[Federal Register Volume 61, Number 14 (Monday, January 22, 1996)]
[Notices]
[Pages 1625-1647]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: 96-676]



      
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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 December 21, 1995, through January 4, 1996. 
The last biweekly notice was published on January 3, 1996 (61 FR 174).

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 Rules Review and 
Directives Branch, Division of Freedom of Information and Publications 
Services, Office of Administration, U.S. Nuclear Regulatory Commission, 
Washington, DC 20555, 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 February 21, 1996, the licensee may file a request for a hearing 
with respect to issuance of the amendment to the subject facility 
operating license and any person whose interest may be affected by this 
proceeding and who wishes to participate as a party in the proceeding 
must file a written request for a hearing and a petition for leave to 
intervene. Requests for a hearing and a petition for leave to intervene 
shall be filed in accordance with the Commission's ``Rules of Practice 
for Domestic Licensing Proceedings'' in 10 CFR Part 2. Interested 
persons should consult a current copy of 10 CFR 2.714 which is 
available at the Commission's Public Document Room, the Gelman 
Building, 2120 L Street, NW., Washington, DC and at the local public 
document room for the particular facility involved. If a request for a 
hearing or petition for leave to intervene is filed by the above date, 
the Commission or an Atomic Safety and Licensing Board, designated by 
the Commission or by the Chairman of the Atomic Safety and Licensing 
Board Panel, will rule on the request and/or petition; and the 
Secretary or the designated Atomic Safety and Licensing Board will 
issue a notice of a hearing or an appropriate order.
    As required by 10 CFR 2.714, a petition for leave to intervene 
shall set forth with particularity the interest of the petitioner in 
the proceeding, and how that interest may be affected by the results of 
the proceeding. The petition should specifically explain the reasons 
why intervention should be permitted with particular reference to the 
following factors: (1) The nature of the petitioner's right under the 
Act to be made a party to the proceeding; (2) the nature and extent of 
the petitioner's property, financial, or other interest in the 
proceeding; and (3) the possible effect of any order which may be 
entered in the proceeding on the petitioner's interest. The petition 
should also identify the specific aspect(s) of the subject matter of 
the proceeding as to which petitioner wishes to intervene. Any person 
who has filed a petition for leave to intervene or who has been 
admitted as a party may amend the petition without requesting leave of 
the Board up to 15 days prior to the first prehearing conference 
scheduled in the proceeding, but such an amended petition must satisfy 
the specificity requirements described above.
    Not later than 15 days prior to the first prehearing conference 
scheduled in the proceeding, a petitioner shall file a supplement to 
the petition to intervene which must include a list of the contentions 
which are sought to be litigated in the matter. Each contention must 
consist of a specific statement of the issue of law or fact to be 
raised or controverted. In addition, the petitioner shall provide a 
brief explanation of the bases of the contention and a concise 
statement of the alleged facts or expert opinion which support the 
contention and on which the petitioner intends to rely in proving the 
contention at the hearing. The petitioner must also provide references 
to those specific sources and documents of which the petitioner is 
aware and on which the petitioner intends to rely to establish those 
facts or expert opinion. Petitioner must provide sufficient information 
to show that a genuine dispute exists with the applicant on a material 
issue of law or fact. Contentions shall be limited to matters within 
the scope of the amendment under consideration. The contention must be 
one which, if proven, would entitle the petitioner to relief. A 
petitioner who fails to file such a supplement which satisfies these 
requirements with respect to at least one contention will not be 
permitted to participate as a party.
    Those permitted to intervene become parties to the proceeding, 
subject to any 

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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, 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. Where petitions are filed during the last 10 days of the notice 
period, it is requested that the petitioner promptly so inform the 
Commission by a toll-free telephone call to Western Union at 1-(800) 
248-5100 (in Missouri 1-(800) 342-6700). The Western Union operator 
should be given Datagram Identification Number N1023 and the following 
message addressed to (Project Director): petitioner's name and 
telephone number, date petition was mailed, plant name, and publication 
date and page number of this Federal Register notice. A copy of the 
petition should also be sent to the Office of the General Counsel, U.S. 
Nuclear Regulatory Commission, Washington, DC 20555, and to the 
attorney for the licensee.
    Nontimely filings of petitions for leave to intervene, amended 
petitions, supplemental petitions and/or requests for a hearing will 
not be entertained absent a determination by the Commission, the 
presiding officer or the Atomic Safety and Licensing Board that the 
petition and/or request should be granted based upon a balancing of 
factors specified in 10 CFR 2.714(a)(1)(i)-(v) and 2.714(d).
    For further details with respect to this action, see the 
application for amendment which is available for public inspection at 
the Commission's Public Document Room, the Gelman Building, 2120 L 
Street, NW., Washington, DC, and at the local public document room for 
the particular facility involved.
    Arizona Public Service Company, et al., Docket Nos. STN 50-528, STN 
50-529, and STN 50-530, Palo Verde Nuclear Generating Station, Units 
Nos. 1, 2, and 3, Maricopa County, Arizona.
    Date of amendments request: December 19, 1995
    Description of amendments request: The proposed amendments would 
allow the implementation of the recently approved Option B to 10 CFR 
Part 50, Appendix J. This new rule allows for a performance-based 
option for determining the test frequency for containment leakage rate 
testing. The proposed amendment would modify Technical Specifications 
(TS) 1.7, 3/4.6.1.1, 3/4.6.1.2, 3/4.6.1.3, and 3/4.6.3 and the Bases of 
TS 3/.6.1.2. It would also create a new TS 6.16.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. The proposed change does not involve a significant increase 
in the probability or consequences of an accident previously 
evaluated.
    The proposed Technical Specification (TS) changes will result in 
generally increased intervals between containment leakage rate tests 
determined through a performance based approach. The interval 
between such tests are not related in any way to conditions which 
cause accidents. Plant structures, systems, and components will not 
be operated in a different manner as a result of the proposed TS 
change, therefore, the proposed changes will not increase the 
probability of an accident previously evaluated.
    Containment leakage may result from accidents which are 
evaluated in the Updated Final Safety Analysis Report. The proposed 
TS changes may result in a small, but acceptable, increase in post-
accident containment leakage. This increase is calculated as a 
statistical expectation using the probability that leakage through a 
penetration will exceed the administrative limit and through the 
increased time needed to detect such excess leakage. NUREG-1493, 
which is the technical basis for 10 CFR Part 50, Appendix J, Option 
B, contains a detailed evaluation of the expected leakage and its 
consequences.
    The increased risk due to the lengthening of the intervals 
between Type A, B, and C leakage rate tests is also evaluated in 
NUREG-1493. Using a statistical approach, NUREG-1493 determined that 
the increase in expected dose to the public, resulting from 
extending the testing interval, is extremely small. NUREG-1493 
concluded that the small increase is justifiable due to the benefits 
which accrue from interval extension. The primary benefit is the 
reduction in occupational exposure. The reduction, on a per person 
basis, is orders of magnitude greater than the marginal, potential 
increase in dose to the public. The reduction in occupational 
exposure is a real reduction, while the small increase in dose to 
the public is statistically derived using conservative assumptions. 
Therefore, the proposed change does not significantly increase the 
consequences of an accident previously evaluated.
    2. The proposed change does not create the possibility of a new 
or different kind of accident from any accident previously 
evaluated.
    The proposed change does not create the possibility of a new or 
different kind of accident from any previously evaluated. The 
proposed change only incorporates the performance based approach 
authorized in the new Option B to Appendix J of 10 CFR Part 50. The 
interval extensions allowed, through this approach, do not have the 
potential for creating the possibility of new or different kinds of 
accidents from those previously evaluated. Plant structures, 
systems, and components will not be operated in a different manner 
as a result of the TS change and, therefore, will not introduce any 
new or different failure modes or initiators.
    3. The proposed change does not involve a significant reduction 
in a margin of safety.
    The proposed Technical Specification does not alter the 
allowable containment leakage rate. The proposed change replaces the 
current, prescriptive testing requirements with a new performance 
based approach for establishing the testing intervals therefore, the 
proposed change does not involve a significant reduction in the 
margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
that review, it appears that the three standards of 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendments request involve no significant hazards consideration.
    Local Public Document Room location: Phoenix Public Library, 1221 
N. Central Avenue, Phoenix, Arizona 85004.
    Attorney for licensee: Nancy C. Loftin, Esq., Corporate Secretary 
and Counsel, Arizona Public Service Company, P.O. Box 53999, Mail 
Station 9068, Phoenix, Arizona 85072-3999.
    NRC Project Director: William H. Bateman.

Baltimore Gas and Electric Company, Docket No. 50-317, Calvert Cliffs 
Nuclear Power Plant, Unit No. 1, Calvert County, Maryland.

    Date of amendment request: December 21, 1995.
    Description of amendment request: The proposed amendment would 
revise the Calvert Cliffs Nuclear Power Plant, 

[[Page 1628]]
Unit No. 1, Technical Specifications (TSs). The requested change would 
allow the use of cladding materials other than Zircaloy or ZIRLO. A 
Temporary Exemption was issued on November 28, 1995 (60 FR 62483) 
approving the loading of four (4) lead fuel assemblies (LFAs) into the 
Unit No. 1 reactor vessel during cycles 13, 14, and 15. The technical 
basis for the Exemption, which is the same basis for the requested TS 
amendment, was provided in the Baltimore Gas and Electric Company (BGE) 
submittal dated July 13, 1995. The submittal addressed the safety 
significance of operating with 4 LFAs in Calvert Cliffs Nuclear Power 
Plant, Unit No. 1, reactor vessel during cycles 13, 14, and 15.
    Specifically, BGE proposes to add a statement to TS 5.2.1, ``Fuel 
Assemblies,'' indicating, for Cycles 13, 14, and 15 only, advanced 
cladding material may be used in 4 lead test assemblies as described in 
a approved Temporary Exemption dated November 28, 1995.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Would not involve a significant increase in the probability 
or consequences of an accident previously evaluated.
    The proposed change is to add an approved temporary exemption to 
the Unit 1 Technical Specifications allowing the installation of 
four lead fuel assemblies. These four assemblies use an advanced 
cladding material which is not specifically permitted by existing 
regulations or Calvert Cliffs' Technical Specifications. A temporary 
exemption to allow the installation of these assemblies was approved 
on November 28, 1995. The addition of this approved temporary 
exemption to Technical Specification 5.2.1 is simply intended to 
allow their installation under the provisions of the temporary 
exemption. The license amendment is effective only as long as the 
exemption is effective. The addition of the approved temporary 
exemption to Unit 1 Technical Specification 5.2.1 does not change 
the probability or consequences of an accident previously evaluated.
    Therefore, this change does not involve a significant increase 
in the probability or consequences of an accident previously 
evaluated.
    2. Would not create the possibility of a new or different type 
of accident from any accident previously evaluated.
    The proposed Technical Specification change adds an approved 
temporary exemption to Technical Specification 5.2.1 for Unit 1. 
This change does not add any new equipment, modify any interfaces 
with existing equipment, change the equipment's function, or change 
the method of operating the equipment. The proposed change does not 
affect normal plant operations or configuration. Since the proposed 
change does not change the design, configuration, or operation, it 
could not become an accident initiator.
    Therefore, the proposed change does not create the possibility 
of a new or different type of accident from any accident previously 
evaluated.
    3. Would not involve a significant reduction in a margin of 
safety.
    The proposed change is to add an approved temporary exemption to 
the Unit 1 Technical Specifications allowing the installation of 
four lead fuel assemblies. These four assemblies use an advanced 
cladding material which is not specifically permitted by existing 
regulations or Calvert Cliffs' Technical Specifications. A temporary 
exemption to allow the installation of these assemblies was approved 
on November 28, 1995. The addition of this approved temporary 
exemption to Technical Specification 5.2.1 is simply intended to 
allow their installation under the provisions of the temporary 
exemption. The license amendment is effective only as long as the 
exemption is effective. This amendment does not change the margin of 
safety by adding a reference to an approved, temporary exemption to 
the Technical Specifications.
    Therefore, the proposed change does not involve a significant 
reduction in the margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 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: Calvert County Library, Prince 
Frederick, Maryland 20678.
    Attorney for licensee: Jay E. Silbert, Esquire, Shaw, Pittman, 
Potts and Trowbridge, 2300 N Street, NW., Washington, DC 20037.
    NRC Project Director: Ledyard B. Marsh.

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: December 7, 1995.
    Description of amendment request: The proposed amendments will 
remove the Technical Specification (TS) requirements for the main 
feedwater pump discharge pressure switch input to the Anticipatory 
Reactor Trip System (ARTS) and the Emergency Feedwater System (EFDW).
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    (1) Involve a significant increase in the probability or 
consequences of an accident previously evaluated:
    No. The accidents addressed within the Oconee Final Safety 
Analysis Report (FSAR) have been reviewed with respect to this 
proposed Technical Specification amendment request. The probability 
or consequences of any accident previously evaluated is not 
significantly increased by the proposed amendment. Emergency 
Feedwater is required for the mitigation of some accidents and the 
availability of this system will be unaffected by this proposed 
revision. Both manual and automatic actuation of the EFDW system on 
a loss of main feedwater will remain.
    (2) Create the possibility of a new or different kind of 
accident from any kind of accident previously evaluated:
    No. This amendment eliminates a portion of the automatic 
actuation circuitry for EFDW and ARTS. This circuitry removal does 
not create the possibility of a new or different kind of accident as 
the design of the circuitry is to sense a loss of main feedwater and 
supply a signal for the initiation of ARTS and EFDW. A loss of main 
feedwater signal will continue to be supplied to ARTS and EFDW; 
however, this loss will be sensed by low hydraulic oil pressure on 
the Main Feedwater Pumps (ARTS and EFDW) and low steam generator 
level (EFDW only) rather than by a low Main Feedwater Pump discharge 
pressure. Since a loss of Main Feedwater will continue to be 
recognized, the system will continue to function as before. Hence, 
no new or different accidents will be created.
    (3) Involve a significant reduction in a margin of safety.
    No. The margin of safety will not be significantly reduced as an 
actuation signal to ARTS and EFDW will continue to be generated by a 
loss of Main Feedwater. Consequently, ARTS and EFDW will continue to 
perform the safety function required for accident mitigation.

    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. 
    
[[Page 1629]]


Florida Power and Light Company, et al., Docket Nos. 50-335 and 50-389, 
St. Lucie Plant, Unit Nos. 1 and 2, St. Lucie County, Florida.

    Date of amendment request: November 22, 1995.
    Description of amendment request: The proposed amendments will 
upgrade existing TS [Technical Specification] 3/4.4.6.1 for the Reactor 
Coolant System Leakage Detection Instrumentation by adapting the 
Standard Technical Specifications for Combustion Engineering Plants 
(NUREG-1432), Specification 3.4.15, to both St. Lucie units. The 
proposal is consistent with the NRC Final Policy Statement on Technical 
Specifications Improvements (58 FR 39132).
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    (1) Operation of the facility in accordance with the proposed 
amendment would not involve a significant increase in the 
probability or consequences of an accident previously evaluated.
    The Reactor Coolant System (RCS) Leakage Detection 
Instrumentation Systems are not accident initiators, and their 
operational status is not a consideration in determining the 
probability of occurrence of accidents previously evaluated. The 
proposed revision to the related Limiting Condition for Operation 
(LCO) 3/4.4.6.1 does not involve a change to the configuration or 
method of operation of any equipment that is used to mitigate the 
consequences of an accident, nor do the changes alter any 
assumptions made involving initial plant conditions in the safety 
analyses. Therefore, operation of the facility in accordance with 
the proposed amendment would not involve a significant increase in 
the probability or consequences of an accident previously evaluated.
    (2) Operation of the facility in accordance with the proposed 
amendment would not create the possibility of a new or different 
kind of accident from any accident previously evaluated.
    The proposed revision to LCO 3/4.4.6.1 is administrative in 
nature and will not result in a change to the physical plant or the 
modes of plant operation defined in the Facility License. The 
revision does not involve the addition or modification of equipment 
nor does it alter the design of plant systems. Therefore, operation 
of the facility in accordance with the proposed amendment would not 
create the possibility of a new or different kind of accident from 
any accident previously evaluated.
    (3) Operation of the facility in accordance with the proposed 
amendment would not involve a significant reduction in a margin of 
safety.
    The RCS Leakage Detection Systems are designed to provide 
diverse methods to assist in the detection and location of 
unidentified leakage that may be associated with potential pressure 
boundary degradation. These systems provide no equipment control or 
accident mitigation functions, and are not associated with the 
safety margin established for protection from analyzed Loss of 
Coolant Accidents. The proposed revision to LCO 3/4.4.6.1 does not 
alter the basis for any technical specification that is related to 
the establishment of, or the maintenance of, a nuclear safety 
margin; and simply adapts the corresponding and previously reviewed 
specification from the Standard Technical Specifications for 
Combustion Engineering Plants, NUREG-1432, to the St. Lucie units. 
Therefore, operation of the facility in accordance with the proposed 
amendment would not involve a significant reduction in a margin of 
safety.
    Based on the above discussions and the supporting Evaluation of 
Technical Specification changes, FPL has determined that the 
proposed license amendment involves no significant hazards 
consideration.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: Indian River Junior College 
Library, 3209 Virginia Avenue, Fort Pierce, Florida 34954-9003.
    Attorney for licensee: Harold F. Reis, Esquire, Newman and 
Holtzinger, 1615 L Street, NW., Washington, DC 20036.
    NRC Project Director: David B. Matthews, Director.

GPU Nuclear Corporation, et al., Docket No. 50-219, Oyster Creek 
Nuclear Generating Station, Ocean County, New Jersey.

    Date of amendment request: December 5, 1995.
    Description of amendment request: The proposed amendment revises 
the submittal date in the Annual Exposure Data Report which brings 
Oyster Creek into conformance with 10 CFR 20.2206 and relaxes an overly 
restrictive administrative requirement.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    . . . The changes do not:
    1. Involve a significant increase in the probability or the 
consequence of an accident previously evaluated.
    This change is administrative in nature and has no effect on the 
operation of the plant. This change will not increase the 
probability or consequence of an accident previously evaluated.
    2. Create the possibility a new or different kind of accident 
from any previously evaluated.
    Operation of the facility in accordance with this proposed 
change will not create the possibility for an accident or 
malfunction of a different type from any accident previously 
evaluated. The proposed amendment does not modify any system 
(component) operation or maintenance activity. The facility will 
continue to be operated within the limits of existing accident 
analysis and margins of safety.
    3. Involve a significant reduction in a margin of safety.
    This change brings the submittal date for the Annual Exposure 
Data Report into conformance with 10 CFR 20.2206 and relaxes an 
overly restrictive administrative requirement. Since the proposed 
change does not alter any system hardware or design basis, the 
margin of safety is not reduced.

    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: Ocean County Library, 
Reference Department, 101 Washington Street, Toms River, NJ 08753.
    Attorney for licensee: Ernest L. Blake, Jr., Esquire. Shaw, 
Pittman, Potts & Trowbridge, 2300 N Street, NW., Washington, DC 20037.
    NRC Project Director: Phillip F. McKee.

IES Utilities Inc., Docket No. 50-331, Duane Arnold Energy Center, Linn 
County, Iowa.

    Date of amendment request: November 15, 1995.
    Description of amendment request: The proposed amendment would 
revise the requirements for the End of Cycle Recirculation Pump Trip 
logic to match more closely the assumptions applicable to the turbine 
trip events for which it was installed. The surveillance requirements 
are also proposed to be revised, based on those same assumptions.
    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 Technical Specification (TS) amendment will not 
significantly increase the probability or consequences of any 
previously evaluated accidents. The [End of Cycle] (EOC) 
[recirculation pump trip] RPT system was installed to preclude 

[[Page 1630]]
violation of reactor fuel limits, and the system will be preserved for 
that purpose. In the event that system is not available, an 
operating penalty will be imposed on the [Minimum Critical Power 
Ratio] MCPR limit to assure sufficient margin to the limit to 
preclude fuel damage during the postulated turbine trip events.
    The change to the ``Minimum Operable Channels per Trip System'' 
will assure that inputs monitoring both the turbine control valve 
fast closure and the turbine stop valve closure will be available to 
initiate (EOC)RPT.
    The change to the ``Applicable Operating Mode'' is an editorial 
change which reflects the existing hardware bypass.
    The change to Action 81 in TS Table 3.2-G will assure that when 
the (EOC)RPT system does not meet the minimum TS availability 
requirements, the [safety limit minimum critical power ratio] SLMCPR 
will not be challenged. By imposing an [operating limit minimum core 
power ratio] OLMCPR penalty for continued operation, the fuel 
thermal limits will not be challenged, since the (EOC)RPT system was 
installed to accomplish the same goal. No increase in the 
consequences of the turbine trip events will result from this 
change. The OLMCPR penalty is dependent on cycle-specific parameters 
and will therefore be included in the cycle-specific [Core Operating 
Limits Report] COLR.
    The change to the surveillance interval results in (EOC)RPT 
logic channel functional tests being performed once per quarter 
instead of once per month. The change also revises the allowed out-
of-service time (AOT) for testing from two hours to six hours. These 
changes are consistent with the Improved Standard Technical 
Specifications, NUREG-1433, Revision 1. The (EOC)RPT is initiated by 
instruments common to the Reactor Protection System (RPS) (i.e., 
turbine stop valve closure and turbine control valve fast closure). 
The surveillance interval and AOT changes for these instruments were 
evaluated in ``Technical Specification Improvement Analysis for BWR 
Reactor Protection System,'' NEDC-30851P-A, March 1988, for the RPS 
function. Although the (EOC)RPT functions were not explicitly 
identified in that document, these changes can be considered bounded 
by that analysis. The basis for this conclusion is similar to the 
basis established for the control rod block instrumentation common 
to the RPS, as documented in ``Technical Specification Improvement 
analysis for BWR Control Rod Block Instrumentation,'' NEDC-30851P-A, 
Supplement 1, October 1988. Failure of the (EOC)RPT function could 
potentially lead to exceeding the SLMCPR, similar to the 
consequences of an unmitigated rod withdrawal error. The slight 
increase in risk of a SLMCPR violation due to extending (EOC)RPT 
surveillance interval and AOT is offset by the same benefits 
associated with the similar approved surveillance interval and AOT 
for the RPS. Both the above referenced reports have been approved 
for application at the DAEC via TS Amendment 193, dated April 14, 
1993.
    The changes to the ``Operating Modes for which Surveillance 
Required'' are clarifications and will result in a more efficient 
utilization of resources. By stating that the surveillance applies 
only when the (EOC)RPT system is OPERABLE, the surveillances will 
not be performed needlessly. During the early part of an OPERATING 
cycle, the (EOC)RPT is not required to mitigate a turbine trip, and 
therefore, may be bypassed. At the time when the (EOC)RPT is assumed 
to be OPERABLE pursuant to the analysis, it will be made OPERABLE 
unless accepting the penalty on the OLMCPR is preferable. The result 
of the proposed change will still be that the (EOC)RPT is 
demonstrated OPERABLE at any time when it is required.
    The change to the acceptance criteria for response time testing 
reflects a recent review of the analytical assumptions and the 
testing methodology. The (EOC)RPT is assumed to interrupt power to 
the recirculation pump motor within 175 milliseconds after 
initiation of either turbine stop valve closure or turbine control 
valve fast closure. The response time test only measures a portion 
of the complete trip (the rest was measured as part of start-up 
testing). The portion measured is dependent on which trip input is 
being tested. The turbine control valve closure is sensed by a 
pressure switch monitoring the hydraulic fluid controlling the valve 
and therefore has no delay between valve motion and initiation of 
the (EOC)RPT logic. The turbine stop valve closure is sensed by 
position switch. Since this switch is set to initiate (EOC)RPT at 
10% valve closed, there is a brief delay between the beginning of 
valve motion and initiation of the (EOC)RPT logic. The respective 
proposed response time tests account for these differences, as 
described in the footnotes on TS page 3.2-36, and demonstrate that 
the measured portions of the action are within allowed time periods.
    None of the proposed changes will significantly increase the 
probability of any accident previously evaluated because the 
(EOC)RPT is not an initiator of any of those events. None of the 
proposed changes will significantly increase the consequences of an 
accident because the (EOC)RPT system serves to prevent a turbine 
trip event from exceeding the fuel SLMCPR, and it will continue to 
perform in that capacity at any time when it is required to assure 
margin to the SLMCPR.
    2. The proposed changes will not add a new or different kind of 
accident because the plant will not be operated in a different way. 
By allowing the implementation of a penalty on OLMCPR in lieu of 
reducing reactor power, the risk of a plant transient is reduced. 
Similarly, the surveillance interval and AOT extensions will also 
result in fewer plant power reductions for testing.
    The (EOC)RPT initiates a trip of the recirculation pumps and any 
TS change affecting that system cannot result in an effect on any 
system other than those pumps. Consequently, no new accidents are 
postulated as a result of this proposed change.
    3. The proposed change will not result in a significant 
reduction in any margin of safety. The (EOC)RPT performs to assure 
adequate margin to the SLMCPR. The proposed change will preserve 
that function and require that additional margin to the SLMCPR be 
imposed for those times when the (EOC)RPT is not OPERABLE. The other 
changes are proposed because they assure correct (EOC)RPT function 
(inputs and response times).

    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: Cedar Rapids Public Library, 
500 First Street, S.E., Cedar Rapids, Iowa 52401.
    Attorney for licensee: Jack Newman, Kathleen H. Shea, Morgan, 
Lewis, & Bockius, 1800 M Street, NW., Washington, DC 20036-5869.
    NRC Project Director: Gail H. Marcus.

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

    Date of amendment request: December 14, 1995.
    Description of amendment request: The proposed amendment would 
modify Technical Specification 3.4.2, ``Flow Control Valves (FCVs),'' 
by deleting the requirement to verify that the average rate of movement 
of each reactor recirculation system FCV is limited to less than or 
equal to 11% per second in the opening and closing directions 
(Surveillance Requirement 3.4.2.2).
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration which is presented below:

    (1) The Clinton Power Station (CPS) Updated Safety Analysis 
Report (USAR) evaluates three specific events related to operation 
of the reactor recirculation flow control valves (FCVs). The impact 
of the proposed change on each of these events is discussed below.
    The loss of coolant accident (LOCA) analysis described in USAR 
Section 6.3.3.7.2 assumes that the FCVs fail ``as is'' in the event 
of a LOCA. This feature is assured by electronic interlocks in the 
FCV control circuitry and periodically verified as required by 
Technical Specification (TS) Surveillance Requirement (SR) 3.4.2.1. 
The design of these interlocks and the testing requirements are not 
affected by this proposed change.
    The Recirculation Flow Controller Failure--Decreasing Flow 
transient analyses are described in USAR Section 15.3.2, and the 
Recirculation Flow Controller Failure--Increasing Flow transient 
analyses are described in USAR Section 15.4.5. Since the 

[[Page 1631]]
control circuitry for the FCVs has been modified such that the 
capability to operate in a master controller mode has been 
eliminated, each FCV is now individually controlled, and the 
possibility that a single failure could affect operation of more 
than one FCV has also been eliminated. As a result, fact closure and 
fast opening of both FCVs are no longer postulated for CPS. Thus, 
the surveillance (SR 3.4.2.2) associated with verifying that FCV 
movement is within the assumptions of the analyses for fast closure 
and fast opening of both FCVs can be deleted.
    With respect to fast closure and fast opening of individual 
FCVs, the modification performed during the fifth refueling outage 
only affected the electronic master control of the FCVs and did not 
affect the hydraulic limitations of the FCVs. Conservative analyses, 
component testing, and the Initial Startup Test program provide 
confidence that individual FCV stroke rates assumed in the transient 
analyses will not be exceeded over the life of the plant. These 
analyses and conditions are sufficient to assure individual FCV 
stroke rates are adequately limited without the periodic performance 
of a specific test.
    In addition to the above, the modification did not add any new 
failure modes to the design of the individual FCV controllers. In 
fact, failure modes associated with misoperation of the common 
master controller have been eliminated from the control circuit 
design. The modification did not alter any of the features 
associated with initiators of any LOCA or features which assure that 
the FCVs fail ``as is'' in the event of a LOCA.
    Based on the above, Illinois Power (IP) has concluded that this 
request does not increase the probability or the consequences of any 
accident (or transient) previously evaluated.
    (2) USAR Sections 15.3.2 and 15.4.5 describe the plant response 
to malfunctions of FCV control failures, and USAR Section 6.3.3.7.2 
describes the assumptions made with respect to FCV failures and 
their impact on the LOCA analysis. The proposed change (and the 
associated modification prompting the proposed change) does not 
affect any other structures, systems, or components beyond the FCVs. 
All associated failure modes thus remain within the scope of the 
failure modes previously considered. As a result, IP has concluded 
that the proposed change cannot create the possibility of an 
accident not previously evaluated.
    (3) This request does not involve any change to the requirements 
or design associated with initiation or mitigation of a LOCA. The 
consequences of transients associated with fast closure and fast 
opening of reactor recirculation system FCVs are bounded by the 
consequences of other transient events and thus are not utilized in 
establishing plant operating limits. Although the control circuitry 
for the FCVs was modified during the fifth refueling outage, that 
modification did not affect the hydraulic failure modes of the FCVs. 
Further, the modification did not add any new failure modes to the 
design of the individual FCV controllers. In fact, failure modes 
associated with misoperation of the common master controller have 
been eliminated from the control circuit design. As a result, 
assumed FCV operation during analyzed accidents and transients has 
not been altered. Conservative analysis, component testing, and the 
Initial Startup Testing program have confirmed that the FCV velocity 
assumed in the transient analyses will not be exceeded over the life 
of the plant. Thus, verification of rate of FCV movement in the 
opening and closing directions need not be performed by periodic 
testing and SR 3.4.2.2 can be deleted without resulting in a 
reduction in a margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: Vespasian Warner Public 
Library, 120 West Johnson Street, Clinton, Illinois 61727.
    Attorney for licensee: Sheldon Zabel, Esq., Schiff, Hardin and 
Waite, 7200 Sears Tower, 233 Wacker Drive, Chicago, Illinois 60606.
    NRC Project Director: Gail H. Marcus.

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

    Date of amendment request: December 14, 1995.
    Description of amendment request: The proposed amendment would 
consist of several changes to the instrumentation sections of the 
Clinton Power Station Technical Specifications. The proposed changes 
are required due to engineering reanalyses or plant modifications. The 
affected instrumentation includes: (1) steam line flow high channels 
for the Reactor Core Isolation Cooling (RCIC) System, (2) ambient 
temperature channels in the Residual Heat Removal (RHR) System heat 
exchanger rooms, (3) reactor vessel pressure channels that provide a 
permissive for operation of the shutdown cooling mode of the RHR 
system, and (4) RCIC storage tank water level instrument channels.
    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) None of the proposed changes involve a significant increase 
in the probability or consequences of any accident previously 
evaluated.
    The changes to Table 3.3.6.1-1 Functions 3.a and 3.i are 
administrative in nature and bring the technical specifications (TS) 
into conformance with the Clinton Power Station (CPS) as-built 
design. The reactor core isolation cooling (RCIC) system steam line 
flow trip Function names have been changed to reflect the 
elimination of the residual heat removal (RHR) steam condensing 
mode. However, these trips have not been physically altered and thus 
will continue to operate as before. As a result of the elimination 
of the RHR steam condensing mode, the possibility of a leak in the 
RCIC steam supply resulting in an increase in the RHR heat exchanger 
room ambient temperature has also been eliminated. Accordingly, the 
RHR ambient temperature isolation trip is changed to only isolate 
the RHR system when the RHR heat exchanger room ambient temperature 
setpoint is exceeded. The Shutdown Cooling System Reactor Vessel 
Pressure--High function is provided to isolate the shutdown cooling 
portion of the RHR system since this piping is designed for 
pressures lower than rated reactor vessel pressure. This interlock 
(RHR cut in permissive) is provided only for equipment protection to 
prevent an intersystem LOCA scenario and credit for the interlock is 
not assumed in the accident or transient analysis in the Updated 
Safety Analysis Report (USAR).
    The proposed change to the setpoint (Allowable Value) is 
conservative with respect to considerations for shutting the RHR 
shutdown cooling motor-operated valves and providing 
overpressurization protection for the low pressure RHR shutdown 
cooling system piping. With respect to the RCIC storage tank water 
level setpoints, no accident or transient analysis takes credit for 
the volume of water in the RCIC storage tank. In addition, the 
setpoint (Allowable Value) has been changed to ensure RCIC system 
operation is not adversely affected by a low level in the storage 
tank.
    The proposed changes do not affect any of the parameters or 
conditions that contribute to initiation of any accidents previously 
evaluated. In addition, the proposed changes do not affect the 
ability of the associated instrumentation to operate as assumed in 
the safety analyses. As a result, the proposed changes will not 
result in a significant increase in the consequences of any accident 
previously evaluated.
    (2) None of the proposed changes create the possibility of a new 
or different kind of accident from any accident previously 
evaluated. The proposed changes for RHR/RCIC Steam Line Flow--High 
[are] administrative in nature and will simply make this item 
description accurate. The RCIC steam supply line no longer supplies 
any steam to the RHR heat exchanger room. As a result, the 
associated isolation of the RCIC system is no longer required. The 
Shutdown Cooling System Reactor Vessel Pressure - High function will 
still perform as designed. The RCIC Storage Tank Level - Low trip 
will continue to perform in accordance with design. None of the 
above listed changes will introduce any new failure modes or changes 
in plant operation.
    As a result, the proposed changes cannot create the possibility 
of a new or different kind of accident from any accident previously 
evaluated.
    (3) None of the proposed changes involve a significant reduction 
in a margin to safety. 

[[Page 1632]]
The proposed changes for RHR/RCIC Steam Line Flow--High do not involve 
a significant reduction in a margin of safety because the change is 
administrative in nature and will simply make the descriptions 
accurate and consistent with completed modifications. The 
elimination of RCIC system isolation in response to a high RHR room 
ambient temperature is no longer required due to the elimination of 
the RHR steam condensing mode. Removing the RHR room ambient 
temperature isolation of the RCIC will reduce the number of 
unnecessary isolations of RCIC. The Shutdown Cooling System Reactor 
Vessel Pressure - High function will still perform as designed. The 
proposed change to the setpoint (Allowable Value) is conservative 
with respect to considerations for shutting the RHR shutdown cooling 
motor-operated valves and providing overpressurization protection 
for the low pressure RHR shutdown cooling system piping. The 
Allowable Value for the RCIC Storage Tank Level - Low Function has 
been changed to be more conservative to ensure the RCIC and HPCS 
systems will perform their system safety function. No credit is 
taken for the volume in the RCIC storage tank for the HPCS or RCIC 
systems in performing their safety-related functions.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: Vespasian Warner Public 
Library, 120 West Johnson Street, Clinton, Illinois 61727
    Attorney for licensee: Sheldon Zabel, Esq., Schiff, Hardin and 
Waite, 7200 Sears Tower, 233 Wacker Drive, Chicago, Illinois 60606
    NRC Project Director: Gail H. Marcus.

Indiana Michigan Power Company, Docket Nos. 50-315 and 50-316, Donald 
C. Cook Nuclear Plant, Unit Nos. 1 and 2, Berrien County, Michigan.

    Date of amendment requests: December 19, 1995 [AEP:NRC:1215B]
    Description of amendment requests: The proposed amendments would 
modify the technical specifications to replace the existing scheduling 
requirements for overall integrated and local containment leakage rate 
testing with a requirement to perform the testing in accordance with 10 
CFR Part 50, Appendix J, Option B. Option B allows test scheduling to 
be adjusted based on past performance.
    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:

Criterion 1

    This amendment request does not involve a significant increase 
in the probability or consequences of an accident previously 
evaluated because the proposed changes to the T/Ss do not affect the 
assumptions, parameters, or results of any UFSAR [updated final 
safety analysis report] accident analysis. The proposed changes do 
not change the acceptance criteria for containment leakage limits 
and do not modify the response of the containment during a design 
basis accident. The proposed amendment does not add or modify any 
existing equipment. The proposed Types A, B, and C testing schedules 
will be consistent with Appendix J Option B to 10 CFR 50 which was 
developed based on analytical efforts documented in NUREG-1493 
[Performance-Based Containment Leak-Test Program]. The analysis 
confirms previous observations of insensitivity of population risks 
from severe reactor accidents to containment leakage rates. Based on 
these considerations, it is concluded that the changes do not 
involve a significant increase in the probability or consequences of 
an accident previously evaluated.

Criterion 2

    The proposed changes do not involve physical changes to the 
plant or changes in plant operating configuration. The proposed 
changes only remove the restrictive schedular requirements for 
conducting Types A, B, and C testing from the T/Ss and substitute 
the schedule specified in Appendix J Option B to 10 CFR 50 and 
Regulatory Guide 1.163 [Performance-Based Containment Leak-Test 
Program]. Thus, it is concluded that the proposed changes do not 
create the possibility of a new or different kind of accident from 
any accident previously evaluated.

Criterion 3

    Based on NUREG-1493, Regulatory Guide 1.163, and the rule 
posting in the Federal Register (60 FR 49495), the margin for safety 
presently provided is not significantly reduced by the proposed 
change to a performance-based test interval for Types A, B, and C 
tests. Although the changes allow more flexibility in scheduling 
tests, the proposed amendment continues to ensure reactor 
containment system reliability by periodic testing in full 
compliance with 10 CFR 50, Appendix J Option B. Based on these 
considerations, it is concluded that the 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 
amendment requests involve no significant hazards consideration.
    Local Public Document Room location: Maud Preston Palenske Memorial 
Library, 500 Market Street, St. Joseph, Michigan 49085.
    Attorney for licensee: Gerald Charnoff, Esq., Shaw, Pittman, Potts 
and Trowbridge, 2300 N Street, NW, Washington, DC 20037.
    NRC Project Director: John N. Hannon.

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

    Date of amendment request: August 15, 1995, as supplemented 
November 14, 1995.
    Description of amendment request: The proposed amendment would 
modify the Monticello Technical Specifications (TS) to: (1) revise the 
main steam line isolation valve leak rate test acceptance criterion to 
be based upon the combined maximum flow path leakage for all four main 
steam lines of 46 standard cubic feet per hour (scfh) in lieu of the 
current limit of 11.5 scfh per valve; (2) revise the operability test 
interval for the drywell spray header and nozzles from 5 years to 10 
years; and (3) revise TS 3/4.7.a.2, Primary Containment Integrity, to 
remove information specific to the primary containment leakage rate 
testing program and replace it with a commitment to abide by the 
requirements of 10 CFR Part 50, Appendix J, Option B, Section III.A, 
for Type A testing.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:
    a. The proposed amendment will not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    The proposed amendment is limited to changes to the surveillance 
testing requirements applicable to the main steam line isolation 
valves [MSIVs] allowable leakage criteria, drywell nozzles test 
interval, and method of applying Appendix J test requirements. With 
respect to monitoring main steam [line] isolation valve performance, 
the proposed criteria are equivalent to the current criteria 
ensuring that leakage past the valves would be within acceptable 
limits under accident conditions. These surveillance tests are 
performed while the plant is in a cold shutdown condition at a time 
when the equipment is not required to be operable. Performance of 
the tests themselves are not input or consideration in any accident 
previously evaluated, thus the proposed change will not increase the 
probability of any such accident occurring.
    The proposed amendment will not adversely affect the function, 
operation, or reliability of the equipment, nor will it diminish the 
capability of the equipment to perform as required during an 
accident. 

[[Page 1633]]
Combining the maximum per valve leak rate into an overall maximum 
leakage limit does not increase the overall permissible leakage and 
thus has no significant impact on the consequences of previously 
analyzed accidents since the combined leak rate of the main steam 
line isolation valves, and thus the contribution of the valves to 
overall primary containment leakage as used for analysis purposes, 
is unchanged. Extending the drywell nozzle test interval has been 
shown by industry experience to not compromise safety, and removing 
the specifics of primary containment leakage testing from the 
Technical Specifications and referencing 10 CFR Part 50 Appendix J 
does not alter either how actual testing is accomplished nor the 
acceptance criteria. It has been shown that adopting longer test 
intervals based on performance, maintains the safety objective for 
containment integrity while at the same time reducing the burden on 
licensees, and provides a greater level of worker safety than that 
provided by the previous rule.
    Therefore, there will be no increase in post accident off-site 
or on-site radiation dose as a result of this amendment. The 
proposed amendment requires compliance with the regulatory 
requirements of 10 CFR Part 50, Appendix J Option B, Section III.A, 
for Type A testing that has previously been reviewed by the NRC and 
found to be acceptable. Therefore, the amendment will not increase 
the consequences of any accident previously evaluated.
    b. The proposed amendment will not create the possibility of a 
new or different kind of accident from any accident previously 
analyzed.
    The proposed amendment does not involve any modification to 
plant equipment or operating procedures, nor will it introduce any 
new equipment failure modes that have not been previously 
considered. The proposed amendment is limited to changes in 
surveillance test frequencies of tests performed while the plant is 
in cold shutdown when the associated equipment is not required to be 
operable. We therefore conclude the proposed changes will not create 
the possibility of a new or different kind of accident from any 
accident previously analyzed.
    c. The proposed amendment will not involve a significant 
reduction in the margin of safety.
    Combining the allowable leak rate for the MSIV's from a per 
valve limit to an overall limit does not change the total allowable 
leakage and therefore post accident dose levels remain unchanged. 
Extending the drywell nozzle surveillance test interval from 5 to 10 
years has been shown by industry experience to be acceptable. 
Extending the intervals between containment integrated leakage tests 
as authorized by 10 CFR Part 50, Appendix J, Option B, does not 
change the acceptance criteria nor how testing is accomplished.
    Based on these considerations, we conclude the proposed 
amendment will not involve a significant reduction in the margin of 
safety.

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

Pacific Gas and Electric Company, Docket Nos. 50-275 and 50-323, Diablo 
Canyon Nuclear Power Plant, Unit Nos. 1 and 2, San Luis Obispo County, 
California.

    Date of amendment requests: December 19, 1995.
    Description of amendment requests: The proposed amendments would 
revise the combined Technical Specifications (TS) for the Diablo Canyon 
Power Plant Unit Nos. 1 and 2 to relocate Technical Specification (TS) 
6.5, ``Review and Audit,'' 6.8, ``Procedures and Programs,'' Sections 
6.8.1c., 6.8.1d., 6.8.2, and 6.8.3, in accordance with guidance in an 
NRC letter dated October 25, 1993, from William T. Russell to the 
chairpersons of industry owners groups and the Commission's Final 
Policy Statement on TS Improvements for Nuclear Power Reactors on 
relocation of TS that do not satisfy the retention criteria. As part of 
the relocation of TS 6.8.2, TS 6.1.1 would be revised to require that 
proposed tests, experiments, or modifications that affect nuclear 
safety be approved by the plant manager or his designee prior to 
implementation.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. The proposed change does not involve a significant increase 
in the probability or consequences of an accident previously 
evaluated.
    The proposed changes simplify the Technical Specifications (TS), 
meet regulatory requirements for relocated TS, and implement the 
recommendations of: (1) the NRC's letter dated October 25, 1993, 
from William T. Russell to the chairpersons of the industry owners 
groups; (2) the Commissions's Final Policy Statement on TS 
Improvements; and (3) the recently revised 10 CFR 50.36. Future 
changes to these requirements will be controlled by 10 CFR 50.54 and 
10 CFR 50.59. Any changes that reduce the effectiveness of the 
Quality Assurance Program will be approved by the NRC prior to 
implementation. The proposed changes are administrative in nature 
and do not involve any modifications to any plant equipment or 
affect plant operation.
    Therefore, the proposed changes do not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    2. The proposed change does not create the possibility of a new 
or different kind of accident from any accident previously 
evaluated.
    The proposed changes are administrative in nature, do not 
involve any physical alterations to any plant equipment, and cause 
no change in the method by which any safety-related system performs 
its function.
    Therefore, the proposed changes do not create the possibility of 
a new or different kind of accident from any accident previously 
evaluated.
    3. The proposed change does not involve a significant reduction 
in a margin of safety.
    The proposed changes do not alter the basic regulatory 
requirements and do not affect any safety analyses. 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 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: California Polytechnic State 
University, Robert E. Kennedy Library, Government Documents and Maps 
Department, San Luis Obispo, California 93407.
    Attorney for licensee: Christopher J. Warner, Esq., Pacific Gas and 
Electric Company, P.O. Box 7442, San Francisco, California 94120.
    NRC Project Director: William H. Bateman.

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

    Date of amendment request: September 15, 1995.
    Description of amendment request: The licensee proposes to extend 
the surveillance test intervals for the auxiliary electrical systems to 
support 24-month operating cycles.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:
    Operation of the James A. Fitzpatrick plant in accordance with 
the proposed 

[[Page 1634]]
Amendment would not involve a significant hazards consideration as 
defined in 10 CFR 50.92, since it would not:
    1. Involve a significant increase in the probability or 
consequences of an accident previously evaluated.
    The proposed changes increase the interval between auxiliary 
electrical system functional tests and also propose additional 
requirements for battery performance testing. These changes are 
consistent with the guidance provided in Generic Letter 91-04. These 
changes do not involve any special changes to the plant, nor do they 
alter the way the auxiliary electrical system functions. Past 
equipment performance indicates that the test acceptance criteria 
has been consistently met, providing additional assurance that the 
longer surveillance interval will not degrade system performance. 
The proposed changes revise Bases section 4.9 to clarify battery 
testing requirements and indicate consistence with the length of the 
24 month operating cycle. Therefore, the proposed changes do not 
involve a significant increase in the probability or consequences of 
an accident previously evaluated.
    2. Create the possibility of a new or different kind of accident 
from any accident previously evaluated.
    The proposed changes increase the interval between auxiliary 
electrical system functional tests and also propose additional 
requirements for battery performance testing. These changes are 
consistent with the guidance provided in Generic Letter 91-04. The 
proposed changes do not change the ability of the auxiliary 
electrical systems to provide electrical power during a design basis 
accident. Past equipment performance indicates that the test 
acceptance criteria has been consistently met, providing additional 
assurance performance. The proposed changes do not modify the design 
or operation of plant equipment, therefore, no new or different 
failure modes are introduced. The proposed changes revise Basis 
section 4.9 to clarify battery testing requirements and indicate 
consistency with the length of the 24 month operating cycle. 
Therefore, the proposed changes do 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 increase the interval between auxiliary 
electrical system functional tests and also propose additional 
requirements for battery performance testing. These changes are 
consistent with the guidance provided in Generic Letter 91-09. The 
proposed changes do not alter the configuration of the auxiliary 
electrical system nor change the manner in which the system 
functions. Operation of the facility remains unchanged by the 
proposed changes. An evaluation of past equipment performance 
indicates that auxiliary electrical system operability is not time 
dependent. The proposed changes revise Bases section 4.9 clarify 
battery testing requirements and indicate consistency with the 
length of the 24 month operating cycle. Therefore, a longer 
surveillance test interval for the station batteries and LPCI [low-
pressure coolant injection] batteries will not degrade performance 
of the auxiliary electrical system and will not involve a 
significant reduction in a margin of safety.

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

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

    Date of amendment request: October 25, 1995.
    Description of amendment request: The licensee proposes to extend 
the surveillance test intervals for the containment systems to support 
24-month operating cycles.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 40.19(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    Operation of the FitzPatrick plant in accordance with the 
proposed Amendment would not involve a significant hazards 
consideration as defined in 10 CFR 50.92, since it would not:
    1. Involve a significant increase in the probability of 
consequences of an accident previously evaluated.
    The proposed changes do not involve any physical changes to the 
plant, do not alter the way the containment systems function, and 
will not degrade the performance of the containment systems. The 
type of testing and the corrective actions required if the subject 
surveillance fail remains the same. The proposed changes do not 
adversely affect the availability of the containment systems or 
affect the ability of the systems to meet their design objectives. A 
historical review of surveillance test results indicated that there 
was no evidence of any failures which would invalidate the above 
conclusions.
    2. Create the possibility of a new or different kind of accident 
from any accident previously evaluated.
    The proposed changes do not modify the design or operation of 
the plant and therefore no new failure modes are introduced. No 
changes are proposed to the type and method of testing performed, 
only to the length of the surveillance interval. Past equipment 
performance and on-line testing indicate that longer test intervals 
will not degrade the containment systems. A historical review of 
surveillance test results indicated that there was no evidence of 
any failure which would invalidate the above conclusions.
    3. Involve a significant reduction in a margin of safety.
    Although the proposed changes will result in an increase in the 
interval between surveillance tests, the impact on system 
reliability is minimal. This is based on more frequent on-line 
testing and the redundant design of the containment systems. A 
review of past surveillance history has shown no evidence of failure 
which would significantly impact the reliability of the containment 
systems. Operation of the plant remains unchanged by the proposed 
containment system surveillance test interval extensions. The 
assumptions in the Plant Licensing Basis are not impacted. Therefore 
the proposed changes do not result in a significant reduction in the 
margin of safety.

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

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

    Date of amendment request: November 30, 1995.
    Description of amendment request: The licensee proposes to extend 
the surveillance test intervals for the standby liquid control (SLC) 
system to support 24 month operating cycles.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.19(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    Operation of the FitzPatrick plant in accordance with the 
proposed Amendment would not involve a significant hazards 
consideration as defined in 10 CFR 50.92 since it would not:
    1. Involve a significant increase in the probability or 
consequences of an accident previously evaluated.

[[Page 1635]]

    The proposed changes do not involve any physical changes to the 
plant, do not alter any SLC system functions, and will not degrade 
the performance of the SLC system. The type of testing and the 
corrective actions required if the subject SLC surveillances fail 
remain the same. The proposed changes do not adversely affect the 
availability of the SLC system or the ability of the system to bring 
the reactor from full power to a cold shutdown condition in the 
unlikely event that control rods cannot be inserted. A historical 
review of SLC surveillance test results indicated that there was no 
evidence of any failures that would invalidate the above 
conclusions.
    2. Create the possibility of a new or different kind of accident 
from any accident previously evaluated.
    The proposed changes do not introduce any failure mechanisms of 
a different type than those previously evaluated since there are no 
physical changes being made to the facility. No changes are proposed 
to the type and method of testing performed, only to the length of 
the surveillance interval. Past equipment performance and on-line 
testing indicate the longer test intervals will not degrade SLC 
equipment. A historical review of surveillance test results 
indicated that there was no evidence of any failures that would 
invalidate the above conclusions.
    3. Involve a significant reduction in a margin of safety.
    Although the proposed changes will result in an increase in the 
interval between surveillance tests, the impact on system 
reliability is minimal. This is based on more frequent on-line 
testing of major system components and the redundant design of the 
SLC system. A review of past SLC surveillance history has shown no 
evidence of failures that would significantly impact the reliability 
of the SLC system. The longer testing intervals do not significantly 
impact the SLC safety margins for SLC normal operation, operation 
with inoperable components, or sodium pentaborate solution as 
described in the bases of the Technical Specifications. Operation of 
the plant remains unchanged by the proposed SLC surveillance 
interval extensions. The assumptions in the Plant Licensing Basis 
are not impacted. Therefore, the proposed changes do not result in a 
significant reduction in the margin of safety.

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

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

    Date of amendment request: December 14, 1995.
    Description of amendment request: The licensee proposes to 
incorporate the inservice testing (IST) requirements of Section XI of 
the American Society of Mechanical Engineers Boiler and Pressure Vessel 
Code (ASME Code). The proposed change adds a new surveillance 
requirement, 4.0.E, which refers to the requirements of Section XI of 
the ASME Code and Addenda established by 10 CFR 50.55a(f). Ancillary 
changes are also required since the proposed specification 4.0.E 
replaces the surveillance testing requirements of safety related pump 
and motor-operated valves and extends the surveillance testing 
frequency of other components from once every month, to coincide with 
the ASME Code Section XI requirements.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    Operation of the FitzPatrick plant in accordance with the 
proposed Amendment would not involve a significant hazards 
consideration as defined in 10 CFR 50.92, since it would not:
    1. Involve a significant increase in the probability or 
consequences of an accident previously evaluated.
    The changes identified in this proposed amendment revise 
surveillance testing for various systems based upon the Section XI 
of the American Society of Mechanical Engineers [***] Boiler and 
Pressure Vessel [***] Code [ASME Code]. None of these changes 
involves a hardware modification to the plant, a change to system 
operation, a change to the manner in which the system is used, or a 
change in the ability of the system to perform its intended 
function.
    The use of Section XI of the ASME [***] Code as a basis for 
establishing surveillance testing and acceptance criteria will not 
alter existing accident analyses. This has been acknowledged and 
accepted by the NRC in the Standard Technical Specifications. The 
change to surveillance testing frequencies reduces testing at power, 
increases the availability of systems important to the mitigation of 
a DBA [design-basis accident], and minimizes component degradation 
due to excessive testing. The ASME [***] Code, Section XI testing 
tracks component performance allowing identification of component 
degradation and the code specifies that if a pump parameter enters 
the alert range, then the testing frequency is doubled until the 
cause of the degradation is determined and the condition corrected. 
Similarly, if a valve stroke time degrades, the valve testing 
frequency is increased to once per month until the cause is 
determined and the condition corrected.
    The editorial changes are strictly non technical in nature with 
no effect on existing analyses. They clarify the Technical 
Specifications by improving the legibility of this document.
    2. Create the possibility of a new or different kind of accident 
from those previously evaluated.
    The proposed changes involve no hardware changes, no changes to 
the operation of the systems, and do not change the ability of the 
systems to perform their intended functions. The use of ASME Section 
XI as the basis for testing involves the same testing alignments and 
practices previously used as part of either the IST program or 
Technical Specification Surveillance Requirements. The editorial 
changes have no effect on plant practices.
    3. Involve a significant reduction in the margin of safety.
    There are no hardware modifications, changes to system 
operations, or effect on the ability of systems to perform their 
intended function associated with the proposed changes. The proposed 
changes to reference pump and valve testing to Section XI of the 
ASME [***] Code and remove individual Surveillance Requirements in 
the Technical Specifications does not relax any controls or 
limitations. The resulting reduction in test frequency, while 
reducing the possibility of detecting a degraded component prior to 
failure, is offset by the increased availability of systems 
important to plant safety and an associated reduction in component 
wear and degradation due to excessive testing. Additionally, the 
ASME testing program evaluates components for degraded performance 
and will identify such degradation early. There are no safety 
margins associated with the editorial corrections.

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

South Carolina Electric & Gas Company (SCE&G), South Carolina Public 
Service Authority, Docket No. 50-395, Virgil C. Summer Nuclear Station, 
Unit No. 1, Fairfield County, South Carolina.

    Date of amendment request: December 8, 1995.
    Description of amendment request: The proposed changes add a new 

[[Page 1636]]
    surveillance requirement to Technical Specification (TS) Section 
4.1.2.2 and deletes TS Sections 3/4.1.2.3 and 3/4.1.2.4 associated with 
the Borations Systems section. TS Section 3/4.9.3 is being revised to 
assure only one charging pump is capable of Reactor Coolant System 
injection in the applicable modes and to add a new surveillance 
requirement to demonstrate this assurance. TS Section 4.5.2.f is being 
revised to delete specific Emergency Core Cooling System pump testing 
acceptance criteria and reference acceptance criteria located in the 
plant Inservice Testing Program. In addition, the licensee has proposed 
changes to the bases.
    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 probability or consequences of an accident previously 
evaluated is not significantly increased.
    The implementation of the above described TS changes will have 
no impact on the probability of an accident occurring. The testing 
of the ECCS pumps at a more appropriate point on their 
characteristic curve is not a precursor to an accident. There is no 
hardware, software, or testing methodology change proposed that 
would decrease confidence in the reliability of these systems/
components.
    The proposed revision to the ECCS Pump testing surveillance will 
allow greater flexibility for testing and will provide more useful 
information about the performance capabilities of those pumps.
    The deletion of the Reactivity Control System Specifications 
(Charging Pumps - Operating and Charging Pumps - Shutdown) will have 
no impact on the capability of the Charging/SI pumps to perform 
their design function. The additional Action Statement and 
Surveillance for low temperature overpressure (LTOP) assure that 
safety analyses remain valid and initial conditions are not changed. 
The additional Surveillance Requirement for Boration Systems assures 
that one charging pump will be operable during Modes 5 and 6.
    2. The proposed license amendment does not create the 
possibility of a new or different kind of accident from any accident 
previously evaluated.
    This proposed TS change does not involve any changes to station 
hardware, software, or operating practices. The changes do provide 
for a revision to the testing methodology used in demonstrating the 
capability of the ECCS pumps.
    This methodology will test the ECCS pumps at a point on the 
pump's characteristic curve that will more reliably indicate the 
pump's continued operability at or near the parameters the pump 
would be required to provide during a postulated accident.
    The deletion of the Reactivity Control System Specifications 
(Charging Pumps - Operating and Charging Pump - Shutdown) will not 
provide additional challenges to the capability of the plant to meet 
normal operational needs or mitigate the conditions of a design 
basis accident. The ECCS Subsystems TS provide similar surveillance 
requirements to insure continued operability of the Charging/SI 
pumps. The LTOP TS will now provide requirements to assure that 
design assumptions are not challenged and RCS integrity is 
maintained.
     Therefore, as the above described change has no impact on plant 
performance, the possibility of a new or different kind of accident 
being created as a result of this change is negligible.
    3. The proposed license amendment does not involve a significant 
reduction in a margin of safety.
    The change in testing philosophy for ECCS pumps should bring an 
increase in margin of safety, since testing will be conducted at 
reference flow points closer to actual pump parameters for accident 
conditions. For the Residual Heat Removal Pumps this will be 
conducted quarterly and for the centrifugal charging pumps, they 
will be tested quarterly on minimum flow and each refueling outage 
at substantial flow per the Inservice Testing Program.
    The surveillance requirements of TS 3/4.1.2.3 and TS 3/4.1.2.4 
are essentially the same as those in 3./4.5.2 and 3/4.5.3 (ECCS 
Subsystems), and the deletion of these requirements will have no 
adverse impact on margin on safety. The addition of the Action 
Statement and Surveillance Requirements to 3/4.4.9.3 (Overpressure 
Protective Systems) provide additional requirements to supplement 
those above to assure RCS integrity is maintained for all 
operational modes. The addition of the Surveillance Requirement to 
3/4.1.2.1 will provide assurance that reactivity control can be 
maintained for Modes 5 and 6 through the charging system flow path.

    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: Fairfield County Library, 300 
Washington Street, Winnsboro, SC 29180.
    Attorney for licensee: Randolph R. Mahan, South Carolina Electric & 
Gas Company, Post Office Box 764, Columbia, South Carolina 29218.
    NRC Project Director: Frederick J. Hebdon.

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

    Date of amendments request: December 19, 1995.
    Description of amendments request: The proposed amendments would 
replace the requirements associated with the Control Room Emergency 
Ventilation System with requirements related to the operation of the 
Control Room Emergency Filtration/Pressurization System and Control 
Room Air Conditioning System. These changes are technically consistent 
with the requirements of NUREG-1431, Revision 1, ``Westinghouse 
Standard Technical Specifications,'' issued on April 7, 1995. Also, a 
one-time extension to the allowable outage time for the control room 
recirculation filtration system is included to facilitate 
implementation of design modifications to enhance the reliability of 
the control room air conditioning system during the spring of 1996.
    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:

    Based on the preceding evaluation, the following conclusions are 
provided with respect to the criteria contained in 10 CFR 50.92.
    (1) The proposed changes do not significantly increase the 
probability or consequences of an accident previously evaluated in 
the FSAR [Final Safety Analysis Report]. The proposed changes have 
no impact on the probability of an accident. The control room 
ventilation systems are support systems which have a role in the 
detection and mitigation of accidents but do not contribute to the 
initiation of any accident previously evaluated. Reorganizing the 
technical specifications by functions have no impact on the course 
of any accidents previously evaluated. The other changes which are 
being made improve the ability to mitigate fuel handling accidents. 
Specifying an allowed outage time (AOT) of 30 days for the cooling 
of recirculated air while one train is inoperable is based on the 
significance of the cooling function but does represent an increase 
in the allowed outage time and thus an increase in the probability 
that the functions could be unavailable. This increase is not 
considered significant based on several factors including: the 
design is based on the worst postulated meteorological conditions; 
generally, less than design cooling is required and a partial 
failure in the system may have no impact; and unavailability failure 
does not create an immediate irreversible impact (i.e., temperature 
will increase slowly over a period of time); the system could be 
restored or its loss mitigated without any impact on the course or 
whatever accident is being considered; and the extended AOT would 
allow more opportunity to perform major required maintenance and 
thus may provide an overall improvement in equipment reliability.
    In addition, the one-time change to the AOT for the 
recirculation filtration will not 

[[Page 1637]]
significantly increase the probability or consequences of an accident 
due to the low probability of an event result[ing] in an airborne 
release of radioactivity. Such an event requires multiple failures 
of safety systems that are governed by technical specifications not 
affected by these changes. In addition, compensatory measures have 
been identified that limit the potential exposure of control room 
operators in response to a postulated release.
    The net effect of these changes is not significant and, as a 
result, the changes do not involve a significant increase in the 
consequences of an accident previously evaluated.
    (2) The proposed changes to the Technical Specifications do not 
increase the possibility of a new or different kind of accident than 
any accident already evaluated in the FSAR. No new limiting single 
failure or accident scenarios have been created or identified due to 
the proposed changes. Safety-related systems are expected to perform 
as designed. Although the changes could have a minor impact on the 
air conditioning system availability, the changes do not create the 
possibility of a new or different kind of accident from any 
previously evaluated.
    (3) The proposed changes do not involve a significant reduction 
in the margin of safety. The changes proposed do not alter the 
environmental conditions which are to be maintained in the control 
room during normal operations and following an accident. As a 
result, the margin of safety for these functions remains the same. 
Although there is a potential impact on the air conditioning 
system's postulated availability, there is no impact on the accident 
analyses. Further, although the one-time AOT extension for the 
recirculation filtration system increases the system unavailability 
during the planned CRACS [Control Room Air Conditioning System] 
design changes, the net effect is a benefit to plant safety due to 
the enhancement to control room cooling capability. Thus, even if 
system availability issues were considered an aspect of margin of 
safety, the proposed changes do not involve a significant reduction 
in the margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the 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: Houston-Love Memorial Library, 
212 W. Burdeshaw Street, Post Office Box 1369, Dothan, Alabama 36302.
    Attorney for licensee: M. Stanford Blanton, Esq., Balch and 
Bingham, Post Office Box 306, 1710 Sixth Avenue North, Birmingham, 
Alabama 35201.
    NRC Project Director: Herbert N. Berkow.

Tennessee Valley Authority, Docket Nos. 50-259, 50-260 and 50-296, 
Browns Ferry Nuclear Plant, Units 1, 2 and 3, Limestone County, 
Alabama.

    Date of amendment request: December 8, 1995 (TS 364).
    Description of amendment request: The licensee proposes revision of 
Units 1, 2, and 3 Technical Specifications (TS) Section 4.7.A to 
implement the revision to 10 CFR 50, Appendix J. The new rule (Option 
B) provides a voluntary performance-based testing option for 
containment leak rate testing. Option B containment leak rate testing 
requirements are based on system and component performance in lieu of 
compliance with the current prescriptive requirements.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    A. The proposed amendment does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    The proposed amendment to TS Section 4.7.A is in accordance with 
Option B to 10 CFR 50, Appendix J. The proposed amendment adds a 
voluntary performance based option for containment leak rate 
testing. The changes being proposed do not affect the precursor for 
any accident or transient analyzed in Chapter 14 of the BFN [Browns 
Ferry Nuclear Plant] Updated Final Safety Analysis Report (UFSAR). 
The proposed change does not increase the total allowable primary 
containment leakage rate. The proposed change does not reflect a 
revision to the physical design and/or operation of the plant. 
Therefore, operation of the facility in accordance with the proposed 
change does not affect the probability or consequences of an 
accident previously evaluated.
    B. The proposed amendment does not create the possibility of a 
new or different kind of accident from any accident previously 
evaluated.
    The proposed amendment to TS Section 4.7.A is in accordance with 
the new performance-based option (Option B) to 10 CFR 50, Appendix 
J. The changes being proposed will not change the physical plant or 
the modes of operation defined in the facility license. The proposed 
changes do not increase the total allowable primary containment 
leakage rate. The changes do not involve the addition or 
modification of equipment, nor do they alter the design or operation 
of plant systems. Therefore, operation of the facility in accordance 
with the proposed change does not create the possibility of a new or 
different kind of accident from any previously evaluated.
    C. The proposed amendment does not involve a significant 
reduction in a margin of safety.
    The proposed change to TS Section 4.7.A is in accordance with 
the new option to 10 CFR 50, Appendix J. The proposed option is 
formulated to adopt performance-based approaches. This option 
removes the current prescriptive details from the TS. The proposed 
changes do not affect plant safety analyses or change the physical 
design or operation of the plant. The proposed change does not 
increase the total allowable primary containment leakage rate. 
Therefore, operation of the facility in accordance with the proposed 
change does not involve a significant reduction in the margin of 
safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: Athens Public Library, South 
Street, Athens, Alabama 35611.
    Attorney for licensee: General Counsel, Tennessee Valley Authority, 
400 West Summit Hill Drive, ET llH, Knoxville, Tennessee 37902.
    NRC Project Director: Frederick J. Hebdon.

Toledo Edison Company, Centerior Service Company, and The Cleveland 
Electric Illuminating Company, Docket No. 50-346, Davis-Besse Nuclear 
Power Station, Unit No. 1, Ottawa County, Ohio.

    Date of amendment request: December 12, 1995.
    Description of amendment request: The proposed amendment would 
revise Technical Specification (TS) 3/4.6.1.1, Containment Systems--
Primary Containment--Containment Integrity; TS 3/4.6.1.2, Containment 
Systems--Containment Leakage; TS 3/4.6.1.6, Containment Systems--
Containment Vessel Structural Integrity; TS 3/4.6.5.3, Containment 
Systems--Shield Building Structural Integrity; and associated Bases. 
The proposed revisions adopt the provisions of Appendix J, Option B for 
Type A containment leakage testing as modified by approved exemptions 
and in accordance with the guidance of Regulatory Guide 1.163. The 
licensee proposes to delete surveillance requirement (SR) 4.6.1.2, SR 
4.6.1.2.b, SR 4.6.1.2.c, and SR 4.6.1.2.i since these requirements 
contain details that are now included in standards that are referenced 
by Regulatory Guide 1.163. TS 3/4.6.1.6 and TS 3/4.6.5.3 which address 
containment building and shield building structural integrity are 
proposed to be deleted since the requirements are addressed in revised 
TS 3.6.1.2.a. The licensee proposes to delete the exemption included in 
Bases 

[[Page 1638]]
3/4.6.1.2 since it is no longer applicable. Additionally, the licensee 
proposes to modify the Action statement associated with TS 3.6.1.2 to 
reflect the action to take if the as-left rather than the as-found 
leakage exceeds 0.75 La.
    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:

    Toledo Edison has reviewed the proposed changes and determined 
that a significant hazards consideration does not exist because 
operation of the Davis-Besse Nuclear Power Station, Unit No. 1, in 
accordance with the changes would:
    1a. Not involve a significant increase in the probability of an 
accident previously evaluated because accident initiators, 
conditions, or assumptions are not affected by the proposed changes.
    The proposed changes to the Technical Specifications implement 
10 CFR 50 Appendix J Option B for Type A testing, including visual 
examinations of the containment vessel and shield building, and make 
various administrative changes to the Technical Specifications and 
associated Technical Specification Bases. Therefore, as stated 
above, these proposed changes do not affect accident initiators, 
conditions, or assumptions.
    1b. Not involve a significant increase in the consequences of an 
accident previously evaluated because the proposed changes do not 
change the source term, containment isolation, or allowable 
releases.
    The proposed changes involve containment leakage testing and 
test frequency. The allowable containment leakage rates presently 
specified in the Technical Specifications remain unchanged.
    2. Not create the possibility of a new or different kind of 
accident from any accident previously evaluated because no new 
accident initiators or assumptions are introduced by the proposed 
changes.
    3. Not involve a significant reduction in a margin of safety, 
for the reasons cited below.
    The proposed changes involve containment leakage testing and 
test frequency. The allowable containment leakage rates presently 
specified in the Technical Specifications remain unchanged. The 
Technical Specifications, under the proposed changes, will continue 
to ensure containment system reliability by periodic testing 
performed in full compliance with 10 CFR 50 Appendix J.
    As stated in the Federal Register publication of the final rule, 
60 FR 49495 dated September 26, 1995, the final rule improves the 
focus of the regulations by eliminating prescriptive requirements 
that are marginal to safety. Further, the final rule allows test 
intervals to be based on system and component performance and 
provides licensees greater flexibility for cost-effective 
implementation methods of regulatory safety objectives. The final 
rule publication also discusses the following specific findings 
documented in NUREG-1493, ``Performance-Based Containment Leak-Test 
Program,'' September, 1995, which justify the proposed change in 
frequency of Type A Integrated Leak Rate Testing (ILRT):
    1. The fraction of leakages detected only by ILRT's is small, on 
the order of a few percent.
    2. Reducing the frequency of ILRT testing from 3 every 10 years 
to one every 10 years leads to a marginal increase in risk.
    3. At a frequency of one test every 10 years, industry-wide 
occupational exposure would be reduced by 0.087 person-sievert (8.7 
person-rem) per year.
    Based on these considerations, it is concluded that 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 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: University of Toledo, William 
Carlson Library, Government Documents Collection, 2801 West Bancroft 
Avenue, Toledo, Ohio 43606.
    Attorney for licensee: Jay E. Silberg, Esquire, Shaw, Pittman, 
Potts and Trowbridge, 2300 N Street, NW., Washington, DC 20037.
    NRC Project Director: Gail H. Marcus.

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

    Date of amendment request: December 13, 1995.
    Description of amendment request: The proposed amendments will 
modify Technical Specification (TS) Sections 15.1, ``Definitions,'' 
15.2, ``Safety Limits and Limiting Safety System Settings,'' 15.3, 
``Limiting Conditions for Operation,'' and 15.6, ``Administrative 
Controls.'' The proposed changes would modify the TSs to account for 
the creation and maintenance of a Core Operating Limits Report.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration which is presented below:

    1. Operation of this facility under the proposed Technical 
Specifications will not create a significant increase in the 
probability or consequences of an accident previously evaluated.
    The relocation of the cycle-specific parameters from the Point 
Beach Nuclear Plant (PBNP) Technical Specifications to the Core 
Operating Limits Report (COLR) has no impact on plant operation or 
accident analyses. The proposed changes are administrative in 
nature. The Technical Specifications will continue to require 
operation within the core operational limits for each cycle reload 
calculated by the NRC-approved reload design methodologies. The 
appropriate actions required if limits are exceeded will remain in 
the Technical Specifications. The reload report presents the results 
of a cycle-specific evaluation of accidents and transients addressed 
in the PBNP Final Safety Analysis Report (FSAR). The cycle-specific 
evaluation demonstrates that changes in the unit's fuel cycle design 
and corresponding COLR parameters do not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated. Therefore, these changes do not involve a 
significant increase in the probability or consequences of any 
accident previously evaluated.
    2. Operation of this facility under the proposed Technical 
Specifications will not create the possibility of a new or different 
kind of accident from any accident previously evaluated.
    The proposed change to relocate the cycle-specific parameters 
from the Technical Specifications to the COLR is administrative in 
nature. No change to the design, configuration, or method of 
operation of the plant is made by this change. The cycle-specific 
parameters will be determined using NRC-approved methodologies. The 
Technical Specifications will continue to require operation within 
the core operating limits and appropriate actions will be taken if 
the limits are exceeded.
    Therefore, these changes do not create the possibility of a new 
or different kind of accident from any accident previously 
evaluated.
    3. Operation of this facility under the proposed Technical 
Specifications will not create a significant reduction in a margin 
of safety.
    Existing Technical Specification operability and surveillance 
requirements are not reduced by the proposed changes to relocate 
cycle-specific parameters from the Technical Specifications to the 
COLR. The cycle-specific COLR limits for reloads will continue to be 
developed based on NRC-approved methodologies, thereby maintaining 
accepted margins of safety. The Technical Specifications will still 
require that the core be operated within these limits and specify 
appropriate actions to be taken if the limits are violated. Each 
reload undergoes a 10 CFR 50.59 safety review to assure that 
operating the unit within the cycle-specific limits will not involve 
a significant reduction in a margin of safety. Therefore, these 
changes do not involve a significant reduction in the margin of 
safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the 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: Joseph P. Mann Library, 1516 

[[Page 1639]]
    Sixteenth Street, Two Rivers, Wisconsin 54241.
    Attorney for licensee: Gerald Charnoff, Esq., Shaw, Pittman, Potts, 
and Trowbridge, 2300 N Street, NW., Washington, DC 20037.
    NRC Project Director: Gail H. Marcus.

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

    Date of amendment request: December 13, 1995.
    Description of amendment request: This license amendment request 
proposes to revise the 125-volt D.C. Sources Technical Specifications 
(3.8.2.1 and 3.8.2.2) to include provisions for installed spare 
chargers, which will be added to the plant design during the next 
refueling outage. The Onsite Power Distribution Technical 
Specifications 3.8.3.1 and 3.8.3.2 would be revised to indicate that 
spare chargers may be connected in place of the primary chargers.
    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.
    These proposed technical specification changes do not alter the 
plant design bases nor do they involve any hardware changes that 
significantly increase the probability of any event initiators. 
There will be no change to normal plant operating parameters or 
accident mitigation capabilities. There will be no increase in the 
consequences of any accident or equipment malfunction.
    2. The proposed change does not create the possibility of a new 
or different kind of accident from any accident previously 
evaluated.
    The proposed technical specification changes do not involve any 
design bases changes nor are there any changes to the method by 
which any safety-related plant system performs its safety function. 
The normal manner of plant operation is unaffected. No new accident 
scenarios, transient precursors, failure mechanisms, or limiting 
single failures are introduced as a result of these changes.
    3. The proposed change does not involve a significant reduction 
in a margin of safety.
    There will be no effect on the manner in which safety limits or 
limiting safety system settings are determined, nor will there be 
any effect in those plant systems necessary to assure the 
accomplishment of protection functions. There will be no impact on 
DNBR [departure from nucleate boiling ratio] limits, FQ, F-
delta-H, LOCA [loss-of-coolant accident] PCT [peak cladding 
temperature], peak local power density or any other 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 locations: Emporia State University, 
William Allen White Library, 1200 Commercial Street, Emporia, Kansas 
66801 and Washburn University School of Law Library, Topeka, Kansas 
66621.
    Attorney for licensee: Jay Silberg, Esq., Shaw, Pittman, Potts and 
Trowbridge, 2300 N Street, NW., Washington, D.C. 20037.
    NRC Project Director: William H. Bateman.

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

    Date of amendment request: December 13, 1995.
    Description of amendment request: This change request proposes 
revising the minimum and maximum flow requirements for the centrifugal 
charging pumps (CCPs) and safety injection pumps (SIPs) specified in 
Technical Specification Surveillance Requirement 4.5.2.h. Specifically, 
the proposed changes would:
    (1) Decrease the minimum limits on the sum of the injection line 
flow rates, excluding the highest flow rate, from 346 gpm to 330 gpm 
for the CCPs and from 459 gpm to 450 gpm for the SIPs.
    (2) Revise the maximum pump flow rate for the SIP from 665 to 670 
gpm, but retain the CCPs maximum pump flow rate at its current value of 
556 gpm.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. The proposed change does not involve a significant increase 
in the probability or consequences of an accident previously 
evaluated.
    The proposed change will not result in a condition where the 
material or construction standards applicable prior to the change 
are altered. The ECCS [emergency core cooling system] system 
integrity is not affected by this change, and this change will not 
affect the ability of the ECCS to fulfill its design functions. This 
change will modify the pump surveillance criteria to prevent pump 
runout during the test, but will not affect the method of operation 
of the system and will not alter the testing method for the pumps. 
This change will slightly alter the acceptance criteria of the test, 
but the changes have been determined to be enveloped by the ECCS 
pump flow and balance criteria assumed in the safety analyses 
described in the USAR [Updated Safety Analysis Report]. This change 
will not affect the ability of the ECCS to mitigate the consequences 
of any previously evaluated accident. The proposed change will not 
alter, degrade or prevent the response of the ECCS to any accident 
scenarios evaluated in the USAR. Therefore, neither the probability 
of occurrence nor the consequences of any accident previously 
evaluated in the USAR will be increased by this change.
    2. The proposed change does not create the possibility of a new 
or different kind of accident from any accident previously 
evaluated.
    The proposed change will alter the existing ECCS pump flow test 
to prevent pump runout during the test by slightly altering the 
acceptance criteria of the test. However, the proposed changes have 
been determined to be enveloped by the ECCS pump flow and balance 
criteria assumed in the safety analyses described in the USAR. This 
change will not create a new type of accident or malfunction, and 
the method and manner of plant operation remains unchanged. This 
change will not alter the safety functions of the ECCS. The safety 
design bases in the USAR have not been altered, and no new or 
different accident scenarios, transient precursors, failure 
mechanisms, or limiting single failures will be introduced as a 
result of this change. Therefore, the possibility of a new or 
different kind of accident other than those already evaluated will 
not be created by this change.
    3. The proposed change does not involve a significant reduction 
in a margin of safety.
    There are no changes being made to any safety limits or safety 
system settings that would adversely impact plant safety. This 
proposed change will have no affect on the availability, operability 
or performance of any safety-related system or component. The 
analysis results and conclusions of the accidents presented in the 
current USAR would not be adversely affected by the revised 
surveillance requirements for the ECCS. This conclusion is drawn 
based on the evaluation that confirms that the actual ECCS flow 
characteristics remain consistent with assumptions used in the WCGS 
[Wolf Creek Generating Station] accident analyses. Specifically, the 
accident analyses which are limiting with minimized ECCS flow have 
already been analyzed using revised ECCS flows that were developed 
based on a more conservative minimum flow than the proposed minimum 
ECCS flow requirement. For the analyses which are limiting with a 
higher ECCS flow, the evaluation indicated that a higher pump runout 
limit proposed for the SIPs would have insignificant effect on the 
results and conclusions of the analyses. The evaluation also 
indicated that the ECCS pump operability would not be a concern as a 
result of increasing the SIPs runout limit because the available 
runout margin is sufficient to accommodate the cumulative effect of 
the ECCS performance issues. Based on these reasons, it is concluded 
that 

[[Page 1640]]
implementation of the proposed changes will have no adverse impact on 
the ECCS subsystems' operability and their intended safety function. 
Therefore, the proposed change would not result in a reduction in a 
margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room locations: Emporia State University, 
William Allen White Library, 1200 Commercial Street, Emporia, Kansas 
66801 and Washburn University School of Law Library, Topeka, Kansas 
66621.
    Attorney for licensee: Jay Silberg, Esq., Shaw, Pittman, Potts and 
Trowbridge, 2300 N Street, N.W., Washington, D.C. 20037.
    NRC Project Director: William H. Bateman.

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

    Date of amendment request: December 13, 1995.
    Description of amendment request: This license amendment request 
proposes revising Surveillance Requirement 4.1.3.1.3 to delete the 
requirement for performing the control rod drop surveillance test with 
Tavg greater than or equal to 551 deg.F. This would allow 
performing this test with Tavg below 551 deg.F. This change will 
also add justification for performing the rod drop test with Tavg 
below 551 deg.F to Bases Section 3/4.1.3, ``Movable Control 
Assemblies.''
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. The proposed change does not involve a significant increase 
in the probability or consequences of an accident previously 
evaluated.
    The proposed change will not result in a condition where the 
material or construction standards applicable prior to the change 
are altered. The rod control system integrity is not affected by 
this change, and this change will not affect the ability of the 
system to fulfill its design function. This change will allow the 
control rod drop test to be performed at lower temperatures than 
currently allowed, but will not affect the method of operation of 
the system and will not alter the drop time criterion of the test. 
This change will not affect any fission product barrier, and will 
not affect the integrity of any fuel assembly or the reactor 
internals. Thus this change will not affect the ability of the rod 
control system to mitigate the consequences of any previously 
evaluated accident. The proposed change will not alter, degrade or 
prevent the response of the rod control system to any accident 
scenarios evaluated in the USAR [Updated Safety Analysis Report]. 
Therefore, neither the probability of occurrence nor the 
consequences of any accident previously evaluated in the USAR will 
be increased by this change.
    2. The proposed change does not create the possibility of a new 
or different kind of accident from any accident previously 
evaluated.
    The proposed change will alter the existing rod drop test to 
allow the test to be performed over a range of temperatures, but 
will not alter the rod drop time criterion of the test. This change 
will not create a new type of accident or malfunction, and the 
method and manner of plant operation remains unchanged. This change 
will not alter the safety functions of the rod control system. The 
safety design bases in the USAR have not been altered, and no new or 
different accident scenarios, transient precursors, failure 
mechanisms, or limiting single failures will be introduced as a 
result of this change. Therefore, the possibility of a new or 
different kind of accident other than those already evaluated will 
not be created by this change.
    3. The proposed change does not involve a significant reduction 
in a margin of safety.
    There are no changes being made to any safety limits or safety 
system settings that would adversely impact plant safety. This 
proposed change will have no affect on the availability, operability 
or performance of any safety-related system or component. The change 
will not prevent inspections or surveillances required by the 
technical specifications, and does not alter the rod drop time 
criterion specified in the technical specifications. Performance of 
the rod drop tests at other temperatures allows an alternative 
method to verify that the rod drop time currently specified in the 
technical specifications and used in the safety analyses continues 
to be valid. Therefore, the proposed change would not result in a 
reduction in a margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room locations: Emporia State University, 
William Allen White Library, 1200 Commercial Street, Emporia, Kansas 
66801 and Washburn University School of Law Library, Topeka, Kansas 
66621.
    Attorney for licensee: Jay Silberg, Esq., Shaw, Pittman, Potts and 
Trowbridge, 2300 N Street, N.W., Washington, D.C. 20037.
    NRC Project Director: William H. Bateman.

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.

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

    Date of amendment request: November 21, 1995
    Brief description of amendment request: The proposed amendments 
would revise surveillance requirements for the high pressure coolant 
injection and reactor core isolation cooling systems and would make an 
administrative change to Section 5.5.7 of the technical specifications 
to eliminate reference to a section which was previously eliminated.
    Date of publication of individual notice in Federal Register: 
December 5, 1995 (60 FR 62271).
    Expiration date of individual notice: January 3, 1996.
    Local Public Document Room location: Government Publications 
Section, State Library of Pennsylvania, (Regional Depository) Education 
Building, Walnut Street and Commonwealth Avenue, Box 1601, Harrisburg, 
Pennsylvania 17105.

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

    Date of amendment request: November 30, 1995.
    
[[Page 1641]]

    Brief description of amendment request: The proposed amendments 
would revise the minimum allowable control rod scram accumulator 
pressure and charging water header pressure from a value of 955 psig to 
a value of 940 psig.
    Date of publication of individual notice in Federal Register: 
December 8, 1995 (60 FR 63073).
    Expiration date of individual notice: January 8, 1996.
    Local Public Document Room location: Government Publications 
Section, State Library of Pennsylvania, (Regional Depository) Education 
Building, Walnut Street and Commonwealth Avenue, Box 1601, Harrisburg, 
Pennsylvania 17105.

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

    Date of amendment request: December 19, 1995.
    Brief description of amendment request: The proposed amendment 
would revise the ventilation filter test program (VFTP) bypass and 
penetration leakage test acceptance criteria from less than 0.05 
percent to less than 1.0 percent. The change corrects an administrative 
error that occurred during the development of the Peach Bottom Improved 
Technical Specifications which were issued as Amendments 210 and 214 to 
the Peach Bottom licenses on August 30, 1995.
    Date of publication of individual notice in Federal Register: 
December 27, 1995 (60 FR 66997).
    Expiration date of individual notice: January 25, 1996.
    Local Public Document Room location: Government Publications 
Section, State Library of Pennsylvania, (REGIONAL DEPOSITORY) Education 
Building, Walnut Street and Commonwealth Avenue, Box 1601, Harrisburg, 
Pennsylvania 17105.

Notice of Issuance of Amendments To Facility Operating Licenses

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

Baltimore Gas and Electric Company, Docket No. 50-317, Calvert Cliffs 
Nuclear Power Plant, Unit No. 1, Calvert County, Maryland.

    Date of application for amendment: October 20, 1995.
    Brief description of amendment: The one-time amendment revises the 
Calvert Cliffs Nuclear Power Plant, Unit No. 1 Technical Specifications 
by extending certain 18-month instrument surveillance intervals by a 
maximum of 39 days to March 31, 1996. This amendment will be superseded 
by Amendment No. 208 when it is implemented prior to restart from the 
Unit No. 1 spring 1996 refueling outage.
    Date of issuance: December 28, 1995.
    Effective date: As of the date of issuance to be implemented within 
30 days.
    Amendment No.: 209.
    Facility Operating License No. DPR-53: Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: November 27, 1995 (60 
FR 58396).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated December 28, 1995.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: Calvert County Library, Prince 
Frederick, Maryland 20678.

Baltimore Gas and Electric Company, Docket No. 50-318, Calvert Cliffs 
Nuclear Power Plant, Unit No. 2, Calvert County, Maryland.

    Date of application for amendment: October 2, 1995.
    Brief description of amendment: The amendment revises the Technical 
Specifications regarding allowable outage time (AOT) associated with 
the control room emergency ventilation system. It extends the AOT for 
one train from 7 days to 30 days on a one-time basis (for the loss of 
the emergency power supply only) to allow for modifications during the 
upcoming Unit No. 1 refueling outage in the spring of 1996.
    Date of issuance: December 19, 1995.
    Effective date: As of the date of issuance to be implemented during 
the Unit No. 1 spring 1996 refueling outage.
    Amendment No.: 187.
    Facility Operating License No. DPR-69: Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: November 8, 1995 (60 FR 
56363).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated December 19, 1995.
    No significant hazards consideration comments received: No
    Local Public Document Room location: Calvert County Library, Prince 
Frederick, Maryland 20678.

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

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

    Date of application for amendments: September 10, 1993, as 
supplemented on June 16, 1995.
    Brief description of amendments: This application upgrades the 
current custom Technical Specifications (TS) for Dresden and Quad 
Cities to the Standard Technical Specifications contained in NUREG-
0123, ``Standard Technical Specification General Electric Plants BWR/
4.'' This application upgrades only Section 3/4.8 (Plant Systems).
    Date of issuance: December 19, 1995. 
    
[[Page 1642]]

    Effective date: Immediately, to be implemented no later than June 
30, 1996.
    Amendment Nos.: 144, 138, 166, and 162.
    Facility Operating License Nos. DPR-19, DPR-25, DPR-29 and DPR-30. 
The amendments revised the Technical Specifications.
    Date of initial notice in Federal Register: July 19, 1995 (60 FR 
37086).
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated December 19, 1995.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: for Dresden, Morris Area 
Public Library District, 604 Liberty Street, Morris, Illinois 60450; 
for Quad Cities, Dixon Public Library, 221 Hennepin Avenue, Dixon, 
Illinois 61021.

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

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

    Date of application for amendments: September 15, 1995.
    Brief description of amendments: The amendments upgrade the current 
custom Technical Specifications (TS) for Dresden and Quad Cities to the 
Standard Technical Specifications contained in NUREG-0123, ``Standard 
Technical Specification General Electric Plants BWR/4.'' The 
application dated September 15, 1995, contains some of the TSUP open 
items from previous Dresden and Quad Cities TS amendments issued by the 
NRC.
    Date of issuance: December 19, 1995.
    Effective date: Immediately, to be implemented no later than June 
30, 1996.
    Amendment Nos.: 145, 139, 167 and 163
    Facility Operating License Nos. DPR-19, DPR-25, DPR-29 and DPR-30. 
The amendments revised the Technical Specifications.
    Date of initial notice in Federal Register: October 5, 1995 (60 FR 
52220).
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated December 19, 1995.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: for Dresden, Morris Area 
Public Library District, 604 Liberty Street, Morris, Illinois 60450; 
for Quad Cities, Dixon Public Library, 221 Hennepin Avenue, Dixon, 
Illinois 61021.

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

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

    Date of application for amendments: September 17, 1993, as 
supplemented July 28, 1995.
    Brief description of amendments: This application upgrades the 
current custom Technical Specifications (TS) for Dresden and Quad 
Cities to the Standard Technical Specifications contained in NUREG-
0123, ``Standard Technical Specification General Electric Plants BWR/
4.'' This application upgrades only Section 3/4.5 (Emergency Core 
Cooling Systems).
    Date of issuance: December 27, 1995.
    Effective date: Immediately, to be implemented no later than June 
30, 1996.
    Amendment Nos.: 146, 140, 168, and 164.
    Facility Operating License Nos. DPR-19, DPR-25, DPR-29 and DPR-30. 
The amendments revised the Technical Specifications.
    Date of initial notice in Federal Register: August 16, 1995 (60 FR 
42599).
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated December 27, 1995.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: for Dresden, Morris Area 
Public Library District, 604 Liberty Street, Morris, Illinois 60450; 
for Quad Cities, Dixon Public Library, 221 Hennepin Avenue, Dixon, 
Illinois 61021.

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: November 14, 1995.
    Brief description of amendments: These amendments change the 
implementation dates of all previous TSUP amendments from December 31, 
1995, to no later than June 30, 1996.
    Date of issuance: December 29, 1995.
    Effective date: December 29, 1995.
    Amendment Nos.: 147 and 141.
    Facility Operating License Nos. DPR-19 and DPR-25: The amendments 
revised the license.
    Date of initial notice in Federal Register: November 29, 1995 (60 
FR 61272).
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated December 29, 1995.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: Morris Area Public Library 
District, 604 Liberty Street, Morris, Illinois 60450.

Commonwealth Edison Company, Docket No. 50-373, LaSalle County Station, 
Unit 1, LaSalle County, Illinois.

    Date of application for amendment: October 2, 1995.
    Brief description of amendment: The amendment revises the safety/
relief valve (SRV) safety function lift setting allowable tolerance 
band from -3/+1% to 3% and includes a requirement for the 
lift settings to be within 1% of the technical 
specification limit following testing.
    Date of issuance: January 3, 1996.
    Effective date: Upon date of issuance; shall be implemented prior 
to the restart of Unit 1 from its seventh refueling outage.
    Amendment No.: 108.
    Facility Operating License No. NPF-11: The amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: November 27, 1995 (60 
FR 58398).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated January 3, 1996.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: Jacobs Memorial Library, 
Illinois Valley Community College, Oglesby, Illinois 61348.
    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 application for amendments: September 5, 1995.
    Brief description of amendments: In Section 5.2.5 of the Catawba 
Safety Evaluation Report (SER, NUREG-0954), the NRC staff identified 
that the air particulate monitors (EMF38, at both Units 1 and 2), are 
designed to seismic Category I requirements. A recent engineering 
review by the licensee determined that documentation did not exist to 
show these monitors are designed to seismic Category I requirements. In 
a submittal dated September 8, 1994, the licensee proposed a technical 
justification for not requiring the subject monitors to be 

[[Page 1643]]
seismic Category I, and by letter dated September 5, 1995, provided 
additional justification and requested amendments to the licenses for 
both Units 1 and 2. The NRC staff has reviewed the licensee's 
justification and concludes that the containment air particulate 
monitors at Catawba do not have to meet seismic Category I 
requirements. The bases for this conclusion are included in the NRC 
staff's Safety Evaluation.
    Date of issuance: December 29, 1995.
    Effective date: As of the date of issuance to be implemented within 
30 days.
    Amendment Nos.: Unit 1--140; Unit 2--134.
    Facility Operating License Nos. NPF-35 and NPF-52: Amendments 
revised the Updated Final Safety Analysis Report.
    Date of initial notice in Federal Register: November 28, 1995 (60 
FR 58690).
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated December 29, 1995 and an Environmental 
Assessment dated December 22, 1995.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: York County Library, 138 East 
Black Street, Rock Hill, South Carolina 29730.

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

    Date of application for amendments: September 1, 1995, as 
supplemented by letters dated October 17 and November 15, 1995.
    Brief description of amendments: The requested changes would revise 
Technical Specification (TS) 6.9.1.9 to include references to updated 
or recently approved methodologies used to calculate cycle-specific 
limits contained in the Core Operating Limits Report (COLR). The 
subject references have previously been reviewed and approved by the 
NRC staff.
    Date of issuance: December 19, 1995.
    Effective date: As of the date of issuance to be implemented within 
30 days.
    Amendment Nos.: Unit 1--160; Unit 2--142.
    Facility Operating License Nos. NPF-9 and NPF-17: Amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: October 25, 1995 (60 FR 
54718).
    The October 17 and November 15, 1995, letters provided clarifying 
information that did not change the scope of the September 1, 1995, 
application and the initial proposed no significant hazards 
consideration determination.
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated December 19, 1995.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: Atkins Library, University of 
North Carolina, Charlotte (UNCC Station), North Carolina 28223.

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

    Date of application for amendments: January 12, 1995, as 
supplemented by letter dated June 29, 1995.
    Brief description of amendments: The amendments would revise and 
clarify portions of Technical Specification Section 6.0, 
``Administrative Controls.''
    Date of issuance: December 19, 1995.
    Effective date: As of the date of issuance to be implemented within 
30 days.
    Amendment Nos.: Unit 1--161; Unit 2--143.
    Facility Operating License Nos. NPF-9 and NPF-17: Amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: March 15, 1995 (60 FR 
14018).
    The June 29, 1995, letter provided clarifying information that did 
not change the scope of the January 12, 1995, application and the 
initial proposed no significant hazards consideration determination. 
The Commission's related evaluation of the amendments is contained in a 
Safety Evaluation dated December 19, 1995.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: Atkins Library, University of 
North Carolina, Charlotte (UNCC Station), North Carolina 28223.

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 application of amendments: July 26, 1995, as supplemented 
by letter dated November 20, 1995.
    Brief description of amendments: The amendments add a footnote to 
Technical Specification 3.7.8 to provide for a one-time extension of 
the allowable outage time from 72 hours to 7 days for the Oconee 
overhead emergency power path to be inoperable, so that proposed 
modifications to the degraded grid protection system and the external 
grid trouble protection system may be performed.
    Date of Issuance: December 27, 1995.
    Effective date: As of the date of issuance to be implemented within 
30 days from the date of issuance.
    Amendment Nos.: Unit 1--213; Unit 2--213; Unit 3--210.
    Facility Operating License Nos. DPR-38, DPR-47, and DPR-55: The 
amendments revised the Technical Specifications.
    Date of initial notice in Federal Register: August 16, 1995 (60 FR 
42601).
    The November 20, 1995, letter provided clarifying information that 
did not change the scope of the July 26, 1995, application and the 
proposed no significant hazards consideration determination. The 
Commission's related evaluation of the amendments is contained in a 
Safety Evaluation dated December 27, 1995.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: Oconee County Library, 501 
West South Broad Street, Walhalla, South Carolina 29691.

Entergy Operations, Inc., Docket No. 50-368, Arkansas Nuclear One, Unit 
No. 2, Pope County, Arkansas.

    Date of application for amendment: July 19, 1995.
    Brief description of amendment: The amendment reduced the 
requirements associated with the exercise frequency of control element 
assemblies from once per 31 days to once per 92 days.
    Date of issuance: December 22, 1995.
    Effective date: December 22, 1995, to be implemented within 30 
days.
    Amendment No.: 173.
    Facility Operating License No. NPF-6. Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: October 11, 1995 (60 FR 
52929).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated December 22, 1995.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: Tomlinson Library, Arkansas 
Tech University, Russellville, AR 72801.

Entergy Operations, Inc., Docket No. 50-368, Arkansas Nuclear One, Unit 
No. 2, Pope County, Arkansas.

    Date of application for amendment: April 4, 1995.
    Brief description of amendment: The amendment revises surveillance 

[[Page 1644]]
    requirements associated with the main turbine steam valves.
    Date of issuance: December 22, 1995.
    Effective date: December 22, 1995, to be implemented within 30 
days.
    Amendment No.: 174.
    Facility Operating License No. NPF-6. Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: July 5, 1995 (60 FR 
35069).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated December 22, 1995.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: Tomlinson Library, Arkansas 
Tech University, Russellville, AR 72801.

Florida Power and Light Company, Docket Nos. 50-250 and 50-251, Turkey 
Point Plant Units 3 and 4, Dade County, Florida.

    Date of application for amendments: September 11, 1995, as 
supplemented by letter dated November 22, 1995.
    Brief description of amendments: These amendments revise the 
emergency diesel generator testing requirements to incorporate the 
recommendations of Generic Letters 93-05 and 94-01.
    Date of issuance: December 28, 1995.
    Effective date: December 28, 1995.
    Amendment Nos. 181 and 175.
    Facility Operating Licenses Nos. DPR-31 and DPR-41: Amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: October 11, 1995 (60 FR 
52930).
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated December 28, 1995.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: Florida International 
University, University Park, Miami, Florida 33199.

Georgia Power Company, Oglethorpe Power Corporation, Municipal Electric 
Authority of Georgia, City of Dalton, Georgia, Docket Nos. 50-321 and 
50-366, Edwin I. Hatch Nuclear Plant, Units 1 and 2, Appling County, 
Georgia.

    Date of application for amendments: December 2, 1994.
    Brief description of amendments: The amendments replace Appendix B, 
``Environmental Technical Specifications,'' with an Environmental 
Protection Plan (Nonradiological) and revise the Operating Licenses to 
reflect these changes.
    Date of issuance: December 19, 1995.
    Effective date: As of the date of issuance to be implemented within 
30 days.
    Amendment Nos.: Unit 1--199; Unit 2--140.
    Facility Operating License Nos. DPR-57 and NPF-5. Amendments 
revised the Technical Specifications and Operating Licenses.
    Date of initial notice in Federal Register: January 4, 1995 (60 FR 
502).
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated December 19, 1995.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: Appling County Public Library, 
301 City Hall Drive, Baxley, Georgia 31513.

Indiana Michigan Power Company, Docket No. 50-315, Donald C. Cook, 
Nuclear Plant, Unit No. 1, Berrien County, Michigan.

    Date of application for amendment: April 13, 1995, as supplemented 
August 28 and October 27, 1995.
    Brief description of amendment: The amendment modifies the 
Technical Specifications to allow use of laser-welded sleeves to repair 
defective steam generator tubes.
    Date of issuance: January 4, 1996.
    Effective date: January 4, 1996, with full implementation within 45 
days.
    Amendment No.: 205.
    Facility Operating License No. DPR-58. Amendment revises the 
Technical Specifications.
    Date of initial notice in Federal Register: June 6, 1995 (60 FR 
29877).
    The August 28 and October 27, 1995, supplements provided clarifying 
information and updated Technical Specification pages. These 
supplements did not change the proposed no significant hazards 
considerations determination.
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated January 4, 1996.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: Maud Preston Palenske Memorial 
Library, 500 Market Street, St. Joseph, Michigan 49085.

Northeast Nuclear Energy Company, Docket No. 50-245, Millstone Nuclear 
Power Station, Unit 1, New London County, Connecticut.

    Date of application for amendment: August 31, 1995, as supplemented 
December 5, 1995.
    Brief description of amendment: The amendment modifies the 
definition of HOT SHUTDOWN and COLD SHUTDOWN to specify that the 
definitions are not applicable during the performance of an inservice 
hydrostatic and leak test (IHLT). Technical Specification Section 3.6.B 
and 4.6.B is modified by adding Section 3.6.B.1.b and 4.6.B.1.b to 
identify the requirements that must be satisfied to consider the 
reactor in COLD SHUTDOWN during the performance of an IHLT. In 
addition, the amendment changes temperature specific requirements on 
several pages to mode or condition specific requirements; makes several 
editorial changes; and changes the associated Bases.
    Date of issuance: December 29, 1995.
    Effective date: As of the date of issuance, to be implemented 
within 30 days.
    Amendment No.: 90.
    Facility Operating License No. DPR-21. Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: September 27, 1995 (60 
FR 49940).
    The December 5, 1995, 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 December 29, 1995.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: Learning Resources Center, 
Three Rivers Community-Technical College, 574 New London Turnpike, 
Norwich, CT 06360.

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

    Date of application for amendment: May 1, 1995.
    Brief description of amendment: The amendment revises the Technical 
Specifications to extend the interval for performance of selected 
surveillances to accommodate a 24-month fuel cycle. Specifically, this 
amendment changes the definition for a refueling interval, changes the 
BASES for surveillances that are performed at least once each fuel 
cycle and changes the surveillance frequencies for:
    (1) The flow path tests of the boron injection system,
    (2) The operability tests of the digital rod position indicatiors,
    (3) The drop time of the full-length shutdown and control rods,
    (4) The channel calibration of the loose-part detection system, 
    
[[Page 1645]]

    (5) The channel calibration of the seismic monitoring 
instrumentation,
    (6) The activation of the pumps and the flow path tests of the 
valves in the containment quench and recirculation spray systems and
    (7) The tests of the intended actuation positions of the 
containment isolation valves.
    Date of issuance: December 28, 1995.
    Effective date: As of the date of issuance, to be implemented 
within 90 days.
    Amendment No.: 122.
    Facility Operating License No. NPF-49. Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: November 27, 1995 (60 
FR 58402).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated December 28, 1995.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: Learning Resources Center, 
Three Rivers Community-Technical College, 574 New London Turnpike, 
Norwich, CT 06360.

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

    Date of application for amendment: July 17, 1995.
    Brief description of amendment: The amendment revises the Technical 
Specifications pertaining to the plant air filtration and ventilation 
systems to extend the surveillance frequencies that are now required to 
be performed at least once per 18 months to specify that the 
surveillances are to be performed at least once each refueling 
interval.
    Date of issuance: December 28, 1995.
    Effective date: As of the date of issuance, to be implemented 
within 90 days.
    Amendment No.: 123.
    Facility Operating License No. NPF-49. Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: November 27, 1995 (60 
FR 58402).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated December 28, 1995.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: Learning Resources Center, 
Three Rivers Community-Technical College, 574 New London Turnpike, 
Norwich, CT 06360.

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

    Date of application for amendment: July 14, 1995.
    Brief description of amendment: The amendment revises the frequency 
of those surveillance requirements for the emergency core cooling 
systems that now require that the surveillances be performed ``at least 
once per 18 months'' to specify that the surveillances be performed 
``at least once each refueling interval.''
    Date of issuance: December 28, 1995.
    Effective date: As of the date of issuance, to be implemented 
within 90 days.
    Amendment No.: 124.
    Facility Operating License No. NPF-49. Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: November 27, 1995 (60 
FR 58402).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated December 28, 1995.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: Learning Resources Center, 
Three Rivers Community-Technical College, 574 New London Turnpike, 
Norwich, CT 06360.

Pacific Gas and Electric Company, Docket Nos. 50-275 and 50-323, Diablo 
Canyon Nuclear Power Plant, Unit Nos. 1 and 2, San Luis Obispo County, 
California.

    Date of application for amendments: September 29, 1995.
    Brief description of amendments: The amendments added a one-time 
footnote to the Technical Specifications related to the diesel 
generator fuel oil storage and transfer system to permit each of the 
existing storage tanks to be removed from service for up to 60 days so 
they can be replaced with double walled tanks and piping that comply 
with new California regulations.
    Date of issuance: January 3, 1996.
    Effective date: January 3, 1996, to be implemented within 30 days 
of issuance.
    Amendment Nos.: Unit 1--Amendment No. 109; Unit 2--Amendment No. 
108.
    Facility Operating License Nos. DPR-80 and DPR-82: The amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: November 27, 1995 (60 
FR 58403).
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated January 3, 1996.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: California Polytechnic State 
University, Robert E. Kennedy Library, Government Documents and Maps 
Department, San Luis Obispo, California 93407.

Pacific Gas and Electric Company, Docket No. 50-133, Humboldt Bay Power 
Plant, Unit 3, Humboldt County, California.

    Date of application for amendment: October 8, 1993, as supplemented 
October 28, 1994.
    Brief description of amendment: This amendment revised the 
Technical Specification by deleting Figure II-2, ``Restricted Area Per 
10 CFR 20.3(a)(14)'' and by deleting the restricted area boundary line 
from Figure V-3, ``HBPP Groundwater Monitoring System Wells.''
    Date of issuance: December 21, 1995.
    Effective date: This license amendment is effective as of the date 
of its issuance and must be fully implemented no later than 30 days 
from the date of issuance.
    Amendment No.: 30.
    Facility License No. DPR-7: This amendment revised the TS.
    Date of initial notice in Federal Register: January 5, 1994 (59 FR 
624).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated December 21, 1995.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: Humboldt County Library, 1313 
3rd Street, Eureka, California 95501.

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

    Date of application for amendments: March 31, 1995.
    Brief description of amendments: The amendments incorporate a 
change in the Station Technical Specifications for both units that 
modifies the requirement in TS 4.4.4.3.a to have the pH of the reactor 
coolant measured every 72 hours. The amendments add the clarification 
that the pH measurement will be performed only when the coolant 
conductivity is greater than 1.0 micro-mho/cm at 25 deg.C ( deg.77).
    Date of issuance: January 3, 1996.
    
[[Page 1646]]

    Effective date: Both units, as of date of issuance and are to be 
implemented within 30 days.
    Amendment Nos.: 156 and 127.
    Facility Operating License Nos. NPF-14 and NPF-22. The amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: April 26, 1995 (60 FR 
20522).
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated January 3, 1996.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: Osterhout Free Library, 
Reference Department, 71 South Franklin Street, Wilkes-Barre, 
Pennsylvania 18701.
    Pennsylvania Power and Light Company, Docket No. 50-388, 
Susquehanna Steam Electric Station, Unit 2, Luzerne County, 
Pennsylvania.
    Date of application for amendment: August 11, 1995.
    Brief description of amendment: The amendment revises the Unit 2 
Technical Specifications (TSs) to reestablish the original operability 
requirements for the Neutron Flux function, and to delete the footnote 
that was added to TS page 3/4 3-71 under Amendment No. 115, regarding 
the length of time that the revised operability values were valid.
    Date of issuance: January 3, 1996.
    Effective date: As of date of issuance, to be implemented within 30 
days.
    Amendment No.: 128.
    Facility Operating License No. NPF-22. This amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: September 13, 1995 (60 
FR 47623).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated January 3, 1996.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: Osterhout Free Library, 
Reference Department, 71 South Franklin Street, Wilkes-Barre, 
Pennsylvania 18701.

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

    Date of application for amendment: May 12, 1995.
    Brief description of amendment: The amendment modifies the 
Technical Specifications (TSs) to extend the surveillance test 
intervals for the emergency service water system to support 24-month 
operating cycles. Surveillance test interval extensions are denoted as 
being performed ``every 24 months'' or ``at least once per 24 months'' 
consistent with the guidance provided in Generic Letter (GL) 91-04, 
``Changes in Technical Specification Surveillance Intervals to 
Accommodate 24-Month Fuel Cycle,'' dated April 2, 1991. The NRC staff 
has determined that the proposed TS changes are in accordance with GL 
91-04, and are therefore acceptable.
    Date of issuance: December 21, 1995.
    Effective date: As of the date of issuance to be implemented within 
30 days.
    Amendment No.: 230.
    Facility Operating License No. DPR-59: Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: September 13, 1995 (60 
FR 47623)
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated December 21, 1995.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: Reference and Documents 
Department, Penfield Library, State University of New York, Oswego, New 
York 13126.

South Carolina Electric & Gas Company, South Carolina Public Service 
Authority, Docket No. 50-395, Virgil C. Summer Nuclear Station, Unit 
No. 1, Fairfield County, South Carolina.

    Date of application for amendment: February 21, 1995, as 
supplemented on August 31, 1995, and December 4, 1995.
    Brief description of amendment: The amendment revises the Technical 
Specifications (TS) support of the licensee's plan to implement the 
revised 10 CFR Part 20, ``Standards for Protection Against Radiation.'' 
Also, several editorial changes to improve the clarity of the TS were 
made.
    Date of issuance: December 28, 1995.
    Effective date: 90 days after issuance.
    Amendment No.: 130.
    Facility Operating License No. NPF-12. Amendment revises the 
operating license.
    Date of initial notice in Federal Register: March 29, 1995 (60 FR 
16200). Renoticed on September 27, 1995 (60 FR 49946) due to changes in 
the licensee's proposed no significant hazards consideration analysis 
that were included in the August 31, 1995 supplemental letter. The 
December 4, 1995 letter provided supplemental information that did not 
change the second proposed no significant hazards consideration. The 
Commission's related evaluation of the amendment is contained in a 
Safety Evaluation dated December 28, 1995.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: Fairfield County Library, 300 
Washington Street, Winnsboro, SC 29180.

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

    Date of application for amendment: June 21, 1994, as supplemented 
by letter dated October 23, 1995.
    Brief description of amendment: The amendment revises Technical 
Specification (TS) 6.5.1, 6.5.2 and 6.5.3 to relocate the review and 
audit requirements of the On-site Review Committee (ORC) and the 
Nuclear Safety Review Board (NSRB) to the Operational Quality Assurance 
Manual (OQAM). In addition, the amendment deletes reference to the 
Manager, Nuclear Safety and Emergency Preparedness, in TS 6.2.3. The 
Index is revised to reflect the relocations.
    Date of issuance: December 26, 1995.
    Effective date: December 26, 1995, to be implemented within 30 days 
from the date of issuance.
    Amendment No.: 107.
    Facility Operating License No. NPF-30. The amendment revises the 
Technical Specifications.
    Date of initial notice in Federal Register: August 31, 1994 (59 FR 
45036) and November 27, 1995 (60 FR 58406). The Commission's related 
evaluation of the amendment is contained in a Safety Evaluation dated 
December 26, 1995.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: Callaway County Public 
Library, 710 Court Street, Fulton, Missouri 65251.

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

    Date of application for amendments: July 20, 1995.
    Brief description of amendments: These amendments establish a new 
setpoint for the steam generator high-high level and provide more 
restrictive setting limits for certain reactor protection system/
engineered safety features actuation system setpoints.
    Date of issuance: December 28, 1995.
    Effective date: December 28, 1995.
    Amendment Nos. 206 and 206.
    Facility Operating License Nos. DPR-32 and DPR-37: Amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: August 30, 1995 (60 FR 
45190).

[[Page 1647]]

    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated December 28, 1995.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: Swem Library, College of 
William and Mary, Williamsburg, Virginia 23185.

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

    Date of application for amendment: September 19, 1995.
    Brief description of amendment: The amendment makes administrative 
changes to the KNPP Technical Specifications (TS) to improve their 
clarity and consistency. The amendment includes changes to reflect 
revisions to 10 CFR Part 20, and changes to correct minor typographical 
and format inconsistencies as part of the licensee's ongoing effort to 
convert the TS to the WordPerfect format.
    Date of issuance: December 21, 1995.
    Effective date: December 21, 1995.
    Amendment No.: 122.
    Facility Operating License No. DPR-43: Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: October 11, 1995 (60 FR 
52936).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated December 21, 1995.
    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 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 application for amendments: April 27, 1995, as supplemented 
by letter dated November 29, 1995.
    Brief description of amendments: The amendments revise TS Table 
15.3.5-1, ``Engineered Safety Features Initiation Instrument Setting 
Limits,'' and TS Table 15.3.5-3, ``Engineered Safety Features.'' 
Setting limits are modified and references are changed. The bases for 
TS Section 15.3.5, ``Instrumentation System,'' are also changed to be 
consistent with the TS changes.
    Date of issuance: December 27, 1995.
    Effective date: December 27, 1995.
    Amendment Nos.: 167 and 171.
    Facility Operating License Nos. DPR-24 and DPR-27: Amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: May 23, 1995 (60 FR 
27346). The November 29, 1995, submittal provided supplemental 
information which did not change the initial no significant hazards 
consideration determination. The Commission's related evaluation of the 
amendments is contained in a Safety Evaluation dated December 27, 1995.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: Joseph P. Mann Library, 1516 
Sixteenth Street, Two Rivers, Wisconsin 54241.

    Dated at Rockville, Maryland, this 11th day January 1996.

    For the Nuclear Regulatory Commission.
Jack W. Roe,
Director, Division of Reactor Projects--III/IV, Office of Nuclear 
Reactor Regulation.
[FR Doc. 96-676 Filed 1-19-96; 8:45 am]
BILLING CODE 7590-01-P 11