[Federal Register Volume 60, Number 196 (Wednesday, October 11, 1995)]
[Notices]
[Pages 52926-52943]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: X95-11011]



=======================================================================
-----------------------------------------------------------------------

[[Page 52927]]


NUCLEAR REGULATORY COMMISSION


Biweekly Notice

Applications and Amendments to Facility Operating LicensesInvolving 
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 September 16, through September 28, 1995. 
The last biweekly notice was published on Septmeber 27, 1995 (60 FR 
49929).

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 November 10, 1995, 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 

[[Page 52928]]
contention will not be permitted to participate as a party.
    Those permitted to intervene become parties to the proceeding, 
subject to any limitations in the order granting leave to intervene, 
and have the opportunity to participate fully in the conduct of the 
hearing, including the opportunity to present evidence and cross-
examine witnesses.
    If a hearing is requested, the Commission will make a final 
determination on the issue of no significant hazards consideration. The 
final determination will serve to decide when the hearing is held.
    If the final determination is that the amendment request involves 
no significant hazards consideration, the Commission may issue the 
amendment and make it immediately effective, notwithstanding the 
request for a hearing. Any hearing held would take place after issuance 
of the amendment.
    If the final determination is that the amendment request involves a 
significant hazards consideration, any hearing held would take place 
before the issuance of any amendment.
    A request for a hearing or a petition for leave to intervene must 
be filed with the Secretary of the Commission, U.S. Nuclear Regulatory 
Commission, Washington, DC 20555, 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.

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

    Date of amendment request: August 10, 1995
    Description of amendment request: The proposed amendment will add a 
footnote to Technical Specification (TS) Section 3/4.4.3, 
``Pressurizer,'' to allow the pressurizer level to be controlled, 
outside of the programmed level, between 25 to 50 percent, plus or 
minus 5 percent in Mode 3 when the reactor coolant system is borated to 
the required Mode 5 concentrations.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:
    ...The proposed change does not involve an SHC because the 
change would not:
    1. Involve a significant increase in the probability or 
consequences of an accident previously evaluated.
    The design basis accidents analyzed in Mode 3 are steam line 
break, control rod withdrawal from subcritical, boron dilution and 
control rod ejection. Of these four analyzed accidents, the relaxing 
of the pressurizer level requirement can only impact the steam line 
break accident analyses. The initial pressurizer level can impact 
the timing of the safety injection signal and the subsequent boron 
addition from the HPSI [high pressure safety injection] system. The 
proposed change requires that the boron concentration be equal to 
the Mode 5 required concentration in order for the pressurizer level 
to be higher than the current requirement. The Mode 5 boron 
concentration ensures that there is sufficient negative reactivity 
in the core due to boron that a steam line break from this condition 
would not need the boron addition from the HPSI system and would be 
bounded by the design basis analyses. Thus the proposed change 
cannot increase the probability or consequences of the design basis 
accidents.
    2. Create the possibility of a new or different kind of accident 
from any previously analyzed.
    The proposed change only modifies the Mode 3 pressurizer level 
requirement. This change does not impact the lower bound but 
provides flexibility to the plant operators in the maximum 
pressurizer level. The upper limit still provides margin to 
pressurizer overfill. This cannot cause an accident nor introduce a 
new type of malfunction. The modified level would allow for a higher 
initial pressurizer level in Mode 3. This higher level is already 
used in the accident analyses which result in an increase in 
pressurizer level. Therefore, the change does not modify the plant's 
response to accidents.
    3. Involve a significant reduction in the margin of safety.
    The proposed change is consistent with or bounded by the design 
basis analyses. The higher shutdown margin required in order to 
relax the upper bound of the pressurizer level assures that a steam 
line break from these conditions is bounded by the design basis 
analyses. Therefore, the proposed change cannot impact 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

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

    Date of amendment request: September 1, 1995
    Description of amendment request: Generic Letter 88-16 provided 
guidance on removing cycle-specific parameters which are calculated 
using NRC-approved methodologies from the Technical Specifications 
(TS). The parameters are replaced in the TS with a reference to a named 
report which contains the parameters, and a requirement that the 
parameters remain within the limits specified in the report. The 
proposed changes incorporate NRC-approved methodologies, approved 
revisions to previously approved methodologies, or republished versions 
of previously approved methodologies into Section 6.9.2 of the Oconee 
TS. The limits to which these methodologies are applied are 1) Axial 
Power Imbalance Protective Limits and Variable Low RCS Pressure 
Protective Limits, 2) Reactor Protective System Trip Setting Limits for 
the Flux/Flow/Imbalance and Variable Low Reactor Coolant System 
Pressure Trip functions, and 3) Power Imbalance Limits. Since the 
proposed changes only incorporate NRC-approved methodologies into the 
TS, the licensee proposed that the changes are administrative in nature 
and can be 

[[Page 52929]]
assumed to have no impact, or potential impact, on the health and 
safety of the public.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:
    The proposed changes will not create a significant hazards 
consideration, as defined by 10 CRF 50.92, because:
    1) The proposed changes will not involve a significant increase 
in the probability or consequences of an accident previously 
evaluated.
    The proposed changes are administrative in nature, and do not 
affect any system, procedure, or manipulation of any equipment which 
could affect the probability or consequences of any accident.
    2) The proposed changes will not create the possibility of any 
new or different kind of accident from any accident previously 
evaluated.
    The proposed changes are administrative in nature, and cannot 
introduce any new failure mode or transient which could create any 
accident.
    3) The proposed changes will not involve a significant reduction 
in a margin of safety.
    The proposed changes are administrative in nature, and will not 
affect any operating parameters or limits which could result in a 
reduction in a margin of safety.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: Oconee County Library, 501 
West South Broad Street, Walhalla, South Carolina 29691
    Attorney for licensee: J. Michael McGarry, III, Winston and Strawn, 
1200 17th Street, NW., Washington, DC 20036
    NRC Project Director: Herbert N. Berkow

Entergy Operations, Inc., et al., Docket No. 50-416, Grand Gulf 
Nuclear Station, Unit 1, Claiborne County, Mississippi

    Date of amendment request: November 9, 1994, as supplemented by 
letter dated August 4, 1995
    Description of amendment request: This supplement revises the 
licensee's November 9, 1994, application by updating the request to 
reflect implementation of the Improved Standard Technical 
Specifications on March 20, 1995, and by deleting the request for a 
definition of the term RECENTLY IRRADIATED FUEL. The proposed amendment 
revises those specifications associated with various engineered safety 
feature systems following a design basis fuel handling accident. The 
proposed changes affect conditions where irradiated fuel is handled in 
the primary or secondary containment and when fuel is handled over the 
reactor vessel with fuel in the vessel. These changes are based on a 
recent re-analysis of the fuel handling accident for Grand Gulf Nuclear 
Station (GGNS).
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:
    1. The proposed changes do not significantly increase the 
probability or consequences of an accident previously evaluated.
    A new term to describe irradiated fuel is used to establish 
operational conditions where specific activities represent 
situations where significant radioactive releases can be postulated. 
These operational conditions are consistent with the design basis 
analysis. Because the equipment affected by the revised operational 
conditions is not considered an initiator to any previously analyzed 
accident, inoperability of the equipment cannot increase the 
probability of any previously evaluated accident. The proposed 
requirements in conjunction with existing administrative controls on 
light loads, bounds the conditions of the current design basis fuel 
handling accident analysis which concludes that the radiological 
consequences are within the acceptance criteria of NUREG 0800, 
Section 15.7.4 and General Design Criteria 19. Therefore, the 
proposed changes do not significantly increase the probability or 
consequences of any previously evaluated accident.
    Based on the above, the proposed changes do not significantly 
increase the probability or consequences of any accident previously 
evaluated.
    2. The proposed changes would not create the possibility of a 
new or different kind of accident from any previous analyzed.
    The new term to describe irradiated fuel is used to establish 
operational conditions where specific activities represent 
situations where significant radioactive releases can be postulated. 
These operational conditions are consistent with the design basis 
analysis. The proposed changes do not introduce any new modes of 
plant operation and do not involve physical modification of the 
plant. Therefore, the proposed changes do not create the possibility 
of a new or different kind of accident from any previous analyzed.
    Therefore, the proposed changes do not create the possibility of 
a new or different kind of accident from any accident previously 
analyzed.
    3. The proposed changes do not involve a significant reduction 
in a margin of safety.
    The new term to describe irradiated fuel is used to establish 
operational conditions where specific activities represent 
situations where significant radioactive releases can be postulated. 
These operational conditions are consistent with the design basis 
analysis and are established such that the radiological consequences 
are at or below the current GGNS licensing limit. Safety margins and 
analytical conservatisms have been evaluated and are well 
understood. Substantial margins are retained to ensure that the 
analysis adequately bounds all postulated event scenarios. The 
proposed change only eliminates the excess margin from the analysis. 
The current margin of safety is retained.
    Specifically, the margin of safety for the fuel handling 
accident is the difference between the 10 CFR 100 limits and the 
licensing limit defined by NUREG 0800, Section 15.7.4. With respect 
to the control room personnel doses, the margin of safety is the 
difference between the 10 CFR 100 limits and the licensing limit 
defined by 10 CFR 50, Appendix A, Criterion 19 (GDC 19). Excess 
margin is the difference between the postulated doses and the 
corresponding licensing limit.
    The proposed applicability continues to ensure that the
    whole-body and thyroid dose at the exclusion area and low 
population zone boundaries as well as control room, doses are at or 
below the corresponding licensing limit. The margin of safety is 
unchanged; therefore, the proposed changes do not involve a 
significant reduction in a margin of safety.
    Therefore, the proposed changes do not result in a significant 
reduction in a margin of safety.
    Based on the above evaluation, operation in accordance with the 
proposed amendment involves no significant hazards considerations.
    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: Judge George W. Armstrong 
Library, 220 S. Commerce Street, Natchez, MS 39120
    Attorney for licensee: Nicholas S. Reynolds, Esquire, Winston and 
Strawn, 1400 L Street, N.W., 12th Floor, Washington, DC 20005-3502
    NRC Project Director: William D. Beckner

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

    Date of amendment request: July 19, 1995
    Description of amendment request: The proposed amendment reduces 
requirements associated with the exercise frequency of control element 
assemblies from once per 31 days to once per 92 days.
    Basis for proposed no significant hazards consideration 
determination: 

[[Page 52930]]
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 not Involve a Significant Increase in the Probability or 
Consequences of an Accident Previously Evaluated.
    Changing the frequency of the control element assemblies (CEA) 
exercise test surveillance introduces no new failure mechanism for 
the system, so the consequences of a postulated stuck CEA are no 
different than those previously evaluated.
    As explained in NUREG-1366, ``Improvements to Technical 
Specifications Surveillance Requirements,'' the purpose of this test 
is to identify immovable CEAs. NUREG-1366 goes on to explain that 
the majority of CEA problems are identified during the performance 
of startup physics testing and during CEA withdrawal for startup, 
not during the exercise test. The incidence of electrical 
malfunctions which will still allow CEA insertion is much greater 
than the incidence of mechanically bound CEAs. As stated in NUREG-
1366, there has only been one incidence of multiple CEAs failing to 
fully insert upon a reactor trip (Point Beach Nuclear Plant, May 
1985) and in this case the two affected CEAs partially inserted. 
Based on this history, simply reducing the test frequency will not 
increase the probability of a stuck CEA.
    Therefore, this change does not involve a significant increase 
in the probability or consequences of any accident previously 
evaluated.
    2. Does Not Create the Possibility of a New or Different Kind of 
Accident from any Previously Evaluated.
    Because the proposed change does not alter the design, 
configuration, or method of operation of the plant, it does not 
create the possibility of a new or different kind of accident from 
any previously evaluated.
    3. Does Not Involve a Significant Reduction in the Margin of 
Safety.
    The proposed change does not alter the acceptance criteria of 
any surveillance requirements, alter any assumptions used in 
accident analysis, change any actuation setpoints, nor allow 
operations in any configuration not previously evaluated. This 
change in surveillance frequency is based on a satisfactory 
operating history of CEAs. Additionally, the number of problems 
created by this test when compared with the number of problems 
identified by this test indicate that reducing the test frequency 
will have no adverse impact on the continued safe operation of the 
unit.
    Therefore, this change does not involve a significant reduction 
in the margin of safety.
    Therefore, based upon the reasoning presented above and the 
previous discussion of the amendment request, Entergy Operations had 
determined that the requested change does not involve a significant 
hazards consideration.
    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: Tomlinson Library, Arkansas 
Tech University, Russellville, AR 72801
    Attorney for licensee: Nicholas S. Reynolds, Esquire, Winston and 
Strawn, 1400 L Street, N.W., Washington, DC 20005-3502
    NRC Project Director: William D. Beckner

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: September 11, 1995
    Description of amendment request: The licensee proposes to change 
Turkey Point Units 3 and 4 Technical Specifications (TS) to incorporate 
line-item improvements to Specifications 3/4.8.1, ``Electrical Power 
Systems-A.C. Sources,'' and the associated BASES. The licensee stated 
that the proposed changes are consistent with the guidance provided by 
the NRC in GL 93-05, ``Line-Item Technical Specifications Improvements 
to Reduce Surveillance Requirements for Testing During Power 
Operation,'' and the corresponding recommendations contained in NUREG-
1366, ``Improvements to Technical Specifications Surveillance 
Requirements.''
    In addition, line-item improvements are proposed following the 
guidance in GL 94-01, ``Removal of Accelerated Testing and Special 
Reporting Requirements for Emergency Diesel Generators.'' The 
implementation of a maintenance program for monitoring and maintaining 
Emergency Diesel Generator (EDG) performance for Turkey Point Units 3 
and 4, consistent with the provisions of 10 CFR 50.65 ``Requirements 
for Monitoring the Effectiveness of Maintenance at Nuclear Power 
Plants'' and the associated guidance of Regulatory Guide (RG) 1.160 
will be met by FPL within 90 days following issuance of the proposed 
amendments.
    The licensee also requested to revise the current wording used in 
the Turkey Point Units 3 and 4 TS to require testing of remaining 
required diesel generators ``[i]f the diesel generator became 
inoperable due to any cause other than planned preventative 
maintenance...''. The licensee requested that TS 3.8.1.1, ACTION 
statements b. and c. be amended such that the word 'preventative' is 
deleted. Deleting this wording will reduce unnecessary testing of 
diesel generators as a result of planned corrective maintenance.
    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 license amendments proposed for Turkey Point Units 3 and 4 
will incorporate line-item Technical Specification (TS) improvements 
for Emergency Diesel Generators (EDG) pursuant to guidance provided 
in Generic Letters (GL) 93-05 and 94-01. The EDGs are not accident 
initiators, the proposed TS changes do not involve any assumptions 
relative to accident initiators in the plant safety analyses, and 
therefore the proposed amendments will not impact the probability of 
occurrence for accidents previously analyzed.
    The EDG line-item TS improvements associated with GL 93-05 are 
based on recommendations designed to remove unwarranted requirements 
for testing during power operation and other factors that are 
counter-productive to safety in terms of equipment degradation and 
availability. These recommendations resulted from a comprehensive 
study of industry-wide EDG surveillance requirements and subsequent 
findings reported by the NRC in NUREG-1366. The proposed amendments 
are consistent with the guidance of GL 93-05 for implementing such 
recommendations as well as contemporary licensing actions by the NRC 
on other light water reactors.
    Similarly, GL 94-01 provides guidance for a line-item TS 
improvement that will remove accelerated testing requirements from 
the TS provided that the licensee commits to a maintenance program 
for monitoring and maintaining EDG performance that includes the 
applicable provisions of the maintenance rule (10 CFR 50.65). Such a 
program will further assure EDG availability. Since the availability 
of EDGs is assumed in certain success paths for mitigating analyzed 
accidents, an improvement in EDG availability will enhance accident 
mitigation capabilities.
    Therefore, 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.
    (2) Operation of the facility in accordance with the proposed 
amendments would not create the possibility of a new or different 
kind of accident from any accident previously evaluated.
    The proposed amendments incorporate line-item TS and other 
improvements to EDG surveillance testing requirements, and will not 
change the physical plant or the modes of plant operation defined in 
the Facility License. The changes do not involve the addition or 
modification of equipment, nor do they alter the design or methods 
of operation of plant systems. Plant configurations that are 
prohibited by TS will 

[[Page 52931]]
not be created by the amendments. Therefore, operation of the facility 
in accordance with the proposed amendment would not create the 
possibility of a new or different kind of accident from any accident 
previously evaluated.
    (3) Operation of the facility in accordance with the proposed 
amendments would not involve a significant reduction in a margin of 
safety.
    The proposed amendments are designed to improve EDG availability 
by eliminating unwarranted surveillance testing. The currently 
specified surveillance intervals are not changed, except to delete 
the requirement for accelerated testing under certain circumstances. 
The proposed changes do not otherwise alter the basis for any 
Technical Specification that is related to the establishment of, or 
the maintenance of a nuclear safety margin. Therefore, operation of 
the facility in accordance with the proposed amendment would not 
involve a significant reduction in a margin of safety.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: Florida International 
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

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

    Date of amendment request: July 24, 1995
    Description of amendment request: The proposed amendment would 
modify Technical Specification 3.6.C to allow up to 7 days to restore 
low pressure safety injection (LPSI) pump subsystem operability, and up 
to 24 hours to restore safety injection tank (SIT) operability.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration. The NRC staff has reviewed the licensee's analysis 
against the standards of 10 CFR 50.92(c). The staff's review is 
presented below:
    1. The proposed amendment does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated. The LPSI system is designed primarily to 
mitigate the consequences of a large loss-of-coolant accident 
(LOCA). Inoperable LPSI components are not accident initiators in 
any accident previously evaluated, and the proposed change does not 
affect any of the assumptions relative to accident initiators in the 
plant's safety analysis. Probabilistic safety analysis (PSA) methods 
were used to fully evaluate the extension of the LPSI system allowed 
outage time (AOT). The licensee asserts that the results of these 
analyses show no significant increase in the consequences of an 
accident previously evaluated. The SITs were designed to mitigate 
the consequences of a LOCA. The proposed amendment does not affect 
any of the assumptions used in the deterministic LOCA analysis. 
Probabilistic safety analysis methods were used to fully evaluate 
the effect of the SIT allowable outage time (AOT). The licensee 
asserts that the results of these analyses show no significant 
increase in the consequences of an accident previously evaluated. 
Thus, there is no significant increase in the probability or 
consequences of an accident previously evaluated.
    2. The proposed amendment does not create the possibility of a 
new or different kind of accident from any accident previously 
evaluated. The proposed amendment does not change the design, 
physical configuration, or modes of operation of the plant. Plant 
configurations that are prohibited by TS will not be created by this 
proposed amendment. Thus, the proposed amendment does not create the 
possibility or consequences of an accident previously evaluated.
    3. The proposed amendment does not involve a significant 
reduction in a margin of safety. The proposed amendment does not 
affect the limiting conditions for operation or the bases used in 
the deterministic analyses to establish the margin of safety. The 
licensee asserts that PSA methods were used to evaluate these 
changes and demonstrate that the changes are either risk neutral or 
risk beneficial. Thus, the proposed amendment does not involve a 
significant reduction in a margin of safety.
    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 this amendment request involves no significant hazards 
determination.
    Local Public Document Room location:  Wiscasset Public Library, 
High Street, P.O. Box 367, Wiscasset, ME 04578
    Attorney for licensee: Mary Ann Lynch, Esquire, Maine Yankee Atomic 
Power Company, 329 Bath Road, Brunswick, ME 04011
    NRC Project Director: Phillip F. McKee

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

    Date of amendment request: August 8, 1995
    Description of amendment request: The proposed amendment would 
modify the definition of Transthermal (Condition 4), Hot Shutdown 
(Condition 5), and Hot Standby (Condition 6) reactor operating 
conditions. The Transthermal and Hot Shutdown conditions are modified 
to establish an applicable range of subcriticality and be consistent 
with other Definitions. The wording of Hot Standby is modified to 
remove reference to control rod position, consistent with NUREG-1432, 
Standard Technical Specifications for Combustion Engineering Plants, 
Revision 1 dated April 1995.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration. The NRC staff has reviewed the licensee's analysis 
against the standards of 10 CFR 50.92(c). The staff's review is 
presented below:
    1. The proposed amendment does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated. The changes to these Definitions are 
administrative in nature. The Transthermal and Hot Shutdown 
conditions are changed by adding ``at least'' to establish a range 
of subcriticality. The current Definitions for the Transthermal and 
Hot Shutdown conditions set one minimum value for subcriticality; 
the change to these two Definitions would allow a range of values 
for subcriticality. All values of subcriticality that may be 
established by this change are below the current Definitions (more 
subcritical). The change to the wording of Hot Standby removes 
confusion about the Conditions during which control rods may be 
withdrawn and is consistent with current NRC guidance. All current 
plant analyses, requirements and acceptance criteria on 
subcriticality conditions remain in effect. The changes to these 
Definitions have no impact on event probabililty. Thus, the proposed 
amendment does not involve a significant increase in the probability 
or consequences of an accident previously evaluated.
    2. The proposed amendment does not create the possibility of a 
new or different kind of accident from any accident previously 
evaluated. The proposed amendment clarifies the subject Definitions. 
Limits on subcriticality requirements are unaffected, as are 
reactivity transients previously evaluated. Plant procedures 
currently require that minimum values for subcriticality be 
established. All values of subcriticality that may be established by 
this change are below the current Definitions (more subcritical). 
Further, the change to the wording of Hot Standbyis consistent with 
current NRC guidance. Thus, the proposed amendment does not create 
the possibility of a new or different kind of accident from any 
accident previously evaluated.
    3. The proposed amendment does not involve a significant 
reduction in a margin of safety. Adding the words ``at least'' to 
the Transthermal and Hot Shutdown conditions establishes a range of 
subcriticality to the 

[[Page 52932]]
Definitions for these terms. All values of subcriticality are below 
(more subcritical) than the current value, thus the margin of safety 
is increased. All current plant analyses, requirements and 
acceptance criteria on subcriticality conditions remain in effect. 
The change to the wording of Hot Standby removes confusion about the 
Conditions during which control rods may be withdrawn and is 
consistent with current NRC guidance. Thus, there is no significant 
reduction in a margin of safety.
    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 this amendment request involves no significant hazards 
consideration.
    Local Public Document Room location: Wiscasset Public Library, High 
Street, P.O. Box 367, Wiscasset, ME 04578
    Attorney for licensee: Mary Ann Lynch, Esquire, Maine Yankee Atomic 
Power Company, 329 Bath Road, Brunswick, ME 04011
    NRC Project Director: Phillip F. McKee

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

    Date of amendment request: August 30, 1995
    Description of amendment request: The proposed amendment would 
change Technical Specification (TS) 1.3.A, Reactor Core, to allow the 
use of fuel rods clad with zirconium alloy, rather than restrict fuel 
rod cladding to Zircaloy-4. In addition, the fuel enrichment limit 
described in this specification would be changed to more closely agree 
with the wording found in NUREG-1432, ``Standard Technical 
Specifications for Combustion Engineering Plants,'' dated April 1995.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration. The NRC staff has reviewed the licensee's analysis 
against the standards of 10 CFR 50.92(c). The staff's review is 
presented below:
    1. The proposed change does not involve a significant increase 
in the probability or consequences of an acident previously 
evaluated. Maine Yankee (MY) reload cores containing fuel rods clad 
with zirconium alloy and having higher fuel enrichments will be 
analyzed using NRC-approved methods and applicable acceptance 
criteria. In addition, the impact of fuel assembly design changes on 
fuel storage will be analyzed using NRC-approved methods and 
acceptance criteria. Compliance with the acceptance criteria for the 
applicable analysis for a given core design must be determined for 
each core prior to reloading. The material used to clad the fuel and 
the fuel enrichment are only two of the factors considered in this 
determinination. The application of approved methods ensures that 
all appropriate variables are addressed and their acceptance 
criteria satisfied. Thus, the proposed change does not involve a 
significant increase in the probability or consequences of an 
accident previously evaluated.
    2. The proposed change does not create the possibility of a new 
or different kind of accident from any accident previously 
evaluated. The determination of compliance with the acceptance 
criteria of the approved safety evaluation for any given core reload 
design is performed for each MY reload core prior to loading. In 
addition, determination of compliance with the acceptance criteria 
of the approved safety evaluation for fuel storage is performed for 
each core prior to receipt of the fuel. The use of approved methods 
and their acceptance criteria ensures that new or different 
accidents will not be encountered by the use of fuel rods clad with 
zirconium alloy and having higher fuel enrichments. Further, the 
proposed change does not involve any altertions to plant equipment 
that would affect any operational modes or accident precursors. 
Finally, the proposed change does not involve, or require secondary 
involvement of, any equipment important to safety. Thus the proposed 
change does not create the possibility of a new or different kind of 
accident from any accident previously evaluated.
    3. The proposed change does not involve a significant reduction 
in a margin of safety. Maine Yankee reload cores containing fuel 
rods clad with zirconium alloy and having higher fuel enrichments 
will be analyzed using NRC-approved methods and applicable 
acceptance criteria. Safety evaluations performed for each core 
reload ensure that the core design meets appropriate safety 
assessment acceptance criteria. In addition, the impact of fuel 
assembly design changes on fuel storage also will be analyzed using 
NRC-approved methods and aceptance criteria. Application of the 
approved methods ensures that the requirements of MY TS 1.1, Fuel 
Storage, are achieved. Because these requirements are not changed, 
the margin of safety remains the same. Thus there is no significant 
reduction in a margin of safety.
    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 this amendment request involves no significant hazards 
consideration.
    Local Public Document Room location: Wiscasset Public Library, High 
Street, P.O. Box 367, Wiscasset, ME 04578
    Attorney for licensee: Mary Ann Lynch, Esquire, Maine Yankee Atomic 
Power Company, 329 Bath Road, Brunswick, ME 04011
    NRC Project Director: Phillip F. McKee

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

    Date of amendment request: August 31, 1995
    Description of amendment request: The proposed amendment would 
relocate fire protection requirements from the Maine Yankee (MY) Atomic 
Power Station Technical Specifications (TS) to other, licensee-
controlled documents. The proposed amendment is consistent with the 
guidance of U.S. NRC Generic Letters 86-10, Implementation of Fire 
Protection Requirements, and 88-12, Removal of Fire Protection 
Requirements from the Technical Specifications.
    Basis for proposed no significant hazards consideration 
etermination: As required by 10 CFR 50.91(a), the licensee has provided 
its analysis if the issue of no significant hazards consideration. The 
NRC staff has reviewed the licensee's analysis against the standards of 
10 CFR 50.92(c). The NRC staff's review 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 is administrative and consistent with 
the guidance provided by the U.S. NRC. Removing fire protection 
requirements from the TS does not affect any fire protection 
equipment, or involve any physical modifications to plant 
structures, systems or components. The proposed change is not 
associated with accident initiation or mitigation and cannot affect 
the probability of occurrence of an accident, or increase the 
consequences of an accident. The licensee's fire protection plan 
contains the relocated requirements.
    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 introduces no new mode of plant 
operation, does not involve physical modification of any structure, 
system or component, and does not affect the function, operation or 
surveillance requirements of any equipment necessary for safe 
operation or shutdown. Further, the proposed change does not involve 
any change to equipment setpoints or operating parameters. The 
proposed change is administrative in nature. Existing plant fire 
protection equipment requirements are retained. Thus, the proposed 
change does not create the possibility for 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. No margins of safety established by system or 
component design, or verified by testing to ensure operability of 
fire protection systems or components, are affected. Fire protection 
requirements currently found in the TS will be relocated in their 
entirety to the Maine Yankee Fire Protection Plan. Any future 

[[Page 52933]]
changes to the Plan will be evaluated in accordance with the 
requirements of 10 CFR 50.59, Changes, tests and experiments. Thus 
the proposed change does not involve a significant reduction in a 
margin of safety.
    Based on this review, it appears that the three standards of 10 CFR 
50.92(c) are satisfied. Therefore, the NRC staff proposes to determine 
that the amendment request involves no significant hazards 
consideration.
    Local Public Document Room location: Wiscasset Public Library, High 
Street, P.O. Box 367, Wiscasset, ME 04578
    Attorney for licensee: Mary Ann Lynch, Esquire, Maine Yankee Atomic 
Power Company, 329 Bath Road, Brunswick, ME 04011
    NRC Project Director: Phillip F. McKee

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

    Date of amendment request: September 11, 1995
    Description of amendment request: The proposed changes affect 
Technical Specification Sections 3.4.8 and 3.9.9, Tables 2.2-1, 3.3-3, 
3.3-5 and 3.3-8, and Bases Sections 3/4.2.1, 3/4.4.8 and 3/4.11.2.1. 
These changes combine several different administrative changes which 
will correct typographical errors, provide clarifications, or make 
editorial changes.
    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:
    Pursuant to 10CFR50.92, NNECO has reviewed the proposed changes. 
NNECO 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:
    1. Involve a significant increase in the probability or 
consequences of an accident previously analyzed.
    The proposed changes are administrative in nature and do not 
result in changes to plant configuration, operation, accident 
mitigation, or analysis assumptions. Thus, it cannot increase the 
probability or consequence of an accident.
    2. Create the possibility of a new or different kind of accident 
from any previously analyzed.
    The proposed changes are administrative in nature and do not 
result in changes to plant configuration, operation, accident 
mitigation, or analysis assumptions. The intent and application of 
the proposed specification will not change. Therefore, the proposal 
does not create the possibility of a new or different kind of 
accident from any previously analyzed.
    3. Involve a significant reduction in the margin of safety.
    Since the proposed change[s] are administrative in nature and do 
not result in changes to plant configuration, operation, accident 
mitigation, or analysis assumptions, there is no reduction in the 
margin of safety.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
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

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

    Date of amendment requests: July 17, 1995
    Description of amendment requests: The proposed amendments would 
revise the Prairie Island Radiological Effluent Technical 
Specifications and other sections relating to radiological controls to 
conform to NUREG-1431, Standard Technical Specifications, Westinghouse 
Plants, Revision 1, and Generic Letter 89-01, ``Implementation of 
Programmatic Controls for Radiological Effluent Technical 
Specifications in the Administrative Controls Section of the Technical 
Specifications and the Relocation of Procedural Details of RETS to the 
Offsite Dose Calculation Manual or to the Process Control Program.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:
    1. The proposed amendment will not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    The proposed changes do not involve a significant increase in 
the probability or consequences of an accident previously evaluated. 
The proposed changes are administrative in nature and alter only the 
format and location of programmatic controls and procedural details 
relative to radioactive effluents, radiological environmental 
monitoring, radioactive source leakage testing, solid radioactive 
wastes, and associated reporting requirements. Existing Technical 
Specifications containing procedural details on radioactive 
effluents, radiological environmental monitoring, radioactive source 
leakage testing, explosive gas monitoring, storage tank radioactive 
content limits, solid radioactive wastes and associated reporting 
requirements are being relocated to the Offsite Dose Calculation 
Manual, Process Control Program or other new programs as 
appropriate. Compliance with applicable regulatory requirements will 
continue to be maintained. In addition, the proposed changes do not 
alter the conditions or the assumptions in any of the previous 
accident analyses. Since the previous accident analyses remain 
bonding, the radiological consequences previously evaluated are not 
adversely affected by the proposed changes.
    Therefore, the probability or consequences of an accident 
previously evaluated are not affected by any of the proposed 
amendments.
    2. The proposed amendment will not create the possibility of a 
new or different kind of accident from any accident previously 
analyzed.
    The proposed changes do not create the possibility of a new or 
different kind of accident from any accident previously evaluated. 
The proposed changes do not involve any change to the configuration 
or method of operation of any plant equipment. Accordingly, no new 
failure modes have been defined for any plant system or component 
important to safety nor has any new limiting single failure been 
identified as a result of the proposed changes. Also, there will be 
no change in types or increase in the amounts of any effluents 
released offsite.
    Therefore, the possibility of a new or different kind of 
accident from any accident previously evaluated would not be 
created.
    3. The proposed amendment will not involve a significant 
reduction in the margin of safety.
    The proposed changes do not involve a significant reduction in a 
margin of safety. The proposed changes do not involve any actual 
change in the methodology used in the control of radioactive 
effluents, radioactive sources, solid radioactive wastes, or 
radiological environmental monitoring. These changes are considered 
administrative in nature and provide for the relocation of 
procedural details outside of the technical specifications but add 
appropriate administrative controls to provide continued assurance 
of compliance to applicable regulatory requirements. These proposed 
changes also comply with the guidance contained in Generic Letter 
89-01 and the Standard Technical Specifications.
    Therefore, it can be concluded a significant reduction in the 
margin of safety would not be involved.
    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.

[[Page 52934]]

    Local Public Document Room location: Minneapolis Public Library, 
Technology and Science Department, 300 Nicollet Mall, Minneapolis, 
Minnesota 55401
    Attorney for licensee: Jay Silberg, Esq., Shaw, Pittman, Potts, and 
Trowbridge, 2300 N Street, NW, Washington, DC 20037
    NRC Project Director: John N. Hannon

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

    Date of amendment request: June 19, 1995
    Description of amendment request: The proposed amendment would 
revise Technical Specification Section 2.1, ``Safety Limits,'' to 
change the Minimum Critical Power Ratio Safety Limit due to the use of 
General Electric 13 fuel product line.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:
    1. The proposed Technical Specifications (TS) change does not 
involve a significant increase in the probability or consequences of 
an accident previously evaluated.
    The derivation of the revised GE13 [General Electric] Minimum 
Critical Power Ratio (MCPR) Safety Limit for incorporation into the 
Technical Specifications, and its use to determine cycle-specific 
thermal limits have been performed using NRC-approved methods within 
the existing design and licensing basis, and cannot increase the 
probability or severity of an accident.
    The basis of the MCPR Safety Limit calculation is to ensure that 
greater than 99.9% of all fuel rods in the core avoid boiling 
transition if the limit is not violated. The new MCPR Safety Limit 
preserves the existing margin to transition boiling and fuel damage 
in the event of a postulated accident.
    All design bases of the MCPR Safety Limit calculation apply to 
GE13 fuel in the same manner that they have applied to previous fuel 
designs. The probability of fuel damage is not increased.
    Therefore, the proposed TS change does not involve an increase 
in the probability or consequences of an accident previously 
evaluated.
    2. The proposed TS change does not create the possibility of a 
new or different kind of accident from any accident previously 
evaluated.
    The MCPR Safety Limit for the GE13 fuel design is a Technical 
Specification numerical value, designed to ensure that fuel damage 
from transition boiling does not occur as a result of the limiting 
postulated accident. It cannot create the possibility of any new 
type of accident. The new Minimum Critical Power Ratio (MCPR) Safety 
Limit is calculated using NRC-approved methods and has the same 
calculational basis as the MCPR Safety Limit for other GE fuel 
designs currently used at LGS [Limerick Generating Station] Unit 1.
    Therefore, the proposed TS change does not create the 
possibility of a new or different kind of accident, from any 
accident previously evaluated.
    3. The proposed TS change does not involve a significant 
reduction in a margin of safety.
    The following TS Bases were reviewed for potential reduction in 
the margin of safety:
    2.1 ``Safety Limits''
    3/4.2.1 ``Average Planar Linear Heat Generation Rate''
    3/4.2.3 ``Minimum Critical Power Ratio''
    3/4.2.4 ``Linear Heat Generation Rate''
    3/4.4.1 ``Recirculation System''
    3/4.9 ``Refueling Operations''
    The margin of safety as defined in the TS Bases will remain the 
same. The new Minimum Critical Power Ratio (MCPR) Safety Limit is 
calculated using NRC approved methods which are in accordance with 
the current fuel design and licensing criteria. The MCPR Safety 
Limit for GE13 fuel remains high enough to ensure that greater than 
99.9% of all fuel rods in the core will avoid boiling transition if 
the limit is not violated, thereby preserving the fuel cladding 
integrity.
    Therefore, the proposed TS change does not involve a reduction 
in a margin of safety.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: Pottstown Public Library, 500 
High Street, Pottstown, Pennsylvania 19464.
    Attorney for licensee: J. W. Durham, Sr., Esquire, Sr. V. P. and 
General Counsel, Philadelphia Electric Company, 2301 Market Street, 
Philadelphia, Pennsylvania 19101
    NRC Project Director: John F. Stolz

Philadelphia Electric Company, Docket Nos. 50-352 and 50-353, 
Limerick Generating Station, Units 1 and 2, Montgomery County, 
Pennsylvania

    Date of amendment request: September 14, 1995
    Description of amendment request: The amendments change the 
Technical Specifications (TS) by removing the Reactor Enclosure and 
Refueling Area Secondary Containment Isolation Valve Tables 3.6.5.2.1-1 
and 3.6.5.2.2-1 from TS in accordance with NRC Generic Letter (GL) 91-
08, ``Removal of Component Lists from 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 TS changes do not involve a significant increase 
in the probability or consequences of an accident previously 
evaluated.
    The proposed changes will remove component tables from TS. The 
component lists will be retained in licensee controlled documents 
(UFSAR [Updated Final Safety Analysis Report] and a plant procedure) 
which will be maintained under the requirements of TS Administrative 
Controls Section 6.0 and the provisions of 10 CFR 50.59. Since any 
changes to licensee controlled documents are required to be 
evaluated per 10 CFR 50.59, no increase (significant or 
insignificant) in the probability or consequences of an accident 
previously evaluated will be allowed.
    In addition, these proposed changes will not affect any 
equipment important to safety, in structure or operation. These 
changes will not alter operation of process variables, structures, 
systems, or components as described in the safety analysis and 
licensing basis. The changes will not increase the probability or 
consequences of occurrence of a malfunction of equipment important 
to safety previously evaluated in the SAR [Safety Analysis Report].
    2. The proposed TS changes do not create the possibility of a 
new or different kind of accident from any accident previously 
evaluated.
    The proposed changes will not alter the plant configuration or 
change the methods governing normal plant operation. The changes 
will not impose different operating requirements and adequate 
control of information will be retained. The changes will not alter 
assumptions made in the safety analysis and licensing basis. Since 
the proposed changes cannot cause an accident, and the plant 
response to the design basis events is unchanged, the changes do not 
create the possibility of a new or different kind of accident from 
any accident previously evaluated.
    3. The proposed TS changes do not involve a significant 
reduction in a margin of safety.
    The proposed changes to remove the component tables from TS have 
been performed under the guidance of NRC GL 91-08. The component 
lists will be retained in licensee controlled documents (UFSAR and a 
plant procedure) which will be maintained under the requirements of 
TS Administrative Controls Section 6.0 and the provisions of 10 CFR 
50.59. These changes will not reduce the margin of safety since they 
have no impact on any safety analysis assumptions. Since any future 
changes to the removed tables will be evaluated under the 
requirements of 10 CFR 50.59, no reduction (significant or 
insignificant) in a margin of safety will be allowed. Therefore, the 
proposed TS 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 


[[Page 52935]]
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: Pottstown Public Library, 500 
High Street, Pottstown, Pennsylvania 19464.
    Attorney for licensee: J. W. Durham, Sr., Esquire, Sr. V. P. and 
General Counsel, Philadelphia Electric Company, 2301 Market Street, 
Philadelphia, Pennsylvania 19101
    NRC Project Director: John F. Stolz

Rochester Gas and Electric Corporation, Docket No. 50-244, R. E. 
Ginna Nuclear Power Plant, Wayne County, New York

    Date of amendment request: May 26, 1995
    Brief description of amendment: The proposed amendment would 
represent a full conversion from the current Technical Specifications 
(TSs) to a set of TS based on NUREG-1431, ``Standard Technical 
Specifications, Westinghouse Plants,'' Revision 0, dated September 
1993, together with approved travellers used in the issuance of 
Revision 1, dated April 1995. NUREG-1431 was developed through working 
groups composed of NRC staff members and industry representatives and 
has been endorsed by the staff as part of an industry-wide initiative 
to standardize and improve the TSs. As part of this submittal, the 
licensee has applied the criteria contained in the Commission's Final 
Policy Statement on Technical Specification Improvements for Nuclear 
Power Reactors of July 22, 1993, to the current Ginna TSs, and using 
NUREG-1431 as a basis, developed a proposed set of improved TSs for 
Ginna.Date of publication of individual notice in Federal Register: 
September 26, 1995 (60 FR 49636)
    Expiration date of individual notice: October 26, 1995
    Local Public Document Room location: Rochester Public Library, 115 
South Avenue, Rochester, New York 14610

Tennessee Valley Authority, Docket No. 50-296, Browns Ferry Nuclear 
Plant, Unit 3, Limestone County, Alabama

    Date of amendment request: September 13, 1995 (TS 368)
    Description of amendment request: The proposed amendment deletes 
requirements for daily checks for certain instruments that do not have 
indications, and provides editorial changes.
    Basis for proposed no significant hazards consideration 
determination:As required by 10 CFR 50.91(a), the licensee has provided 
its analysis of the issue of no significant hazards consideration, 
which is presented below:
    A. The proposed amendment does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    The proposed changes are administrative in nature and correct 
errors that were introduced by previous changes to the TSs. These 
changes do not affect any of the design basis accidents nor do they 
involve an increase in the probability or consequences of an 
accident previously evaluated.
    B. The proposed amendment does not create the possibility of a 
new or different kind of accident from any accident previously 
evaluated.
    The proposed changes are administrative in nature. These changes 
do not change the operation or function of the affected 
instrumentation. The deletion of the RCIC and HPCI instrument checks 
reflects the actual installed configuration of this instrumentation 
(no indication) and the change to Table 4.2.C corrects the 
referenced note for the SRM Upscale function. Therefore, the 
possibility for an accident or malfunction of a different type than 
any evaluated previously is not created by this change.
    C. The proposed amendment does not involve a significant 
reduction in a margin of safety.
    The proposed changes are administrative in nature. The proposed 
changes to TS Tables 4.2.B and 4.2.C do not affect any acceptable 
limit of operation, instrument setpoint, or analysis assumption in 
the TS or Bases. Therefore, this change does not reduce the margin 
of safety as defined in the basis for any TS.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: Athens Public Library, South 
Street, Athens, Alabama 35611
    Attorney for licensee: General Counsel, Tennessee Valley Authority, 
400 West Summit Hill Drive, ET llH, Knoxville, Tennessee 37902
    NRC Project Director: Frederick J. Hebdon

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

    Date of amendment request: August 15, 1995
    Brief description of amendments: The proposed amendment would 
relocate the Shutdown Margin limits from the Technical Specifications 
(TSs) to the Core Operating Limits Report. The proposed changes are 
consistent with the intent of Generic Letter (GL) 88-16 which provides 
guidelines for the removal of cycle-specific parameter limits from the 
TSs.
    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. Do the proposed changes involve a significant increase in the 
probability or consequences of an accident previously evaluated?
    The proposed changes remove cycle-specific parameter limits from 
the Technical Specifications, add them to the list of limits 
contained in the Core Operating Limits Report (COLR), and revise the 
Administrative Controls section of the Technical Specifications. The 
changes do not, by themselves, alter any of the parameter limits. 
The changes are administrative in nature and have no adverse effect 
on the probability of an accident or on the consequences of an 
accident previously evaluated. The removal of parameter limits from 
the Technical Specifications does not eliminate the requirement to 
comply with the parameter limits.
    The parameter limits in the COLR may be revised without prior 
NRC approval. However, Specification 6.9.1.6c continues to ensure 
that the parameter limits are developed using NRC-approved 
methodologies and that applicable limits of the safety analyses are 
met. While future changes to the COLR parameter limits could result 
in event consequences which are either slightly less or slightly 
more severe than the consequences for the same event using the 
present parameter limits, the differences would not be significant 
and would be bounded by the requirement of specification 6.9.1.6c to 
meet the applicable limits in the safety analysis.
    Based on the above, removal of the parameter limits from the 
Technical Specifications and the addition of these limits the list 
of limits in the COLR, thus allowing revision of the parameter 
limits without prior NRC approval, has no significant effect on the 
probability or consequences of an accident previously evaluated.
    2. Do the proposed changes create the possibility of a new or 
different kind of accident from any accident previously evaluated?
    The proposed changes remove certain parameter limits from the 
Technical Specifications and add these limits to the list of limits 
in the COLR, removing the requirement for prior NRC approval of 
revisions to those parameters. The changes do not add new hardware 
or change plant operations and therefore cannot initiate an event 
nor cause an analyzed event to progress differently. Thus, the 
possibility of a new or different kind of accident is not created.
    3. Do the proposed changes involve a significant reduction in a 
margin of safety?
    The margin of safety, as it relates to a parameter limit, is the 
difference between the 

[[Page 52936]]
acceptance criterion for that parameter and its failure value. The 
proposed changes do not affect the failure values for any system. 
Through the accident analyses, all relevant event acceptance 
criteria (as described in the NRC-approved analysis methodologies) 
are shown to be satisfied; therefore, there is no impact on an event 
acceptance criteria. Because neither the failure values nor the 
acceptance criteria are affected, the proposed change has no effect 
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: University of Texas at 
Arlington Library, Government Publications/Maps, 702 College, P.O. Box 
19497, Arlington, TX 76019
    Attorney for licensee: George L. Edgar, Esq., Morgan, Lewis and 
Bockius, 1800 M Street, N.W., Washington, DC 20036
    NRC Project Director: William D. Beckner

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

    Date of amendment request: September 19, 1995
    Description of amendment request: The proposed amendment would make 
administrative changes to the Kewaunee Nuclear Power Plant (KNPP) 
Technical Specifications (TS) to improve their clarity and consistency. 
The proposed amendment includes changes to reflect revisions to 10 CFR 
Part 20, and changes to correct minor typographical and format 
inconsistencies as part of an ongoing effort to convert the TS to the 
WordPerfect format.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:
    The proposed changes were reviewed in accordance with the 
provisions of 10 CFR 50.92 to show no significant hazards exist. The 
proposed changes will not:
    1. involve a significant increase in the probability or 
consequences of an accident previously evaluated.
    The likelihood that an accident will occur is neither increased 
or decreased by these TS changes. These TS changes will not impact 
the function or method of operation of plant equipment. Thus, there 
is not a significant increase in the probability of a previously 
analyzed accident due to these changes. No systems, equipment, or 
components are affected by the proposed changes. Thus, the 
consequences of the malfunction of equipment important to safety 
previously evaluated in the Updated Safety Analysis Report (USAR) 
are not increased by these changes.
    The proposed changes are administrative in nature and, 
therefore, have no impact on accident initiators or plant equipment, 
and thus, do not affect the probabilities or consequences of an 
accident.
    2. create the possibility of a new or different kind of accident 
from any accident previously evaluated.
    Operation of the facility in accordance with the proposed TS 
changes would not create the possibility of a new or different kind 
of accident from any accident previously evaluated.
    The proposed changes do not involve changes to the physical 
plant or operations. Since these administrative changes do not 
contribute to accident initiation, they do not produce a new 
accident scenario or produce a new type of equipment malfunction. 
Also, these changes do not alter any existing accident scenarios; 
they do not affect equipment or its operation, and thus, do not 
create the possibility of a new or different kind of accident.
    3. involve a significant reduction in the margin of safety.
    Operation of the facility in accordance with the proposed TS 
would not involve a significant reduction in a margin of safety. The 
proposed changes do not affect plant equipment or operation. Safety 
limits and limiting safety system settings are not affected by these 
proposed changes.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: University of Wisconsin, 
Cofrin Library, 2420 Nicolet Drive, Green Bay, Wisconsin 54311-7001.
    Attorney for licensee: Bradley D. Jackson, Esq., Foley and Lardner, 
P. O. Box 1497, Madison, Wisconsin 53701-1497.
    NRC Project Director: Gail H. Marcus

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

    Date of amendment request: September 14, 1995
    Description of amendment request: The proposed amendment would 
revise Technical Specification 3/4.5.5 to increase the outage time 
allowed for adjusting the boron concentration of the refueling water 
storage tank (RWST) from 1 hour to 8 hours.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:
    1. The proposed change does not involve a significant increase 
in the probability or consequences of an accident previously 
evaluated.
    The increase in the RWST allowed outage time does not alter the 
plant configuration or operation. The potential for the RWST boron 
concentration to be outside the technical specification limits is 
small because the RWST and its contents are not involved with normal 
plant operation and are not subject to process variations associated 
with plant operation.
    The potential causes of boron concentration deviation have been 
evaluated with the conclusion that any deviation in RWST boron 
concentration would not be expected to increase significantly during 
the proposed 7 hour allowed outage time increase.
    Therefore, the proposed change does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    2. The proposed change does not create the possibility of a new 
or different kind of accident from any accident previously 
evaluated.
    Increasing the RWST allowed outage time from 1 hour to 8 hours 
for reasons directly related to boron concentration does not require 
physical alteration to any plant system and does not change the 
method by which any safety related system performs its functions. 
Therefore, the proposed change does not create the possibility of a 
new or different kind of accident from any accident previously 
evaluated.
    3. The proposed change does not involve a significant reduction 
in a margin of safety.
    Increasing the RWST allowed outage time for reasons directly 
related to boron concentration does not affect any accident analysis 
assumptions, initial conditions, or results. The margins of safety 
reflected in the Wolf Creek Generating Station Technical 
Specifications are not compromised by the 7 hour allowed outage time 
increase. Therefore, the proposed change does not involve a 
significant reduction in a margin of safety.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room 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

[[Page 52937]]


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. STN 50-454 and STN 50-455, 
Byron Station, Unit Nos. 1 and 2, Ogle County, Illinois, Docket 
Nos. STN 50-456 and STN 50-457, Braidwood Station, Unit Nos. 1 and 
2, Will County, Illinois

    Date of amendment request: September 1, 1995
    Description of amendment request: The proposed amendments would 
revise the present voltage-based repair criteria in the Byron 1 and 
Braidwood 1 Technical Specifications (TSs). These proposed revisions 
would raise the lower voltage limit from its present value of 1.0 volt 
to 3.0 volts; there would no longer be an upper voltage limit.
    The Braidwood 1 TSs were revised by License Amendment No. 54, 
issued on August 18, 1994, to add voltage-based repair criteria to the 
existing steam generator (SG) tube repair criteria. The Byron 1 TSs 
were revised in a similar manner by License Amendment No. 66, issued on 
October 24, 1994.
    The voltage-based repair criteria in the subject TSs are applicable 
only to a specific type of SG tube degradation which is predominantly 
axially-oriented outer diameter stress corrosion cracking (ODSCC). This 
particular form of SG tube degradation occurs entirely within the 
intersections of the SG tubes with the tube support plates (TSPs).
    The present voltage values for the ODSCC repair criteria are based 
on the assumption of a ``free span'' exposure of the SG tube flaw; 
i.e., no credit is given for any constraint against burst or leakage, 
which may be provided by the presence of the TSPs. This approach is, in 
turn, based on the assumption that under postulated accident 
conditions, the TSPs may be displaced sufficiently by blowdown 
hydrodynamic loads such that a SG tube flaw which was fully confined 
within the thickness of the TSP prior to the accident would then be 
fully exposed. This approach was first advanced by the NRC staff in a 
draft generic letter issued on August 12, 1994, which was subsequently 
modified slightly and issued as Generic letter (GL) 95-05, ``Voltage-
Based Repair Criteria For Westinghouse Steam Generator Tubes Affected 
by Outside Diameter Stress Corrosion Cracking,'' dated August 3, 1995. 
The previous license amendments related to the issue of ODSCC were 
based to a large extent on the draft generic letter cited above.
    The fundamental difference between the pending proposal to raise 
the lower voltage repair limit to 3.0 volts and the methodology 
contained in GL 95-05, is that the licensee proposes to install certain 
modifications to the SG internal structures, thereby limiting to a 
small value, the maximum displacement of the TSPs under accident 
conditions. The proposed structural modifications consist of expanding 
a limited number of SG tubes only on the hot leg side of the TSP, at 
each of the intersections of the tubes with the TSPs. The purpose of 
this approach would be to greatly reduce the probability of SG tube 
burst under postulated accident conditions by several orders of 
magnitude. There would be a negligible impact on the primary-to-
secondary SG tube leakage under accident conditions.
    While the voltage-based repair criteria for ODSCC flaws are 
applicable only to Byron 1 and Braidwood 1, the pending request for 
license amendments involves all four units in that both stations have a 
common set of TSs. Date of publication of individual notice in Federal 
Register: September 27, 1995 (60 FR 49963)
    Expiration date of individual notice: October 27, 1995
    Local Public Document Room location: For Byron, the Byron Public 
Library District, 109 N. Franklin, P.O. Box 434, Byron, Illinois 61010; 
for Braidwood, the Wilmington Public Library, 201 S. Kankakee Street, 
Wilmington, Illinois 60481

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

    Date of amendment request: September 1, 1995
    Description of amendment request: The proposed amendment would 
upgrade the Dresden 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 
Letters (GL), and (4) relocating specific items to more appropriate TS 
locations. The September 1, 1995, application proposed to upgrade only 
Section 6.0 (Administrative Controls) of the Dresden TS.Date of 
publication of individual notice in Federal Register: September 20, 
1995 (60 FR 48728)
    Expiration date of individual notice: October 20, 1995
    Local Public Document Room location: Morris Area Public Library 
District, 604 Liberty Street, Morris, Illinois 60450

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: September 13, 1995
    Brief description of amendment request: The proposed amendments 
would revise the Administrative Controls section and the Bases section 
of the Beaver Valley Power Station, Unit Nos. 1 and 2 (BVPS-1 and BVPS-
2), technical specifications to be consistent with the requirements of 
the Offsite Dose Calculation Manual (ODCM). The ODCM was recently 
updated to reflect the radioactive liquid and gaseous effluent release 
limits and the liquid holdup tank activity limit of BVPS-1 License 
Amendment No. 188 and BVPS-2 License Amendment No. 70 which were issued 
June 12, 1995.Date of publication of individual notice in Federal 
Register: September 22, 1995 (60 FR 49292)
    Expiration date of individual notice: October 23, 1995
    Local Public Document Room location: B. F. Jones Memorial Library, 
663 Franklin Avenue, Aliquippa, Pennsylvania 15001

[[Page 52938]]


PECO Energy Company, Public Service Electric and Gas Company, 
Delmarva Power and Light Company, and Atlantic City Electric 
Company, Docket No. 50-278, Peach Bottom Atomic Power Station, Unit 
No. 3, York County, Pennsylvania

    Date of amendment request: September 1, 1995
    Brief description of amendment request: The proposed amendment 
would delete License Condition 2.C.(5) from Facility Operating License 
DPR-56 which restricts power levels to no less than seventy percent in 
the coastdown condition.
    Date of publication of individual notice in Federal Register: 
September 19, 1995 (60 FR 48530)
    Expiration date of individual notice: October 18, 1995
    Local Public Document Room location: Government Publications 
Section, State Library of Pennsylvania, (REGIONAL DEPOSITORY) Education 
Building, Walnut Street and Commonwealth Avenue, Box 1601, Harrisburg, 
Pennsylvania 17105

Notice Of Issuance Of Amendments To Facility Operating Licenses

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

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: December 7, 1994, as 
supplemented by letter dated August 1, 1995.
    Brief description of amendments: The amendments change Note 5 to 
Table 4.3-1 of Technical Specification 3/4.3.1 to allow verification of 
the shape-annealing matrix elements used in the core protection 
calculators. This provides the option of using generic shape-annealing 
matrix elements in the core protection calculators. Presently, cycle-
specific shape-annealing elements are determined during startup testing 
after each core reload. Use of a generic shape-annealing matrix 
eliminates several hours of critical path work during startup after a 
refueling outage.
    Date of issuance: September 20, 1995
    Effective date: September 20, 1995
    Amendment Nos.: Unit 1 - Amendment No. 100; Unit 2 - Amendment No. 
88; Unit 3 - Amendment No. 71
    Facility Operating License Nos. NPF-41, NPF-51, and NPF-74: The 
amendments revised the Technical Specifications.
    Date of initial notice in Federal Register: January 4, 1995 (60 FR 
495). The August 1, 1995, supplemental letter provided clarifying 
information and did not change the original no significant hazards 
consideration determination. The Commission's related evaluation of the 
amendments is contained in a Safety Evaluation dated September 20, 
1995.No significant hazards consideration comments received: No
    Local Public Document Room location: Phoenix Public Library, 1221 
N. Central Avenue, 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: March 26, 1993, as 
supplemented May 15, 1995
    Brief description of amendments: These amendments upgrade the 
current custom Technical Specifications (TS) for Dresden and Quad 
Cities to the Standard Technical Specifications contained in NUREG-
0123, ``Standard Technical Specification General Electric Plants BWR/
4.'' These amendments upgrade only Section 3/4.9 (Electrical Power 
Systems). These amendments include the relocation of some TS 
requirements to licensee-controlled documents.
    Date of issuance: September 18, 1995
    Effective date: Immediately, to be implemented no later than 
December 31, 1995, for Dresden Nuclear Power Station and June 30, 1996, 
for Quad Cities Nuclear Power Station.
    Amendment Nos.: 138, 132, 160, 156
    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: January 19, 1994 (59 FR 
2864) The Commission's related evaluation of the amendments is 
contained in a Safety Evaluation dated September 18, 1995. No 
significant hazards consideration comments received: No
    Local Public Document Room location: for Dresden, Morris Area 
Public Library District, 604 Liberty Street, Morris, Illinois 60450; 
for Quad Cities, Dixon Public Library, 221 Hennepin Avenue, Dixon, 
Illinois 61021

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

    Date of application for amendments: December 8, 1992, as 
supplemented September 10, 1993, and May 17, 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 3/4.1 (Reactor 
Protection System). Date of issuance: 

[[Page 52939]]
September 20, 1995Effective date: Immediately, to be implemented no 
later than December 31, 1995, for Dresden Station and June 30, 1996, 
for Quad Cities Station.
    Amendment Nos.: 139, 133, 161, and 157
    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: June 6, 1995 (60 FR 
29872) The Commission's related evaluation of the amendments is 
contained in a Safety Evaluation dated September 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, Illinois 
Docket Nos. 50-254 and 50-265, Quad Cities Nuclear Power Station, 
Units 1 and 2, Rock Island County, Illinois

    Date of application for amendments: September 17, 1993, as 
supplemented June 30, 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 3/4.6.
    Date of issuance: September 21, 1995
    Effective date: Immediately, to be implemented no later than 
December 31, 1995, for Dresden Station and June 30, 1996, for Quad 
Cities Station.
    Amendment Nos.: 140, 134, 162, and 158
    Facility Operating License Nos. DPR-19, DPR-25, DPR-29 and DPR-30. 
The amendments revised the Technical Specifications.
    Date of initial notice in Federal Register: July 19, 1995 (60 FR 
37087) The Commission's related evaluation of the amendments is 
contained in a Safety Evaluation dated September 21, 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. 50-373 and 50-374, LaSalle 
County Station, Units 1 and 2, LaSalle County, Illinois

    Date of application for amendments: April 11, 1995
    Brief description of amendments: The amendments allow a one-time 
extension of specific LaSalle, Units 1 and 2, 18-month Technical 
Specification Surveillance Requirements to allow surveillance testing 
to coincide with the LaSalle, Unit 1, seventh refueling outage (L1R07). 
The shutdown for L1R07 has been rescheduled from September 1995 until 
early 1996. The proposed extensions apply to calibrations and 
functional testing of isolation actuation instrumentation, emergency 
core cooling system actuation instrumentation, and recirculation pump 
trip actuation instrumentation; leakage testing of reactor coolant 
system isolation valves; inspection of fire-rated seals; functional 
testing of mechanical snubbers; inspections of emergency diesel 
generators; and testing of batteries, battery chargers, and other 
electrical components.
    Date of issuance: September 27, 1995
    Effective date: Immediately, to be implemented within 30 days.
    Amendment Nos.:  106 and 92
    Facility Operating License Nos. NPF-11 and NPF-18: The amendments 
revised the Facility Operating Licenses.
    Date of initial notice in Federal Register: July 5, 1995 (60 FR 
35066) The Commission's related evaluation of the amendments is 
contained in a Safety Evaluation dated September 27, 1995. No 
significant hazards consideration comments received: No
    Local Public Document Room location: Jacobs Memorial Library, 
Illinois Valley Community College, Oglesby, Illinois 61348.

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

    Date of application for amendments: June 17, 1993, as supplemented 
July 5, 1995
    Brief description of amendments: The amendments revise Technical 
Specification Section 5.3.1 ``Fuel Assemblies'' in accordance with 
Generic Letter 90-02, Supplement 1, ``Alternative Requirements For Fuel 
Assemblies in The Design Features Section of Technical 
Specifications.''
    Date of issuance: September 18, 1995
    Effective date: As of the date of issuance to be implemented within 
30 days from the date of issuance
    Amendment Nos.: 135 and 129
    Facility Operating License Nos. NPF-35 and NPF-52: Amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: July 21, 1993 (58 FR 
39048) and ReNoticed August 16, 1995 (60 FR 42601) The Commission's 
related evaluation of the amendments is contained in a Safety 
Evaluation dated September 18, 1995. No significant hazards 
consideration comments received: No
    Local Public Document Room location: York County Library, 138 East 
Black Street, Rock Hill, South Carolina 29730

Duquesne Light Company, et al., Docket No. 50-334, Beaver Valley 
Power Station, Unit No. 1, Shippingport, Pennsylvania

    Date of application for amendment: July 11, 1995
    Brief description of amendment: This amendment revised the required 
area of the reactor coolant system overpressure protection system vent 
from 3.14 square inches to 2.07 square inches which is equal to the 
relief area of a single power-operated relief valve.
    Date of issuance: September 26, 1995
    Effective date: As of the date of issuance, to be implemented 
within 60 days.
    Amendment No.: 193
    Facility Operating License No. DPR-66. Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: August 16, 1995 (60 FR 
42603) The Commission's related evaluation of the amendment is 
contained in a Safety Evaluation dated September 26, 1995. No 
significant hazards consideration comments received: No
    Local Public Document Room location: B. F. Jones Memorial Library, 
663 Franklin Avenue, Aliquippa, Pennsylvania 15001

Duquesne Light Company, et al., Docket No. 50-412, Beaver Valley 
PowerStation, Unit 2, Shippingport, Pennsylvania

    Date of application for amendment: July 24, 1995
    Brief description of amendment: This amendment revises TS 3/4.4.11, 
``Relief Valves,'' and associated Bases to make Unit 2 TS 3/4.4.11 
consistent with Unit 1 TS 3/4.4.11 which was revised by Unit 1 License 
Amendment No. 187 issued on May 15, 1995. The amendment generally 
reflects the guidance provided in NRC Generic Letter 90-06 and in the 
NRC's Improved Standard Technical Specifications (NUREG-1431).
    Date of issuance: September 18, 1995

[[Page 52940]]

    Effective date: As of the date of issuance, to be implemented 
within 60 days.
    Amendment No.: 76
    Facility Operating License No. NPF-73: Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: August 16, 1995 (60 FR 
42604) The Commission's related evaluation of the amendment is 
contained in a Safety Evaluation dated September 18, 1995. No 
significant hazards consideration comments received: No
    Local Public Document Room location: B. F. Jones Memorial Library, 
663 Franklin Avenue, Aliquippa, Pennsylvania 15001

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

    Date of application for amendment: March 17, 1995
    Brief description of amendment: The amendment revises requirements 
associated with the frequency of containment post-entry visual 
inspections.
    Date of issuance: September 15, 1995
    Effective date: September 15, 1995
    Amendment No.: 162
    Facility Operating License No. NPF-6. Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: July 19, 1995 (60 FR 
37089) The Commission's related evaluation of the amendment is 
contained in a Safety Evaluation dated September 15, 1995. No 
significant hazards consideration comments received: No
    Local Public Document Room location: Tomlinson Library, Arkansas 
Tech University, Russellville, AR 72801

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

    Date of application for amendment: October 27, 1993
    Brief description of amendment: The amendment relocated reactor 
incore detector requirements from the TSs to the safety analysis 
report.
    Date of issuance: September 15, 1995
    Effective date: September 15, 1995
    Amendment No.: 163
    Facility Operating License No. NPF-6. Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: December 8, 1993 (58 FR 
64606) The Commission's related evaluation of the amendment is 
contained in a Safety Evaluation dated September 15, 1995. No 
significant hazards consideration comments received: No
    Local Public Document Room location: Tomlinson Library, Arkansas 
Tech University, Russellville, AR 72801

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

    Date of application for amendment: March 17, 1995
    Brief description of amendment: The amendment transfers 
requirements for cycle specific core operating limits from the 
Technical Specifications to the Core Operating Limits Report. 
Additionally, a reference to a statistical methodology for determining 
uncertainties is being changed to reference a methodology that was 
recently approved by the NRC.
    Date of issuance: September 19, 1995
    Effective date: September 19, 1995
    Amendment No.: 164
    Facility Operating License No. NPF-6. Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: July 19, 1995 (60 FR 
37088) The Commission's related evaluation of the amendment is 
contained in a Safety Evaluation dated September 19, 1995. No 
significant hazards consideration comments received: No
    Local Public Document Room location: Tomlinson Library, Arkansas 
Tech University, Russellville, AR 72801

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

    Date of application for amendment: April 4, 1995, as supplemented 
August 25, 1995
    Brief description of amendment: The amendment provides a one-time 
extension of the reactor coolant pump flywheel inservice inspection.
    Date of issuance: September 22, 1995
    Effective date: September 22, 1995
    Amendment No.: 165
    Facility Operating License No. NPF-6. Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: July 5, 1995 (60 FR 
35069) The August 25, 1995, submittal did not change the original no 
significant hazards consideration determination.The Commission's 
related evaluation of the amendment is contained in a Safety Evaluation 
dated September 22, 1995. No significant hazards consideration comments 
received: No
    Local Public Document Room location: Tomlinson Library, Arkansas 
Tech University, Russellville, AR 72801

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

    Date of application for amendment: May 19, 1995 as supplemented 
July 21, 1995.
    Brief description of amendment: The amendment revises the 
specifications to permit the containment personnel airlock doors to 
remain open during fuel handling.
    Date of issuance: September 28, 1995
    Effective date: September 28, 1995
    Amendment No.: 166
    Facility Operating License No. NPF-6. Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: August 2, 1995 (60 FR 
39437) The July 22, 1995, supplement provided clarifying information 
and did not change the original no significant hazards consideration 
determination. The Commission's related evaluation of the amendment is 
contained in a Safety Evaluation dated September 28, 1995. No 
significant hazards consideration comments received: No
    Local Public Document Room location: Tomlinson Library, Arkansas 
Tech University, Russellville, AR 72801

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

    Date of application for amendment: April 4, 1995, as 
supplementedSeptember 28, 1995
    Brief description of amendment: The amendment removes the 
requirement to maintain water level 23 feet above irradiated fuel 
assemblies in the reactor while latching and unlatching control element 
assemblies.
    Date of issuance: September 28, 1995
    Effective date: September 28, 1995
    Amendment No.: 167
    Facility Operating License No. NPF-6. Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: August 16, 1995 (60 FR 
42604) The September 28, 1995, submittal provided clarifying 
information and did not change the original no significant hazards 
consideration determination. The Commission's related evaluation of the 
amendment is contained in a Safety Evaluation dated September 28, 1995. 
No significant hazards consideration comments received: No.
    Local Public Document Room location: Tomlinson Library, Arkansas 
Tech University, Russellville, AR 72801

[[Page 52941]]


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

    Date of amendment request: June 22, 1994, as supplemented by 
letters dated June 28, 1995 and August 22, 1995
    Brief description of amendment: The amendment changes the Appendix 
A TSs by increasing the control room radiation monitor setpoint (CRRMS) 
to a fixed value of 5.45E-6 micro curies per cubic centimeters instead 
of being set at two times the background.
    Date of issuance: September 27, 1995
    Effective date: Septembe 27, 1995
    Amendment No.: 114
    Facility Operating License No. NPF-38. Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: August 3, 1994 (59 FR 
39586) The June 28, 1995 and August 22, 1995, letters provided 
clarifying information that did not change the originial proposed no 
significant hazards consideration determination. The Commission's 
related evaluation of the amendment is contained in a Safety Evaluation 
dated September 27, 1995. No significant hazards consideration comments 
received: No
    Local Public Document Room location: University of New Orleans 
Library, Louisiana Collection, Lakefront, New Orleans, LA 70122

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: August 11, 1995
    Brief description of amendment: The amendment removes the Technical 
Specifications for the Makeup, Purification, and Chemical Addition 
Systems from the Technical Specifications (Section 3.2) and relocates 
the pertinent design information, including tank volume and boron 
concentrations, to the TMI-1 Updated Final Safety Analysis Report.
    Date of issuance: September 19, 1995
    Effective date: September 19, 1995
    Amendment No.: 196
    Facility Operating License No. DPR-50. Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: August 18, 1995 (60 FR 
43172) The Commission's related evaluation of this amendment is 
contained in a Safety Evaluation dated September 19, 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

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

    Date of application for amendment: June 9, 1995
    Brief description of amendment: The amendment modifies Technical 
Specification 4.1, ``Site Location,'' to incorporate a description of 
the exclusion area boundary. The change is necessary to ensure the 
content of the technical specifications conform to Section 182 of the 
Atomic Energy Act of 1954.
    Date of issuance: September 14, 1995
    Effective date:  September 14, 1995
    Amendment No.: 101
    Facility Operating License No. NPF-62: The amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: July 19, 1995 (60 FR 
37093) The Commission's related evaluation of the amendment is 
contained in a Safety Evaluation dated September 14, 1995. No 
significant hazards consideration comments received: No
    Local Public Document Room location: The Vespasian Warner Public 
Library, 120 West Johnson Street, Clinton, Illinois 61727

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: July 21, 1995
    Brief description of amendment: The amendment revised Technical 
Specifications Section 6.0 (Administrative Controls) to replace the 
title-specific list of members on the Plant Operating Review Committee 
(PORC) with a more general statement of membership requirements. The 
scope of disciplines represented on the PORC was also expanded to 
include nuclear licensing and quality assurance. The amendment also 
changed the title ``Resident Manager'' to ``Site Executive Officer.'' 
This title change was an administrative change that did not affect the 
reporting relationship, authority, or responsibility of the position.
    Date of issuance: September 20, 1995
    Effective date: As of the date of issuance to be implemented within 
30 days.
    Amendment No.: 163
    Facility Operating License No. DPR-64: Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: August 16, 1995 (60 FR 
42606) The Commission's related evaluation of the amendment is 
contained in a Safety Evaluation dated September 20, 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: April 25, 1994
    Brief description of amendment: This amendment revises TS Section 
3.8.1.1, ``A.C. Sources - Operating,'' TS Section 3.8.1.2, ``A.C. 
Sources - Shutdown,'' and associated Bases, to increase the required 
quantity of fuel in the Emergency Diesel Generator Fuel Oil Day Tanks 
from 200 to 360 gallons.
    Date of issuance: September 15, 1995
    Effective date: As of the date of issuance and shall be implemented 
within 60 days.
    Amendment No.: 79
    Facility Operating License No. NPF-57: This amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: June 8, 1994 (59 FR 
29632)The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated September 15, 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. 50-354, Hope 
Creek Generating Station, Salem County, New Jersey

    Date of application for amendment: January 20, 1995
    Brief description of amendment: This amendment changes Technical 
Specification (TS) 4.1.3.1.2.b, ``Control Rods - Surveillance 
Requirement'' to change the required action to be taken when a control 
rod becomes immovable due to excessive friction from ``at least once 
per'' 24 hours to ``within'' 24 hours.
    Date of issuance: September 20, 1995
    Effective date: As of its date of issuance, to be implemented 
within 60 days.
    Amendment No.: 80
    Facility Operating License No. NPF-57: This amendment revised the 
Technical Specifications. 

[[Page 52942]]

    Date of initial notice in Federal Register: August 2, 1995 (60 FR 
39452) The Commission's related evaluation of the amendment is 
contained in a Safety Evaluation dated September 20, 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. 50-354, Hope 
Creek Generating Station, Salem County, New Jersey

    Date of application for amendment: January 11, 1995
    Brief description of amendment: This amendment changes Technical 
Specification (TS) 3/4.3.8, ``Turbine Overspeed Protection System,'' 
removing these requirements from the TS and relocating the Bases to the 
Hope Creek Updated Final Safety Analysis Report (UFSAR) and the 
Surveillance Requirements to the applicable surveillance procedures. 
The Limiting Conditions for Operation (LCOs) are eliminated.
    Date of issuance: September 25, 1995
    Effective date: As of the date of issuance and shall be implemented 
within 60 days.
    Amendment No.: 81
    Facility Operating License No. NPF-57: This amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: August 2, 1995 (60 FR 
39451). The Commission's related evaluation of the amendment is 
contained in a Safety Evaluation dated September 25, 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. 50-354, Hope 
Creek Generating Station, Salem County, New Jersey

    Date of application for amendment: September 29, 1994
    Brief description of amendment: This amendment changes Technical 
Specification (TS) Sections 3/4.3.7.2, ``Seismic Monitoring 
Instrumentation,'' and 3/4.3.7.3, ``Meteorological Instrumentation,'' 
to remove the requirements from the TS and relocate the appropriate 
descriptive information and testing requirements to the Hope Creek 
Updated Final Safety Analysis Report.
    Date of issuance: September 25, 1995
    Effective date: As of the date of issuance, to be implemented 
within 60 days.
    Amendment No.: 82
    Facility Operating License No. NPF-57: This amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: August 2, 1995 (60 FR 
39449). The Commission's related evaluation of the amendment is 
contained in a Safety Evaluation dated September 25, 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 Nos. 50-272 and 50-
311, Salem Nuclear Generating Station, Unit Nos. 1 and 2, Salem 
County, New Jersey

    Date of application for amendments: September 20, 1994
    Brief description of amendments: The amendments change the Channel 
Functional Test surveillance frequency for the Manual Reactor Trip 
Switches and Reactor Trip Breakers (RTB) and relocate the RTB 
maintenance requirements from the Technical Specifications to the Salem 
Updated Final Safety Analysis Report.
    Date of issuance: September 18, 1995
    Effective date: Both units, as of the date of issuance, to be 
implemented within 60 days.
    Amendment Nos.: 176 and 157
    Facility Operating License Nos. DPR-70 and DPR-75. The amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: November 9, 1994 (59 FR 
55890 The Commission's related evaluation of the amendments is 
contained in a Safety Evaluation dated September 18, 1995. No 
significant hazards consideration comments received: No
    Local Public Document Room location: Salem Free Public Library, 112 
West Broadway, Salem, New Jersey 08079

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

    Date of application for amendments: January 21, 1994, as 
supplemented June 28 and September 13, 1994, and April 4, 1995.
    Brief description of amendments: Revised Technical Specifications 
3.8.2.3, ``125-Volt D.C. DISTRIBUTION - OPERATING.''
    Date of issuance: September 19, 1995
    Effective date: Both units, as of the day of issuance and shall be 
implemented within 60 days.
    Amendment Nos.: 177 and 158
    Facility Operating License Nos. DPR-70 and DPR-75. The amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: April 28, 1994 (58 FR 
22012) The June 28 and September 13, 1994, and April 4, 1995 letters 
provided clarifying information that did not change the scope of the 
January 21, 1994 application and initial proposed no significant 
hazards consideration determination, nor go beyond the scope of the 
Federal Register notice. The Commission's related evaluation of the 
amendments is contained in a Safety Evaluation dated September 19, 
1995. No significant hazards consideration comments received: No
    Local Public Document Room location: Salem Free Public Library, 112 
West Broadway, Salem, New Jersey 08079

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

    Date of application for amendment: June 19, 1995, as supplemented 
on August 21, 1995.
    Brief description of amendment: The amendment revises the Technical 
Specifications to change the required test frequency for the reactor 
building spray nozzle flow test from once per five years to once per 
ten years.
    Date of issuance: September 18, 1995
    Effective date: September 18, 1995
    Amendment No.: 127
    Facility Operating License No. NPF-12: Amendment revises the 
Technical Specifications.
    Date of initial notice in Federal Register: July 19, 1995 (60 FR 
37100). The August 21, 1995 letter provided supplemental 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 September 18, 1995. No 
significant hazards consideration comments received: No
    Local Public Document Room location: Fairfield County Library, 300 
Washington Street, Winnsboro, SC 29180

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

    Date of application for amendment: July 28, 1995

[[Page 52943]]

    Brief description of amendment: The amendment revises the Technical 
Specifications to exclude the requirement to perform the slave relay 
test of the 36-inch containment purge supply and exhaust valves on a 
quarterly basis while the plant is in Modes 1, 2, 3, or 4.
    Date of issuance: September 18, 1995
    Effective date: September 18, 1995
    Amendment No.: 128
    Facility Operating License No. NPF-12. Amendment revises the 
Technical Specifications.
    Date of initial notice in Federal Register: August 16, 1995 (60 FR 
42608) The Commission's related evaluation of the amendment is 
contained in a Safety Evaluation dated September 18, 1995. No 
significant hazards consideration comments received: No
    Local Public Document Room location: Fairfield County Library, 300 
Washington Street, Winnsboro, SC 29180

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

    Date of application for amendment: June 19, 1995, as supplemented 
on August 21, 1995.
    Brief description of amendment: The amendment revises the Technical 
Specifications to change the required test frequency for the reactor 
building spray nozzle flow test from once per five years to once per 
ten years.
    Date of issuance: September 18, 1995
    Effective date: September 18, 1995
    Amendment No.: 129
    Facility Operating License No. NPF-12. Amendment revises the 
Technical Specifications.
    Date of initial notice in Federal Register: July 19, 1995 (60 FR 
37100). The August 21, 1995 letter provided supplemental 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 September 18, 1995. No 
significant hazards consideration comments received: No
    Local Public Document Room location: Fairfield County Library, 300 
Washington Street, Winnsboro, SC 29180

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: April 3, 1995
    Brief description of amendment: The amendment revised the Technical 
Specifications (TS) to relocate radiological effluent and radiological 
environmental monitoring TS to the Offsite Dose Calculation Manual or 
to the Process Control Program. Programmatic controls for radioactive 
effluent and radiological environmental monitoring were included in TS 
6.8.4.
    Date of issuance: September 15, 1995
    Effective date: September 15, 1995
    Amendment No.: 72
    Facility Operating License No. NPF-58: This amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: May 10, 1995 (60 FR 
24921) The Commission's related evaluation of the amendment is 
contained in a Safety Evaluation dated September 15, 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: June 1, 1995
    Brief description of amendment: The amendment revised the Technical 
Specifications to make them more restrictive regarding control rod 
drive scram time testing. CRD scram time testing would be required 
following maintenance prior to considering the CRD operable, and could 
be performed at any reactor pressure. Additional testing would be 
required when reactor coolant pressure is greater than or equal to 950 
psig and prior to 40 percent rated thermal power.
    Date of issuance: September 26, 1995
    Effective date: September 26, 1995
    Amendment No. 73
    Facility Operating License No. NPF-58: This amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: August 2, 1995 (60 FR 
39452) The Commission's related evaluation of the amendment is 
contained in a Safety Evaluation dated September 26, 1995. No 
significant hazards consideration comments received: No
    Local Public Document Room location: Perry Public Library, 3753 
Main Street, Perry, Ohio 44081

Washington Public Power Supply System, Docket No. 50-397, Nuclear 
Project No. 2, Benton County, Washington

    Date of application for amendment: January 14, 1992, as 
supplemented by letters dated February 10, 1995, and August 16, 1995.
    Brief description of amendment: The amendment revises technical 
specification surveillance requirements regarding demonstration of jet 
pump operability and corrects several administrative discrepancies.
    Date of issuance: September 18, 1995
    Effective date: September 18, 1995, to be implemented within 30 
days of issuance
    Amendment No.: 141
    Facility Operating License No. NPF-21: The amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: May 27, 1992 (57 FR 
22272) and March 29, 1995 (60 FR 16204). The August 16, 1995, 
supplemental letter provided additional clarifying information and did 
not change the initial no significant hazards consideration 
determination. The Commission's related evaluation of the amendment is 
contained in a Safety Evaluation dated September 18, 1995.No 
significant hazards consideration comments received: No
    Local Public Document Room location: Richland Public Library, 955 
Northgate Street, Richland, Washington 99352
    Dated at Rockville, Maryland, this 3rd day of October 1995.
    For the Nuclear Regulatory Commission
Elinor G. Adensam,
Deputy Director, Division of Reactor Projects - III/IV, Office of 
Nuclear Reactor Regulation
[Doc. 95-25006 Filed 10-10-95; 8:45 am]
BILLING CODE 7590-01-F