;
    The proposed changes do not cause the initiation of any accident 
nor create any new limiting single failures. The OTDT and OPDT 
protection functions are used for accident mitigation and do not 
initiate any accidents. Also, the affected systems and components 
will still perform their intended design functions.
* * * 
    The proposed changes do not create any new failure modes for 
safety related equipment. The changes do not result in any original 
design specification, such as seismic requirements, electrical 
separation requirements or equipment qualification being altered. 
The OTDT and OPDT setpoint changes do not require any hardware 
modifications and only require adjustments to the setpoint values. 
The setpoints are modeled in accident analyses which are used to 
demonstrate equipment and structural qualification during a SLB. 
With the setpoint changes and the TDF of 93,100 gpm, the current SLB 
break M&E releases inside containment remain bounding and thus there 
is no effect on the qualification of the equipment inside 
containment during a SLB. The SLB M&E releases outside containment 
have been re-calculated. The analysis of the impacts on equipment 
qualification outside containment has been completed by generating 
new temperature profiles. The application addresses and provides for 
continued qualification of equipment through the normal EQ program.
    Also, with the reduced TDF of 93,100 gpm, the current LOCA M&E 
releases are still bounding, and thus there is no effect on the 
qualification of equipment inside containment during a LOCA. The 
OTDT and OPDT functions are not modeled in the LOCA analyses. 
Furthermore, all of the applicable compartments and subcompartments 
will maintain their integrity during the LOCA and the SLB since the 
mass and energy releases for these compartments and subcompartments 
remain unaffected.
    In addition, the LOCA hydraulic forcing functions remain 
bounding for the TDF of 93,100 gpm. Thus, the applicable NSSS 
systems and components will still perform their structural functions 
during a LOCA.
    Finally, the changes in the tolerances for DTo, T 
prime and T double prime do not require any hardware modifications 
and only require changes to the Technical Specification Allowable 
Values for the OPDT and OTDT setpoints and for the vessel DT 
equivalent to power functions. Thus, there is no increase in the 
probability of an accident different than any previously evaluated 
since the appropriate Allowable Values have been modified to 
determine channel operability for these functions.
    (3) or involve a significant reduction in a margin of safety.
    The margin of safety for the applicable safety analyses has not 
been reduced. The OPDT and OTDT setpoints have been incorporated 
into the affected safety analyses and all safety analysis criteria 
continue to be met. All of the applicable DNB limits continue to be 
met for the non-LOCA analyses. The LBLOCA input parameters do not 
require adjustment for the TDF of 93,100 gpm. The SBLOCA has been 
re-analyzed for the TDF of 93,100 gpm, and the SBLOCA PCT is well 
below the 2200+F limit. The affected NSSS systems and 
components will still meet the applicable design limits and perform 
their intended safety functions with the TDF of 93,100 gpm. Also, 
the SLB and LOCA M&E releases are still within the applicable 
equipment qualification limits. The SGTR doses remain within the 
applicable 10 CFR 100 limits, and the steam generator margin to 
overfill is maintained.
    Summary--Parts I and II. Based on the above, TVA has determined 
that operation of

[[Page 35855]]

Watts Bar in accordance with the proposed amendment would not: (1) 
involve a significant increase in the probability or consequences of 
an accident previously evaluated, (2) create the possibility of a 
new or different kind of accident from any accident previously 
evaluated, or (3) involve a significant reduction in a margin of 
safety. Therefore, operation of Watts Bar in accordance with the 
proposed amendment would not involve a significant hazards 
consideration as defined in 10 CFR 50.92.

Part 3--Addition to Core Operating Limit Report Methodologies

    (1) involve a significant increase in the probability or 
consequences of an accident previously evaluated;
    The use of ZIRLOTM is already permitted by TS section 
4.2.1. Accordingly, the addition of the NRC approved Westinghouse 
COLR methodology reference is administrative in nature. Therefore, 
there is no increase in the probability or consequences of an 
accident previously evaluated.
    (2) or create the possibility of a new or different kind of 
accident from any accident previously evaluated;
    Since the use of ZIRLOTM is already permitted by TS 
section 4.2.1, the addition of the NRC approved Westinghouse COLR 
methodology reference is administrative in nature. Accordingly, no 
new or different kind of accident has been created from those 
previously evaluated.
    (3) or involve a significant reduction in a margin of safety.
    The use of ZIRLOTM is already permitted by TS section 
4.2.1. The addition of the NRC approved Westinghouse COLR 
methodology reference is administrative in nature. Therefore, there 
is no significant reduction in the margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: Chattanooga-Hamilton County 
Library, 1001 Broad Street, Chattanooga, TN 37402.
    Attorney for licensee: General Counsel, Tennessee Valley Authority, 
400 West Summit Hill Drive, ET l0H, Knoxville, Tennessee 37902.
    NRC Project Director: Frederick J. Hebdon.

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.

Wisconsin Public Service Corporation, Docket No. 50-305, Kewaunee 
Nuclear Power Plant, Kewaunee County, Wisconsin

    Date of application for amendment: May 2, 1997.
    Brief description of amendment request: The proposed amendment 
would change the main steam isolation valve (MSIV) closure time 
assumption used in the main steam line break accident analysis and 
referenced in the Basis for Technical Specification 4.7.
    Date of individual notice in Federal Register: May 15, 1997 (62 FR 
26829).
    Expiration date of individual notice: June 16, 1997.
    Local Public Document Room location: University of Wisconsin, 
Cofrin Library, 2420 Nicolet Drive, Green Bay, Wisconsin.

Wisconsin Electric Power Company, Docket Nos. 50-266 and 50-301, Point 
Beach Nuclear Plant, Unit Nos. 1 and 2, Town of Two Creeks, Manitowoc 
County, Wisconsin

    Date of amendment request: September 30, 1996, as supplemented 
November 26, and December 12, 1996, February 13, March 5, April 2, 
April 16, May 9, and June 3, 1997 (TSCR 192).
    Description of amendment request: The proposed amendments would 
change Technical Specification requirements related to the service 
water system, component cooling water system, containment cooling and 
iodine removal systems, auxiliary electrical systems, and the control 
room emergency filtration system. The supplemental applications dated 
April 2, April 16, May 9, and June 3, 1997, would eliminate separate 
requirements for the component cooling water system for single-unit and 
two-unit operation, revise the acceptance criteria for laboratory 
testing of the control room emergency filtration system charcoal 
adsorber banks from 90 percent to 99 percent, and supplement additional 
information on the basis for acceptability of equipment qualification 
analyses and dose assessments resulting from a loss-of-coolant 
accident. The June 3, 1997, submittal requested the proposed amendments 
be handled on an exigent basis based on the current schedule which 
indicates that Unit 2 restart is scheduled for June 25, 1997, and Unit 
1 restart is scheduled for July 1, 1997, and failure of the issuance of 
the amendments by these dates would result in prevention of Point 
Beach's resumption of operation.
    Date of individual notice in the Federal Register: June 10, 1997 
(62 FR 31636).
    Expiration date of individual notice: July 10, 1997.
    Local Public Document Room location: Joseph P. Mann Library, 1516 
Sixteenth Street, Two Rivers, Wisconsin 54241.

Notice of Issuance of Amendments to Facility Operating Licenses

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

[[Page 35856]]

local public document rooms for the particular facilities involved.

Commonwealth Edison Company, Docket Nos. 50-237 and 50-249, Dresden 
Nuclear Power Station, Units 2 and 3, Grundy County, Illinois

    Date of application for amendments: June 20, 1996, as supplemented 
by letters dated December 30, 1996, and March 5, 1997.
    Brief description of amendments: The amendments would change the 
Technical Specifications (TS) by incorporating NRC-approved thermal 
limit licensing methodology in the list of approved methodologies used 
in establishing the fuel cycle-specific thermal limits. In addition, 
the proposed amendment will change the TS to reflect the use of Siemens 
Power Corporation (SPC) ATRIUM-9B fuel for all operating Modes at 
Dresden, Unit 3. The proposed amendment would also correct minor 
editorial items in the TS.
    Date of issuance: June 12, 1997.
    Effective date: Immediately, to be implemented within 30 days.
    Amendment Nos.: 160 and 155.
    Facility Operating License Nos. DPR-19 and DPR-25: The amendments 
revised the licenses and the Technical Specifications.
    Date of initial notice in Federal Register: April 9, 1997 (62 FR 
17227). The Commission's related evaluation of the amendments is 
cotained in a Safety Evaluation dated June 12, 1997.
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated June 12, 1997.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: Morris Area Public Library 
District, 604 Liberty Street, Morris, Illinois 60450.Consolidated 
Edison Company of New York, Docket No. 50-247, Indian PointNuclear 
Generating Unit No. 2, Westchester County, New York
    Date of application for amendment: March 31, 1997.Brief description 
of amendment: The amendment revises Technical Specifications (TSs) to 
remove the reference of Valve 863 from TS Table 3.6-1. This revision 
would allow for the installation of a proposed modification for 
automatic closure of Valve 863 upon receipt of a Phase A containment 
Isolation signal.
    Date of issuance: June 19, 1997.
Effective date: As of the date of issuance to be implemented within 30 
days.
    Amendment No.: 193.
    Facility Operating License No. DPR-26: Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: May 15, 1997 (62 FR 
26823)The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated June 19, 1997.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: White Plains Public Library, 
100 Martine Avenue, White Plains, New York 10610.

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

    Date of amendment request: March 27, 1997, as supplemented by 
letter dated May 6, 1997.
    Brief description of amendment: The amendment changes the Technical 
Specification 3/4.5.2, ``ECCS Subsystems--Modes 1, 2, and 3.'' The 
proposed changes add a surveillance requirement to verify the Emergency 
Core Cooling System (ECCS) piping is full of water at least once per 31 
days, and clarifies wording of surveillance requirement 4.5.2.j. The 
amendment also revises the TS Bases 3/4.5.2 and 3/4.5.3 to reflect 
surveillance requirement.

    Date of issuance: June 11, 1997.
Effective date: June 11, 1997, to be implemented within 60 days.
Amendment No.: 130.
    Facility Operating License No. NPF-38: Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: April 9, 1997 (62 FR 
17234). The Commission's related evaluation of the amendment is 
contained in a Safety Evaluation dated June 11, 1997.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: University of New Orleans 
Library, Louisiana Collection, Lakefront, New Orleans, LA 70122.

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

    Date of application for amendment: December 20, 1996, and 
supplemented February 13, and April 17, 1997.
    Brief description of amendment: This amendment modifies the 
Technical Specifications (TS) to delete a footnote associated with TS 
2.1.1, ``Reactor Core Safety Limits'' which requires reactor thermal 
power to be limited to 90% of 2700 Megawatts thermal for Cycle 14 
operation beyond 7000 Effective Full Power Hours.
    Date of Issuance: May 16, 1997.
    Effective Date: May 16, 1997.
    Amendment No.: 151.
    Facility Operating License No. DPR-67: Amendment revised the TS.
    Date of initial notice in Federal Register: January 15, 1997 (62 FR 
2190).
    The February 13, and April 17, 1997, letters provided clarifying 
information that did not change the scope of the December 20, 1996, 
application and the initial proposed no significant hazards 
consideration determination.
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated May 16, 1997.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: Indian River Junior College 
Library, 3209 Virginia Avenue, Fort Pierce, Florida 34954-9003.

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

    Date of application for amendments: July 28, 1995, as revised 
February 21, 1997.
    Brief description of amendments: The amendments revise the 
Technical Specifications for the Prairie Island Nuclear Generating 
Plant to allow credit for soluble boron in spent fuel criticality 
analyses. The request is based on the NRC approval of the Westinghouse 
Owners Group generic methodology for crediting soluble boron given in 
Topical Report WCAP-14416-NP-A, ``Westinghouse Spent Fuel Rack 
Criticality Analysis Methodology,'' Revision 1, November 1996.
    Date of issuance: June 12, 1997.
    Effective date: June 12, 1997, with full implementation within 30 
days.
    Amendment Nos.: 129 and 121.
    Facility Operating License Nos. DPR-42 and DPR-60. Amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: March 26, 1997 (62 FR 
14464).
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated June 12, 1997.
    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.

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: January 13, 1997, as 
supplemented March 24, 1997, May 13, 1997, and May 23, 1997.

[[Page 35857]]

    Brief description of amendment: The amendment revises Technical 
Specifications Requirements for containment leakage testing to add 
several containment isolation valves and to implement the requirements 
of 10 CFR Part 50, Appendix J, Option B for performance-based primary 
reactor containment leakage testing.
    Date of issuance: June 17, 1997.
    Effective date: As of the date of issuance to be implemented within 
30 days.
    Amendment No.: 174.
    Facility Operating License No. DPR-64: Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: March 19, 1997 (62 FR 
13173).
    The March 24, May 13, and May 23, 1997, supplemental letters 
provided clarifying information that did not change the initial 
proposed no significant hazards consideration.
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated June 17, 1997.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: White Plains Public Library, 
100 Martine Avenue, White Plains, New York 10610.

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

    Date of application for amendments: January 31, 1997.
    Brief description of amendments: The amendments revise Technical 
Specification 3/4.6.1.5, and its associated Bases section, to ensure 
that a representative average containment air temperature is measured.
    Date of issuance: June 13, 1996.
    Effective date: Both units, as of the date of issuance, to be 
implemented within 60 days.
    Amendment Nos. 195 and 178.
    Facility Operating License Nos. DPR-70 and DPR-75. The amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: March 12, 1997 (62 FR 
11497).
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated June 13, 1997.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: Salem Free Public Library, 112 
West Broadway, Salem, NJ 08079.

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

    Date of application for amendments: August 22, 1996, as 
supplemented March 28, 1997.
    Brief description of amendments: Revise Technical Specifications 
(TS) 3.6.5 and associated Bases to lower the minimum TS ice basket 
weight. Also extend the chemical analysis surveillance interval for the 
ice condenser ice bed from 12 months to 18 months.
    Date of issuance: June 10, 1997.
    Effective date: June 10, 1997.
    Amendment Nos.: 224, 215.
    Facility Operating License Nos. DPR-77 and DPR-79: Amendments 
revise the TS.
    Date of initial notice in Federal Register: April 23, 1997 (62 FR 
19835).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated June 10, 1997.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: Chattanooga-Hamilton County 
Library, 1101 Broad Street, Chattanooga, Tennessee 37402.

Wisconsin Public Service Corporation, Docket No. 50-305, Kewaunee 
Nuclear Power Plant, Kewaunee County, Wisconsin

    Date of application for amendment: April 22, 1997, as supplemented 
on May 15, and June 2, 1997. The April 22, 1997, submittal superseded a 
previous submittal on this subject dated September 6, 1996 (61 FR 
53769), as supplemented on October 30, October 31, November 7, November 
15, and November 27, 1996, and January 23 and January 29, 1997.
    Brief description of amendment: The amendment revises TS Section 
4.2.b, ``Steam Generator Tubes,'' and its associated Basis, by allowing 
a laser-welded repair of Westinghouse hybrid expansion joint (HEJ) 
sleeved steam generator tubes.
    Date of issuance: June 7, 1997.
    Effective date: June 7, 1997.
    Amendment No.: 135.
    Facility Operating License No. DPR-43: Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: May 7, 1997 (62 FR 
24988).
    The May 15, and June 2, 1997, submittals provided supplemental 
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 June 7, 1997.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: University of Wisconsin, 
Cofrin Library, 2420 Nicolet Drive, Green Bay, Wisconsin 54311-7001.

Wisconsin Public Service Corporation, Docket No. 50-305, Kewaunee 
Nuclear Power Plant, Kewaunee County, Wisconsin

    Date of application for amendment: April 24, 1997, as supplemented 
on May 15 and 28, and June 5, 1997.
    Brief description of amendment: The amendment revises TS Section 
4.2.b, ``Steam Generator Tubes,'' to allow repair of steam generator 
(SG) tubes with Combustion Engineering (CE) leak-tight sleeves in 
accordance with CE generic topical report CEN-629-P, Revision 2, 
``Repair of Westinghouse Series 44 and 51 Steam Generator Tubes Using 
Leak-Tight Sleeves.'' The TS are also revised to allow re-sleeving of 
tubes with existing sleeve joints in accordance with KNPP specific 
topical report CEN-632-P, ``Repair of Kewaunee Steam Generator Tubes 
Using a Re-Sleeving Technique.''
    Date of issuance: June 7, 1997.
    Effective date: June 7, 1997.
    Amendment No.: 134.
    Facility Operating License No. DPR-43: Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: May 7, 1997 (62 FR 
24989).
    The May 15 and 28, and June 5, 1997, submittals provided 
supplemental 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 June 7, 1997.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: University of Wisconsin, 
Cofrin Library, 2420 Nicolet Drive, Green Bay, Wisconsin 54311-7001.

Wisconsin Public Service Corporation, Docket No. 50-305, Kewaunee 
Nuclear Power Plant, Kewaunee County, Wisconsin

    Date of application for amendment: April 28, 1997, as supplemented 
on May 19, 1997.
    Brief description of amendment: The amendment establishes a new 
design basis flow rate for the auxiliary feedwater (AFW) pumps 
consistent with the assumptions used in the reanalysis of the limiting 
design basis event for the

[[Page 35858]]

AFW system. The Basis for TS 3.4.b, ``Auxiliary Feedwater System,'' has 
been revised to reflect the change in AFW flow and to clarify the 
requirements for the AFW cross-connect valves.
    Date of issuance: June 7, 1997.
    Effective date: June 7, 1997.
    Amendment No.: 133.
    Facility Operating License No. DPR-43: Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: May 7, 1997 (62 FR 
24977).
    The May 19, 1997, submittal 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 June 7, 1997.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: University of Wisconsin, 
Cofrin Library, 2420 Nicolet Drive, Green Bay, Wisconsin 54311-7001.

    Dated at Rockville, Maryland, this 25th day of June, 1997.

    For the Nuclear Regulatory Commission.
Jack W. Roe,
Director, Division of Reactor Projects III/IV, Office of Nuclear 
Reactor Regulation.
[FR Doc. 97-17140 Filed 7-1-97; 8:45 am]
BILLING CODE 7590-01-P