[Federal Register Volume 60, Number 244 (Wednesday, December 20, 1995)]
[Notices]
[Pages 65671-65694]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: X95-21220]



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[[Page 65672]]



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 November 27, 1995, through December 8, 1995. 
The last biweekly notice was published on December 6, 1995 (60 FR 
62485).

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 January 19, 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 

[[Page 65673]]
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: November 7, 1995
    Description of amendments request: The proposed amendment would 
adopt the improved Standard Technical Specifications (NUREG-1432) 
format and content of Section 5.0, ``Design Features,'' as modified by 
approved changes to the improved Standard Technical Specifications.
    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 amendment does not change the Design Features, only 
relocates the information to other documents. This is consistent 
with the NRC Policy Statement and NUREG-1432. Therefore, relocating 
existing information, eliminating information which duplicates 
information found in other licensee documents, and making 
administrative improvements provide Technical Specifications which 
are easier to use. Because information is relocated to established 
programs where changes to those programs are controlled by 
regulatory requirements, there is no reduction in commitment and 
adequate control is still maintained. Likewise, the elimination of 
information which duplicates information in other licensee 
documents, enhances the useability of the Technical Specifications 
without reducing commitments. The administrative improvements being 
proposed neither add nor delete requirements, but merely clarify and 
improve the understanding and readability of the Technical 
Specifications. Since the requirements remain the same, these 
changes only affect the method of presentation and are considered 
administrative, and as such, would not affect possible initiating 
events for accidents previously evaluated or any system functional 
requirement.
    Therefore, the proposed changes would 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 relocation of existing requirements, the elimination of 
requirements which duplicate existing information, and making 
administrative improvements are all changes that are administrative 
in nature. The proposed changes will not affect any plant system or 
structure, nor will they affect any system functional or operability 
requirements. Consequently, no new failure modes are introduced as a 
result of the proposed changes. The proposed changes are consistent 
with the improved Standard Technical Specifications, for the most 
part, as plant specific information is included in this section. 
Therefore, the proposed change would not create the possibility of a 
new or different type 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 are administrative in nature in that no 
change[s] to the design features of the facility are being made. The 
Design Features Section is being reformatted to be consistent, for 
the most part, with NUREG-1432, ``Standard Technical Specifications, 
Combustion Engineering Plants,'' Revision 1. The proposed changes do 
not affect the UFSAR design bases, accident assumptions, or 
Technical Specification Bases. In addition, the proposed changes do 
not affect release limits, monitoring equipment, or practices. 
Consequently, the proposed changes would not involve a significant 
reduction in a 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

Connecticut Yankee Atomic Power Company, Docket No. 50-213, Haddam 
Neck Plant, Middlesex County, Connecticut

    Date of amendment request: October 20, 1995
    Description of amendment request: The proposed amendment would 
revise the Electrical Power Systems Surveillance Intervals from 18 
months to once per refueling (i.e., nominal 24 months).
    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 

[[Page 65674]]
below. The no significant hazards consideration analysis has been 
divided into three parts: AC Sources Operating, DC Sources Operating, 
and On-Site Power distribution:

    In accordance with 10CFR50.92, CYAPCO has reviewed the proposed 
changes and concluded that they do not involve an SHC. The basis for 
this conclusion is that the three criteria of 10CFR50.92(c) are not 
compromised. The proposed changes do not involve an SHC because the 
changes would not:
    AC Sources Operating
    1. Involve a significant increase in the probability or 
consequences of an accident previously evaluated.
    The proposed change will increase the interval between a 
surveillance that is performed during plant shutdown from once per 
18 months to a maximum of once per 30 months (i.e., 24 months 
nominal + 25% as allowed by Specification 4.0.2). The proposed 
change to Surveillance Requirement 4.8.1.1.2.f does not alter the 
intent or the method by which the surveillance is conducted. In 
addition, the acceptance criterion for the surveillance is 
unchanged. As such, the proposed change will not degrade the ability 
of the EDG [emergency diesel generator] to perform its intended 
function.
    A review of the past surveillances, and preventive maintenance 
of the diesel generators indicates that the appropriate acceptance 
criterion was met in each case. Additional assurance of the diesel 
generator's operability is provided by Surveillance Requirement 
4.8.1.1.2.a.4 and the performance of other on-line testing as 
described above. As such, the proposed changes do not adversely 
affect the probability of an accident previously analyzed.
    2. Create the possibility of a new or different kind of accident 
than any accident previously evaluated.
    The proposed change regarding the testing frequency of the 
diesel generators [i.e., from once per 18 months to a maximum of 
once per 30 months (i.e., 24 months + 25 percent as allowed by 
Specification 4.0.2)] does not affect the operation or response of 
any plant equipment, including the diesel generators, or introduce 
any new failure mechanism. The proposed change does not affect the 
test acceptance criteria of the EDGs. The plant equipment will 
respond per design and analyses, and there will not be a malfunction 
of a new or different type introduced by the testing frequency 
revision to the EDG surveillance requirements. As such, the changes 
do not create the possibility of a new or different kind of accident 
from any previously analyzed.
    3. Involve a significant reduction in a margin of safety.
    The Bases Section of Technical Specification Section 3/4.8, 
``Electrical Power Systems,'' states that the operability of the AC 
and DC power systems and associated distribution systems ensure that 
sufficient power will be available to supply the safety-related 
equipment required for safe shutdown and mitigation and control of 
accident conditions. Bases Section 3/4.8 also states that the 
surveillance requirements for determining the operability of the 
EDGs are in accordance with the recommendations of Regulatory Guide 
1.108, Revision 1. The revision of surveillance requirements will 
continue to verify that the EDGs are operable. Operable EDGs ensure 
that the assumptions in the Bases of the Technical Specifications 
are not affected and ensure that the margin of safety is not 
reduced. Therefore, the assumptions in the Bases of the Technical 
Specifications are not affected and the change does not result in a 
significant reduction in the margin of safety.
    DC Sources Operating
    1. Involve a significant increase in the probability or 
consequences of an accident previously evaluated.
    CYAPCO is proposing to modify the frequency of Surveillance 
Requirements 4.8.2.1.c, d, and f of the Haddam Neck Plant Technical 
Specifications from at least once per 18 months to at least once 
each refueling interval. These surveillance requirements verify the 
operability of components of the Class 1E DC power system. CYAPCO is 
also proposing to delete the term ``during shutdown'' contained in 
Surveillance Requirements 4.8.2.1.d, 4.8.2.1.e, and 4.8.2.1.f.
    Additional assurance of the operability of the Class 1E DC power 
system is provided by Surveillance Requirements 4.8.2.1.a, b, and e.
    The proposed changes do not alter the intent or method by which 
the surveillances are conducted, do not involve any physical changes 
to the plant, do not alter the way any structure, system, or 
component functions, and do not modify the manner in which the plant 
is operated. As such, the proposed changes in the frequency of 
Surveillance Requirements 4.8.2.1.c, d, and f will not degrade the 
ability of the Class 1E DC power system to perform its intended 
safety function. Also, the Class 1E DC power system is designed to 
perform its intended safety function even in the event of a single 
failure.
    Equipment performance over the last four operating cycles was 
evaluated to determine the impact of extending the frequency of 
Surveillance Requirements 4.8.2.1.c, d and f. cThis evaluation 
included a review of surveillance results, preventive maintenance 
associated with normal surveillance activities, and corrective 
maintenance records. It concluded that the Class 1E DC power system 
is highly reliable, and that there is no indication that the 
proposed extension could cause deterioration in the condition or 
performance of any of the subject Class 1E DC power system 
components.
    The deletion of the phrase ``during shutdown'' in Surveillance 
Requirement 4.8.2.1.d, e, and f is acceptable. The terms ``Cold 
Shutdown'' and ``Hot Shutdown'' are defined in the Haddam Neck Plant 
Technical Specifications as operating modes or conditions. The 
proposed deletion of the term ``during shutdown'' is intended to 
prevent possible misinterpretations and is consistent with the 
recommendations of GL 91-04.
    Based on the above, the proposed changes to Surveillance 
Requirements 4.8.2.1.c, d, e, and f of the Haddam Neck Plant 
Technical Specifications do not involve a significant increase in 
the probability or consequences of an accident previously analyzed.
    2. Create the possibility of a new or different kind of accident 
than any accident previously evaluated.
    CYAPCO is proposing to modify the frequency of Surveillance 
Requirements 4.8.2.1.c, d, and f of the Haddam Neck Plant Technical 
Specifications from at least once per 18 months to at least once 
each refueling interval. CYAPCO is also proposing to delete the term 
``during shutdown'' contained in Surveillance Requirements 
4.8.2.1.d, 4.8.2.1.e, and 4.8.2.1.f. These surveillance requirements 
verify the operability of components of the Class 1E DC power 
system.
    The proposed changes do not alter the intent or method by which 
the surveillances are conducted, do not involve any physical changes 
to the plant, do not alter the way any structure, system, or 
component functions, and do not modify the manner in which the plant 
is operated. As such, the proposed changes to Surveillance 
Requirements 4.8.2.1.c, d, e, and f will not introduce a new failure 
mode.
    Based on the above, the proposed changes to Surveillance 
Requirements 4.8.2.1.c, d, e, and f of the Haddam Neck Plant 
Technical Specifications will not create the possibility of a new or 
different kind of accident from any previously evaluated.
    3. Involve a significant reduction in a margin of safety.
    CYAPCO is proposing to modify the frequency of Surveillance 
Requirements 4.8.2.1.c, d, and f of the Haddam Neck Plant Technical 
Specifications from at least once per 18 months to at least once 
each refueling interval. CYAPCO is also proposing to delete the term 
``during shutdown'' contained in Surveillance Requirements 
4.8.2.1.d, 4.8.2.1.e, and 4.8.2.1.f. These surveillance requirements 
verify the operability of components of the Class 1E DC power 
system.
    Equipment performance over the last four operating cycles was 
evaluated to determine the impact of extending the frequency of 
Surveillance Requirements 4.8.2.1.c, d and f. This evaluation 
included a review of surveillance results, preventive maintenance 
associated with normal surveillance activities, and corrective 
maintenance records. It concluded that the Class 1E DC power system 
is highly reliable, and that there is no indication that the 
proposed extension could cause deterioration in the condition or 
performance of any of the subject Class 1E DC power system 
components.
    Additional assurance of the operability of the Class 1E DC power 
system is provided by Surveillance Requirements 4.8.2.1.a, b, and e.
    Since decreasing the surveillance frequency does not involve a 
significant increase in the consequences of a design basis accident 
previously analyzed, the proposed changes to Surveillance 
Requirements 4.8.2.1.c, d, e, and f of the Haddam Neck Plant 
Technical Specifications do not involve a significant reduction in 
the margin of safety.

[[Page 65675]]

    On-Site Power Distribution
    1. Involve a significant increase in the probability or 
consequences of an accident previously evaluated.
    The proposed change to Surveillance Requirement 4.8.3.1.2 will 
increase the surveillance interval from once each refueling outage 
(once per 18 months) to a maximum of once per 30 months (i.e., 24 
months nominal + 25% as allowed by Specification 4.0.2.). The 
proposed change to Surveillance Requirement 4.8.3.1.2 does not alter 
the intent or the method by which the surveillance is conducted. In 
addition, the acceptance criterion for the surveillance is 
unchanged. As such, the proposed changes will not degrade the 
ability of the MCC-5 ABT scheme to perform its intended function.
    The successful past surveillance results, and the simpler re-
design of the MCC-5 ABT provide assurance of system operability up 
to a maximum of 30 months. As such, the proposed changes do not 
adversely affect the probability or consequences of an accident 
previously analyzed.
    2. Create the possibility of a new or different kind of accident 
than any accident previously evaluated.
    The proposed change does not alter the intent or method by which 
the surveillance is conducted, does not involve any physical changes 
to the plant, does not alter the way any structure, system, or 
component functions, and does not modify the manner in which the 
plant is operated. As such, the proposed change to Surveillance 
Requirement 4.8.3.1.2 will not introduce a new failure mode.
    Based on the above, the proposed change to Surveillance 
Requirement 4.8.3.1.3 of the Haddam Neck Plant Technical 
Specifications will not create the possibility of a new or different 
kind of accident from any previously evaluated.
    3. Involve a significant reduction in a margin of safety.
    The proposed change to Surveillance Requirement 4.8.3.1.2 
extends the frequency for verifying the operability of the MCC-5 ABT 
scheme from at least once per 18 months to at least once per 
refueling interval (i.e., 24 months nominal + 25% as allowed by 
Specification 4.0.2).
    The proposed change does not alter the intent or method by which 
the surveillance is conducted, does not involve any physical changes 
to the plant, does not alter the way any structure, system, or 
component functions, and does not modify the manner in which the 
plant is operated. As such, the proposed change in the frequency of 
Surveillance Requirement 4.8.3.1.2 will not degrade the ability of 
the MCC-5 ABT to perform its safety function and 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: Russell Library, 123 Broad 
Street, Middletown, CT 06457.
    Attorney for licensee: Lillian M. Cuoco, Esq., Senior Nuclear 
Counsel, Northeast Utilities Service Company, P.O. Box 270, Hartford, 
CT 06141-0270.
    NRC Project Director: Phillip F. McKee

Connecticut Yankee Atomic Power Company, Docket No. 50-213, Haddam 
Neck Plant, Middlesex County, Connecticut

    Date of amendment request: October 27, 1995
    Description of amendment request: The proposed amendment will 
revise Technical Specification (TS) Section 3.6.3, ``Containment 
Isolation Valves.'' These changes will clarify the action statement for 
when a penetration has only one containment isolation valve (CIV) and 
that valve is inoperable.
    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 (SHC), which is presented below:

    ...The proposed change does not involve an SHC because the 
change would not:
    1. Involve a significant increase in the probability or 
consequences of an accident previously evaluated.
    The containment isolation system is an engineered safety feature 
that functions to allow normal or emergency passage of fluids 
through the containment boundary, while preserving the ability of 
the boundary to prevent or limit the escape of fission products that 
may result from postulated accidents.
    All fluid system pipelines that penetrate the containment are 
provided with one or more valves that can be closed remotely, either 
electrically or pneumatically, or are locked manual valves. Most of 
the piping penetrations connect to equipment inside the reactor 
containment. Thus, they are not open to the reactor containment 
atmosphere and will not pass radioactive contamination to the CIV 
unless the pipe is ruptured inside containment during an accident.
    Lines that penetrate the reactor containment and are not in 
service during operation are isolated with one or more locked closed 
CIVs. Lines that are in service and that pass fluids during 
operation are provided with one or more motor-operated valves, 
positive closure trip valves, or check-valves.
    The lack of guidance contained in Technical Specification 
Section 3.6.3 for a penetration that has only one CIV in it, does 
not increase the probability or consequences of an accident 
previously evaluated. This design, and the consequences that could 
result from this configuration have been evaluated previously and 
found acceptable. The proposed modification simply provides guidance 
to the operators should a penetration with only one CIV becomes 
inoperable. This proposed technical specification will, as do other 
technical specification action statements, provide a reasonable time 
to correct the situation before a required shutdown must commence. 
In addition, this proposed Action Statement was developed to be 
consistent with Technical Specification Section 3.0.3.
    2. Create the possibility of a new or different kind of accident 
from any accident previously evaluated.
    The proposed modification provides guidance to the operators 
should a penetration which has only one CIV be inoperable. This 
design has been previously evaluated and found to be acceptable from 
both a deterministic and probabilistic standpoint. The proposed 
modification will provide the operators specific guidance to restore 
the penetration to an operable state or to isolate it. With this 
guidance, they can avert the risk associated with a plant shutdown, 
which would be mandated without this guidance. Should a CIV be 
inoperable and not capable of being restored, the proposed technical 
specification provides additional options. However, a probabilistic 
risk assessment review has determined that these additional options 
are not risk significant. Finally, the containment isolation system 
cannot be an accident initiator, rather it is designed to respond to 
accidents. The inability of the CIVs to operate cannot create a new 
or different kind of accident from any previously evaluated.
    3. Involve a significant reduction in a margin of safety.
    The proposed modification provides the requirement to the 
operators should a penetration which has only one CIV be inoperable. 
The effects of this design have been previously evaluated and found 
to be acceptable from both a deterministic and probabilistic 
standpoint.
    The current Haddam Neck Plant containment isolation system has 
been previously reviewed by the NRC. CYAPCO is not making any 
changes to the containment isolation system. CYAPCO is however, 
providing guidance in the technical specifications should a 
penetration which has only one CIV be inoperable. This guidance will 
allow CYAPCO to correct the event associated with the penetration 
with an NRC approved alternative, in a set time. This provision is 
safe especially when compared to the alternative which is a plant 
shutdown under Technical Specification Section 3.0.3.

    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: Russell Library, 123 Broad 
Street, Middletown, CT 06457.
    Attorney for licensee: Lillian M. Cuoco, Esq., Senior Nuclear 
Counsel, Northeast Utilities Service Company, P.O. Box 270, Hartford, 
CT 06141-0270.

[[Page 65676]]

    NRC Project Director: Phillip F. McKee

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

    Date of amendment request: August 17, 1995
    Description of amendment request: The Commission issued Amendment 
Nos. 128 and 122 to the Facility Operating Licenses for Catawba Units 1 
and 2 on February 17, 1995, which revised Technical Specification (TS) 
Table 2.2-1 and TS Surveillance Requirement (SR) 4.2.5 to allow a 
change in the method for measuring reactor coolant system (RCS) 
flowrate from the calorimetric heat balance method to a method based on 
a one-time calibration of the RCS cold leg elbow differential pressure 
taps. In its application submitted on January 10, 1994, for the above 
listed amendments, Duke Power (the licensee) neglected to modify SR 
4.2.5.2 to delete that portion of the SR that specifies that the 
measurement instrumentation shall be calibrated within 7 days prior to 
the performance of the flowrate measurement. The licensee states that 
the requirement to calibrate the measurement instrumentation within 7 
days prior to the performance of the flowrate measurement is 
impractical based on utilization of the cold leg elbow pressure tap 
method of RCS flowrate measurement. Accordingly, the licensee proposes 
to modify SR 4.2.5.2 to reflect the deletion of the subject 
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:

    Criterion 1
    The requested amendments will not involve a significant increase 
in the probability or consequences of an accident previously 
evaluated. This change is considered administrative in nature and 
should have been requested in Duke Power Company's January 10, 1994 
application, as amended. The instrumentation which was subject to 
the requirement is no longer utilized in the fulfillment of the TS 
required RCS flowrate determination. The proposed changes will not 
result in any impact upon accident probabilities, since the RCS 
flowrate measurement instrumentation is not accident initiating 
equipment. Likewise, they will not result in any impact upon 
accident consequences, since no change to any method or frequency of 
calibration of the RCS flowrate transmitters will result. The plant 
response to accidents will not be affected.
    Criterion 2
    The requested amendments will not create the possibility of a 
new or different kind of accident from any accident previously 
evaluated. No change is being made to any plant design feature, or 
to the manner in which the plant will be operated. Therefore, no new 
accident causal mechanisms can be generated. As noted above, the 
proposed changes are considered administrative in nature, and should 
have been requested in the January 10, 1994 application, as amended.
    Criterion 3
    The requested amendments will not involve a significant 
reduction in a margin of safety. No impact upon any fission product 
barriers will occur as a result of the approval of the proposed 
changes. No change to plant design, operating, maintenance, or test 
characteristics will result from the proposed amendments. No impact 
upon any plant safety margins will result.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: York County Library, 138 East 
Black Street, Rock Hill, South Carolina 29730
    Attorney for licensee: Mr. Albert Carr, Duke Power Company, 422 
South Church Street, Charlotte, North Carolina 28242
    NRC Project Director: Herbert N. Berkow

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

    Date of amendment request: November 15, 1995
    Description of amendment request: The proposed amendments modify 
Technical Specification (TS) 3/4.7.1 and the associated Bases to 
increase the setpoint tolerance of the main steam safety valves (MSSVs) 
from plus or minus one percent to plus or minus three percent, to 
incorporate a requirement to reset as-left MSSV lift settings to within 
plus or minus one percent following surveillance testing, and to delete 
two obsolete footnotes.
    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
    The requested amendments will not involve a significant increase 
in the probability or consequences of an accident previously 
evaluated. As demonstrated previously, all applicable licensing 
basis safety analyses were evaluated with a MSSV setpoint drift of 
plus or minus 3%. The results of the evaluations were within all 
appropriate accident analysis acceptance criteria. No significant 
impact on DNBR results, peak primary or secondary pressures, peak 
fuel cladding temperature, dose, or any other accident analysis 
acceptance criterion was involved. No impact on the probability of 
any accident occurring exists as a result of the increased MSSV 
setpoint tolerance.
    Criterion 2
    The requested amendments will not create the possibility of a 
new or different kind of accident from any accident previously 
evaluated. No change is being made to any plant design feature, or 
to the manner in which the plant will be operated. Therefore, no new 
accident causal mechanisms can be generated. The MSSV setpoint 
tolerance only affects the time at which the valve opens following 
or during a transient, and is not a contributor to the probability 
of an accident.
    Criterion 3
    The requested amendments will not involve a significant 
reduction in a margin of safety. As stated above, all relevant 
accident analyses were examined to determine the effect of the wider 
MSSV setpoint tolerances. All analysis results are within applicable 
acceptance criteria. Finally, the NRC has previously approved TS 
changes for other plants seeking to use the [plus or minus] 3 
[percent] setpoint tolerance, including McGuire Nuclear Station 
(reference Amendment Nos. 146 and 128 for Units 1 and 2, 
respectively).

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: York County Library, 138 East 
Black Street, Rock Hill, South Carolina 29730
    Attorney for licensee: Mr. Albert Carr, Duke Power Company, 422 
South Church Street, Charlotte, North Carolina 28242
    NRC Project Director: Herbert N. Berkow

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

    Date of amendment request: November 15, 1995
    Description of amendment request: The proposed amendments modify 
Technical Specification (TS) Limiting Condition for Operation 3.7.5 to 
raise the minimum nuclear service water system's (RN) water level in 
the standby nuclear service water pond (SNSWP) from 570 to 571 feet 
mean sea level. 

[[Page 65677]]
This change will increase the volume of water that will be available 
for use of the SNSWP as the ultimate heat sink for postulated accidents 
under all meteorological conditions.
    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
    The requested amendments will not involve a significant increase 
in the probabililty or consequences of an accident previously 
evaluated. The proposed amendments will have no impact upon any 
accident probabilities, since the RN system is not a accident 
initiating system. It is an accident mitigating system. Accident 
consequences will not be affected, since the proposed amendments 
will require a greater surface area for heat transfer from the SNSWP 
water to the environment. It has been determined that with the 
required TS minimum water level of 571 feet and with the required TS 
temperature limit of 91.5F [degrees Farenheit], the SNSWP will be 
capable of fulfilling all design basis requirements pertaining to 
accident mitigation.
    Criterion 2
    The requested amendments will not create the possibility of a 
new or different kind of accident from any accident previously 
evaluated. As stated previously, the RN system is not an accident 
initiator. No change is being made to the plant which would cause 
the RN system to become an accident initiator. All relevant 
procedures will be changed as required, commensurate with the NRC 
issuance of the requested amendments. No accident causal mechanisms 
will be affected. The effect of the increased SNSWP level on the 
SNSWP dam was evaluated and found to be negligible.
    Criterion 3
    The requested amendments will not involve a significant 
reduction in a margin of safety. As noted above, the SNSWP was 
evaluatd with the new TS level requirement and was determined to be 
operable and capable of meeting all design basis requirements. No 
impact on any fission product barriers is created by the proposed 
changes. The proposed changes will ensure that the RN system remains 
capable of fulfilling its required accident mitigating functions. 
SNSWP temperature will continue to be monitored at an elevation of 
568 feet, which is considered to be the highest elevation at which 
the average SNSWP surface temperature is accurately represented and 
minimally influenced by daily temperature swings due to variations 
in solar heat input, air temperature, and rainfall temperature.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: York County Library, 138 East 
Black Street, Rock Hill, South Carolina 29730
    Attorney for licensee: Mr. Albert Carr, Duke Power Company, 422 
South Church Street, Charlotte, North Carolina 28242
    NRC Project Director: Herbert N. Berkow

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

    Date of amendment request: August 20, 1992, as supplemented 
December 5, 1995.
    Description of amendment request: The proposed amendments, would 
revise the Technical Specifications (TS) related to the 60-month 120-
volt battery surveillance requirement. The proposed change is to delete 
the words ``during shutdown'' from SR 4.8.2.1.2.e (performance 
discharge test). The licensee contends that the ``during shutdown'' 
provision in the TS is an impractical requirement because both units 
would have to be shutdown to perform the performance discharge test 
(PDT).
    In the licensee's supplement dated December 5, 1995, proposed 
changes were made to TS 3/4 8.2 Bases to support the frequency of the 
PDT on the other batteries in the system after a battery that had its 
PDT performed is returned to service.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. The proposed amendment seeks to change the surveillance 
requirements to allow the performance with the units on line. The 
surveillance can be safely completed as proposed without affecting 
unit operation. The equipment would not be removed from service for 
a time that would exceed the current allowed outage time. The 
probability or consequences of any accident previously evaluated 
will not be increased because the removal of a battery from service 
can be performed while on line, and the loads of each battery can be 
assumed by another same-train battery which is the case for the 
battery being inoperable for any other reason. During the allowed 
outage time, even a single failure of any component (including 
Emergency Diesel Generator) will still leave a full capacity train 
available to provide instrumentation and control power for both 
units. Train redundancy is maintained at all times. Compensatory 
action is taken to prohibit discharge testing of the other remaining 
batteries within 10 days following a battery performance discharge 
test to ensure that the tested battery is fully recharged. 
Probabilistic Risk Analysis shows that the increase in Core Damage 
Frequency due to this operation is negligible.
    2. The proposed amendment will not change any actual 
surveillance requirements, the change would simply allow the 
requirements to be met at different unit conditions. The performance 
of the surveillance with the units on line does not require any new 
component configurations that would reduce the ability of any 
equipment to mitigate an accident. The station would not be in any 
degraded status beyond that which has previously been evaluated. 
Therefore the proposed change will not create the possibility of a 
new accident.
    3. The change would allow a battery to be removed from service 
for testing. However, the testing must be completed within the 
current allowed outage time. As the allowed outage time defines the 
required margin of safety for equipment operability, removing 
equipment from service for testing and returning it to service 
within the allowed time does not affect a margin of safety. 
Compensatory action is taken to prohibit discharge testing of the 
other remaining batteries within 10 days following a battery 
performance discharge test to ensure that the tested battery is 
fully recharged.

    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: Atkins Library, University of 
North Carolina, Charlotte (UNCC Station), North Carolina 28223
    Attorney for licensee: Mr. Albert Carr, Duke Power Company, 422 
South Church Street, Charlotte, North Carolina 28242
    NRC Project Director: Herbert N. Berkow

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

    Date of amendment request: November 6, 1995
    Description of amendment request: The proposed amendment would 
revise the alarm setpoints for the noble gas and in-containment high 
range area radiation monitors listed in Table 3.3-6 of Beaver Valley 
Power Station, Unit 1 Technical Specification (TS) 3.3.3.1. The 
proposed revisions would make these alarm setpoints consistent with the 
criteria in the Emergency Action Levels (EALs) which were revised and 

[[Page 65678]]
approved by the NRC in August 1994. The revised EALs use the noble gas 
radiation monitors as indications of effluent releases and are based on 
dose to the public. The revised EALs use the in-containment high range 
area radiation monitors as indication of fission product barrier 
challenges or failures rather than as indications of effluent release.
    The proposed amendment would also revise Action Statement 36 of 
Table 3.3-6 of TS 3.3.3.1 for both BVPS-1 and BVPS-2 to reflect a 
previously approved change in reporting frequency for effluent 
releases. BVPS-1 License Amendment No. 188 and BVPS-2 License Amendment 
No. 70 (both issued on June 12, 1995) approved a change in the 
reporting frequency for effluent releases from semi-annual to annual. 
The proposed change would make Action Statement 36 consistent with this 
previously approved change.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Does the change involve a significant increase in the 
probability or consequences of an accident previously evaluated?
    The proposed monitor alarm setpoint changes and editorial 
changes are administrative in nature. Should the radiation alarm 
fail to annunciate or give a false alarm, there would be no affect 
on any other plant equipment or systems. The noble gas monitors are 
not safety related and do not interface with any safety related 
system. The containment area monitors are safety related; however, 
they do not initiate any safety function, nor do they interface with 
any other safety related system.
    The monitors' alarm as a visual (lighted icon) and audible alarm 
in the control room. The operator is then responsible for taking any 
corrective actions necessary, based on the alarm and Emergency 
Action Level (EAL) guidelines. The monitors do not provide for any 
automatic actions of other equipment or systems when an alarm 
condition occurs.
    The operating and design parameters of the radiation monitors 
will not change. The proposed change affects only the radiation 
level at which an alarm condition is created and does not affect any 
accident assumptions or radiological consequences of an accident.
    Therefore, the proposed change will not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    2. Does the change create the possibility of a new or different 
kind of accident from any accident previously evaluated?
    The proposed radiation monitor alarm revisions cannot initiate a 
new type of accident. A failure of the monitor itself cannot serve 
as the initiating event of an accident and has no effect on the 
operation of a safety system. Operator action is not made solely on 
a radiation monitor alarm; other plant condition indicators are also 
evaluated.
    Therefore, the proposed change does not create the possibility 
of a new or different kind of accident from any accident previously 
evaluated.
    3. Does the change involve a significant reduction in a margin 
of safety?
    The referenced radiation monitoring channels have no capability 
to mitigate the consequences of an accident. Also, they do not 
interface with any safety related system. The containment area 
monitors are safety related channels which provide indication to the 
operator of the integrity of the fission product barriers in 
containment. This indication, combined with other indications of 
plant conditions may direct an operator to take action to mitigate 
the consequences of an accident. The alarm setpoint itself does not 
perform any specific safety related function and the trip value is 
not referenced in the Updated Final Safety Analysis Report (UFSAR), 
nor does any site design basis document take credit for this 
setpoint. Safety limits and limiting safety system settings are not 
affected by this proposed change. Also, the site will continue to 
meet the requirements of 10 CFR Part 100 which limits offsite dose 
following a postulated fission product release.
    Therefore, use of the proposed technical specification would 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: B. F. Jones Memorial Library, 
663 Franklin Avenue, Aliquippa, Pennsylvania 15001.
    Attorney for licensee: Jay E. Silberg, Esquire, Shaw, Pittman, 
Potts & Trowbridge, 2300 N Street, NW., Washington, DC 20037.
    NRC Project Director: John F. Stolz

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

    Date of amendment request: November 22, 1995
    Description of amendment request: The licensee proposes to change 
Turkey Point Units 3 and 4 Technical Specifications (TS) Index to 
delete reference to the BASES. The proposed revisions to Turkey Point 
Units 3 and 4 TS are administrative in nature. Changes to the TS BASES 
will be controlled by a plant procedure under administrative controls 
and reviews. Proposed changes to the TS BASES will be evaluated in 
accordance with 10 CFR 50.59.
    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 
amendments would not involve a significant increase in the 
probability or consequences of an accident previously evaluated.
    The proposed amendments are administrative in nature and do not 
affect assumptions contained in plant safety analyses, the physical 
design and operation of the plant, nor do they affect Technical 
Specifications that preserve safety analysis assumptions. The 
Technical Specification BASES, per 10 CFR 50.36(a), are not a part 
of the Technical Specifications. Changes to the TS BASES will be 
controlled by a plant procedure under administrative controls and 
reviews. Proposed changes to the TS BASES will be evaluated in 
accordance with 10 CFR 50.59. Therefore, the proposed change does 
not affect the probability or consequences of accidents previously 
analyzed.
    (2) The proposed license amendments do not create the 
possibility of a new or different kind of accident from any accident 
previously evaluated.
    The proposed amendments are administrative in nature. The 
proposed amendments will not create the possibility of a new or 
different kind of accident from any accident previously evaluated 
since the proposed amendments will not change the physical plant or 
the modes of plant operation defined in the facility operating 
license. No new failure mode is introduced due to the administrative 
change, since the proposed change does not involve the addition or 
modification of equipment nor does it alter the design or operation 
of affected plant systems, structures, or components.
    (3) The proposed license amendments do not involve a significant 
reduction in a margin of safety.
    The operating limits and functional capabilities of the affected 
systems, structures, and components are unchanged by the proposed 
amendments. The BASES information, per 10 CFR 50.36(a), is not a 
part of the Technical Specifications. Changes to the TS BASES will 
be controlled by a plant procedure under administrative controls and 
reviews. Proposed changes to the TS BASES will be evaluated in 
accordance with 10 CFR 50.59. Therefore, the proposed change does 
not reduce any margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 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: Florida International 

[[Page 65679]]
    University, University Park, Miami, Florida 33199
    Attorney for licensee: J. R. Newman, Esquire, Morgan, Lewis & 
Bockius, 1800 M Street, NW., Washington, DC 20036
    NRC Project Director: David B. Matthews

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

    Date of amendment request: November 3, 1995
    Description of amendment request: The proposed amendment would 
revise the technical specifications (TS) to delay for one cycle the 
volumetric and surface examinations of the Reactor Coolant Pump (RCP) 
motor flywheels required by Regulatory Guide (RG) 1.14, Regulatory 
position C.4.b, incorporated by reference in Technical Specification 
5.6.2.8.c, to coincide with Crystal River Unit 3 (CR-3) Refueling 
Outage 11, scheduled for Spring 1998.
    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 will not significantly increase the 
probability or consequences of an accident previously evaluated.
    The safety function of the RCP flywheels is to provide a 
coastdown period during which the RCPs would continue to provide 
reactor coolant flow to the reactor after loss of power to the RCPs. 
The maximum loading on the RCP motor flywheel results from overspeed 
following a large LOCA [loss-of-coolant accident]. The estimated 
maximum obtainable speed in the event of a Reactor Coolant System 
piping break was established conservatively. The proposed one time 
change does not affect that analysis. Reduced coastdown times due to 
a single failed flywheel would not place the plant in an unanalyzed 
condition since a locked rotor (instantaneous coastdown) is analyzed 
in the FSAR [Final Safety Analysis Report]. The proposed change does 
not increase the amount of radioactive material available for 
release or modify any systems used for mitigation of such releases 
during accident conditions. Therefore, the proposed change does not 
involve a significant increase in the consequences of any accident 
previously evaluated.
    2. The proposed change will not create the possibility of a new 
or different kind of accident from any accident previously 
evaluated.
    The proposed change will not change the design, configuration, 
or method of operation of the plant. Therefore, the proposed change 
will not create the possibility of a new or different kind of 
accident from any previously evaluated.
    3. The proposed change will not involve a significant reduction 
to any margin of safety.
    FPC [Florida Power Corporation] has performed two full 
volumetric examinations in excess of those recommended in RG 1.14, 
Revision 1 during the Second ISI [inservice inspection] Interval. 
The margins of safety defined in RG 1.14, Revision 1 used in the 
analysis are not significantly changed.

    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: Coastal Region Library, 8619 
W. Crystal Street, Crystal River, Florida 32629
    Attorney for licensee: A. H. Stephens, General Counsel, Florida 
Power Corporation, MAC - A5D, P. O. Box 14042, St. Petersburg, Florida 
33733
    NRC Project Director: David B. Matthews

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 amendment request: November 10, 1995
    Description of amendment request: The proposed amendments would 
revise the Technical Specifications (TS) for containment systems to 
reflect the adoption of the requirements of 10 CFR Part 50, Appendix J, 
Option B, and the implementation of a performance-based containment 
leak-rate testing program at the Edwin I. Hatch Nuclear Plant, Units 1 
and 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 proposed changes do not involve a significant increase in 
the probability of consequences of an accident previously evaluated. 
The proposed changes do not involve any physical or operational 
changes to structures, systems or components. The proposed changes 
provide a mechanism within the TS for implementing a performance-
based leakage rate test program which was promulgated by the 
revision to 10 CFR 50 to incorporate Option B to Appendix J. The TS 
Limiting Conditions for Operation (LCO) remain unaffected by these 
changes. Thus, the safety design basis for the accident mitigation 
functions of the primary containment, the airlocks, and the primary 
containment isolation valves is maintained. Therefore, these changes 
will not increase the probability or consequences of an accident 
previously evaluated.
    2. The proposed changes do not create the possibility of a new 
or different kind of accident from any accident previously analyzed. 
Revising Surveillance Requirement acceptance criteria and 
frequencies does not physically modify the plant and does not modify 
the operation of any existing equipment.
    3. The proposed changes do not involve a significant reduction 
in the margin of safety, nor do they affect a safety limit, an LCO, 
or the manner in which plant equipment is operated. The NRC letter 
dated November 2, 1995, recognizes that changes similar to the 
proposed changes are required to implement Option B of 10 CFR 50, 
Appendix J. In NUREG-1493, ``Performance-Based Containment Leak-Test 
Program,'' which forms the basis for the Appendix J revision, the 
NRC concludes that adoption of performance-based test intervals for 
Appendix J testing will not significantly reduce the margin of 
safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location:Appling County Public Library, 
301 City Hall Drive, Baxley, Georgia 31513
    Attorney for licensee: Ernest L. Blake, Jr., Esquire, Shaw, 
Pittman, Potts and Trowbridge, 2300 N Street, NW., Washington, DC 20037
    NRC Project Director: Herbert N. Berkow

GPU Nuclear Corporation, Docket No. 50-320, Three Mile Island 
Nuclear Station, Unit No. 2 (TMI-2), Dauphin County, Pennsylvania

    Date of amendment request: January 16, 1995
    Description of amendment request: The proposed amendment would 
revise TMI-2 Operating License No. DPR-73 by modifying Section 6.5.1.7 
of the administrative controls portion of the technical specifications. 
The revision would change Section 6.5.1.7 to delete the requirement for 
personnel in the internal GPU Nuclear (GPUN) Review and Approval matrix 
to render an unreviewed safety question (USQ) determination regarding 
(1) proposed changes to unit technical specifications and (2) 
investigations of violations of technical specifications. Both of these 
activities involve docketed correspondence with the NRC in which the 
USQ determination is made and justified. This obviates the need for a 
requirement for the licensee to perform and document an internal USQ 

[[Page 65680]]
determination. This change would make the TMI-2 Technical 
Specifications consistent with the Standard Technical Specifications 
for B&W Plants (NUREG 1430).
    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:

    10 CFR 50.92 provides the criteria which the Commission uses to 
perform a no significant hazards consideration. 10 CFR 50.92 states 
that an amendment to a facility license involves no significant 
hazards if operation of the facility in accordance with the proposed 
amendment would not:
    1. Involve a significant increase in the probability or 
consequences of an accident previously evaluated, 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 proposed change to the technical specifications is 
administrative and does not involve any physical changes to the 
facility. No changes are made to operating limits or parameters, nor 
to any surveillance activities. Based on this, GPU Nuclear has 
concluded that the proposed change does not:
    1. Involve a significant increase in the probability of 
occurrence of the consequences of an accident previously evaluated. 
The proposed amendment is purely administrative and affects only the 
review of activities that involve considerable review by the NRC. 
This change will not degrade the performance of review for either of 
the two activities that are affected. This proposed technical 
specification change does not involve changes to hardware 
configuration, operation, or testing. Therefore, this change does 
not increase the probability of occurrence or the consequences of an 
accident previously evaluated.
    2. Create the possibility of a new or different kind of accident 
since the change is administrative and no new failure modes are 
created.
    3. Involve a change in the margin of safety. This change is 
administrative in nature; compatible with standard technical 
specifications; and does not affect any safety settings, equipment, 
or operational parameters.
    Based on the above analysis it is concluded that the proposed 
changes involve no significant safety hazards considerations as 
defined by 10 CFR 50.92.

    The NRC staff has reviewed the analysis of the licensee 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: Government Publications 
Section, State Library of Pennsylvania, Walnut Street and Commonwealth 
Avenue, Box 1601, Harrisburg, Pennsylvania 17105
    Attorney for licensee: Ernest L. Blake, Jr., Esquire, Shaw, 
Pittman, Potts & Trowbridge, 2300 N Street, NW, Washington, DC 20037
    NRC Project Director: Seymour H. Weiss

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: October 27, 1995
    Description of amendment request: The proposed amendment would 
revise Technical Specification (TS) 3.1.3, ``Control Rod OPERABILITY,'' 
to include the 25% surveillance overrun allowed by Limiting Condition 
for Operation (LCO) 3.0.2 into the allowances of the surveillance Notes 
for control rod ``notch'' testing per Surveillance Requirement (SR) 
3.1.3.2 and SR 3.1.3.3. The proposal also includes a clarification to 
the description of TS Table 3.3.3.1-1, ``Post Accident Monitoring 
Instrumentation,'' Function 7, to indicate that the Function's 
requirements apply to the position indication for only automatic 
primary containment isolation valves, rather than all primary 
containment isolation valves. Finally, the proposal includes changes to 
correct a number of editorial and typographical errors inadvertently 
contained in TS 3.3.4.1, ``End of Cycle Recirculation Pump Trip (EOC-
RPT) Instrumentation,'' TS 3.3.6.1, ``Primary Containment and Drywell 
Isolation Instrumentation,'' TS 3.3.8.2, ``Reactor Protection System 
(RPS) Electric Power Monitoring,'' and TS 3.6.5.2, ``Drywell Air 
Lock.''
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration which is presented below:

    (1) The proposed changes associated with Limiting Condition for 
Operation (LCO) 3.1.3 are being made to make the surveillance 
requirement (SR) Notes agree with their original intent. The Notes 
were originally intended to allow the testing of control rods to be 
tracked as a group, i.e., partially withdrawn and fully withdrawn. 
In the event that a control rod(s) has changed from one test group 
to another, the Notes were intended to allow performance of the next 
surveillance on that control rod(s) to be delayed to coincide with 
the next regularly scheduled performance of the test of the new 
group. However, these Notes failed to include the 25% surveillance 
extension allowances of SR 3.0.2. This proposed change merely adds 
the 25% extension to the time allowed by the Notes to make them 
agree with the Frequency plus the extension allowance of SR 3.0.2. 
The addition of the word ``fully'' to the Note for SR 3.1.3.2 is to 
provide for clarification only. These changes are consistent with 
changes approved for the Grand Gulf Nuclear Station (GGNS) and River 
Bend Station and are being proposed for the Clinton Power Station 
(CPS) for consistency. The proposed changes do not involve a change 
to the control rods or control rod drive system design or operation. 
Further, the proposed change does not affect the way in which the 
associated control rod test is performed, only the ``triggers'' for 
performance of the test are affected. These triggers are being 
revised to make them consistent with their original intent. As a 
result, the proposed change cannot increase the probability or the 
consequences of any accident previously evaluated.
    The proposed change to the description of LCO 3.3.3.1 Function 7 
to include ``automatic'' is provided for clarification only. As 
described in the Bases for this Function, the requirements for 
operability are currently only associated with automatic primary 
containment isolation valves (PCIVs). As a result, this change does 
not involve a change to the scope of this LCO. In addition, these 
changes are consistent with changes approved for GGNS and are being 
proposed for CPS for consistency. Since this request does not affect 
the design or operation of this equipment, nor does it alter the 
scope of this Technical Specification (TS) requirement, this 
proposed change cannot increase the probability or the consequences 
or any accident previously evaluated.
    The remaining proposed changes are purely editorial and do not 
affect the design or operation of any equipment or alter the 
technical requirements of any TS. As a result, these proposed 
changes cannot increase the probability or the consequences of any 
accident previously evaluated.
    (2) The proposed changes do not affect the design or operation 
of any equipment. In addition, the proposed changes do not affect 
the manner in which any test is performed or involve a change to any 
plant operating mode or configuration. As a result, Illinois Power 
has concluded that the proposed changes cannot create the 
possibility of an accident not previously evaluated.
    (3) The proposed changes to the SRs for LCO 3.1.3 are being made 
to make the SR Notes agree with their original intent and thus 
permit control rods to be tested as originally intended. The 
proposed changes do not involve a change to the control rods or 
control rod drive system design or operation. Further, the proposed 
change does not affect the way in which this test is performed or 
the routine Frequency of performing the test, only the ``triggers'' 
are affected. Since these triggers are being revised to make them 
consistent with their original intent, Illinois Power has determined 
that this change does not result in a reduction in the margin of 
safety.
    The proposed change to the description of LCO 3.3.3.1 Function 7 
to include ``automatic'' is provided for clarification only. As 
described in the Bases for this Function, the requirements for 
operability are 

[[Page 65681]]
currently only associated with automatic PCIVs. As a result, this 
change does not involve a change to the current scope of this LCO. 
Since this request does not affect the design or operation of this 
equipment, nor does it alter the scope of this TS requirement, this 
proposed change does not result in a reduction in the margin of 
safety.
    The remaining changes are purely editorial and do not affect the 
design or operation of any equipment or alter the technical 
requirements of any TS. As a result, these proposed changes do not 
result in a reduction in the margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: 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: October 27, 1995
    Description of amendment request: The proposed amendment would 
revise Technical Specification 5.2.2.e, ``Unit Staff,'' to revise the 
requirements for controls on the working hours of unit staff who 
perform safety related functions. The proposal would clarify the 
approval requirements for deviations from the overtime guidelines and 
eliminate the requirement for a monthly review of individual overtime, 
consistent with GL 82-12, ``Nuclear Power Plant Staff Working Hours,'' 
dated June 15, 1982.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration which is presented below:

    (1) The proposed changes do not involve a change to the plant 
design or operation. The proposed changes do not affect the level of 
approval required for deviations from the overtime guidelines. As 
the Technical Specifications will continue to require deviations 
from the guidelines for overtime control to be approved and 
documented, the proposed changes do not adversely affect the level 
of alertness for the unit staff who perform safety-related 
functions. The current requirement for the plant manager (or his 
designee) to perform a monthly review of individual overtime is an 
after the fact review that has not been proven to provide any 
significant benefit with respect to the control of individual 
overtime. In addition, the proposed changes do not directly affect 
the automatic operation of equipment or systems assumed to mitigate 
the consequences of previously evaluated accidents. As a result, the 
proposed changes do not affect any of the parameters or conditions 
that contribute to initiation of an accident previously evaluated, 
and thus, the proposed changes cannot increase the probability or 
the consequences of any accident previously evaluated.
    (2) The proposed changes do not involve a change to the plant 
design or operation. The proposed changes do not affect the level of 
approval required for deviations from the overtime guidelines and do 
not adversely affect the level of alertness for the unit staff who 
perform safety-related functions. As a result, the proposed changes 
do not affect any of the parameters or conditions that could 
contribute to initiation of an accident, and thus cannot create the 
possibility of an accident not previously evaluated.
    (3) The proposed changes do not involve a significant reduction 
in a margin of safety. As noted previously, the proposed changes do 
not change the level of approval required for deviations from the 
overtime guidelines. Only the requirement for an after-the-fact 
monthly review is proposed to be deleted. To the extent that 
personnel alertness may be regarded as a margin of safety, deleting 
this requirement will not result in a significant reduction in a 
margin of safety since overtime controls consistent with the 
guidelines and requirements of GL 82-12 will continue to remain in 
place.

    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: May 19, 1995, as supplemented October 
20, 1995 (AEP:NRC:1213A)
    Description of amendment requests: The proposed amendments would 
modify the Technical Specification (TS) action statement associated 
with the main steam safety valves (MSSVs). The action statement would 
reflect different requirements based on operating mode and the power 
range neutron flux high setpoint with inoperable MSSVs would be revised 
in response to an issue raised in Westinghouse Nuclear Safety Advisory 
Letter 94-001. The supplement also requested the addition of an 
exemption to TS 4.0.4 in the surveillance requirements for the MSSVs.
    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
    Correction of the setpoint methodology does not represent a 
credible accident initiator. The new methodology reduces the 
allowable power level setpoints and is conservative compared to the 
presently evaluated setpoints. The consequences of any previously 
evaluated accident are not adversely affected by this action because 
the decrease in the setpoints resulting from the new calculational 
methodology will ensure that the MSSVs are capable of relieving the 
pressure at the allowable power levels. 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.
    Correcting the overly restrictive action statements of T/S 3.7.1 
does not involve a significant increase in the probability of an 
accident. The proposed changes modify existing text to more 
accurately reflect the intention of the restrictions imposed by the 
action statements. The changes do not create any situation that 
would initiate a credible accident sequence.
    The proposed 4.0.4 exemption is necessary to make the T/Ss 
accurately reflect limitations associated with conduct of the 
surveillance in Mode 3. Additionally, the change is needed to 
address the fact that unscheduled outages can and do occur and, when 
they do, surveillances can expire with no way to correct the 
situation until the unit returns to power. Since the purpose of the 
4.0.4 exemption is to allow surveillances to be conducted after an 
extended period of reactor shutdown, the decay heat to be removed by 
the MSSVs will be less than (and therefore conservative compared to) 
the conditions experienced when the surveillances are already 
allowed by the T/Ss. These allowed conditions include conduct of the 
surveillance during power operation or immediately after shutdown. 
Therefore, we believe that any increase in the probability of 
occurrence or consequences of an accident previously analyzed would 
be insignificant.
    Criterion 2
    The change in Table 3.7-1 reduces the allowable power levels 
that can be achieved in the event that one or more main steam safety 
valve(s) is inoperable. This change is a result of vendor guidance 
to correct an error in the existing methodology used to determine 
the setpoints for the power level. 

[[Page 65682]]
Changing the methodology used to determine the setpoints, and lowering 
the setpoints themselves, do not create a new condition that could 
lead to a credible accident. Therefore, 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.
    The action statements remain in effect to perform the intended 
function of protecting the plant's secondary side when the main 
steam safety valves are inoperable. They have only been modified to 
correct the overly restrictive language that specifies when, in each 
mode, specific actions must be taken. Therefore, the proposed change 
does not create a new or different type of accident.
    Because the proposed 4.0.4 exemption requires neither physical 
changes to the plant nor changes to the safety analyses, we believe 
that they will not create the possibility of a new or different kind 
of accident from any previously evaluated.
    Criterion 3
    The margin of safety presently provided is not reduced by the 
proposed change in the setpoints. The change will correct the 
limiting power levels that are to be implemented when MSSVs are 
inoperable. This action does not adversely affect the margin that 
was previously allocated for the ability of the MSSVs to relieve 
secondary side pressure. Based on these considerations, it is 
concluded that the changes do not involve a significant reduction in 
a margin of safety.
    The margin of safety is also not significantly reduced by the 
proposed change to the action statements of the T/S. The proposed 
revision clarifies when specific actions are to be taken in response 
to inoperable main steam safety valves. The changes do not decrease 
the effectiveness of the actions to be taken; therefore, they do not 
significantly reduce any margin of safety.
    The margin of safety is not adversely affected by the proposed 
exemption to T/S 4.0.4, since the surveillance conditions allowed by 
the exemption are bounded by the normal surveillance conditions seen 
immediately after shutdown or during power operation.

    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. The 
initial application was noticed in the Federal Register on June 21, 
1995 (60 FR 32368).
    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: Brian E. Holian, Acting

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: November 10, 1995 (AEP:NRC:0896X) 
(Supersedes application dated June 15, 1995.)
    Description of amendment requests: The proposed amendments would 
change the 18-month emergency diesel generator (EDG) surveillance test 
from a 24-hour run to an 8-hour run and would add voltage and frequency 
measurement and power factor monitoring.
    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
    The safety function of the EDGs is to supply AC electrical power 
to plant safety systems whenever the preferred AC power supply is 
unavailable. Through surveillance requirements, the ability of the 
EDGs to meet their load and timing requirements is tested and the 
quality of the fuel and the availability of the fuel supply are 
monitored. Reduction of the 24 hour run to 8 hours will not reduce 
the surveillance effectiveness and will sufficiently exercise the 
EDG and its support systems to identify potential conditions that 
could lead to performance degradation (See Attachment 4 [of 
amendment request]). Further, monthly full-load testing will provide 
confidence in diesel reliability and performance capability. Based 
on these considerations, it is concluded that the proposed 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 changes only 
involve EDG surveillance test requirements. These changes will not 
affect EDG operability and are designed to improve surveillance 
effectiveness. Also, paralleling the diesel to the system grid 
during normal operations has been performed to fulfill monthly 
surveillance requirements when the resistive load banks were not 
available.
    It is recognized that, during the 1 hour monthly surveillance 
test period, the diesel could be exposed to electrical system 
transients (e.g., transients induced by inclement weather 
conditions) which could cause the paralleled diesel output breaker 
to trip open. Such a scenario, although unlikely, is mitigated by 
the availability of the alternate EDG which is placed in the auto 
start mode prior to the surveillance. In addition, during testing, 
an operator is continuously monitoring the diesel control panel and 
can, if necessary, reset the affected EDG lockout relays to restore 
EDG availability. Therefore, 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
    Although the duration of the EDG 18 month 24 hour surveillance 
test would be reduced, the EDG components will continue to be 
sufficiently exercised such that the ability to detect incipient and 
degraded conditions will be maintained (See Attachment 4, Figure 2 
[of amendment request]). Also, the added review of diesel reactive 
loading ensures that test conditions closely match potential 
emergency conditions. In addition, the monthly full-load testing 
will provide confidence in diesel reliability and performance 
capability without impacting diesel operability. During the monthly 
test, the impact on plant safety due to potential exposure to 
transient grid conditions is considered to be insignificant based on 
the likelihood of such transients coincident with the testing and 
the mitigating factors discussed in Criterion 2 above.
    Based on the above 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 requests involve no significant hazards consideration. This 
notice supersedes the staff's notice published in the Federal Register 
on July 19, 1995 (60 FR 37096).
    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: Brian E. Holian, Acting

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

    Date of amendment request: October 25, 1995
    Description of amendment request: The amendment request would 
revise the Technical Specifications (TS) to relocate the flow-biased 
average power range monitor (APRM) scram and rod block setpoint 
requirements for reactor operation with excessive core peaking, which 
will also include surveillance requirements to verify the setpoints. 
The amendment would also delete TS Figure 2.1.2, and any references to 
the figure. APRM meter setting adjustments would be changed to allow 
setpoint adjustment to be made at power levels less than or equal to 
90% of the rated, and the 

[[Page 65683]]
requirement that the scram setting adjustment be <10% would be further 
defined as <10% of the rated thermal power. The amendment would 
incorporate several editorial changes and renumbered pages, the removal 
of blank pages, a revised Table of Contents, and a modified Bases 
section for the APRM setpoint 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 (SHC), which is presented below:

    The proposed changes do not involve an SHC because the changes 
would not:
    1. Involve a significant increase in the probability or 
consequences of an accident previously analyzed.
    The proposed changes do not cause the APRM scram and rod block 
setpoints or APRM meter readings to be manipulated differently. The 
change limiting the scram and rod block setting adjustments to less 
than 10% of rated thermal power is more conservative than the 
current specification in that it allows the APRM meter indication to 
be set closer to the flow-biased scram or rod block setpoint. There 
are no other changes to the basic function of any plant equipment. 
The proposed changes to technical specifications will not decrease 
the margin to the fuel thermal-mechanical design limits, so the 
potential for any fuel failure from the LHGR [linear heat generation 
rate] transient overpower condition is not increased. Therefore, the 
consequences of a transient overpower are also not increased. Based 
on the above, these changes will not significantly increase the 
probability or consequences of an accident previously evaluated.
    2. Create the possibility of a new or different kind of accident 
from any accident previously analyzed.
    Moving the APRM setpoint adjustment from Section 2 to Section 3/
4.11 does not reduce or eliminate any requirements. The requirements 
for the APRM setpoint adjustment are more clearly defined in the LCO 
[limiting conditions for operation] and Surveillance Requirements 
with specific applicability and corrective action requirements. The 
proposed changes do not affect the basic function of any plant 
equipment. The basic process for performing the APRM setpoint 
adjustment is not significantly changed, so the proposed changes do 
not create a new process and do not involve any new failure that 
would cause a new or different kind of accident to occur.
    The elimination of redundant information in the technical 
specifications and the relocation of information pertinent to the 
operators for performing the APRM setdown determination does not 
create a new or different kind of accident.
    3. Involve a significant reduction in the margin of safety.
    Allowing APRM setpoint adjustment during power operation at off-
rated conditions improves the flexibility to make control rod 
pattern or core flow adjustments, but will still preserve the 
required setdown factor that must be maintained in that flux shape 
and power level. The change to set up the APRM meter reading up to 
10% above the nominal power indication (instead of setting up only 
to the current MFLPD [maximum fraction of limiting power density] 
percentage) allows a higher APRM meter setting to be made. This 
allows the conservative setting, but eliminates frequent setting 
changes each time a new value of FRP/MFLPD [fraction of rated power] 
is calculated provided the APRM setting remains conservatively 
greater than or equal to MFLPD/FRP multiplied by percent core 
thermal power. Thus, the margins to the fuel thermal and mechanical 
design limits are not reduced. The fuel remains adequately protected 
from failure due to a transient LHGR overpower condition. There is 
no reduction in any margin of safety.
    The time requirements imposed are consistent with the current 
fuel thermal limit LCO actions and are more conservative than STS, 
therefore, the proposed action time requirement provides the same 
margin of safety as currently exists in the MP1 [Millstone Unit 1] 
Technical Specifications. The margins to the fuel thermal and 
mechanical design limits are not reduced. There is no reduction in 
any margin of safety and the fuel remains adequately protected from 
failure due to a transient LHGR overpower condition.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: Learning Resources Center, 
Three Rivers Community-Technical College, 574 New London Turnpike, 
Norwich, CT 06360.
    Attorney for licensee: Lillian M. Cuoco, Esq., Senior Nuclear 
Counsel, Northeast Utilities Service Company, P.O. Box 270, Hartford, 
CT 06141-0270.
    NRC Project Director: Phillip F. McKee

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

    Date of amendment request: November 20, 1995
    Description of amendment request: The first of the proposed changes 
provides clarification to the applicability statement for the steam 
generator blowdown monitor in Table 3.3-12. The applicability is 
changed to be for Modes 1-4 only. The second proposed change involves 
the action statement for the steam generator blowdown monitor in Table 
3.3-12, Action 2. The action required when the monitor is not operable 
is clarified to state that if discharges are suspended, no sampling is 
required. The last proposed change involves the applicability statement 
for the condensate polishing facility waste neutralizing sump radiation 
monitor. It is clarified to state that the monitor is only required 
when the pathway is in use.
    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 (SHC), which is presented below:

    ... NNECO [Northeast Nuclear Energy Company] concludes that 
these changes do not involve a significant hazards consideration 
since the proposed changes satisfy the criteria in 10CFR50.92(c). 
That is, the proposed changes do not:
    1. Involve a significant increase in the probability or 
consequences of an accident previously evaluated.
    The proposed changes clarify the modes and conditions for which 
the radiation monitors are utilized, as well as the required actions 
when the monitors are not operable. These changes are administrative 
in nature, therefore, the changes will not increase the probability 
or consequences of an accident previously evaluated.
    2. Create the possibility of a new or different kind of accident 
from any accident previously evaluated.
    The proposed changes have no [e]ffect on the ability of the 
monitors to perform their design function. The clarifications do not 
involve any physical modifications to any equipment, structures, or 
components. The proposed changes have no impact on design basis 
accidents, and the changes will not modify plant response or create 
a new or unanalyzed event.
    3. Involve a significant reduction in a margin of safety.
    These changes do not have any impact on the protective 
boundaries and, therefore, have no impact on the safety limits for 
these boundaries. The instrumentation associated with these changes 
do not provide a safety function and only serve to provide 
radiological information to plant operators. The instrumentation has 
no [e]ffect on the operation of any safety-related equipment. No 
hardware, software, or setpoint changes are involved in this 
proposed change. These changes provide clarification of modes and 
conditions for which the radiation monitors are utilized. As such, 
these changes have no impact on the margin of safety.

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

[[Page 65684]]
    Three Rivers Community-Technical College, 574 New London Turnpike, 
Norwich, CT 06360.
    Attorney for licensee: Lillian M. Cuoco, Esq., Senior Nuclear 
Counsel, Northeast Utilities Service Company, P.O. Box 270, Hartford, 
CT 06141-0270.
    NRC Project Director: Phillip F. McKee

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

    Date of amendment request: November 21, 1995
    Description of amendment request: The proposed amendment would 
clarify the reactor containment building temperature as ``an 
equilibrium liner temperature,'' and the affected Bases will be updated 
to reflect the results of the most recent main steam line break (MSLB) 
analysis. The changes to the Bases also identify that the limiting 
event affecting containment temperature and pressure now includes the 
MSLB in addition to a Loss of Coolant Accident.
    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 (SHC), which is presented below:

    ... NNECO [Northeast Nuclear Energy Company] concludes that 
these changes do not involve a significant hazards consideration 
since the proposed change satisfies the criteria in 10CFR50.92(c). 
That is, the proposed changes do not:
    Involve a significant increase in the probability or 
consequences of an accident previously evaluated.
    These changes are clarifications that are administrative in 
nature. The changes only incorporate the revised containment 
analysis as approved by the NRC. There are no hardware changes and 
no change to the functioning of any equipment which could affect any 
operational modes or accident precursors. Therefore, there is no way 
that the probability of previously evaluated accidents could be 
affected.
    There are no hardware modifications associated with these 
changes and no change to the functioning of any equipment which 
could affect radiological releases. The safety analysis of the plant 
is unaffected by the changes. Therefore, there is no effect on the 
consequences of previously evaluated accidents.
    Create the possibility of a new or different kind of accident 
from any accident previously evaluated.
    These changes are clarifications that are administrative only. 
There are no hardware changes and no change to the functioning of 
any equipment which could introduce new or unique operational modes 
or accident precursors. Therefore, there is no possibility of an 
accident of a new or different type than previously evaluated.
    Involve a significant reduction in a margin of safety.
    These changes are clarifications that are administrative in 
nature. They do not increase or decrease any plant operating 
requirements or limits. Therefore, they have no effect on any safety 
analysis and no impact on the margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: Learning Resources Center, 
Three Rivers Community-Technical College, 574 New London Turnpike, 
Norwich, CT 06360.
    Attorney for licensee: Lillian M. Cuoco, Esq., Senior Nuclear 
Counsel, Northeast Utilities Service Company, P.O. Box 270, Hartford, 
CT 06141-0270.
    NRC Project Director: Phillip F. McKee

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: November 14, 1995
    Description of amendment request: The proposed amendment would 
remove the Technical Specification (TS) for motor operated valves with 
thermal overload protection and bypass devices (TS 3/4.8.4.2) to follow 
the guidance of the improved Westinghouse Standardized TS (NUREG-1431, 
Rev. 1).
    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 
in the FSAR [Final Safety Analysis Report] is not significantly 
increased.
    The removal of TS 3/4.8.4.2 from TS in no way impacts the 
accident analysis of the FSAR. Compliance of 10 CFR 50, as applies 
to Regulatory Guide 1.106, will be maintained and controlled through 
plant procedures with changes evaluated through 10 CFR 50.59 rather 
than through TS amendments. Therefore, the probability or 
consequences of a previously evaluated accident has not been 
increased.
    2. The possibility of an accident or a malfunction of a 
different type than any previously evaluated is not created.
    The proposed TSCR [TS change request] does not necessitate 
physical alteration of the plant nor changes in parameters governing 
normal plant operation. Therefore, the change does not create the 
possibility of a new or different kind of accident or malfunction.
    3. The margin of safety has not been significantly reduced.
    The removal of TS 3/4.8.4.2 and Table 3.8-2 will not diminish 
the existing thermal overload protection and/or bypass devices 
operability and testing requirements. They will be maintained and 
controlled in plant procedures, and changes will be subject to 10 
CFR 50.59 review. Therefore, the margin of safety has not decreased.

    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

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: November 21, 1995
    Description of amendment request: The proposed amendment would 
change Technical Specification (TS) 3/4.5.2 by allowing a one time 
extension of the allowable outage time from 72 hours to 7 days for each 
residual heat removal (RHR) train. The one time extension is needed to 
allow maintenance and modification to the RHR system while the plant is 
in Mode 1.
    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 probability of an accident previously evaluated in the FSAR 
[Final Safety Analysis Report] does not change. A one time extension 
to increase the allowed outage time for each train of RHR from 72 
hours to 7 days affects only RHR train availability which does not 
contribute to the probability of a LOCA [loss-of-coolant accident]. 
The proposed change to TS 3/4.5.2 has been shown to have only a 
small increase in Core Damage Frequency. The consequences of a 

[[Page 65685]]
LOCA does not change from those currently resulting from a LOCA 
initiated while in TS 3.5.2 ACTION statement (a.), thus, there is no 
change in consequences of an accident previously evaluated in the 
FSAR.
    2. The possibility of an accident or a malfunction of a 
different type than any previously evaluated is not created.
    The proposed TSCR [TS change request] only results in a one time 
increase in the allowable outage time for each train of RHR. It does 
not result in an operational condition different from that which has 
already been considered by TS. Therefore, the change does not create 
the possibility of a new or different kind of accident or 
malfunction.
    3. The margin of safety has not been significantly reduced.
    The effects of increasing the allowed outage time on the 
calculated core damage frequency has been evaluated and determined 
to be small.

    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

The Cleveland Electric Illuminating Company, Centerior Service 
Company, Duquesne Light Company, Ohio Edison Company, Pennsylvania 
Power Company, Toledo Edison Company, Docket No. 50-440, Perry 
Nuclear Power Plant, Unit No. 1, Lake County, Ohio

    Date of amendment request: November 22, 1995
    Description of amendment request: The proposed amendment would 
change the operating license to reflect the license transfer for part 
of Ohio Edison Company's ownership interest in the Perry Nuclear Power 
Plant (PNPP), Unit No. 1 to its wholly owned subsidiary, OES Nuclear 
Inc.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration which is presented below:

    1. The proposed changes do not involve a significant increase in 
the probability or consequences of an accident previously evaluated.
    The proposed changes to the PNPP Operating License are 
administrative and have no effect on the PNPP facility, programs, 
personnel or any plant systems. All Limiting Conditions for 
Operation, Limiting Safety Systems Settings, and Safety Limits 
specified in the Technical Specifications will remain unchanged. 
This change meets one of the examples of a change not likely to 
involve a significant hazards consideration in that it is a purely 
administrative change. 48 Fed. Reg. 14,864 (1983).
    2. The proposed changes do not create the possibility of a new 
or different kind of accident from any accident previously 
evaluated.
    The proposed changes to the PNPP Operating License are 
administrative and have no effect on the PNPP facility, programs, 
personnel or any plant systems. PNPP's design and design bases will 
remain unchanged as will All Limiting Conditions for Operation, 
Limiting Safety Systems Settings, and Safety Limits specified in the 
Technical Specifications. This change meets one of the examples of a 
change not likely to involve a significant hazards consideration in 
that it is a purely administrative change. 48 Fed. Reg. 14,864 
(1983).
    3. The proposed changes do not involve a significant reduction 
in the margin of safety.
    The proposed changes to the PNPP Operating License are 
administrative and have no effect on the PNPP facility, programs, 
personnel or any plant systems. All Limiting Conditions for 
Operation, Limiting Safety Systems Settings, and Safety Limits 
specified in the Technical Specifications will remain unchanged. 
This change meets one of the examples of a change not likely to 
involve a significant hazards consideration in that it is a purely 
an administrative change. 48 Fed. Reg. 14,864 (1983).

    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:Perry Public Library, 3753 Main 
Street, Perry, Ohio 44081
    Attorney for licensee: Jay Silberg, Esq., Shaw, Pittman, Potts & 
Trowbridge, 2300 N Street, NW., Washington, DC 20037
    NRC Project Director: Gail H. Marcus

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

    Date of amendment request: November 20, 1995
    Description of amendment request: The proposed changes would revise 
the Technical Specifications (TS) for the North Anna Power Station, 
Units No. 1 and No. 2 (NA-1&2). Specifically, the change would permit 
the use of 10 CFR Part 50, Appendix J, Option B, Performance-Based 
Containment Leakage Rate Testing.
    The Nuclear Regulatory Commission (NRC) has amended its regulations 
to provide a performance-based option for leakage-rate testing of 
containments. This testing option is available in lieu of compliance 
with the prescriptive requirements contained in Appendix J regulations. 
In order to implement the performance-based leakage-rate testing option 
the TS must be changed to eliminate reference to the prescriptive 
Appendix J requirements. Therefore, the licensee is proposing a change 
to the NA-1&2 TS to eliminate the current prescriptive requirements for 
leakage rate testing of the containment and reference Option B to 10 
CFR 50 Appendix J and NRC Regulatory Guide 1.163, ``Performance-Based 
Containment Leakage-Test Program.'' This change will permit use of the 
performance-based surveillance testing, Option B, of 10 CFR 50 Appendix 
J.
    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:

    Specifically, operation of North Anna Power Station with the 
proposed change will not:
    1. Involve a significant increase in either the probability of 
occurrence or consequences of any accident or equipment malfunction 
scenario which is important to safety and which has been previously 
evaluated in the Updated Final Safety Analysis Report (UFSAR).
    Plant systems and components will not be operated in a different 
manner as a result of the proposed Technical Specifications change. 
The proposed change permits a performance-based approach to 
determining the leakage-rate test frequency for the containment and 
containment penetrations (Type A, B, and C tests). Since the 
proposed change only affects the test frequency for containment and 
containment penetrations, the probability of occurrence of an 
accident is not affected by the proposed changes in the leak-rate 
test interval.
    The proposed change increases the probability of a malfunction 
due to the longer intervals between leakage tests. It has been 
estimated that the longer test intervals will increase the overall 
accident risk to the public by approximately 0.7% and 2.2% (for 
changes in the frequency of Type A tests and Type B and C tests, 
respectively). However, this increase in accident risk has been 
judged to be insignificant. This increase has been reviewed and 
judged to be acceptable by the NRC as documented in NUREG-1493 and 
the recent rulemaking to 10 CFR 50 Appendix J.
    The Limiting Conditions for Operation are not being changed for 
the containment or any other safety system. The containment and 
other safety system remain operable as 

[[Page 65686]]
assumed in the accident analysis. Since the proposed change does not 
affect the Limiting Conditions for Operation for the containment, 
the containment penetrations, or the other safety systems, the 
consequences of an accident are not affected by the changes in test 
frequency.
    2. Create the possibility of a new or different type of accident 
than those previously evaluated in the UFSAR.
    Implementing the proposed Technical Specifications change to 
remove the prescriptive testing requirements and permit use of 
Appendix J, Option B, performance-based testing of containment and 
its penetrations do not create the possibility of an accident of a 
different type than was previously evaluated in the UFSAR. Plant 
systems and components will not be operated in a different manner as 
a result of the proposed Technical Specifications change. Thus, the 
proposed Technical Specifications change in leakage-rate test 
frequency does not introduce any new accident precursors or modes of 
operation. The containment and containment penetrations will not be 
operated any differently as a result of the proposed change.
    Therefore, the possibility for an accident of a different type 
than was previously evaluated in the Safety Analysis Report is not 
created by the proposed Technical Specifications change.
    3. Involve a significant reduction in a margin of safety.
    The proposed change, which replace[s] the present prescriptive 
testing requirements with Appendix J, Option B, performance-based 
testing of containment and its penetrations, will continue to ensure 
that the existing accident analysis assumptions are maintained. The 
containment and containment penetrations will not be operated or 
tested any differently. Only the leakage rate test frequency is 
being changed as a result of the proposed change. The operational 
leakage-rate test acceptance criteria and the operability 
requirements are not being changed.

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

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

    Date of amendment request: November 20, 1995
    Description of amendment request: The proposed changes to the Surry 
Technical Specifications would eliminate the existing prescriptive 
testing requirements for leakage rate testing of the containment and 
instead reference the Nuclear Regulatory Commission (NRC) Regulatory 
Guide 1.163,'' Performance-Based Containment Leak-Test Program,'' which 
would permit use of the performance-based leakage rate testing, Option 
B of 10 CFR Part 50 Appendix J.
    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:

    Specifically, operation of Surry Power Station with the proposed 
change will not:
    1. Involve a significant increase in either the probability of 
occurrence or consequences of any accident or equipment malfunction 
scenario which is important to safety and which has been previously 
evaluated in the Updated Final Safety Analysis Report (UFSAR).
    Plant systems and components will not be operated in a different 
manner as a result of the proposed Technical Specifications change. 
The proposed change permits a performance-based approach to 
determining the leakage-rate test frequency for the containment and 
containment penetrations (Type A, B, and C tests). There are no 
plant modifications, or changes in methods of operation. Therefore, 
the changes in testing intervals for the containment and containment 
penetrations have no [e]ffect on the probability of occurrence of a 
LOCA [loss-of-coolant accident]. Since the proposed change only 
affects the test frequency for containment and the containment 
penetrations, and the as-found test acceptance criteria at Surry the 
probability of occurrence and the consequences of an accident are 
not affected by the proposed changes in the leak-rate test interval.
    The proposed change increases the probability of a malfunction 
of equipment important to safety due to the longer intervals between 
leakage tests. It has been estimated that the longer test intervals 
will increase the overall accident risk to the public by 
approximately 0.7% and 2.2% (for changes in the frequency of Type A 
tests and Type B and C tests, respectively). However, this increase 
in accident risk has been judged to be insignificant. This increase 
has been reviewed and judged to be acceptable by the NRC as 
documented in NUREG-1493 and the recent rulemaking to 10 CFR 50 
Appendix J.
    The containment and other safety system remain operable as 
assumed in the accident analysis. Changing the as-found acceptance 
criterion to 1.0 La at Surry does not increase the probability or 
consequences of an accident, since the accident analysis assume[s] a 
leakage rate of La for Design Basis Accidents. The as-left Type A 
test acceptance criterion remains at less than [or equal to] 0.75 
La. Since the proposed changes do not affect the Limiting Conditions 
for Operation for the containment, the containment penetrations, or 
the other safety systems, the consequences of an accident are not 
affected by the changes in test frequency.
    Therefore, the probability of an accident or consequences of an 
accident are not adversely affected as a result of this change.
    2. Create the possibility of a new or different type of accident 
than those previously evaluated in the UFSAR.
    Implementing the proposed Technical Specifications change to 
remove the prescriptive testing requirements and permit use of 
Appendix J, Option B, performance-based testing of containment and 
its penetrations does not create the possibility of an accident of a 
different type than was previously evaluated in the UFSAR. Plant 
systems and components will not be operated in a different manner as 
a result of the proposed Technical Specifications changes. Thus, the 
proposed Technical Specifications changes in leakage-rate test 
frequency do not introduce any new accident precursors or modes of 
operations. The containment and containment penetrations will not be 
operated any differently as a result of the proposed changes. 
Therefore, the possibility for an accident of a different type than 
was previously evaluated in the Safety Analysis Report is not 
created by the proposed Technical Specifications change.
    3. Involve a significant reduction in a margin of safety.
    The proposed Technical Specifications change, which replace[s] 
the present prescriptive testing requirements with Appendix J, 
Option B, performance-based testing of containment and its 
penetrations, will continue to ensure that the existing accident 
analysis assumptions are maintained. The containment and containment 
penetrations will not be operated or tested any differently. The 
leakage rate test frequency is being changed as a result of the 
proposed change. Changing the as-found acceptance criterion to 1.0 
La at Surry does not increase the consequences of an accident, since 
the accident analysis assume[s] a leakage rate of La for Design 
Basis Accidents. The as-left Type A test acceptance criterion 
remains at less than [or equal to] 0.75 La, which maintains the 
operating margin. The operational leakage-rate test acceptance 
criteria and the operability requirements are not being changed. 
Therefore, the margin of safety as defined in the Technical 
Specifications bases is unaffected.

    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.

[[Page 65687]]

    Local Public Document Room location: Swem Library, College of 
William and Mary, Williamsburg, Virginia 23185.
    Attorney for licensee: Michael W. Maupin, Esq., Hunton and 
Williams, Riverfront Plaza, East Tower, 951 E. Byrd Street, Richmond, 
Virginia 23219
    NRC Project Director: David B. Matthews

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: October 23, 1995
    Description of amendment request: The proposed amendment would 
change the name of the licensee from Wisconsin Electric Power Company 
to Wisconsin Energy Company.
    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 Point Beach Nuclear Plant in accordance with 
the proposed amendments will not involve a significant increase in 
the probability or consequences of an accident previously evaluated.
    As a result of the proposed license amendment, there will be no 
physical change to the facilities and all Limiting Conditions for 
Operations, Limiting Safety System Settings, and Safety Limits 
specified in the Technical Specifications will remain unchanged. 
Also, the facilities' Quality Assurance Program, Emergency Plan, 
Security Plan, and Operator Training and Requalification Program 
will be unaffected. Therefore, this amendment will not cause a 
significant increase in the probability or consequences of an 
accident previously evaluated.
    2. Operation of the Point Beach Nuclear Plant in accordance with 
the proposed amendments will not create the possibility of a new or 
different kind of accident from any accident previously evaluated.
    The proposed amendment will have no effect on the physical 
configuration of the facilities or the manner in which they will 
operate. The design and design basis of the plants will remain the 
same. The current plant safety analysis will therefore remain 
complete and accurate in addressing the design basis events and in 
analyzing plant response and consequences for the facilities. The 
Limiting Conditions for Operations, Limiting Safety System Settings, 
and Safety Limits specified in the Technical Specifications for the 
facilities are not affected by the proposed license amendment. The 
plant conditions for which the design basis accident analysis have 
been performed will remain valid. Therefore, the proposed license 
amendment cannot create the possibility of a new or different kind 
of accident from any accident previously evaluated.
    3. Operation of the Point Beach Nuclear Plant in accordance with 
the proposed amendments will not involve a significant reduction in 
the margin of safety.
    Plant safety margins are established through Limiting Conditions 
for Operation, Limiting Safety System Settings, and Safety Limits 
specified in the Technical Specifications. Since there will be no 
change to the physical design or operation of the plant, there will 
be no change to any of these margins. Thus, the proposed license 
amendment will not involve a reduction in any margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: Joseph P. Mann Library, 1516 
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

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: November 17, 1995
    Description of amendment request: The proposed amendment would 
revise Technical Specification (TS) 15.6.3, ``Facility Staff 
Qualifications.'' The position of Health Physics Manager would be 
renamed Health Physicist. This change would provide additional staffing 
flexibility.
    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 Point Beach Nuclear Plant in accordance with 
the proposed amendments does not result in a significant increase in 
the probability or consequences of an accident previously evaluated.
    The proposed changes separate the qualifications requirements of 
the Technical Specifications from the Health Physics Manager, while 
requiring that the same qualifications be fulfilled by a designated 
Health Physicist position within the organization. This change 
maintains the present knowledge requirements of the PBNP staff. The 
personnel holding the health physics qualifications are not 
considered in the probability of any accident. By ensuring the 
appropriate expertise remains on the staff to advise management on 
issues related to radiological safety, appropriate action is assured 
during analyzed events to assess and mitigate the radiological 
consequences. Therefore, this change does not affect the probability 
or consequences of any accident previously evaluated.
    2. Operation of the Point Beach Nuclear Plant in accordance with 
the proposed amendments will not result in a new or different kind 
of accident from any accident previously evaluated.
    The proposed change separates the Health Physics Manager 
qualifications from the position while maintaining the requirements 
for that expertise to be maintained within the organization. This is 
an administrative change only and does not affect any plant 
structures, systems and components. Therefore, a new or different 
kind of accident from any accident previously evaluated cannot 
result.
    3. Operation of the Point Beach Nuclear Plant in accordance with 
the proposed amendments will not result in a significant reduction 
in a margin of safety.
    The proposed changes are administrative only. The required 
levels of expertise and experience will be maintained within the 
Health Physics organization. Therefore, there is no 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: Joseph P. Mann Library, 1516 
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. MarcusWolf Creek Nuclear Operating 
Corporation, Docket No. 50-482, Wolf Creek Generating Station, Coffey 
County, Kansas
    Date of amendment request: November 22, 1995
    Description of amendment request: The amendment would revise 
Technical Specification 3.9.4, ``Containment Building Penetrations,'' 
and its associated Bases section to allow the containment personnel 
airlock doors to be open during core alterations and movement of 
irradiated fuel in containment provided that a minimum of one door in 
the emergency airlock is closed and one door in the personnel airlock 
is capable of being closed. Also, Surveillance Requirement 4.9.4 would 
be revised to specify that each containment penetration should be in 
its ``required condition,'' instead of ``closed/isolated condition.''

[[Page 65688]]

    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. The proposed change does not involve a significant increase 
in the probability or consequences of an accident previously 
evaluated.
    The proposed change to Technical Specification 3.9.4 would allow 
the containment personnel airlock to be open during fuel movement 
and core alterations. The containment personnel airlock is currently 
closed during fuel movement and core alterations to prevent the 
escape of radioactive material in the event of a fuel handling 
accident. The containment personnel airlock is not an initiator of 
any accident. Whether the containment personnel airlock doors are 
open or closed during fuel movement and core alterations has no 
affect on the probability of any accident previously evaluated.
    The proposed change does alter assumptions previously made in 
evaluating the radiological consequences of the fuel handling 
accident inside the containment building. The proposed change allows 
for the containment personnel airlock to be open during refueling. 
The radiological consequences described in this change are bounded 
by those given in the Wolf Creek Generating Station Safety 
Evaluation Report and General Design Criteria 19. All doses for the 
proposed change are less than the acceptance criteria, therefore, 
there is no significant increase in the consequences of an accident 
previously analyzed.
    The proposed change would significantly reduce the dose to 
workers in the containment in the event of a fuel handling accident 
by accelerating the containment evacuation process. The proposed 
change would also significantly decrease the wear on the containment 
personnel airlock doors and, consequently, increase the reliability 
of the containment personnel airlock doors in the event of an 
accident.
    Since the probability of a fuel handling accident is unaffected 
by the airlock door positions, and the increased doses do not exceed 
acceptance limits, operation of the facility in accordance with the 
proposed amendment would not affect the probability or consequences 
of an accident previously analyzed.
    2. The proposed change does not create the possibility of a new 
or different kind of accident from any accident previously 
evaluated.
    The proposed change affects a previously evaluated accident, 
e.g., a fuel handling accident inside containment. The existing 
accident has been modified to account for the containment personnel 
airlock doors being opened at the time of the accident. It does not 
represent a significant change in the configuration or operation of 
the plant. 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. The proposed change does not involve a significant reduction 
in a margin of safety.
    The margin of safety is reduced when the offsite and control 
room doses exceed the acceptance criteria in the Wolf Creek 
Generating Station Safety Evaluation Report. As previously discussed 
in the response to Standard I, the offsite and control room doses 
are below the acceptance criteria. Therefore, operation of the 
facility in accordance with the proposed amendment would 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 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.

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 amendment request: November 14, 1995
    Description of amendment request: The proposed amendment would 
close out additional open items identified in the NRC staff's review of 
the upgrade of the Dresden and Quad Cities Technical Specifications 
(TS) to the standard Technical Specifications (STS) contained in NUREG-
0123. The Technical Specification Upgrade Program (TSUP) is not a 
complete adaption of the STS. The TS upgrade focuses on (1) integrating 
additional information such as equipment operability requirements 
during shutdown conditions, (2) clarifying requirements such as 
limiting conditions for operation and action statements utilizing STS 
terminology, (3) deleting superseded requirements and modifications to 
the TS based on the licensee's responses to Generic Letter (GL), and 
(4) relocating specific items to more appropriate TS locations. The 
November 14, 1995, application proposed to close out all open items 
identified during the NRC's review as noted in previous NRC staff 
Safety Evaluations for previously provided submittals regarding the 
TSUP project.
    Date of publication of individual notice in Federal Register: 
November 29,1995 (60 FR 61272).
    Expiration date of individual notice: December 28, 1995
    Local Public Document Room location: for Dresden, the Morris Area 
Public Library District, 604 Liberty Street, Morris, Illinois; and for 
Quad Cities Station, the Dixon Public Library, 221 Hennepin Avenue, 
Dixon, Illinois.

Commonwealth Edison Company, Docket Nos. 50-237 and 50-249, Dresden 
Nuclear Power Station, Units 2 and 3, Grundy County, 
Illinois;Docket Nos. 50-373 and 50-374, LaSalle County Station, 
Units 1 and 2, LaSalle County, Illinois; Docket Nos. 50-254 and 50-
265, Quad Cities Nuclear Power Station, Units 1 and 2, Rock Island 
County, Illinois

    Date of amendment request: November 14, 1995
    Description of amendment request: The proposed amendment would 
change the technical specifications of these plants to incorporate 10 
CFR Part 50, Appendix J, ``Primary Reactor Containment Leakage Testing 
For Water-Cooled Power Reactors'', Option B.
    Date of publication of individual notice in Federal Register: 
December 7, 1995 (60 FR 62896)
    Expiration date of individual notice: January 8, 1996
    Local Public Document Room location: for Dresden Station, Morris 
Area Public Library District, 604 Liberty Street, Morris, Illinois; for 
LaSalle County Station, Jacobs Memorial Library, Illinois Valley 
Community 

[[Page 65689]]
College, Oglesby, Illinois; and for Quad Cities Station, Dixon Public 
Library, 221 Hennepin Avenue, Dixon, Illinois.

Connecticut Yankee Atomic Power Company, Docket No. 50-213, Haddam 
Neck Plant, Middlesex County, Connecticut

    Date of amendment request: November 14, 1995
    Description of amendment request: The notice relates to your 
November 14, 1995, application to amend the Technical Specifications to 
provide a one-time exception to the Technical Specification 3.9.12, 
``Fuel Building Storage Air Cleanup System,'' to allow the fuel storage 
building air cleanup system to be inoperable during intervals in which 
new fuel rack modules will be moved into and old fuel modules will be 
moved out of the fuel storage building.
    Date of pulbication of individual notice in Federal Register: 
November 28, 1995 (60 FR 58688)
    Expiration date of individual notice: December 28, 1995
    Local Public Document Room location: Russell Library, 123 Broad 
Street, Middletown, CT 06457.
    Attorney for licensee: Lillian M. Cuoco, Esq., Senior Nuclear 
Counsel, Northeast Utilities Service Company, P.O. Box 270, Hartford, 
CT 06141-0270.
    NRC Project Director: Phillip F. McKee

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

    Date of amendment request: January 12, 1995, as supplemented by 
letter dated June 29, 1995
    Description of amendment request: The proposed amendments would 
modify portions of Technical Specification Section 6.0, 
``Administrative Controls.''
    Date of publication of individual notice in Federal Register: 
November 24, 1995, (60 FR 58109)
    Expiration date of individual notice: December 26, 1995
    Local Public Document Room location: York County Library, 138 East 
Black Street, Rock Hill, South Carolina 29730

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

    Date of amendment request: September 5, 1995
    Description of amendment request: The proposed amendments would 
revise the Updated Final Safety Analysis Report.
    Date of publication of individual notice in Federal Register: 
November 28, 1995 (60 FR 58690)
    Expiration date of individual notice: December 28, 1995
    Local Public Document Room location: York County Library, 138 East 
Black Street, Rock Hill, South Carolina 29730

Niagara Mohawk Power Corporation, Docket Nos. 50-220 and 50-410, 
Nine Mile Point Nuclear Station, Unit Nos. 1 and 2, Oswego County, 
New York

    Date of amendments request: October 25, 1995
    Description of amendments request: The proposed amendments would 
change position titles and reassign responsibilites at the upper 
management level to reflect a restructuring of Niagara Mohawk's upper 
management organization.
    Date of publication of individual notice in Federal Register: 
November 16, 1995 (60 FR 57605)
    Expiration date of individual notice: December 18, 1995
    Local Public Document Room location: Reference and Documents 
Department, Penfield Library, State University of New York, Oswego, New 
York 13126.

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

    Date of amendment request: October 3, 1995
    Description of amendment request: The notice relates to your 
October 3, 1995, application to amend the Technical Specifications to 
remove the Limiting Condition for Operation (LCO) and Surveillance 
Requirements for the loss-of-normal power (LNP) trip function from 
Tables 3.2.2 and 4.2.1 and insert new LCO 3.2.F and Surveillance 
Requirement 4.2.F. In addition, the proposed amendment will add a new 
table to specify the required LNP instrumentation for each bus, will 
update the Table of Contents, will make some editorial changes, and 
will revise the associated Bases section.
    Date of publication of individual notice in Federal Register: 
December 4, 1995 (60 FR 62111).
    Expiration date of individual notice: January 3, 1996
    Local Public Document Room location: Learning Resources Center, 
Three Rivers Community-Technical College, 574 New London Turnpike, 
Norwich, CT 06360.

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.

Arizona Public Service Company, et al., Docket Nos. STN 50-528, STN 
50-529, and STN 50-530, Palo Verde Nuclear Generating Station, 
Units 1, 2, and 3, Maricopa County, Arizona

    Date of application for amendments: July 3, 1995
    Brief description of amendments: The amendment temporarily adds new 
Action Statements 3.8.1.1.f and 3.8.1.1.g to Technical Specification 
3.8.1.1, ``A.C. Sources - Operating,'' to provide a method of 
responding to sustained degraded voltage. Also, Bases 3/4.8.1, 3/4.8.2, 
and 3/4.8.3 (A.C. Sources,'' ``D.C. Sources,'' and ``Onsite 
Distribution 

[[Page 65690]]
Systems,'' respectively) are being revised to provide guidance on how 
and why degraded offsite power voltage and the number of startup 
transformers in service affect compliance to GDC 17 and to give the 
basis for the additional action statements.
    Date of issuance: November 28, 1995
    Effective date: November 28, 1995
    Amendment Nos.:  Unit 1 - Amendment No. 102; Unit 2 - Amendment No. 
90; Unit 3 - Amendment No. 73
    Facility Operating License Nos. NPF-41, NPF-51, and NPF-74: The 
amendments revised the Technical Specifications.
    Date of initial notice in Federal Register: August 2, 1995 (60 FR 
39431) The Commission's related evaluation of the amendments is 
contained in a Safety Evaluation dated November 28, 1995.No significant 
hazards consideration comments received: No.
    Local Public Document Room location:Phoenix Public Library, 12 East 
McDowell Road, Phoenix, Arizona 85004

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: August 30, 1994, as 
supplemented August 4, 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 3/4.2, 
``Instrumentation.''Date of issuance: November 20, 1995
    Effective date: Immediately, to be implemented no later than June 
30, 1996.
    Amendment Nos.:  142, 136, 164, and 160
    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 30, 1995 (60 FR 
45177) The Commission's related evaluation of the amendments is 
contained in a Safety Evaluation dated November 20, 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, IllinoisDocket 
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 20, 1995.
    Brief description of amendments: This application upgrades the 
current custom Technical Specifications (TS) for Dresden and Quad 
Cities to the Standard Technical Specifications (STS) contained in 
NUREG-0123, ``Standard Technical Specification General Electric Plants 
BWR/4.'' This application upgrades only Section 3/4.7, ``Containment 
Systems.''
    Date of issuance: November 27, 1995
    Effective date: Immediately, to be implemented no later than June 
30, 1996, for Dresden Station and June 30, 1996, for Quad Cities 
Station.
    Amendment Nos.: 143, 137, 165, 161
    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 2, 1995 (60 FR 
39433) The Commission's related evaluation of the amendments is 
contained in a Safety Evaluation dated November 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.

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: 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 1, 1995
    Effective date: As of the date of issuance to be implemented within 
30 days
    Amendment Nos.: 211, 211, and 208
    Facility Operating License Nos. DPR-38, DPR-47, and DPR-55: The 
amendments revised the Technical Specifications.
    Date of initial notice in Federal Register: March 15, 1995 (60 FR 
14020) 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 signficant hazards consideration determination. The 
Commission's related evaluation of the amendments is contained in a 
Safety Evaluation dated December 1, 1995No significant hazards 
consideration comments received: No
    Local Public Document Room location: Oconee County Library, 501 
West South Broad Street, Walhalla, South Carolina 29691

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: September 1, 1995, as 
supplemented by letter dated November 15, 1995
    Brief description of amendments: The amendments revise Technical 
Specification (TS) 6.9.2 to include references to updated or recently 
approved mathodologies used to calculate cycle-specific limits 
contained in the Core Operating Limits Report. The subject references 
have previously been reviewed and approved by the NRC staff.
    Date of Issuance: December 4, 1995
    Effective date: As of the date of issuance to be implemented within 
30 days
    Amendment Nos.: 212, 212, 209
    Facility Operating License Nos. DPR-38, DPR-47, and DPR-55: The 
amendments revised the Technical Specifications.
    Date of initial notice in Federal Register: October 11, 1995 (60 FR 
52928) The November 15, 1995, letter 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 4, 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

[[Page 65691]]


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: July 26, 1995, as supplemented 
by letter dated October 4, 1995
    Brief description of amendments: These amendments concern revising 
certain surveillance intervals and allowable outage times for the RPS 
and ESFAS equipment.
    Date of issuance: November 29, 1995
    Effective date: November 29, 1995Amendment Nos. 179 and 173Facility 
Operating Licenses Nos. DPR-31 and DPR-41: Amendments revised the 
Technical Specifications.
    Date of initial notice in Federal Register: October 25, 1995 (60 FR 
54720) The Commission's related evaluation of the amendments is 
contained in a Safety Evaluation dated November 29, 1995 No significant 
hazards consideration comments received: No
    Local Public Document Room location:Florida International 
University, University Park, Miami, Florida 33199.
GPU Nuclear Corporation, et al., Docket No. 50-289, Three Mile 
Island Nuclear Station, Unit No. 1, Dauphin County, Pennsylvania

    Date of application for amendment: May 24, 1995, as supplemented 
July 24, 1995
    Brief description of amendment: The amendment revises the Technical 
Specifications to extend the test interval for the source range neutron 
flux instrumentation from 7 days prior to startup to 6 months prior to 
startup.
    Date of Issuance: November 24, 1995
    Effective date: As of its date of issuance, to be implemented 
within 30 days.
    Amendment No.: 199
    Facility Operating License No. DPR-50. Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: June 21, 1995 (60 FR 
32365) The July 24, 1995, letter provided clarifying information that 
did not change the initial proposed no significant hazards 
consideration determination. The Commission's related evaluation of 
this amendment is contained in a Safety Evaluation dated November 24, 
1995.No significant hazards consideration comments received: No.
    Local Public Document Room location: Law/Government Publications 
Section, State Library of Pennsylvania, (REGIONAL DEPOSITORY) Walnut 
Street and Commonwealth Avenue, Box 1601, Harrisburg, PA 17105.

Nebraska Public Power District, Docket No. 50-298, Cooper Nuclear 
Station, Nemaha County, Nebraska

    Date of amendment request: June 14, 1993, supplemented April 12, 
1994
    Brief description of amendment: The amendment revised the technical 
specifications (TSs) to include wording consistent with 10 CFR Part 20, 
and to deleted TSs governing miscellaneous radioactive material sealed 
sources.
    Date of issuance: November 28, 1995
    Effective date: November 28, 1995
    Amendment No.: 174
    Facility Operating License No. DPR-46. Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: September 1, 1993 (58 
FR 46237) The Commission's related evaluation of the amendment is 
contained in a Safety Evaluation dated November 28, 1995.No significant 
hazards consideration comments received: No
    Local Public Document Room location:Auburn Public Library, 118 15th 
Street, Auburn, NE 68305.

North Atlantic Energy Service Corporation, Docket No. 50-443, 
Seabrook Station, Unit No. 1, Rockingham County, New Hampshire

    Date of amendment request: July 24, 1995, as supplemented by letter 
dated October 30, 1995.
    Description of amendment request: The amendment revises the 
Appendix A Technical Specifications (TS) relating to reactor coolant 
system leakage. Specifically, the amendment deletes Table 3.4-1, 
``Reactor Coolant System Pressure Isolation Valves'' from the Seabrook 
Station, Unit No. 1 TS section 3.4.6.2. Also, reference to Table 3.4-1 
is deleted from Limiting Condition for Operation 3.4.6.2 f and from 
Surveillance Requirement 4.4.6.2.2. The information contained in Table 
3.4-1 is to be relocated to the Technical Requirements Manual. 
Additionally, a footnote providing certain exceptions from the 
requirements of SR 4.4.6.2.2d for the RHR Pump A and RHR Pump B Suction 
Isolation Valves previously located on Table 3.4-1 is relocated as a 
footnote to SR 4.4.6.2.2d.
    Date of issuance: November 28, 1995
    Effective date: As of its date of issuance, to be implemented 
within 60 days.
    Amendment No.: 44
    Facility Operating License No. NPF-86. Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: August 30, 1995 (60 FR 
45180). The licensee's letter dated October 30, 1995, provided a minor 
revision to the application that was within the scope of the original 
notice and did not change the initial proposed no significant hazards 
consideration determination. The October 30, 1995, letter also 
contained a request for an additional change that will be addressed 
separately.The Commission's related evaluation of the amendment is 
contained in a Safety Evaluation dated November 28, 1995.No significant 
hazards consideration comments received: No.
    Local Public Document Room location:Exeter Public Library, Founders 
Park, Exeter, NH 03833.

North Atlantic Energy Service Corporation, Docket No. 50-443, 
Seabrook Station, Unit No. 1, Rockingham County, New Hampshire

    Date of amendment request: September 20, 1995
    Description of amendment request: The amendment modifies the 
Appendix A Technical Specifications for the Engineered Safety Features 
Actuation System Instrumentation. Specifically, the amendment revises 
the Seabrook Station Technical Specifications to relocate Functional 
Unit 6.b, ``Feedwater Isolation - Low RCS Tavg Coincident with a 
Reactor Trip'' from Technical Specification 3.3.2. ``Engineered Safety 
Features Actuation System Instrumentation'' to the Technical 
Requirements Manual which is a licensee controlled document.
    Date of issuance: November 29, 1995
    Effective date: As of its date of issuance, to be implemented 
within 60 days.
    Amendment No.: 45
    Facility Operating License No. NPF-86. Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: October 24, 1995 (60 FR 
54524). The Commission's related evaluation of the amendment is 
contained in a Safety Evaluation dated November 29, 1995.No significant 
hazards consideration comments received: No.
    Local Public Document Room location:Exeter Public Library, Founders 
Park, Exeter, NH 03833.

North Atlantic Energy Service Corporation, Docket No. 50-443, 
Seabrook Station, Unit No. 1, Rockingham County, New Hampshire

    Date of amendment request: June 7, 1995.
    Description of amendment request: The amendment increases the 
temperature limit, as specified by the 

[[Page 65692]]
footnotes to Technical Specification Surveillance Requirement 4.4.7 and 
to Table 3.4-2, above which reactor coolant sampling and analysis for 
dissolved oxygen is required and dissolved oxygen limits apply.
    Date of issuance: November 29, 1995
    Effective date: November 29, 1995
    Amendment No.: 46
    Facility Operating License No. NPF-86. Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: July 19, 1995 (60 FR 
37098). The Commission's related evaluation of the amendment is 
contained in a Safety Evaluation dated November 29, 1995.No significant 
hazards consideration comments received: No.
    Local Public Document Room location:Exeter Public Library, Founders 
Park, Exeter, NH 03833.

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

    Date of application for amendment: June 8, 1995
    Brief description of amendment: The amendment modifies Surveillance 
Requirement (SR) 4.5.1.c and deletes Technical Specification (TS) 3/
4.8.4.3, ``AC Circuits Inside Containment.'' The changes to SR 4.5.1.c 
clarify the requirements for securing the safety injection accumulator 
isolation valve breakers (3SIL*MV8808A, B, C, and D) in the tripped 
position for the applicable modes. The amendment also deletes TS 3/
4.8.4.3 since reasonable assurance is provided to protect the 
electrical penetrations and penetration conductors against an 
overcurrent condition and single failure of a circuit breaker.
    Date of issuance: November 29, 1995
    Effective date: As of the date of issuance, to be implemented 
within 60 days.
    Amendment No.: 121
    Facility Operating License No. NPF-49. Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: August 2, 1995 (60 FR 
39444) The Commission's related evaluation of the amendment is 
contained in a Safety Evaluation dated November 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.

Philadelphia Electric Company, Docket No. 50-353, Limerick 
Generating Station, Unit 2, Montgomery County, Pennsylvania

    Date of application for amendment: June 23, 1995
    Brief description of amendment: This amendment involves a one-time 
change affecting the Allowed Outage Time (AOT) for the Emergency 
Service Water (ESW) system, Residual Heat Removal Service Water (RHRSW) 
System, the Suppression Pool Cooling, the Suppression Pool Spray, and 
Low Pressure Coolant Injection modes of the Residual Heat Removal 
System, and Core Spray System to be extended from 3 and 7 days to 14 
days during the Unit 2 refueling outage scheduled to begin in January 
1996. This proposed extended AOT allows adequate time to install 
isolation valves and cross-ties on the ESW and RHRSW Systems to 
facilitate future inspections or maintenance.
    Date of issuance: November 30, 1995
    Effective date: November 30, 1995
    Amendment No. 70
    Facility Operating License No. NPF-85. This amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: August 2, 1995 (60 FR 
39448) The Commission's related evaluation of the amendment is 
contained in a Safety Evaluation dated November 30, 1995No significant 
hazards consideration comments received: No
    Local Public Document Room location:Pottstown Public Library, 500 
High Street, Pottstown, Pennsylvania 19464.

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

    Date of application for amendment: March 3, 1995, as supplemented 
April 12, 1995, and November 20, 1995.
    Brief description of amendment: The amendment revises the TS to 
extend the calibration frequency for the following:
    (1) Containment water level monitor instrumentation (specified in 
TS Table 4.1-1)
    (2) Containment building ambient temperature sensors (specified in 
TS Table 4.1-1)
    (3) Seismic monitoring instrumentation (specified in TS Table 4.10-
2)
    In addition, the amendment added a new surveillance requirement to 
TS Table 4.1-1 for testing the core exit thermocouples.
    These changes allow operation on a 24-month fuel cycle and follow 
the guidance provided in Generic letter 91-04, ``Changes in Technical 
Specification Surveillance Intervals to Accommodate a 24-Month Fuel 
Cycle,'' as applicable.
    Date of issuance: December 1, 1995
    Effective date: As of the date of issuance to be implemented within 
30 days.
    Amendment No.: 164
    Facility Operating License No. DPR-64: Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: May 10, 1995 (60 FR 
24917) The April 12 and November 20, 1995, letters provided clarifying 
information that did not change the initial proposed no significant 
hazards consideration. The Commission's related evaluation of the 
amendment is contained in a Safety Evaluation dated December 1, 1995.No 
significant hazards consideration comments received: No
    Local Public Document Room location:White Plains Public Library, 
100 Martine Avenue, White Plains, New York 10610.

Public Service Electric & Gas Company, Docket No. 50-354, Hope 
Creek Generating Station, Salem County, New Jersey

    Date of application for amendment: July 27, 1995
    Brief description of amendment: The amendment changes the Technical 
Specifications to incorporate updated pressure vs. temperature 
operating limit curves.
    Date of issuance: November 28, 1995
    Effective date: As of the date of issuance to be implemented within 
60 days
    Amendment No.: 88
    Facility Operating License No. NPF-57: This amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: September 13, 1995 (60 
FR 47624) The Commission's related evaluation of the amendment is 
contained in a Safety Evaluation dated November 28, 1995. No 
significant hazards consideration comments received: No
    Local Public Document Room location:Pennsville Public Library, 190 
S. Broadway, Pennsville, New Jersey 08070

Public Service Electric & Gas Company, Docket No. 50, Hope Creek 
Generating Station, Salem County, New Jersey

    Date of application for amendment: March 31, 1994, supplemented by 
letters dated August 29, and October 16, 1995.
    Brief description of amendment: This amendment changes Technical 
Specification (TS) 3.5.1, ``ECCS - 

[[Page 65693]]
Operating,'' and associated Bases, to establish a new allowed out-of-
service time. Action c.2 for TS 3.5.1 allows any one Low Pressure 
Coolant Injection subsystem, or one Core Spray subsystem, to be 
inoperable in addition to an inoperable High Pressure Coolant Injection 
system, for 72 hours.
    Date of issuance: November 30, 1995
    Effective date: As of the date of issuance and shall be implemented 
within 60 days.
    Amendment No.: 89
    Facility Operating License No. NPF-57: This amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: June 8, 1994 (59 FR 
29631).The supplemental letters did not change the NRC staff's proposed 
no significant hazards consideration determination.The Commission's 
related evaluation of the amendment is contained in a Safety Evaluation 
dated November 30, 1995.No significant hazards consideration comments 
received: No
    Local Public Document Room location:Pennsville Public Library, 190 
S. Broadway, Pennsville, New Jersey 08070
The Cleveland Electric Illuminating Company, Centerior Service 
Company, Duquesne Light Company, Ohio Edison Company, Pennsylvania 
Power Company, Toledo Edison Company, Docket No. 50-440, Perry 
Nuclear Power Plant, Unit No. 1, Lake County, Ohio
    Date of application for amendment: March 24, June 9, and June 30, 
1995
    Brief description of amendment: The amendment revised the Technical 
Specifications to allow a one-time extension for the performance of 
certain Surveillance Requirements (SRs). Affected SRs include 
penetration leak rate testing, valve operability testing, instrument 
calibration, response time testing, and logic system functional tests. 
The proposed changes are to support refueling outage 5 scheduled to 
begin no later than February 15, 1996.
    Date of issuance: November 29, 1995
    Effective date: November 29, 1995
    Amendment No. 75
    Facility Operating License No. NPF-58: This amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: May 10, 1995 (60 FR 
24919) and August 16, 1995 (60 FR 42612)The Commission's related 
evaluation of the amendment is contained in a Safety Evaluation dated 
November 29, 1995. No significant hazards consideration comments 
received: No
    Local Public Document Room location:Perry Public Library, 3753 Main 
Street, Perry, Ohio 44081

The Cleveland Electric Illuminating Company, Centerior Service 
Company, Duquesne Light Company, Ohio Edison Company, Pennsylvania 
Power Company, Toledo Edison Company, Docket No. 50-440, Perry 
Nuclear Power Plant, Unit No. 1, Lake County, Ohio

    Date of application for amendment: October 21, 1994
    Brief description of amendment: The amendment revised Technical 
Specification 3/4.6.1.2, ``Primary Containment Leakage,'' and its 
associated Bases to reflect the partial exemptions to the requirements 
of 10 CFR Part 50, Appendix J, Sections III.A.5(b)(2), III.B.3, 
III.C.3, III.A.1(d), III.D.1(a), and III.D.3 that were granted by the 
NRC on December 4, 1995.
    Date of issuance: December 8, 1995
    Effective date: ]December 8, 1995
    Amendment No.: 76
    Facility Operating License No. NPF-58: This amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: August 16, 1995 (60 FR 
42611) The Commission's related evaluation of the amendment is 
contained in a Safety Evaluation dated December 8, 1995.No significant 
hazards consideration comments received: No
    Local Public Document Room location: Perry Public Library, 3753 
Main Street, Perry, Ohio 44081

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 application for amendment: July 28, 1995
    Brief description of amendment: The amendment clarifies the 
limiting condition for operation for TS 3.8.1.1 and 3.8.1.2 from 
``independent'' circuit to ``qualified'' circuit; explains in the Bases 
the requirements for operability of an offsite circuit; deletes the 
STAGGERED TEST BASIS scheduling requirement to perform emergency diesel 
generatorsurveillances; explains in the Bases an acceptable method for 
verification of Emergency Diesel Generator speed for surveillance 
requirements (SR) 4.8.1.1.2.a.4 and 4.8.1.1.2.c.4; removes a 
surveillance test extension that has expired for SR 4.8.1.1.1.b; adds 
an exception for SR 4.8.1.1.2.c.5 and 4.8.1.1.2.c.7 to SR 4.8.1.2; and 
revises Bases 3.0.5 to reflect the clarification from ``independent'' 
circuit to ``qualified'' circuit.
    Date of issuance: December 8, 1995
    Effective date: December 8, 1995
    Amendment No.: 203
    Facility Operating License No. NPF-3. Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: November 8, 1995 (60 FR 
56370) The Commission's related evaluation of the amendment is 
contained in a Safety Evaluation dated December 8, 1995.No significant 
hazards consideration comments received: No
    Local Public Document Room location:University of Toledo, William 
Carlson Library, Government Documents Collection, 2801 West Bancroft 
Avenue, Toledo, Ohio 43606.

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 application for amendment: October 2, 1995
    Brief description of amendment: The amendment revises Technical 
Specification (TS) Section 5.0, ``Design Features,'' by adding a site 
location description, removing site area maps, removing containment and 
reactor coolant system design parameters, removing the description of 
the meteorological tower location, removing component cyclic or 
transient limits, and revising the fuel assembly description to include 
the use of ZIRLO clad fuel rods.
    Date of issuance: December 8, 1995
    Effective date: December 8, 1995
    Amendment No.: 204
    Facility Operating License No. NPF-3. Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: November 8, 1995 (60 FR 
56371) The Commission's related evaluation of the amendment is 
contained in a Safety Evaluation dated December 8, 1995.No significant 
hazards consideration comments received: No
    Local Public Document Room location:University of Toledo, William 
Carlson Library, Government Documents Collection, 2801 West Bancroft 
Avenue, Toledo, Ohio 43606.

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

    Date of application for amendment: June 23, 1995
    Brief description of amendment: The amendment revises Technical 

[[Page 65694]]
    Specification (TS) 4.1.3.1.2, 4.4.6.2.2.b, 4.4.3.2, 4.6.2.1.d, 4.6.4.2, 
and Table 4.3-3 in accordance with guidance provided in NRC Generic 
Letter (GL) 93-05, ``Line Item Technical Specification Improvements to 
Reduce Surveillance Requirements for Testing During Power Operations.'' 
Additionally, the amendment revises TS 4.1.1.1.1, 4.1.1.2, 3/4.1.3.1 
and the associated Bases to implement portions of NUREG-1431, 
``Standard Technical Specifications - Westinghouse Plants.''
    Date of issuance: December 7, 1995
    Effective date: December 7, 1995
    Amendment No.: 105
    Facility Operating License No. NPF-30. The amendment revises the 
Technical Specifications.
    Date of initial notice in Federal Register: August 30, 1995 (60 FR 
45187). The Commission's related evaluation of the amendment is 
contained in a Safety Evaluation dated December 7, 1995.No significant 
hazards consideration comments received: No.
    Local Public Document Room location:Callaway County Public Library, 
710 Court Street, Fulton, Missouri 65251.
    Dated at Rockville, Maryland, this 13th day of December 1995.For 
the Nuclear Regulatory Commission
Steven A. Varga,
Director, Division of Reactor Projects - I/II, Office of Nuclear 
Reactor Regulation
[Doc. 95-30755 Filed 12-19-95; 8:45 am]
BILLING CODE 7590-01-F