[Federal Register Volume 60, Number 80 (Wednesday, April 26, 1995)]
[Notices]
[Pages 20513-20539]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: 95-10127]



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NUCLEAR REGULATORY COMMISSION

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

I. Background

    Pursuant to Public Law 97-415, the U.S. Nuclear Regulatory 
Commission (the Commission or NRC staff) is publishing this regular 
biweekly notice. Public Law 97-415 revised section 189 of the Atomic 
Energy Act of 1954, as amended (the Act), to require the Commission to 
publish notice of any amendments issued, or proposed to be issued, 
under a new provision of section 189 of the Act. This provision grants 
the Commission the authority to issue and make immediately effective 
any amendment to an operating license upon a determination by the 
Commission that such amendment involves no significant hazards 
consideration, notwithstanding the pendency before the Commission of a 
request for a hearing from any person.
    This biweekly notice includes all notices of amendments issued, or 
proposed to be issued from March 31, 1995, through April 14, 1995. The 
last biweekly notice was published on Wednesday, April 12, 1995 (60 FR 
18621).

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 [[Page 20514]] 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 May 26, 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 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. [[Page 20515]] 
    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.

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

    Date of amendment request: March 20, 1995.
    Description of amendment request: The licensee proposes a revision 
to Technical Specification (TS) 2.2.1, Reactor Trip System 
Instrumentation Setpoints, and to relocate cycle specific Overpower and 
Overtemperature Delta T trip setpoint parameters to the Core Operating 
Limits Report (COLR).
    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:

    This change does not involve a significant hazards consideration 
for the following reasons:
    1. The proposed amendment does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    The proposed change of relocating Overtemperature Delta T * * * 
and the Overpower Delta T * * * trip setpoint parameters to the COLR 
has no influence or impact to the probability or consequences of an 
accident. The revised TS will continue to implement the Reactor Trip 
System Instrumentation [Overtemperature Delta T] and [Overpower 
Delta T] setpoint limits through reference to the parameters in the 
COLR. In addition, the COLR is subject to the existing controls of 
TS 6.9.1.6, including the establishment of the parameter values 
using an NRC approved methodology. Given that this change 
administratively relocates the selected trip setpoint parameter 
values to another TS-controlled document, there would be no 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.
    No safety-related equipment, safety function, or plant operation 
will be altered as a result of this proposed change. The limits are 
simply being relocated to another TS-controlled document. The TS 
will continue to require operation within the required limits as 
established per NRC approved methodologies. Therefore, the proposed 
changes do not create the possibility of a new or different kind of 
accident from any accident previously evaluated.
    3. The proposed amendment does not involve a significant 
reduction in the margin of safety.
    Relocation of the Reactor Trip System Instrumentation 
[Overtemperature Delta T] and [Overpower Delta T] setpoint limits to 
the TS-controlled COLR has no effect on the trip system setpoints 
currently in force in TS 2.2.1. Future revisions to the trip 
setpoint parameters are governed by TS 6.9.1.6. TS 6.9.1.6 lists 
each TS that references values in the COLR and the NRC approved 
methodologies utilized in developing those values. Since this change 
is only an administrative relocation of the selected trip setpoint 
parameter values to another TS controlled document, the proposed 
changes do not involve a significant reduction in a margin of 
safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: Cameron Village Regional 
Library, 1930 Clark Avenue, Raleigh, North Carolina 27605.
    Attorney for licensee: R. E. Jones, General Counsel, Carolina Power 
& Light Company, Post Office Box 1551, Raleigh, North Carolina 27602.
    NRC Project Director: David B. Matthews.

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

    Date of amendment request: March 30, 1995.
    Description of amendment request: The licensee proposes to revise 
the Emergency Diesel Generator (EDG) surveillance requirements 
contained in Technical Specification (TS) 4.8.1.1.2 to be consistent 
with NUREG-1431, Standard Technical Specifications for Westinghouse 
Plants, and to eliminate the need for duplicate EDG testing that has 
already been implemented to satisfy the requirements of the Station 
Blackout Rule and the Maintenance Rule.
    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:

    This change does not involve a significant hazards consideration 
for the following reasons:
    1. The proposed amendment does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    A failure of the Emergency Diesel Generator (EDG) is not an 
initiator for any previously evaluated FSAR Chapter 15 accident 
scenario. By committing to and implementing an EDG reliability 
program that satisfies the requirements of the Station Blackout Rule 
and the Maintenance Rule, the Shearon Harris Nuclear Power Plant 
(SHNPP) will continue to ensure that target EDG reliability and 
availability is being achieved by conducting appropriate monitoring, 
testing, and maintenance activities. This program will be developed 
and controlled as a Plant Operating Manual procedure and will 
incorporate industry, vendor, and TDI Owners Group recommendations. 
Therefore, with commensurate levels of testing and inspection in 
place to provide assurance that the EDGs will perform their intended 
safety function in the event of an accident, the proposed changes 
will have no effect on the probability or consequences of such an 
accident.
    2. The proposed amendment does not create the possibility of a 
new or different kind of accident from any accident previously 
evaluated.
    A failure of the EDG is not an initiator for any previously 
evaluated FSAR Chapter 15 accident scenario nor would the proposed 
changes to the EDG surveillance requirements result in the 
possibility of a new or different kind of accident from any accident 
previously evaluated. By committing to and implementing an EDG 
reliability program that satisfies the requirements of the Station 
Blackout Rule and the Maintenance Rule, SHNPP will continue to 
ensure that target EDG reliability and availability is being 
achieved by conducting appropriate monitoring, testing, and 
maintenance activities. This program will be developed and 
controlled as a Plant Operating Manual procedure and will 
incorporate industry, vendor, and TransAmerica Delaval Inc. Owners 
Group recommendations. Therefore, with commensurate levels of 
testing and inspection in place to provide assurance that the EDGs 
will perform their intended safety function in the event of an 
accident, the proposed changes would not increase the possibility of 
a new or different kind of accident from any previously evaluated.
    3. The proposed amendment does not involve a significant 
reduction in the margin of safety.
    The proposed changes will not affect any parameters which relate 
to the margin of safety as defined in the Technical Specifications 
or the FSAR. Testing, inspection and maintenance necessary to verify 
the EDGs' ability to perform their intended safety function will 
continue to be [[Page 20516]] performed. Therefore, the proposed 
change does not involve a significant reduction in a margin of 
safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: Cameron Village Regional 
Library, 1930 Clark Avenue, Raleigh, North Carolina 27605.
    Attorney for licensee: R. E. Jones, General Counsel, Carolina Power 
& Light Company, Post Office Box 1551, Raleigh, North Carolina 27602.
    NRC Project Director: David B. Matthews.

Commonwealth Edison Company, Docket Nos. 50-295 and 50-304, Zion 
Nuclear Power Station, Units 1 and 2, Lake County, Illinois

    Date of amendment request: March 24, 1995.
    Description of amendment request: The proposed amendments would 
acknowledge the acceptability of performing containment leakage rate 
testing in accordance with 10 CFR Part 50, Appendix J, and all approved 
exemptions.
    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 does not involve a significant 
increase in the probability of occurrence or consequences of any 
accident previously evaluated.
    The proposed changes to Technical Specifications to add an 
allowance to test in accordance with approved exemptions to 10 CFR 
50 Appendix J are administrative in nature and will not affect any 
accident initiators or precursors. 10 CFR 50 Appendix J provides the 
requirements to periodically test the primary reactor containment. 
The objective of these requirements is to ensure that leakage from 
the primary reactor containment structure and systems and components 
that penetrate the containment is maintained below the limits 
established for containment leakage. The performance of periodic 
integrated leakage rate testing (Type A) and local penetration 
testing (Type B and C) during containment life provides a current 
assessment of potential leakage from containment during accident 
conditions.
    10 CFR 50.12 allows the Commission to grant specific exemptions 
to the requirements of 10 CFR 50 Appendix J when those exemptions 
are authorized by law, will not present undue risk to the public, 
and are consistent with the common defense and security. In 
addition, special circumstances must exist as described in Section 
50.12. Since all exemptions to 10 CFR 50 Appendix J receive NRC 
review and approval prior to being implemented, all containment 
leakage rate testing will continue to be performed in accordance 
with NRC approved methodologies when relying upon the allowance that 
is added to the Technical Specifications by the proposed amendment. 
The proposed changes are consistent with the requirements provided 
in NUREG-1431, ``Standardized Technical Specifications, Westinghouse 
Plants'' which has been approved by the NRC.
    The proposed changes will not affect any accident initiators or 
precursors and will not change or alter the design assumptions for 
the systems used to mitigate the consequences of an accident. The 
proposed changes do not involve the addition of any new or different 
type of equipment, nor do they involve the operation of equipment 
required for safe operation of the facility in a manner different 
from those addressed in the UFSAR. There are no changes to 
parameters governing plant operation as a result of the proposed 
changes. The results and conclusions in the Zion Updated Final 
Safety Analysis Report (UFSAR) are unaffected by this proposed 
License Amendment.
    Based on the previous discussion, the proposed changes do not 
involve a significant increase in the probability of occurrence or 
consequences of any accident previously evaluated.
    2. The proposed amendment does not create the possibility of a 
new or different kind of accident from any previously analyzed.
    The proposed changes to Technical Specifications to add an 
allowance to perform containment leakage rate testing in accordance 
with approved exemptions to 10 CFR 50 Appendix J are administrative 
in nature and do not involve the addition of any new or different 
types of safety related equipment, nor does it involve the operation 
of equipment required for safe operation of the facility in a manner 
different from those addressed in the safety analyses. The proposed 
changes may only affect the methods used to perform containment 
leakage rate testing while in a shutdown condition. No safety 
related equipment or function will be altered as a result of the 
proposed changes. Also, the procedures governing normal plant 
operation and recovery from an accident are not changed by the 
proposed Technical Specification changes. Since no new failure modes 
or mechanisms are added by the proposed changes, the possibility of 
a new or different kind of accident is not created.
    3. The proposed changes do not involve a significant reduction 
in a margin of safety.
    Given the proposed changes to Technical Specifications, testing 
would be allowed in accordance with approved exemptions to Appendix 
J. Exemptions are allowed by the Commission in accordance with 10 
CFR 50.12 when it is shown that the exemption is authorized by law, 
will not present an undue risk to the public health and safety, and 
is consistent with the common defense and security. In addition, 
special circumstances must exist.
    The proposed changes will not impact any margin of safety and 
testing in accordance with approved exemptions will not involve a 
significant reduction in a level of safety since containment leakage 
testing is performed while in a shutdown condition. In addition, it 
is likely that any test methodology that significantly reduces a 
margin of safety would not be approved by the NRC.
    The ability to safely shut down the operating unit and mitigate 
the consequences of all accidents previously evaluated will be 
maintained. Therefore, the margin of safety is not significantly 
affected.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
requested amendments involve no significant hazards consideration.
    Local Public Document Room location: Waukegan Public Library, 128 
N. County Street, Waukegan, Illinois 60085.
    Attorney for licensee: Michael I. Miller, Esquire; Sidley and 
Austin, One First National Plaza, Chicago, Illinois 60690.
    NRC Project Director: Robert A. Capra.

Consumers Power Company, Docket No. 50-155, Big Rock Point Plant, 
Charlevoix County, Michigan

    Date of amendment request: December 15, 1994.
    Description of amendment request: The proposed amendment would 
revise Technical Specification 11.3.1.5 ACTION a. to eliminate the need 
to demonstrate that the actuation circuitry of the unaffected reactor 
depressurization system channels is operable. In addition, an editorial 
change correcting a typographical error is also proposed.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Will the proposed change involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    The proposed change will eliminate the probability of a 
subsystem failure caused by additional testing (which unnecessarily 
introduces the potential for human and equipment problems), 
therefore eliminating the probability that the facility would have 
to be challenged and brought to the SHUTDOWN condition within 12 
hours and to the COLD SHUTDOWN condition within the following 24 
hours.
[[Page 20517]]

    2. Will the proposed change create the possibility of a new or 
different kind of accident from any accident previously evaluated?
    The proposed change does not alter the plant configuration, 
systems, components, or operation; and does not create the 
possibility of a new or different kind of accident from any accident 
previously evaluated. The proposed change is expected to eliminate 
unnecessary challenges to a safety system that has already been 
determined to be operable by routine surveillance testing; therefore 
contributing to the overall safe operation of the facility.
    3. Will the proposed change involve a significant reduction in 
the margin of safety?
    The RDS [Reactor Depressurization System] provides for both 
manual and automatic depressurization of the primary system to allow 
injection of the core spray following a small-to-intermediate size 
break in the primary system. This will allow core cooling with the 
objective of preventing excessive fuel clad temperatures. The design 
of the system is based on the specified initiation set points 
described in the Technical Specifications. Transient analysis 
demonstrated that these conditions result in adequate safety margins 
for both the fuel and the system pressure. The proposed change does 
not affect these setpoints, therefore the margin of safety is not 
changed.

    In addition, the proposed editorial change to correct a 
typographical error is administrative in nature and, therefore, would 
have no effect on the three standards of 10 CFR 50.92 discussed above.
    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: North Central Michigan 
College, 1515 Howard Street, Petoskey, Michigan 49770.
    Attorney for licensee: Judd L. Bacon, Esquire, Consumers Power 
Company, 212 West Michigan Avenue, Jackson, Michigan 49201.
    NRC Project Director: Cynthia A. Carpenter, Acting.

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

    Date of amendment request: January 26, 1995, as supplemented March 
9, 1995.
    Description of amendment request: The proposed amendment would 
revise the technical specifications (TS) to increase the allowable 
nominal fuel enrichment from 4.2 to 5.0 weight percent for reload fuel 
assemblies. TS impose a limit on fuel enrichment of stored fuel 
assemblies to prevent inadvertent criticality. Presently, the Crystal 
River Unit 3 (CR3) TS specify a maximum enrichment of 4.5 weight 
percent for storage pool A and dry fuel (new fuel) storage racks, and 
4.2 weight percent for fuel pool B. The licensee proposed to revise TS 
3.7.15, 4.2, and 4.3, and associated TS bases to allow increasing the 
enrichment limits from 4.2 to 5.0 weight percent for the dry fuel 
storage racks and for A and B fuel pools. Additionally, a typographical 
error in TS 4.3.1.2.b will also be corrected.
    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. This amendment will not involve a significant increase in the 
probability or consequences of an accident previously evaluated.
    An increase in fuel enrichment will not by itself affect the 
mixture of fission product nuclides. A change in fuel cycle design 
which makes use of an increased enrichment may result in fuel burnup 
consisting of a somewhat different mixture of nuclides. The effect 
in this instance is insignificant because:
    a. The isotopic mixture of the irradiated assembly is relatively 
insensitive to the assembly's initial enrichment.
    b. Most accident doses are such a small fraction of 10 CFR 100 
limits, a large margin exists before any change becomes significant.
    c. The change in Pu content which would result from an increase 
in burnup would produce more of some fission product nuclides and 
less of other nuclides. Small increases in some doses are offset by 
reductions in other doses. The radiological consequences of 
accidents are not significantly changed.
    2. This amendment will not create the possibility of a new or 
different kind of accident from any accident previously evaluated.
    An unplanned criticality event will not occur as Keff 
[effective neutron multiplication factor] will not exceed 0.95 with 
the maximum allowable enriched fuel in Pool A and Pool B, when 
flooded with unborated water, and Keff will not exceed 0.98 in 
the new fuel storage racks assuming conditions of optimum 
hypothetical low density moderation. The new fuel storage racks have 
two rows of storage cells physically blocked to ensure reactivity 
limits are not exceeded. Administrative controls assure fuel is 
stored in configurations which meet the requirements of the safety 
analysis.
    3. This amendment will not involve a significant reduction in a 
margin of safety.
    While the increased enrichment in Pool A, Pool B, and the dry 
storage racks may lessen the margin to criticality, this reduction 
is not significant because the overall safety margin is within NRC 
criteria of Keff [less than or equal to] 0.95 (NRC Standard 
Review Plan, Section 9.1.2.)
    Therefore, this amendment request satisfies the criteria 
specified in 10 CFR 50.92 for amendments which do 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 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: Coastal Region Library, 8619 
W. Crystal Street, Crystal River, Florida 32629.
    Attorney for licensee: A. H. Stephens, General Counsel, Florida 
Power Corporation, MAC-A5D, P. O. Box 14042, St. Petersburg, Florida 
33733.
    NRC Project Director: David B. Matthews.

Houston Lighting & Power Company, City Public Service Board of San 
Antonio, Central Power and Light Company, City of Austin, Texas, Docket 
Nos. 50-498 and 50-499, South Texas Project, Units 1 and 2, Matagorda 
County, Texas

    Date of amendment request: March 16, 1995.
    Description of amendment request: The proposed amendment would 
revise Technical Specification 4.6.1.2, regarding the overall 
integrated containment leakage rate tests, so that it would reference 
10 CFR Part 50, Appendix J directly, rather than paraphrase the 
regulation, and allow approved exemptions to the test frequency 
requirements. In addition, there is an associated proposed exemption, 
from the requirements of 10 CFR Part 50, Appendix J, to provide a one-
time interval extension for the Unit 2 Type A test (containment 
integrated leak rate test) from the current scheduled 48 months to 
approximately 66 months.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

Criterion 1--Does Not Involve a Significant Increase in the Probability 
or Consequences of an Accident Previously Evaluated

    The proposed change specific to Unit 2 will provide a onetime 
exemption from the 10 CFR 50, Appendix J Section III.D.I.(a) leak 
rate test schedule requirement. This change will allow for a one-
time test interval for Type A Integrated Leak Rate Tests of 
approximately 66 months.
    Leak rate testing is not an initiating event in any accident; 
therefore, this proposed [[Page 20518]] change does not involve a 
significant increase in the probability of a previously evaluated 
accident.
    Type A tests are capable of detecting both local leak paths and 
gross containment failure paths. Experience at South Texas Project 
Unit 2 demonstrates that excessive containment leakage paths are 
local leakage detected by Type B and C Local Leak Rate Tests.
    Administrative controls govern maintenance and testing of 
containment penetrations such that the probability of excessive 
penetration leakage due to improper maintenance or valve 
misalignment is very low. Following maintenance on any containment 
penetration, a Local Leak Rate Test is performed to ensure 
acceptable leakage levels. Following a Local Leak Rate Test on a 
containment isolation valve, an independent valve alignment check is 
performed. Therefore, Type A testing is not necessary to ensure 
acceptable leakage rates through containment penetrations.
    While Type A testing is not necessary to ensure acceptable 
leakage rates through containment penetrations, Type A testing is 
necessary to demonstrate that there are no gross containment 
failures. Structural failure of the containment is considered to be 
a very unlikely event, and in fact, since South Texas Project Unit 2 
has been in operation, it has successfully passed each Type A 
Integrated Leak Rate Test. Therefore, a one-time exemption 
increasing the interval for performing an Integrated Leak Rate Test 
results [sic] in a significant decrease in the confidence in the 
leak tightness of the containment structure. Therefore, this change 
does not involve a significant increase in the consequences of an 
accident previously evaluated.
    The proposed amendment revised Technical Specification 4.6.1.2 
to reference the testing frequency requirements of 10 CFR 50, 
Appendix J, and to state that Nuclear Regulatory Commission approved 
exemptions to the applicable regulatory requirements are permitted. 
This portion of the proposed change is applicable to Unit 1 and Unit 
2. The current language of Technical Specification 4.6.1.2 
paraphrases the requirements of Section III,D.I.(a) [sic] of 
Appendix J. The proposed administrative revision simply deletes the 
paraphrased language and directly references Appendix J. No new 
requirements are added, nor are any existing requirements deleted. 
Any specific changes to the requirements of Section III.D.I.(a) will 
require a submittal from Houston Lighting & Power under 10 CFR 50.12 
and subsequent review and approval by the Nuclear Regulatory 
Commission prior to implementation.
    The proposed amendment, in itself, does not affect reactor 
operations or accident analysis and has no radiological 
consequences. The change provides clarification so that future 
Technical Specification changes will not be necessary to correspond 
to applicable Nuclear Regulatory Commission-approved exemptions from 
the requirements of Appendix J.
    Therefore, this proposed amendment does not involve a 
significant increase in the probability or consequences of any 
accident previously evaluated.

Criterion 2--Does Not Create the Possibility of a New or Different Kind 
of Accident From Any Previously Evaluated

    The proposed Unit 2 exemption request does not affect normal 
plant operations or configuration, nor does it affect leak rate test 
methods. The proposed change allows a one-time test interval of 
approximately 66 months for the Integrated Leak Rate Test. Because 
the test history of South Texas Project Unit 2 demonstrates no Type 
A test failures during plant lifetime, the relaxation in schedule 
should not significantly decrease the confidence in the leak 
tightness of the containment.
    The proposed Technical Specification amendment for Units 1 and 2 
provides clarification to a specification that paraphrases a 
codified requirement.
    Since the proposed change and amendment would not change the 
design, configuration or method of operation of the plant, they 
would not create the possibility of a new or different kind of 
accident from any previously evaluated.

Criterion 3--Does Not Involve a Significant Reduction in the Margin of 
Safety

    The purpose of the existing schedule for Integrated Leak Rate 
Tests is to ensure that release of radioactive materials will be 
restricted to those leak paths and leak rates assumed in accident 
analyses. The relaxed schedule for Integrated Leak Rate Tests does 
not allow for relaxation of Type B and C Local Leak Rate Tests. 
Therefore, methods for detecting local containment leak paths and 
leak rates are unaffected by this proposed change. A one-time 
increase of the South Texas Project Unit 2 test interval does not 
leak to a significant probability of creating a new leakage path or 
increased leakage rates because the test history for Integrated Leak 
Rate Tests shows no failure during plant life. The margin of safety 
inherent in existing accident analyses is maintained.
    The proposed Technical Specification amendment for Units 1 and 2 
is administrative and clarifies the relationship between the 
requirements of Technical Specification 4.6.1.2, Appendix J, and any 
approved exemptions to Appendix J. It does not, in itself, change a 
safety limit, a Limiting Condition of Operation, or a surveillance 
requirement on equipment required to operate the plant. Nuclear 
Regulatory Commission approval of any proposed change or exemption 
to III.D.1.(a) of Appendix J will be required prior to 
implementation.
    Therefore, this change and amendment do not involve a 
significant reduction in the margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
request for amendments involves no significant hazards consideration.
    Local Public Document Room location: Wharton County Junior College, 
J. M. Hodges, Learning Center, 911 Boling Highway, Wharton, TX 77488.
    Attorney for licensee: Jack R. Newman, Esq., Newman & Holtzinger, 
P.C., 1615 L Street, NW., Washington, DC 20036.
    NRC Project Director: William D. Beckner.

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

    Date of amendment request: March 10, 1995.
    Description of amendment request: The proposed amendment would 
remove redundant Limiting Conditions of Operation and Surveillance 
Requirements for the containment hydrogen and oxygen monitors in the 
Technical Specifications (TS).
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. The proposed amendment does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated. No physical changes will result from this 
amendment. This change deletes requirements that are redundant and 
unduly restrictive. The annual surveillance deleted by this 
amendment is redundant to the semi-annual surveillance required in 
Table 4.2-H. The Limiting Conditions for Operation are not changed 
by the proposed amendment.
    2. The proposed amendment does not create the possibility of a 
new or different kind of accident from any accident previously 
evaluated. No physical changes will result from this amendment. 
Functional tests are performed on the hydrogen and oxygen analyzers 
semiannually as required in TS Table 4.2-H. Deleting the annual 
requirement for a functional test of the same equipment will not 
reduce the amount of testing performed or increase the possibility 
of degraded equipment being undetected.
    3. The proposed amendment does not involve a significant 
reduction in a margin of safety. No physical changes will result 
from this amendment. The existing requirement for a semi-annual test 
of the hydrogen and oxygen analyzer in Table 4.2-H exceeds the 
requirements to be deleted in Section 3.7/4.7-H. The frequency of 
testing of the hydrogen and oxygen analyzers will not be reduced as 
a result of this amendment.

    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. 
[[Page 20519]] 
    Local Public Document Room location: Cedar Rapids Public Library, 
500 First Street, S.E., Cedar Rapids, Iowa 52401.
    Attorney for licensee: Jack Newman, Kathleen H. Shea, Morgan, Lewis 
& Bockius, 1800 M Street, NW., Washington, DC 20036-5869.
    NRC Project Director: Gail H. Marcus.

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

    Date of amendment request: March 28, 1995.
    Description of amendment request: The proposed amendment would 
revise and clarify Technical Specification Table 3.2-A that lists 
allowable out-of-service times and surveillance test intervals for 
instrumentation.
    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 to TS Table 3.2-A will not significantly 
increase the probability or consequences of an accident previously 
evaluated. The changes do not alter the physical design or operation 
of the plant and serve to describe more accurately and clearly the 
actual logic configurations. The existing logic designs are in 
conformance with the Architect/Engineer's design documentation since 
plant startup. These changes will assure that the information in the 
tables is clearer and more consistent with the column headings of 
the table. The proposed changes do not affect assumptions contained 
in the plant safety analysis.
    The Bases changes provide additional information about the logic 
arrangements as appropriate to identify unique or different logic 
configurations. Changes to the Allowed Outage Time (AOT) 
descriptions for the MSL Flow--High and MSL Tunnel Temperature--High 
provide clarification regarding application of the AOT to these 
logic arrangements, since multiple instrument channels provide input 
into multiple logic channels. This application conforms to the 
single failure criterion of the design basis (NEDO-10139, Compliance 
of Protection Systems to Industry Criteria: General Electric BWR 
Nuclear Steam Supply System, dated June 1970) and to the analytical 
basis for the TS (NEDC-31677P-A, Technical Specification Improvement 
Analysis for BWR Isolation Actuation Instrumentation, dated July 
1990).
    2. The proposed changes to Table 3.2-A will not introduce a new 
or different kind of accident from any accident previously 
evaluated. The changes do not alter the physical design of the plant 
or affect any modes of operation. The proposed changes serve to 
clarify the existing information to better assure that the trip 
instrumentation will be maintained as assumed in the accident 
analyses contained in the Updated Final Safety Analysis Report.
    3. The proposed changes do not involve a significant reduction 
in a margin of safety. Clarification of the logic arrangements in 
both Table 3.2-A and the TS Bases and how the AOT is applied does 
not affect the ability of the isolation logic to perform its 
intended function. No physical changes to the plant are being made 
as part of this amendment.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: Cedar Rapids Public Library, 
500 First Street, S.E., Cedar Rapids, Iowa 52401.
    Attorney for licensee: Jack Newman, Kathleen H. Shea, Morgan, Lewis 
& Bockius, 1800 M Street, NW., Washington, DC 20036-5869.
    NRC Project Director: Gail H. Marcus.

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

    Date of amendment request: March 17, 1995.
    Description of amendment request: The proposed amendment would 
defer performance of the Type A containment integrated leakage rate 
test until the next refueling outage.
    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:

    Per 10 CFR 50.92, a proposed change does not involve a 
significant hazards consideration if the change does not:
    1. Involve a significant increase in the probability or 
consequences of an accident previously evaluated,
    2. Create the possibility of a new or different kind of accident 
from any accident previously evaluated, or
    3. Involve a significant reduction in a margin of safety.

Criterion 1

    The Cook Nuclear Plant Type A test history provides substantial 
justification for the proposed test schedule. Three Type A tests 
were performed over a seven year period with successful results. The 
tests indicate that the Cook Nuclear Plant has a low leakage 
containment. In addition, there are no adverse trends in the results 
from the previous Types A, B, and C tests or visual inspections that 
indicate a gradual degradation of the containment boundary. Further, 
there are no structural modifications planned which would adversely 
affect the structural capability of the containment and that would 
be a factor in deferring the Type A test one refueling outage. 
Containment leak rate testing is not an initiator of any accident. 
The proposed interval extension does not affect reactor operations 
or the accident analysis and has no radiological consequences, 
except for the dose savings associated with not performing the test. 
There will be no changes to 10 CFR 100 dose limits or the control 
room dose limits. Extending the test interval will not increase the 
probability of a malfunction of equipment important to safety. Based 
on these considerations, it is concluded that the change does not 
involve a significant increase in the probability or consequences of 
an accident previously evaluated.

Criterion 2

    The proposed change does not involve physical changes to the 
plant or changes in plant operating configuration. The proposed 
change only relaxes the schedular requirements for conducting one 
Type A test from the T/Ss and defers performance of the test one 
cycle. The purpose of the test is to provide periodic verification 
of the leak-tight integrity of the primary reactor containment, and 
systems and components which penetrate containment. The tests assure 
that leakage through containment and systems and components 
penetrating containment will not exceed the allowable leak rate 
values established in 10 CFR 50, Appendix J. Thus, it is concluded 
that the proposed change does not create the possibility of a new or 
different kind of accident from any accident previously evaluated.

Criterion 3

    The proposed change to the schedule for performing the Type A 
test does not reduce the margin of safety assumed in the accident 
analysis for any release of radioactive materials or reduce any 
margin of safety preserved by the technical specifications. The 
methodology, acceptance criteria, and the technical specification 
leak rate limits for the performance of the Type A test will not 
change. Type A tests will continue to be performed in accordance 
with 10 CFR 50, Appendix J and the applicable Cook Nuclear Plant 
Technical Specifications beginning in 1997. In addition, there are 
no adverse trends in the results from the previous Type A, B, and C 
tests or visual inspections that indicate a gradual degradation of 
the containment boundary. Based on these considerations, it is 
concluded that the change does not involve a significant reduction 
in a margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: Maud Preston Palenske Memorial 
Library, 500 Market Street, St. Joseph, Michigan 49085.
    Attorney for licensee: Gerald Charnoff, Esq., Shaw, Pittman, Potts 
and [[Page 20520]] Trowbridge, 2300 N Street, NW, Washington, DC 20037.
    NRC Project Director: Cynthia A. Carpenter, Acting.

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

    Date of amendment requests: March 31, 1995.
    Description of amendment requests: The proposed amendments would 
modify the Containment Ventilation System Technical Specifications (and 
associated Bases) to allow limited containment purge operation in Modes 
1, 2, 3, and 4 for pressure control, ALARA [as low as is reasonably 
achievable], and respirable air quality considerations.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

Criterion 1

    The purpose of this amendment is to allow flexibility in the use 
of the containment purge system during MODES 1, 2, 3, and 4. The use 
of this system during these modes of operation has previously been 
approved (Amendment No. 66). Therefore, this amendment request does 
not involve a significant increase in the probability or 
consequences of an accident previously evaluated because the 
proposed change to the T/Ss does not affect the assumptions, 
parameters, or results of any UFSAR [Updated Final Safety Analysis 
Report] accident analysis. Based on the existing system design and 
demonstrated closure capability it is concluded that the proposed 
changes do not modify the response of the containment during a 
design basis accident. The proposed amendment does not add or modify 
any existing equipment. Based on these considerations, it is 
concluded that the changes do not involve a significant increase in 
the probability or consequences of an accident previously evaluated.

Criterion 2

    The proposed change does not involve physical changes to the 
plant or changes in the plant operating configuration. Thus, it is 
concluded that the proposed change does not create the possibility 
of a new or different kind of accident from any accident previously 
evaluated.

Criterion 3

    The margin for safety presently provided is not reduced by the 
proposed change. As discussed previously, the containment purge 
valves have been designed and demonstrated capable of closure 
against the dynamic forces resulting from a loss of coolant 
accident. The proposed amendment does not impact the ability of the 
purge valves to perform their intended function (i.e. achieve 
closure) in the event of an accident. Based on these considerations, 
it is concluded that the changes do not involve a significant 
reduction in a margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment requests involve no significant hazards consideration.
    Local Public Document Room location: Maud Preston Palenske Memorial 
Library, 500 Market Street, St. Joseph, Michigan 49085.
    Attorney for licensee: Gerald Charnoff, Esq., Shaw, Pittman, Potts 
and Trowbridge, 2300 N Street, NW, Washington, DC 20037.
    NRC Project Director: Cynthia A. Carpenter, Acting.

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

    Date of amendment request: March 31, 1995.
    Description of amendment request: The proposed amendment would 
revise the Technical Specifications (TS) to increase the as-found 
setpoint tolerance of the safety/relief valves (SRVs) from plus or 
minus 1% to plus or minus 3%. In addition, the proposed amendment (1) 
would allow the as-found condition of one SRV to be inoperable, (2) 
clarifies the 1325 psig safety limit wording, (3) increases the number 
of SRVs to be tested during each refueling outage, (4) makes editorial 
changes to reflect the TS changes, and (5) revises the bases for the 
applicable sections.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    NNECO has reviewed the proposed changes in accordance with 10 
CFR 50.92 and concluded that the changes do not involve a 
significant hazards consideration (SHC). The basis for this 
conclusion is that the three criteria of 10 CFR 50.92(c) are not 
compromised. The proposed changes do not involve an SHC because the 
changes would not:
    1. Involve a significant increase in the probability or 
consequences of an accident previously analyzed.
    The safety function of the SRVs is to mitigate the effects of a 
RPV [reactor pressure vessel] overpressurization, therefore a 
failure to open until the upper setpoint limit (+3%) is reached 
cannot affect the probability of an accident. The lowest allowable 
limit (-3%) is still above normal operating pressure and therefore 
does not significantly increase the probability of an inadvertent 
opening.
    Should the SRVs open in response to an RCS [reactor coolant 
system] overpressure event, opening of the SRVs below the nominal 
setpoints does not adversely affect the consequences of an accident. 
The fuel reload analysis demonstrates that actuation of five valves 
at or below 103% of nominal provides sufficient pressure reduction 
to maintain peak RCS pressure below the safety limit of 1375 psig 
and to maintain vessel steam space pressure below 1325 psig. The 
hydrodynamic loads on the SRV discharge pipe (i.e., tail pipe) and 
the torus remain within the design limits.
    The performance of the high pressure systems; FWCI [feedwater 
coolant injection], SLC [standby liquid control] and IC [isolation 
condenser] remain acceptable. There is also no adverse impact on the 
operability of the APR [automatic pressure relief] system.
    The SRV setpoints will continue to be required to be within 
[plus or minus] 1% prior to plant startup from a refueling outage. 
This ensures that the SRVs are restored to the optimal conditions at 
the start of each fuel cycle.
    Therefore, increasing the ``as-found'' tolerance from [plus or 
minus] 1% to [plus or minus] 3% does not result in a significant 
increase in the probability or consequences of a previously analyzed 
accident.
    2. Create the possibility of a new or different kind of accident 
from any previously analyzed.
    Revising the acceptable as-found setpoint tolerance from [plus 
or minus] 1% to [plus or minus] 3% does not change the type of 
action that these valves are expected to perform, nor does it change 
the initial ``as-left'' requirements for the valves. Plant operating 
parameters have not changed. Therefore, this change cannot create 
the possibility of a new or different kind of accident.
    3. Involve a significant reduction in the margin of safety.
    The margin of safety established and stated in the Millstone 
Unit No. 1 Technical Specifications, is a peak RCS pressure of 1375 
psig and a peak vessel steam space pressure of 1325 psig. While 
allowing the SRV setpoint tolerance to increase to [plus or minus] 
3% would allow peak pressures from an MSIV [main steam isolation 
valve] closure event to approach that safety limit, the safety limit 
will not be exceeded. Therefore, this change does not involve a 
significant reduction in the margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: Learning Resource Center, 
Three Rivers Community-Technical College, Thames Valley Campus, 574 New 
London Turnpike, Norwich, CT 06360. [[Page 20521]] 
    Attorney for licensee: Ms. L. M. Cuoco, Senior Nuclear Counsel, 
Northeast Utilities Service Company, Post Office Box 270, Hartford, CT 
06141-0270.
    NRC Project Director: Phillip F. McKee.

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

    Date of application for amendments: February 10, 1995.
    Description of amendment request: The proposed changes provide for 
the correction of administrative errors made in the past during the 
processing of technical specification changes related to control room 
ventilation filter surveillance testing.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. The proposed changes do not involve a significant increase in 
the probability or consequences of an accident previously evaluated 
because the changes are purely administrative and do not involve any 
physical changes to plant SSC [systems, structures, or components]. 
Therefore, these changes will not involve a significant increase in 
the probability or consequences of an accident previously evaluated.
    2. The proposed changes do not create the possibility of a new 
or different kind of accident from any accident previously evaluated 
because the changes will not alter the plant or the manner in which 
the plant is operated. The changes do not allow plant operation in 
any mode that is not already evaluated in the safety analysis. The 
changes will not alter assumptions made in the safety analysis and 
licensing bases. Therefore, these changes will not create the 
possibility of a new or different kind of accident from any accident 
previously evaluated.
    3. The proposed changes do not involve a significant reduction 
in a margin of safety because they are purely administrative and 
have no impact on any safety analysis assumptions.

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

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

    Date of amendment request: March 15, 1994.
    Description of amendment request: This amendment would reflect an 
exemption from 10 CFR Part 50, Appendix J, Section II.H.4, concerning 
the scope of Type `C' testing on specified emergency core cooling 
system and reactor core isolation cooling containment isolation valves 
by revising Technical Specification Table 3.6.3-1, Primary Containment 
Isolation Valves. The subject valves on systems which terminate below 
the minimum water level of the suppression pool and are associated with 
closed systems would be tested using requirements of the American 
Society of Mechanical Engineers' Section XI Code.
    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:

    I. This proposal does not involve a significant increase in the 
probability or consequences of an accident previously evaluated.
    The proposed change to the scope of Type `C' testing for the 
subject valves does not affect the probability of the design basis 
accidents. The valves will continue to be maintained in an operable 
state, and in their current design configuration. There is no 
correlation between the scope of the Type `C' testing and accident 
probability.
    PP&L reviewed the postulated consequences of design basis events 
on primary containment isolation under the proposed change. GDC 50 
design conformance states that the primary containment structure, 
including access openings, penetrations and the containment heat 
removal system, is designed so that the containment structure and 
its internal compartments can withstand, without exceeding the 
design leakage rate (1.0% per day), the peak accident pressure and 
temperature that could occur during any postulated LOCA.
    For the purposes of considering the consequences of LOCAs under 
the proposed change, a single active failure of a CIV or a passive 
failure of the closed system were reviewed, within the limits of the 
existing licensing basis. Under the existing licensing basis, a pipe 
rupture of seismically qualified ECCS piping does not have to be 
assumed concurrent with the LOCA, except if it is a consequence of 
the LOCA. Consequential failures can be eliminated, since a LOCA 
inside containment is separated from the ECCS piping by the 
containment structure. Consequential failures of the ECCS piping 
from LOCA's outside containment are outside the Appendix J design 
considerations, although they are adequately addressed through the 
redundancy and separation of the ECCS design. A single active 
failure of the CIV, under the LOCA condition, can be accommodated 
since the closed and filled system piping remains as the leakage 
barrier. The ECCS passive failure criterion does require 
consideration of system leaks, but not pipe breaks, beyond the 
initiating LOCA. Pipe leakage, equivalent to the leakage from a 
valve or pump seal failure, should be considered at 24 hours or 
greater post-LOCA. The capability to make-up inventory to the 
suppression pool is adequate to ensure that postulated seat leakage 
and pipe leakage does not result in a condition that jeopardizes 
pool level. Make-up capability exists to the suppression pool via 
the Condensate Storage Tank and Spray Pond. Actions to make-up to 
the suppression pool are delineated in Emergency Operating 
Procedures.
    Therefore, the proposal to eliminate the subject Type `C' tests 
does not involve a significant increase in the probability or 
consequences of an accident previously evaluated.
    II. This proposal does not create the possibility of a new or 
different kind of accident from any accident previously evaluated.
    The acceptability of the proposed change to the scope of Type 
`C' testing for the subject valves is based on maintaining the 
existing barriers to primary containment leakage, and ensuring that 
the suppression pool level is assured for 30 days during all design 
basis, post-accident modes of operation. By meeting these dual 
objectives, the plant response to the design basis events will be 
unchanged, and no new accident scenarios will be encountered. These 
two objectives are related, in that, the suppression pool inventory 
creates a passive barrier to primary containment atmospheric leakage 
for penetrations which are located below the minimum water level of 
the pool. The subject valve lines terminate below the minimum 
suppression pool water level.
    The subject valves are all single isolation valves associated 
with lines that penetrate the primary containment, but are not 
connected directly to the primary containment atmosphere or the 
reactor coolant pressure boundary. The redundant isolation boundary 
for each of the affected valves is the closed system associated with 
the valve. This configuration is described in General Design 
Criteria (GDC) 57. The proposed exemption, and Technical 
Specification change, does not alter the configuration of these 
systems. The valves will continue to be tested and maintained to 
ensure their operability. The closed system [[Page 20522]] piping 
meets PP&L's design conformance to GDC 56 and is verified via a 
10CFR50 Appendix J Type `A' test. The integrity of the closed 
systems is also monitored and controlled via Technical Specification 
6.8.4.a.
    The subject valves may be open, or change state, postaccident to 
support the design function of their associated ECCS systems (HPCI, 
Core Spray, RHR) or RCIC. The subject valves function as system 
valves during the periods when they are open or in an intermediate 
state, not as containment isolation valves. Reliance is placed on 
the suppression pool seal and the closed system piping to maintain 
the barrier between primary and secondary containment atmospheres.
    Therefore, with the valve and closed system configuration 
unaffected by the proposed change, the existing barriers to primary 
containment atmospheric leakage are maintained, so long as the 
suppression pool level is ensured.
    The suppression pool is designed and operated so that it is 
filled with water in accordance with Technical Specifications 3/
4.5.3, ``Suppression Chamber,'' 3/4.6.2, ``Depressurization 
Systems--Suppression Chamber,'' and the associated Bases. The supply 
of water in the suppression pool is assured for 30 days during all 
design basis, post-accident modes of operation. Type `C' leak rate 
testing has historically been performed on valves associated with 
lines that connect to the suppression pool. The acceptance criteria 
for combined leakage from these penetrations is 3.3 gpm. This 
leakage rate is at a level which ensures the 30 day post-accident 
suppression pool level. However, for the valves discussed in this 
change, seat leakage past the CIV is into a closed and filled 
system. Thus ``leakage'' from the suppression pool, past the CIV, is 
a function of closed system leakage.
    As mentioned above, the integrity of the closed system piping is 
verified via a 10CFR50 Appendix J Type `A' test and is monitored and 
controlled via Technical Specification 6.8.4.a. TS 6.8.4.a 
establishes a program to monitor and control leakage from systems 
located outside containment that could contain highly radioactive 
fluids during a serious transient or accident. This program applies 
to the ECCS systems and RCIC affected by the proposed change and 
ensures that leakage into secondary containment via packing, 
flanges, seals, etc., is controlled. Leakage from these systems, 
plus the Scram Discharge Volume, Reactor Water Clean-up, and PASS, 
has been found to be very low, and well below the 5 gpm limit 
established for these systems. Current leakage for Unit 1 is 0.14 
gpm and for Unit 2, 0.043 gpm. The proposed change is not expected 
to contribute to higher levels of system leakage. Any leakage from 
these systems is processed via Standby Gas Treatment and the 
radwaste system to maintain ALARA and comply with regulatory 
guidance. The closed systems are maintained filled, so that a supply 
of water exists on both sides of the isolation valves.
    While suppression pool leakage is a function of closed system 
leakage for the subject penetrations, a review of Type `C' test data 
for the subject CIVs showed that the valves have had low leakage 
rates during previous tests. This leakage is on the order of 0.6 
gpm, per unit. Proposed testing of the valves under Section XI and 
the current requirements of the Generic Letter 8910 program will 
ensure valve operability.
    Therefore, leakage past the CIV and out of the closed system is 
expected to be low and in keeping with the design basis for the 
suppression pool. However, the capability does exist to make-up 
water to the suppression pool from the Condensate Storage Tank or 
Spray Pond if necessary. Existing Emergency Operating Procedures 
require actions if suppression pool level is less than 22 feet or 
greater than 24 feet. Thus, the level of the suppression pool is 
ensured, independent of the current CIV Type `C' testing 
requirement.
    The proposed change to the scope of Type `C' testing for the 
subject valves maintains the existing barriers to primary 
containment leakage, and ensures that the suppression pool level is 
assured for 30 days during all design basis, post-accident modes of 
operation. Therefore, the plant response to the design basis events 
is unchanged, and the proposal does not create the possibility of a 
new or different kind of accident from any accident previously 
evaluated.
    III. This change does not involve a significant reduction in a 
margin of safety.
    As discussed in questions I and II, the proposed change does not 
alter the plant response to existing accident scenarios, and does 
not introduce new or different scenarios. So the margin of safety 
from a design basis accident standpoint is maintained.
    Historically, the leakage rate through the subject valves has 
been determined under the Type `C' testing program. This leakage 
rate has been found to be very low, and is currently on the order of 
0.6 gpm. Quantifying leakage past the CIVs has been used to ensure 
that the suppression pool level is assured for 30 days post-
accident. Under the proposed change, this leakage rate will not be 
quantified. This is acceptable since leakage from the suppression 
pool is in reality a function of closed system leakage, not solely 
CIV leakage. Closed system leakage is monitored and controlled by an 
existing Technical Specification program. Closed system leakage has 
been found to be very low on both units, and is currently a small 
fraction of a gallon per minute compared with a 5 gpm allowable. 
Therefore, leakage past the CIV and out of the closed system is 
expected to be low and in keeping with the design basis for the 
suppression pool. However, the capability does exist, and is 
proceduralized, to make-up water to the suppression pool from the 
Condensate Storage Tank or Spray Pond if necessary. Thus the current 
capability to maintain adequate suppression pool level for 30 days 
postaccident is assured under the proposed change.
    Therefore the proposed change to the scope of Type `C' testing 
for the subject valves does not involve a significant reduction in a 
margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: Osterhout Free Library, 
Reference Department, 71 South Franklin Street, Wilkes-Barre, 
Pennsylvania 18701.
    Attorney for licensee: Jay Silberg, Esquire, Shaw, Pittman, Potts 
and Trowbridge, 2300 N Street NW., Washington, DC 20037.
    NRC Project Director: John F. Stolz.

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

    Date of amendment request: March 31, 1995.
    Description of amendment request: These amendments would modify the 
surveillance requirement for reactor coolant system pH analysis in 
section 4.4.4 of the Technical Specifications (TS) for each unit. Also, 
they would clarify in the TS that the pH analysis would be taken at 
least every 72 hours whenever reactor coolant conductivity exceeds 1.0 
mho/cm.
    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:

    I. Involve a significant increase in the probability or 
consequences of an accident previously evaluated.
    The pH limits on reactor coolant are not affected by this 
change. The pH will be measured whenever it is theoretically 
possible for it to be outside the Tech Spec [Technical 
Specifications] limits of <5.6 or >8.6 (i.e., whenever the 
conductivity is greater than 1.0 mho/cm). Because of the 
theoretical relationship between pH and conductivity as shown in 
Attachment A [see application dated March 31, 1995, for this 
reference], it is possible to establish pH limits on the reactor 
coolant by limiting the conductivity. As shown in this figure, the 
pH must be >5.6 and <8.6 if the conductivity is less than or equal 
to 1.0 mho/cm. Attachment A was taken from Regulatory Guide 
1.56 Revision 1, July 1978 ``Maintenance of Water Purity in Boiling 
Water Reactors''. As noted in both FSAR final safety analysis report 
and Technical Specification Bases, the pH and conductivity limits 
for OPERATIONAL CONDITION 1 are consistent with this theoretical 
relationship. The Bases for Section 3/4.4.4 of the Tech Specs 
[Technical Specifications] contains [contain] the following 
statement: ``When the conductivity is within limits, the pH, 
[[Page 20523]] chlorides and other impurities affecting conductivity 
must also be within their acceptable limits[''].
    Since the conductivity is measured by grab sampling at least 
every 72 hours to verify that it is within limits, this will also 
verify that pH is within limits every 72 hours. If the conductivity 
should exceed 1.0 mho/cm, pH measurements will be made to 
determine if the Tech Spec [Technical Specifications] pH limits have 
been exceeded. It should also be noted that inline conductivity 
instrumentation is very stable and reliable and is used to 
continuously monitor the reactor coolant per Tech Spec [Technical 
Specifications] requirements, with instrumentation connected to 
redundant sources (reactor water cleanup influent and reactor 
recirculation loop). Therefore, the proposed change will not involve 
a significant increase in the probability or consequences of an 
accident previously evaluated.
    II. Create the possibility of a new or different kind of 
accident from any accident previously evaluated.
    As stated above, the pH limits on reactor coolant are not 
affected by this change. Since the conductivity is monitored 
continuously, to verify that it is within limits, this will also 
verify that pH is within limits. If the conductivity should exceed 
1.0 mho/cm, pH measurements will be made to determine if 
the Tech Spec [Technical Specifications] pH limits have been 
exceeded. Therefore, the incorporation of this change will not 
create the possibility of a new or different kind of accident from 
any accident previously evaluated.
    III. Involve a significant reduction in a margin of safety.
    The in-line conductivity instrumentation has been determined to 
be very stable and reliable in its use to continuously monitor the 
reactor coolant per Tech Spec [Technical Specifications] 
requirements. To maintain this reliability, this instrumentation is 
connected to redundant sources (reactor water cleanup influent and 
reactor recirculation loop). Based on this continuous monitoring of 
reactor coolant conductivity, as provided by this instrumentation, 
the incorporation of this change will have no impact on current 
safety margins, nor will it involve a significant reduction in the 
margin [of] safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: Osterhout Free Library, 
Reference Department, 71 South Franklin Street, Wilkes-Barre, 
Pennsylvania 18701.
    Attorney for licensee: Jay Silberg, Esquire, Shaw, Pittman, Potts 
and Trowbridge, 2300 N Street NW., Washington, DC 20037.
    NRC Project Director: John F. Stolz.

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

    Date of amendment request: March 31, 1995.
    Description of amendment request: These amendments would delete 
from the Technical Specifications of each unit, the operational 
condition restriction in Surveillance Requirement 4.8.1.1.2.d.7 which 
requires that 24-hour emergency diesel generator testing be performed 
with at least one unit in operational condition 4 or 5 (cold shutdown 
or refueling).
    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 do not:

    I. Involve a significant increase in the probability or 
consequences of an accident previously evaluated.
    The proposed change to permit the 24 hour testing of the 
emergency diesel generators during power operation does not increase 
the chances for a previously analyzed accident to occur. The 
function of the EDGs [emergency diesel generators] is to supply 
emergency power in the event of a loss of offsite power. As stated 
above [,] the diesel generator being tested has been determined to 
remain operable and available to supply the emergency loads within 
the required times. In addition, the three remaining EDGs will be 
operable during this test. Operations [Operation] of an EDGs [EDG] 
is not a precursor to any accident. If, however, an offsite 
disturbance were to occur that affected the operability of the DG 
[emergency diesel generator] being tested, the remaining EDGs are 
capable of feeding the loads necessary for safe shutdown of the 
plant. In summary, the proposed change does not adversely affect the 
performance or the ability of the diesel generators to perform their 
intended safety function. Therefore, the proposed change will not 
involve a significant increase in the probability or consequences of 
an accident previously evaluated.
    II. Create the possibility of a new or different kind of 
accident from any accident previously evaluated.
    The proposed change to the 24 hour surveillance requirement will 
not affect the operation of any safety system or alter its response 
to any previously evaluated accident. The diesel generator will 
automatically transfer from test mode if necessary to supply 
emergency loads in the required time. The test mode is used for the 
monthly surveillances of these diesel generators, resulting in no 
new plant operating modes being introduced. In the event the EDG 
fails the functional test[,] it will be declared inoperable and the 
actions required for an inoperable diesel will be performed. The 
remaining three EDGs will be operable and are capable of feeding the 
loads necessary for safe shutdown of the plant. Therefore, the 
incorporation of this change will not create the possibility of a 
new or different kind of accident from any accident previously 
evaluated.
    III. Involve a significant reduction in a margin of safety.
    Changing the EDG test timing results is no reduction in the 
safety margin as defined in the design basis. Because loss of an EDG 
is not expected as a result of LOOP [loss of offsite power] or LOCA/
LOOP [loss-of-coolant accident with a loss of offsite power] during 
the 24 hour test, SSES [Susquehanna Steam Electric Station] remains 
within its design basis. In fact, because the test EDG loads the ESS 
[engineered safety system] bus 8.5 seconds earlier than the non-test 
EDGs during LOCA with LOOP, plant response is actually improved. 
Risk of operation during the 24 hour EDG test is certainly less than 
during the current 84 hour allowed outage time (AOT) because both 
the impact of the initiating events evaluated (EDG in test is not 
actually failed) and the frequency of the limiting plant condition 
(loss of two EDGs) are less. No increase in frequency or impact of 
design basis events, and no reduction in the safety margin occurs 
during the 24 hour EDG test. Therefore, the incorporation of this 
change will have no impact on current safety margins, nor will it 
involve a significant reduction in the margin to [of] safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: Osterhout Free Library, 
Reference Department, 71 South Franklin Street, Wilkes-Barre, 
Pennsylvania 18701.
    Attorney for licensee: Jay Silberg, Esquire, Shaw, Pittman, Potts 
and Trowbridge, 2300 N Street NW., Washington, DC 20037.
    NRC Project Director: John F. Stolz.

Pennsylvania Power and Light Company, Docket No. 50-388, Susquehanna 
Steam Electric Station, Unit 2, Luzerne County, Pennsylvania

    Date of amendment request: March 31, 1995.
    Description of amendment request: This amendment would change 
Susquehanna Unit 2 Technical Specifications (TS) by incorporating the 
General Electric (GE) NRC approved methodology for GE-12 type lead use 
fuel assemblies (NEDE-24011-P-A-10) into the list of references in 
Section 6.9.3.2. The licensee plans to insert four of these fuel 
assemblies into the Unit 2 core during the fall of 1995. The addition 
of the reference to the TS would allow the use of the GE methodology to 
document that all [[Page 20524]] applicable requirements of the safety 
analysis are met by the assemblies.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:
    The proposed change does not:

    I. Involve a significant increase in the probability or 
consequences of an accident previously evaluated.
    Incorporation of this proposed change of adding reference NEDE-
24011-P-A-10, ``General Electric Standard Application for Reactor 
Fuel'' to the list of references in [the] Unit 2 Technical 
Specifications will allow the use of the GE methodology to calculate 
the operating limits for the four GE Lead Use Assemblies which are 
of a different mechanical design than the Siemens 9X9 fuel 
[currently installed in the reactor core]. This NRC approved 
methodology will be referenced as the approved methodology in 
showing that all applicable safety limits of the safety analysis are 
met by the four GE-12 LUAs. Results of incorporating this change 
will not significantly increase the probability or the consequences 
of an accident previously evaluated.
    II. Create the possibility of a new or different kind of 
accident from any accident previously evaluated.
    As stated above, the incorporation of this change will allow the 
use of the GE methodology to be referenced as the approved 
methodology to show that all applicable limits of the safety 
analysis are met by the four GE-12 LUAs. Therefore, the 
incorporation of this change will not create the possibility of a 
new or different kind of accident from any accident previously 
evaluated.
    III. Involve a significant reduction in a margin of safety.
    The use of the GE methodology will not result in a change in 
safety margin, but will ensure that the safety margin is maintained 
with the insertion of the four GE LUAs of the GE-12 type in Unit 2 
Cycle 8. Therefore, the incorporation of these changes will have no 
impact on current safety margins, nor will they involve a 
significant reduction in the margin to [of] safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: Osterhout Free Library, 
Reference Department, 71 South Franklin Street, Wilkes-Barre, 
Pennsylvania 18701.
    Attorney for licensee: Jay Silberg, Esquire, Shaw, Pittman, Potts 
and Trowbridge, 2300 N Street NW., Washington, DC 20037.
    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: December 2, 1994.
    Description of amendment request: The proposed change to Limerick 
Generating Station, Units 1 and 2, Technical Specifications (TS) 
relocates the TS Fire Protection Requirements to Licensee controlled 
documents consistent with NRC Generic Letter (GL) 86-10 
``Implementation of Fire Protection Requirements,'' and GL 88-12, 
``Removal of Fire Protection Requirements 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 Technical Specifications 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 are 
consistent with NRC GL 86-10 and GL 88-12. Removal of Fire 
Protection Program (FPP) requirements does not affect any fire 
protection equipment nor plant equipment important to safety, or 
involve any physical modifications to plant structures, systems or 
components, and therefore is not associated with an accident 
initiator or accident mitigator and can not affect the probability 
of occurrence of an accident or increase the consequences of an 
accident. The licensee controlled Technical Requirements Manual 
(TRM) containing the relocated requirements will be maintained in 
accordance with TS Section 6.0. ``Administrative Controls'' and 
subject to review in accordance with 10 CFR 50.59. Since future 
changes to the FPP (i.e., Updated Final Safety Analysis Report and 
the TRM) will 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. Therefore, these 
changes do not involve a significant increase in the probability or 
consequences of an accident previously evaluated.
    2. The proposed TS changes do not create the possibility of a 
new or different kind of accident from any accident previously 
evaluated.
    The changes will not alter the plant configuration (no new or 
different type of equipment will be installed) or create changes in 
methods governing normal plant operation that will introduce new 
failure modes. These changes will not impose different requirements 
and proper control of information will be maintained. These changes 
will not alter assumptions made in the safety analysis and licensing 
basis. Therefore, these changes do not create the possibility of a 
new or different kind of accident previously evaluated.
    3. The proposed TS changes do not involve a significant 
reduction in a margin of safety.
    The proposed changes are administrative in nature and are 
consistent with NRC GL 86-10 and GL 88-12. The changes will not 
reduce the margin of safety since they have no impact on any safety 
analysis assumptions or sequence of events used in any accident 
analysis. In addition, any future changes to the FPP will be 
evaluated per the requirements of 10 CFR 50.59. Therefore, the 
changes do not involve a significant reduction in a margin of 
safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment 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: January 27, 1995.
    Description of amendment request: The proposed change to Limerick 
Generating Station (LGS) Units 1 and 2 Technical Specifications (TS) 
will eliminate the TS active safety function designation of eight 
(i.e., four per unit) Drywell Chilled Water System (DCWS) valves.
    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 changes do not involve 
a significant increase in the probability or consequences of an 
accident previously evaluated.
    The proposed changes will eliminate the TS active safety 
function designation of eight (i.e., four per unit) DCWS valves. The 
DCWS motor operated valves (MOVs) are designated in TS as Primary 
Containment Isolation Valves (PCIVs), during operational conditions 
(OPCONS) 1, 2, and 3, which mitigate the consequences of design 
basis accidents. The proposed changes will prohibit the subject DCWS 
valves from opening during OPCONs 1, 2, and 3, thereby, 
[[Page 20525]] eliminating the active safety function, and 
maintaining a passive safety function. The postulated accidents 
which require the Primary Containment to act as a barrier in order 
to mitigate the release of radioactivity described in the LGS 
Updated Final Safety Analysis Review [Report] (UFSAR) Section 15, 
are not affected by these changes. Therefore, the previously 
evaluated postulated on-site and off-site radiological effects of 
these accidents will not change.
    The DCWS valves will be prohibited from opening during OPCONs 1, 
2, and 3 by physical changes made to the electrical control 
circuitry and administrative controls. Therefore, the probability of 
the valves to fail in the open position will diminish, and the 
required Primary Containment isolation safety function will be 
maintained.
    Therefore, these proposed changes will not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    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 remove the affected automatic isolation 
relays from the DCWS MOVs' circuitry. These changes eliminate any 
postulated relay failure effects on the associated control circuits 
and electrical power supplies. The proposed changes do not introduce 
any new accident initiators or any new valve failure modes not 
previously evaluated.
    Therefore, these changes will not create the possibility of a 
new or different kind of accident from any accidents previously 
evaluated.
    3. The proposed TS changes do not involve a significant 
reduction in a margin of safety.
    The proposed changes will prohibit the opening of the DCWS 
valves which provide backup cooling from RECW [reactor enclosure 
cooling water] during OPCONs 1, 2, and 3. The RECW System is not the 
normal DCWS cooling alignment, is not required as a backup safety 
related drywell cooling system, and this backup alignment is not an 
automatic function. The proposed changes do not affect the function 
or operation of DCWS, and since the proposed changes and 
administrative controls ensure the valves will remain closed during 
OPCONs 1, 2, and 3, the capability for Primary Containment isolation 
is not affected. Therefore, the changes will not involve a 
significant reduction in a margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: 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: February 22, 1995.
    Description of amendment request: The proposed change to Limerick 
Generating Station (LGS) Units 1 and 2 Technical Specifications (TS) 
revises various TS Surveillance Requirements to clarify the Emergency 
Diesel Generator acceptable steady state voltage range.
    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 changes do not involve 
a significant increase in the probability or consequences of an 
accident previously evaluated.
    The proposed Emergency Diesel Generator steady state voltage 
range of 4280 [plus or minus] 120 volts provides voltages through 
the 4160V and 480V distribution systems which are within the 
operating range of the connected equipment and power system 
components. Therefore, the reduced steady state voltage range will 
not cause the malfunction of any equipment or affect the operation 
of any equipment in a manner which would increase the probability of 
occurrence of an accident previously evaluated in the [Safety 
Analysis Report] SAR.
    Reducing the Emergency Diesel Generator steady state voltage 
range in the Technical Specifications maintains the capability of 
the Emergency Diesel Generator to start and attain rated voltage and 
frequency within 10 seconds and to accept the engineered safeguard 
loads in the required time in order to mitigate the consequences of 
an accident. The Emergency Diesel Generator automatic voltage 
regulator setting is calibrated to within a range of 4266.5 volts to 
4308.5 volts. A review of results from recent monthly Emergency 
Diesel Generator Surveillance Tests has confirmed that the voltage 
regulators currently maintain the Emergency Diesel Generator steady 
state voltage within the 4280 [plus or minus] 120 volt range to be 
included in the Technical Specifications. Establishing, via 
Technical Specification surveillance requirements and administrative 
limits within Station Surveillance Test Procedures, that the 
Emergency Diesel Generator voltage regulator is maintaining the 
steady state voltage within a narrower range (within the existing 
range) provides increased assurance that connected equipment 
required to mitigate the consequences of an accident will operate as 
required.
    Therefore, the proposed TS changes do not involve an increase in 
the probability or consequences of an accident previously evaluated.
    2. The proposed TS changes do not create the possibility of a 
new or different kind of accident from any accident previously 
evaluated.
    Reducing the Emergency Diesel Generator steady state voltage 
range in the Technical Specifications to a range of 4280 [plus or 
minus] 120 volts does not create any new accident initiators or 
affect any existing accident initiators such that a different type 
of accident than previously evaluated could result. The function and 
operation of the Emergency Diesel Generators and their connected 
loads are not changed in a manner which would create the possibility 
of an accident of a different type than any previously evaluated.
    Therefore, the proposed TS changes do not create the possibility 
of a new or different kind of accident from any previously 
evaluated.
    3. The proposed TS changes do not involve a significant 
reduction in a margin of safety.
    Reducing the Emergency Diesel Generator steady state voltage 
range in the Technical Specifications to a range of 4280 [plus or 
minus] 120 volts does not reduce the margin of safety. The reduced 
range provides increased assurance that the equipment powered by the 
Emergency Diesel Generators will start and operate as designed in 
order to perform their design basis functions.
    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 
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: March 1, 1995.
    Description of amendment request: The proposed changes will clarify 
the concentrations of calibration gas required to calibrate the 
Hydrogen and Oxygen Analyzers, and support the requirements of Limerick 
Generating Station (LGS) Transient Response Implementation Plant (TRIP) 
T-102, ``Primary Containment Control Bases.''
    Basis for proposed no significant hazards consideration 
determination: [[Page 20526]] 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 TS [Technical Specification] changes remove 
calibration of the H2/O2 Analyzers using zero volume percent 
hydrogen (H2) and 100% bottled Nitrogen (N2). A calibration gas 
containing zero volume percent H2 and 100% bottled N2 is not 
required for calibration of the analyzers to the required accuracy. 
Calibration of the H2/O2 Analyzers is done in accordance with the 
manufacturer's instructions. The proposed TS changes also revise the 
span gas concentration from 5% to 7% to support the requirements of 
TRIP T-120. The H2/O2 Analyzers provide indication of the 
concentrations of combustible gases in the primary containment and 
provide annunciation when combustible gas concentrations reach 
unacceptable levels. Failure of the analyzers is not an accident 
initiator. The analyzers do not connect to the reactor coolant 
pressure boundary; therefore, they do not increase the probability 
of a LOCA [loss-of-coolant accident]. The proposed TS changes do not 
involve any design changes to analyzers. Therefore, these TS changes 
will not involve a significant increase in the probability or 
consequences of an accident previously evaluated.
    2. The proposed TS changes do not create the possibility of a 
new or different kind of accident from any accident previously 
evaluated.
    The H2/O2 Analyzers provide indication and alarms for H2 and O2 
concentrations in containment. No physical or design changes to the 
analyzers are being made by these TS changes. During normal 
operations, the potential for an explosive atmosphere is negligible 
due to the absence of H2 sources. For Post-LOCA, conditions the 
levels of H2 and O2 in containment have already been evaluated in 
LGS UFSAR [Updated Final Safety Analysis Report] Section 6.2.5. No 
physical or design changes which could introduce a new analyzer 
failure mode are being made. The failure modes of the analyzers are 
evaluated in UFSAR Table 6.2-21. Therefore, these TS changes will 
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.
    These TS changes will clarify statements in the LGS UFSAR and TS 
concerning calibrated ranges of the analyzers. The change of the 
span gas from 5% to 7% falls within conditions previously analyzed. 
The Bases for TS 3/4.3.7.5 and 3/4.6.6 require operable H2/O2 
Analyzers to ensure the analyzers will be available for monitoring, 
assessing and controlling H2 and O2 in containment following a LOCA. 
These TS changes do not adversely affect operability of the 
analyzers or their availability for use during Post-LOCA conditions; 
therefore, the margin of safety is not reduced.

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

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

    Date of amendment request: December 15, 1994.
    Description of amendment request: In accordance with 10CFR50.90, 
PSE&G proposes to remove Technical Specification Surveillance 
Requirement 4.8.1.1.2.h.1, and utilize plant- controlled programs to 
govern diesel generator maintenance. To ensure procedural consistency 
and reduce the impact of this change on Hope Creek procedures, the 
remaining Surveillance Requirements of Technical Specification 
4.8.1.1.2.h are not renumbered.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Will not involve a significant increase in the probability or 
consequences of an accident previously evaluated.
    The proposed change is consistent with the improved Standard 
Technical Specifications (NUREG-1433) and does not result in any 
changes to the existing plant design. The Hope Creek preventative 
maintenance program will utilize diesel generator performance 
history, engineering analyses and manufacturer's recommendations as 
appropriate for determining diesel generator inspection 
requirements. Since the changes do not impact the ability of the 
diesel generators and the AC electrical power sources to perform 
their function, the changes do not result in a significant increase 
in the consequences of any accident previously evaluated. The diesel 
generators will continue to function as designed. Therefore, the 
proposed change will not impact the probability of occurrence of any 
accident previously evaluated.
    2. Will not create the possibility of a new or different kind of 
accident from any accident previously evaluated.
    This request does not result in any change to the plant design 
nor does it involve a significant change in current plant operation. 
The diesel generators will be inspected utilizing diesel generator 
operating history, engineering analyses and manufacturer's 
recommendations as appropriate, and the remaining surveillance 
requirements will not be changed. As a result, no new failure modes 
will be introduced, and the proposed changes will not create the 
possibility of a new or different kind of accident from any accident 
previously evaluated.
    3. Will not involve a significant reduction in a margin of 
safety. The proposed request does not adversely impact the 
reliability of the diesel generators. As stated above, the diesel 
generator operating history, engineering analyses and the 
manufacturer's recommendations will be utilized as appropriate to 
perform the diesel generator inspections. In addition, the diesel 
generators will continue to perform their design functions. This 
request does not involve an adverse impact on diesel generator 
operation or reliability. Since the diesel generator function is not 
affected by the proposed changes, this request does not involve a 
significant reduction in a margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: Pennsville Public Library, 190 
S. Broadway, Pennsville, New Jersey 08070.
    Attorney for licensee: M. J. Wetterhahn, Esquire, Winston and 
Strawn, 1400 L Street, NW., Washington, DC 20005-3502.
    NRC Project Director: John F. Stolz.

Tennessee Valley Authority, Docket No. 50-327, Sequoyah Nuclear Plant, 
Unit 1, Hamilton County, Tennessee

    Date of amendment request: April 6, 1995 (TS 95-09).
    Description of amendment request: The proposed change would revise 
Operating Condition 2.C.(25) to extend the ice condenser Surveillance 
4.6.5.1.d to October 1, 1995, to coincide with the Unit 1 Cycle 7 
refueling outage.
    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:

    TVA has evaluated the proposed technical specification (TS) 
change and has determined that it does not represent a significant 
hazards consideration based on criteria established in 10 CFR 
50.92(c). Operation of Sequoyah Nuclear Plant (SQN) in accordance 
with the proposed amendment will not: [[Page 20527]] 
    1. Involve a significant increase in the probability or 
consequences of an accident previously evaluated.
    The proposed change is temporary and allows a one-time extension 
of the ice condenser Surveillance Requirement 4.6.5.1.d for Cycle 7 
to allow surveillance testing to coincide with the seventh refueling 
outage. The proposed surveillance interval extension will not cause 
a significant reduction in system reliability nor affect the ability 
of the system to perform the design function. Current monitoring of 
plant conditions and continuation of the surveillance testing 
required during normal plant operation will continue to be performed 
to ensure conformance with TS operability requirements. Therefore, 
this change does not involve a significant increase in the 
probability or consequences of an accident previously evaluated.
    2. Create the possibility of a new or different kind of accident 
from any previously analyzed.
    Extending the surveillance interval for the performance of ice 
condenser testing will not create the possibility of a new or 
different kind of accident. No change is required to any system 
configuration, plant equipment, or analyses.
    3. Involve a significant reduction in a margin of safety.
    The safety limits assumed in the accident analyses and the 
design function of the equipment required to mitigate the 
consequences of postulated accidents will not be changed 
significantly. Existing analysis indicates that the potential 
reduction in ice weight resulting from the proposed extension will 
continue to maintain the maximum containment accident pressure below 
12 pounds per square inch gauge. The ice condenser will continue to 
support accident mitigation functions although some Row 1 baskets 
could drop slightly below the required 993-pound analysis limit. 
Therefore, the plant will be maintained with acceptable ice weights 
for accident mitigation and the proposed extension will not 
significantly reduce the margin of safety.

    The NRC has reviewed the licensee's analysis and, based on this 
review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: Chattanooga-Hamilton County 
Library, 1101 Broad Street, Chattanooga, Tennessee 37402.
    Attorney for licensee: General Counsel, Tennessee Valley Authority, 
400 West Summit Hill Drive, ET 11H, Knoxville, Tennessee 37902.
    NRC Project Director: Frederick J. Hebdon.

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

    Date of amendment request: April 6, 1995 (TS 95-11).
    Description of amendment request: The proposed change would 
relocate the constant numerical value found in the overtemperature 
delta temperature and overpower delta temperature equations of 
Technical Specifications Table 2.2-1 and place them in the Core 
Operating Limits Report (COLR). This would be accomplished by revising 
notes 1 and 2 in Table 2.2-1 to state that the values are located in 
the COLR. The values of the constants, however, would not be changed. 
Also, the ``Overtemperature and Overpower Delta Temperature Setpoint 
Parameter Values for Specification 2.2.1'' would be added to the list 
of core operating limits specified in Section 6.9.1.14 that are 
required to be included in the COLR. In addition, a reference to WCAP-
8745-P-A, ``Design Bases for the Thermal Overpower delta-T and Thermal 
Overtemperature delta-T Trip Functions,'' would be added to Section 
6.9.1.14.a.
    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:

    TVA has evaluated the proposed technical specification (TS) 
change and has determined that it does not represent a significant 
hazards consideration based on criteria established in 10 CFR 
50.92(c). Operation of Sequoyah Nuclear Plant (SQN) in accordance 
with the proposed amendment will not:
    1. Involve a significant increase in the probability or 
consequences of an accident previously evaluated.
    The proposed changes will allow changes to the constant 
numerical values for the overtemperature delta temperature 
P(OT[delta-]T) and overpower delta temperature (OP[delta-]T) 
equations in accordance with the 10 CFR 50.59 requirements. This 
revision does not revise these constants, but relocates them to the 
core operating limits report (COLR) that is governed by the 10 CFR 
50.59 requirements. The addition of the lag compensator functions 
for measured [delta-]T and average temperature in these equations 
does not alter the setpoint because this lag function has a value of 
unity. Therefore, the proposed revision does not alter plant 
functions or setpoints, but does allow for a more timely revision 
process for parameters that may require changes due to specific fuel 
cycle requirements or updated plant analyses. The use of the lag 
functions and revisions to the constant numerical values will be 
maintained within the safety analysis for the plant by the 10 CFR 
50.59 process requirements. The probability of an accident is not 
increased because the plant functions are not altered by the 
proposed revision and future changes will be in accordance with 10 
CFR 50.59. Additionally, the consequences of an accident are not 
increased because the mitigation functions of the OT[delta-]T and 
OP[delta-]T functions are not changed and revisions to the equations 
that derive the setpoints will be processed under the requirements 
of the 10 CFR 50.59 program.
    2. Create the possibility of a new or different kind of accident 
from any previously analyzed.
    The proposed revision will not change plant functions and future 
revisions will continue to be controlled in accordance with the 10 
CFR 50.59 requirements. The addition of the lag functions does not 
create a new accident potential because these functions have already 
been considered in the analysis as shown in NUREG 1431. Therefore, 
the possibility of a new or different kind of accident is not 
created by the proposed revision.
    3. Involve a significant reduction in a margin of safety.
    Plant parameters are not altered by the proposed revision and 
the OT[delta-]T and OP[delta-]T functions will not reflect a change 
in setpoint generation or value. The proposed change will allow 
revision of the constant numerical values and use of the lag 
compensator functions in accordance with the 10 CFR 50.59 provisions 
to ensure the design basis of the plant is maintained. This revision 
does not result in changes that reduce the margin of safety because 
the OT[delta-]T and OP[delta-]T functions remain unchanged and 
future revisions to these functions will be performed in accordance 
with 10 CFR 50.59.

    The NRC has reviewed the licensee's analysis and, based on this 
review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: Chattanooga-Hamilton County 
Library, 1101 Broad Street, Chattanooga, Tennessee 37402.
    Attorney for licensee: General Counsel, Tennessee Valley Authority, 
400 West Summit Hill Drive, ET 11H, Knoxville, Tennessee 37902.
    NRC Project Director: Frederick J. Hebdon.

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

    Date of amendment request: April 6, 1995 (TS 94-15).
    Description of amendment request: The proposed change would modify 
the Technical Specifications associated with the Post Accident Sampling 
(PAS) system by deleting License Condition 2.C.(23)F for Unit 1 and 
2.C.(16)g for Unit 2 that require operation of the PAS system in 
accordance with referenced letters no later than startup from the 
second refueling outage. The submittal also includes a revised 
description of operation of the PAS system for insertion into the 
Updated Final Safety [[Page 20528]] Analysis Report for staff approval. 
This information supersedes the information contained in the letters 
referenced in the License Conditions listed above and would be 
maintained in accordance with the 10 CFR 50.59 process.
    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:

    TVA has evaluated the proposed technical specification (TS) 
change and has determined that it does not represent a significant 
hazards consideration based on criteria established in 10 CFR 
50.92(c). Operation of Sequoyah Nuclear Plant (SQN) in accordance 
with the proposed amendment will not:
    1. Involve a significant increase in the probability or 
consequences of an accident previously evaluated.
    The proposed change involves the deletion of license conditions 
that authorized TVA to operate SQN's postaccident sampling (PAS) 
system. TVA proposed change establishes programmatic control of 
SQN's PAS system under SQN TS 6.8.4.e and the SQN Final Safety 
Analysis Report. Any future changes to SQN's PAS Program would be 
governed by the 10 CFR 50.59 process. PAS and analysis will continue 
at SQN through grab sample acquisition and laboratory analysis and 
will continue to meet the PAS objectives in NUREG-0737, Item II.B.3 
and Regulatory Guide 1.97, Revisions 2. Accordingly, the proposed 
change does not affect the probability or consequences of an 
accident.
    2. Create the possibility of a new or different kind of accident 
from any previously analyzed.
    The proposed change involves improvements in the operational 
reliability of SQN's PAS system by using more reliable laboratory 
analysis methods, reducing sampling personnel radiation dose, and 
incorporating practical methods for sample acquisition and analysis. 
Because the proposed change involves license conditions and sampling 
methods that are utilized for postaccident recovery, the potential 
for an unanalyzed accident is not created. Consequently, no new 
failure modes are introduced.
    3. Involve a significant reduction in a margin of safety.
    Plant safety margins are established through limiting conditions 
of operation, limiting safety system settings, and safety limits 
specified in the TSs. As a result of the proposed amendment, there 
will be no change to either the physical design of the plant or to 
any of these settings and limits. The proposed changes do not affect 
the safe operation of SQN. Therefore, there are no changes to any of 
the margins of safety.

    The NRC has reviewed the licensee's analysis and, based on this 
review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: Chattanooga-Hamilton County 
Library, 1101 Broad Street, Chattanooga, Tennessee 37402.
    Attorney for licensee: General Counsel, Tennessee Valley Authority, 
400 West Summit Hill Drive, ET 11H, Knoxville, Tennessee 37902.
    NRC Project Director: Frederick J. Hebdon.

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

    Date of amendment request: April 6, 1995 (TS 94-18).
    Description of amendment request: The proposed change would revise 
Surveillance Requirement (SR) 4.0.5 by replacing the current Inservice 
Inspection program requirements with the requirements stated in the 
Standard Technical Specifications (NUREG-1431). As a result, SR 4.0.5 
would more clearly specify the inservice inspection (ISI) program 
requirements and the inservice testing (IST) program requirements of 
the American Society of Mechanical Engineers (ASME) Code Class 1, 2, 
and 3 components. The licensee also proposed deletion of Technical 
Specification 3/4.4.10, ``Structural Integrity ASME Code Class 1, 2 and 
3 Components,'' and its related Bases information.
    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:

    TVA has evaluated the proposed technical specification (TS) 
change and has determined that it does not represent a significant 
hazards consideration based on criteria established in 10 CFR 
50.92(c). Operation of Sequoyah Nuclear Plant (SQN) in accordance 
with the proposed amendment will not:
    1. Involve a significant increase in the probability or 
consequences of an accident previously evaluated.
    Operation of the facility in accordance with the proposed 
amendment would not involve any increase in the probability of 
occurrence or consequences of an accident previously evaluated. The 
Inservice Inspection and Testing Programs, pursuant to 10 CFR 50.55a 
are described in the TSs. The proposed amendment, in accordance with 
NUREG-1431 and draft NUREG-1482 permits relief from an American 
Society of Mechanical Engineers (ASME) code requirement in the 
interim between the time of submittal of a relief request and NRC 
approval of the relief. The changes being proposed do not affect 
assumptions contained in plant safety analyses or change the 
physical design and/or operation of the plant, nor do they affect 
TSs that preserve safety analysis assumptions. Any relief from the 
approved ASME Section XI code requirements that is implemented prior 
to NRC review and approval will require evaluation under the 10 CFR 
50.59 process to determine that no TS changes or unreviewed safety 
questions exist. This evaluation process will ensure that the impact 
of any code relief is thoroughly evaluated and that the structures, 
systems, and components remain in conformance with assumptions made 
in the safety analysis. The proposed change to delete SQN TS 3/
4.4.10, Structural Integrity, does not affect plant safety analyses 
or change the physical design or operation of the plant. The 
proposed amendment relocates the structural integrity requirements 
under the existing TS Surveillance Requirement (SR) 4.0.5 to allow 
these requirements to be governed and controlled within the 
inservice inspection (ISI) program. Therefore, operation of the 
facility in accordance with the proposed amendment would not affect 
the probability or consequences of an accident previously evaluated.
    2. Create the possibility of a new or different kind of accident 
from any previously analyzed.
    Operation of the facility in accordance with the proposed 
amendment would not create the possibility of a new or different 
kind of accident from any accident previously evaluated. The 
Inservice Inspection and Testing Programs, pursuant to 10 CFR 50.55a 
are described in the TSs. The proposed amendment, in accordance with 
NUREG-1433 and draft NUREG-1482, permits relief from an ASME code 
requirement in the interim between the time of submittal of a relief 
request and NRC approval of the relief. The changes being proposed 
will not change the physical plant or the modes of operation defined 
in the Facility License. The changes do not involve the addition or 
modification of equipment nor do they alter the design or operation 
of plant systems. Any relief from the approved ASME Section XI code 
requirements that is implemented prior to NRC review and approval 
will require evaluation under the 10 CFR 50.59 process to determine 
that no TS changes or unreviewed safety questions exist. This 
evaluation process will ensure that the impact of the code relief is 
thoroughly evaluated and that the structures, systems, and 
components remain in conformance with assumptions made in the safety 
analysis. The proposed change to delete SQN TS 3/4.4.10 does not 
affect plant safety analyses or change the physical design or 
operation of the plant. The proposed amendment relocates the 
structural integrity requirements under the existing TS SR 4.0.5 to 
allow these requirements to be governed and controlled within the 
ISI program. 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. Involve a significant reduction in a margin of safety.
    Operation of the facility in accordance with the proposed 
amendment would not involve any reduction in a margin of safety. 
[[Page 20529]] The Inservice Inspection and Testing Programs, 
pursuant to 10 CFR 50.55a, are required by the SQN TSs. The proposed 
amendment, in accordance with NUREG-1431 and draft NUREG-1482 
permits relief from an ASME code requirement in the interim between 
the time of submittal of a relief request and NRC approval of the 
relief. Any relief from the ASME Section XI code is required to be 
evaluated under the 10 CFR 50.59 process to determine that no TS 
changes or unreviewed safety questions exist. This evaluation 
process will ensure that code relief does not affect the ability of 
structures, systems, or components to perform their design function, 
affect compliance with any TS requirements or reduce the margin of 
safety. The proposed change to delete SQN TS 3/4.4.10 does not 
affect plant safety analyses or change the physical design or 
operation of the plant. The proposed amendment relocates the 
structural integrity requirements under the existing TS SR 4.0.5 to 
allow these requirements to be governed and controlled within SQN's 
ISI program. Therefore, operation of the facility in accordance with 
the proposed amendment would not involve a reduction in the margin 
of safety.

    The NRC has reviewed the licensee's analysis and, based on this 
review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: Chattanooga-Hamilton County 
Library, 1101 Broad Street, Chattanooga, Tennessee 37402.
    Attorney for licensee: General Counsel, Tennessee Valley Authority, 
400 West Summit Hill Drive, ET 11H, Knoxville, Tennessee 37902.
    NRC Project Director: Frederick J. Hebdon.

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

    Date of amendment request: April 6, 1995 (TS 95-06).
    Description of amendment request: The proposed change would delete 
the technical specification requirement that limits and controls loads 
traveling over the spent fuel pool (Specification 3.9.7), the graph 
that relates the Load Carried Over the Shield to the Allowable Height 
Above the Shield Surface (Figure 3.9-1), the crane interlocks and 
physical stops surveillance requirements (Specifications 4.9.7.1 and 
4.9.7.2), and the related Bases information. These requirements would 
be relocated to administratively controlled procedures.
    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:

    TVA has evaluated the proposed technical specification (TS) 
change and has determined that it does not represent a significant 
hazards consideration based on criteria established in 10 CFR 
50.92(c). Operation of Sequoyah Nuclear Plant (SQN) in accordance 
with the proposed amendment will not:
    1. Involve a significant increase in the probability or 
consequences of an accident previously evaluated.
    The proposed TS change involves the relocation of a requirement 
that does not pertain to limitations or conditions of reactor 
operation or to equipment to mitigate design basis accidents or 
transients. SQN is proposing to relocate this TS based on NRC's 
final policy statement on TS improvement (58 FR 39132, dated July 
22, 1993). Based on this criteria, the spent fuel pit (SFP) crane 
travel is not important to operational safety and may be relocated 
to administratively controlled procedures. By placing the crane 
travel requirements in administratively controlled procedures, 
adequate controls will remain in place to prevent heavy loads from 
traveling over fuel assemblies in the SFP. The administratively 
controlled procedure that controls the by-passing of the interlocks 
and physical stops is subject to the requirements of TS 6.5.1A. 
Therefore, the relocation of this TS will not involve an increase in 
the probability or consequences of an accident previously evaluated.
    2. Create the possibility of a new or different kind of accident 
from any previously analyzed.
    The proposed change involves relocating TS requirements to 
another administratively controlled document. No modifications to 
the plant are involved. Additionally, there are no changes to the 
operation of the plant or equipment proposed. Based on this, the 
relocation of this TS will not create the possibility of a new or 
different kind of accident from any previously analyzed.
    3. Involve a significant reduction in a margin of safety.
    The proposed change involves the relocation of TS requirements 
to administratively controlled procedures. The relocation of this 
requirement is based on the criteria endorsed in the Commission's 
Policy Statement on TS improvements as it pertains to 10 CFR 50.36. 
Additionally, this change does not alter the basic design and safety 
analysis requirements, as discussed in the Updated Final Safety 
Analysis Report. Therefore, the deletion of this TS will not involve 
a reduction in the margin of safety.

    The NRC has reviewed the licensee's analysis and, based on this 
review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: Chattanooga-Hamilton County 
Library, 1101 Broad Street, Chattanooga, Tennessee 37402.
    Attorney for licensee: General Counsel, Tennessee Valley Authority, 
400 West Summit Hill Drive, ET 11H, Knoxville, Tennessee 37902.
    NRC Project Director: Frederick J. Hebdon.

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

    Date of amendment request: April 6, 1995 (TS 94-19).
    Description of amendment request: The proposed change would revise 
the Action statement for Technical Specification 3.8.1.1 by inserting a 
new Action a, relabeling and modifying existing Action a to become 
Action b, adding a footnote referenced to Action b, renumbering the 
subsequent action statements, and adding information to the Bases that 
amplifies the action statements. The proposed new Action a would no 
longer address required actions for diesel generator testing. It would 
require that, should one of the AC electrical power sources listed be 
inoperable, then operability of the remaining offsite AC circuit be 
demonstrated by performing Surveillance Requirement 4.8.1.1.1.a within 
1 hour and at least once per 8 hours thereafter. If two offsite 
circuits cannot be restored within 72 hours, the unit must be placed in 
hot standby within the next 6 hours and in cold shutdown within the 
following 30 hours.
    The proposed change to Action b would address the testing 
requirements should a diesel generator become inoperable. It would 
require testing of operable diesel generators if the inoperability of 
the affected diesel generator has the potential to be the result of a 
common cause failure. A footnote would clarify that the common cause 
determination must be completed regardless of how long the diesel 
generator inoperability persists or Surveillance 4.8.1.1.2.a.4 must be 
completed to verify diesel generator operability. The proposed change 
to the Bases would provide clear guidance on the use of common cause 
failure determinations to eliminate unnecessary diesel generator 
testing and would define when testing is required to verify diesel 
generator operability.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    TVA has evaluated the proposed technical specification (TS) 
change and has determined [[Page 20530]] that it does not represent 
a significant hazards consideration based on criteria established in 
10 CFR 50.92(c). Operation of Sequoyah Nuclear Plant (SQN) in 
accordance with the proposed amendment will not:
    1. Involve a significant increase in the probability or 
consequences of an accident previously evaluated.
    The proposed revisions do not alter the plant features or 
operating practices. This revision will reduce unnecessary starts of 
the diesel generators (D/Gs) when a common cause failure is not 
involved or for an inoperable offsite circuit. This change will not 
affect the accident mitigation capabilities of the D/Gs, but should 
improve the reliability by reducing the wear and tear associated 
with starting the D/Gs. The D/Gs are not the source of a postulated 
accident and because this change does not alter plant functions or 
operating practices the probability of an accident is not increased. 
The D/G's operability will continue to be verified for conditions 
that indicate a potential common-cause failure to ensure accident 
mitigation capabilities are not affected. Therefore, this revision 
will continue to provide actions that will support alternating-
current (ac) electrical power source safety functions without 
unnecessary degradation of the D/Gs and will not increase the 
consequences of an accident.
    2. Create the possibility of a new or different kind of accident 
from any previously analyzed.
    The D/Gs are not the source of accidents and the proposed 
revision will not alter plant functions or actions by more 
appropriately limiting the conditions when a D/G must be verified 
operable. Therefore, the possibility of a new or different accident 
is not created
    3. Involve a significant reduction in a margin of safety.
    This revision does not alter plant functions that provide the 
margin of safety. The reduction of D/G testing will only be allowed 
for situations where the operable D/Gs are not affected by the 
conditions resulting in the ac power source inoperability. This 
reduced testing should improve D/G reliability for accident 
mitigation functions and further ensure the margin of safety 
provided by the D/Gs. Therefore, the margin of safety is not reduced 
by the proposed revision to limit unnecessary D/G starts.

    The NRC has reviewed the licensee's analysis and, based on this 
review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: Chattanooga-Hamilton County 
Library, 1101 Broad Street, Chattanooga, Tennessee 37402.
    Attorney for licensee: General Counsel, Tennessee Valley Authority, 
400 West Summit Hill Drive, ET 11H, Knoxville, Tennessee 37902.
    NRC Project Director: Frederick J. Hebdon.

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

    Date of amendment request: April 6, 1995 (TS 95-02).
    Description of amendment request: The proposed change would add 
Limiting Condition for Operation 3.0.6 to allow equipment that has been 
removed from service or declared inoperable to be returned to service 
under administrative control in order to perform testing required to 
demonstrate operability. It would be applicable for operability testing 
of the inoperable equipment or other equipment that requires the 
operability feature to be in service in order to perform the test. A 
related change to the Bases would provide amplifying explanation on the 
use of this new provision. In addition, a proposed change to Action 18 
of Table 3.3-3, ``Engineered Safety Feature Actuation System 
Instrumentation,'' would clarify the time interval that an instrument 
channel may be in the bypass condition. For those instruments that 
reference Action 18, the change would allow the bypass for 6 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:

    TVA has evaluated the proposed technical specification (TS) 
change and has determined that it does not represent a significant 
hazards consideration based on criteria established in 10 CFR 
50.92(c). Operation of Sequoyah Nuclear Plant (SQN) in accordance 
with the proposed amendment will not:
    1. Involve a significant increase in the probability or 
consequences of an accident previously evaluated.
    The addition of the exception to TSs 3.0.1 and 3.0.2 and the 
definition for the time to place a channel in bypass will not change 
plant equipment or the operating practices at SQN. The exception 
will allow testing to be performed with inoperable equipment 
returned to service under administrative controls, but will not 
change functions. The function will be available from other 
redundant channels during the brief durations that the new provision 
would be utilized. The specified time interval to achieve a bypass 
condition will clarify the implementation of the action requirement 
with the affected functions remaining available through the 
redundant operable channels. This clarification does not change the 
intent of the action but does set the previously undefined time 
interval.
    The proposed change affects actions associated with the 
actuation of functions to mitigate accidents and are not the source 
of an accident. Therefore, the probability of an accident is not 
increased. The affected functions provide accident mitigation 
functions and the proposed revisions serve to ensure equipment can 
be maintained in required conditions within acceptable time 
intervals and administrative controls. The brief periods utilized 
for the TSs 3.0.1 and 3.0.2 exception will not significantly affect 
the accident mitigation capabilities because of the availability of 
redundant equipment. In addition, the benefit of performing 
operability testing to return equipment permanently to service or to 
maintain the operability of other equipment outweighs the slight 
reduction in safety function actuation redundancy. Therefore, the 
proposed change will not significantly increase the consequences of 
an accident.
    2. Create the possibility of a new or different kind of accident 
from any previously analyzed.
    The proposed changes affect functions utilized to mitigate an 
accident and are not the source of an accident. The exception 
provides reasonable flexibility to maintain equipment operability 
and the bypass time interval reduces the potential for damage to 
safety related equipment. Because plant functions are not changed as 
a result of this request the possibility of a new or different kind 
of accident is not created.
    3. Involve a significant reduction in a margin of safety.
    The proposed change does not alter setpoints or operating 
considerations that maintain the margin of safety for SQN. These 
changes provide needed flexibility to perform TS required testing 
and clarifications for implementing action requirements. These 
changes will slightly affect the redundancy of the affected safety 
functions but provide greater benefit for maintaining equipment in 
an operable condition. Therefore, the margin of safety provided by 
the affected equipment has not changed and the proposed change will 
not result in a reduction.

    The NRC has reviewed the licensee's analysis and, based on this 
review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: Chattanooga-Hamilton County 
Library, 1101 Broad Street, Chattanooga, Tennessee 37402.
    Attorney for licensee: General Counsel, Tennessee Valley Authority, 
400 West Summit Hill Drive, ET 11H, Knoxville, Tennessee 37902.
    NRC Project Director: Frederick J. Hebdon.

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

    Date of amendment request: April 6, 1995 (TS 94-04).
    Description of amendment request: The proposed amendment would 
change the power range neutron flux channel calibration frequency 
[[Page 20531]] surveillance requirement from monthly to every 31 
effective full power days and delay the requirement to perform the 
surveillance for 96 hours after reaching 15 percent power. A proposed 
change to the Bases would provide amplifying information.
    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:

    TVA has evaluated the proposed technical specification (TS) 
change and has determined that it does not represent a significant 
hazards consideration based on criteria established in 10 CFR 
50.92(c). Operation of Sequoyah Nuclear Plant (SQN) in accordance 
with the proposed amendment 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 this TS change, which only affects when the first 
surveillance is performed following an outage and changes the 
frequency of performance of the surveillance. Before start-up 
following refueling outage, the power range high trip setpoint is 
set below 85 percent power, typically 60 percent, for conservatism. 
The power range low trip setpoint is set at 22 percent power, TS 
requires the setpoint to be less than or equal to 25 percent power. 
These settings are in addition to the conservatism built into start-
up following a refueling outage. Therefore, delaying the first 
performance for 96 hours will not impact on the operation of the 
plant since the setpoints are set conservatively. Also, the change 
of the frequency to every 31 effective full power days (EFPD) only 
delays the surveillance when the plant is operated at reduced power. 
During operation at reduced power changes in the neutron flux are 
also reduced. Therefore, changing the frequency from monthly to 
every 31 EFPD allows slow changes in neutron flux during the fuel 
cycle to be more accurately detected and evaluated.
    This TS change 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 
this change. No systems, equipment, or components are affected by 
the proposed change. Thus, the consequences of a malfunction of 
equipment important to safety previously evaluated in the Updated 
Final Safety Analysis Report are not increased by this change.
    The proposed changes provide TS improvements that ensure the 
system operates within the bounds of SQN's accident analysis as 
contained in the Final Safety Analysis Report (FSAR) and only 
affects when a surveillance is performed. This change has no impact 
on accident initiators and does not involve a physical modification 
to the plant. Accordingly, the proposed changes do not involve an 
increase in the probability or consequences of an accident 
previously evaluated.
    2. Create the possibility of a new or different kind of accident 
from any previously analyzed.
    This revision will not change any plant equipment, system 
configurations, or accident assumptions. This change will more 
accurately monitor changes in the condition of the core.
    Fuel burn-up is necessary to change the relationship between the 
incore axial power and the excore detectors response. At reduced 
levels the effectiveness of the monitoring activity is reduced. 
Therefore, changing the frequency to 31 EFPD allows slow changes in 
neutron flux during the fuel cycle to be more accurately detected 
and evaluated. Delaying the first performance of the surveillance 
requirement, until 96 hours after reaching 15 percent rated thermal 
power, will allow the unit to be in a more stable condition. 
Therefore, this change will not affect the safety function of any 
components and will create the possibility of a new or different 
kind of accident.
    3. Involve a significant reduction in a margin of safety.
    The proposed changes provide TS improvements for SQN's power 
range monitoring system that ensure the system operates within the 
bounds of SQN's accident analysis as contained in the FSAR since 
only the time interval between performances of the surveillance is 
being extended. This change does not involve a physical modification 
to SQN's power range monitoring system. Accordingly, the margin of 
safety has not been reduced.

    The NRC has reviewed the licensee's analysis and, based on this 
review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: Chattanooga-Hamilton County 
Library, 1101 Broad Street, Chattanooga, Tennessee 37402.
    Attorney for licensee: General Counsel, Tennessee Valley Authority, 
400 West Summit Hill Drive, ET 11H, Knoxville, Tennessee 37902.
    NRC Project Director: Frederick J. Hebdon.

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

    Date of amendment request: April 6, 1995 (TS 95-08).
    Description of amendment request: The proposed change would (1) 
change the core alteration definition to limit the term to reactor 
vessel internal activities that could have an affect on core 
reactivity, (2) change the quadrant power tilt ratio definition to 
eliminate the conflict in the definition of the term and its use in 
Surveillance Requirement 4.2.4.2, and (3) revise the Unit 1 Operational 
Modes parameters in Table 1.1 to be consistent with the description in 
Table 1.1 for Unit 2.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    TVA has evaluated the proposed technical specification (TS) 
change and has determined that it does not represent a significant 
hazards consideration based on criteria established in 10 CFR 
50.92(c). Operation of Sequoyah Nuclear Plant (SQN) in accordance 
with the proposed amendment will not:
    1. Involve a significant increase in the probability or 
consequences of an accident previously evaluated. The proposed 
changes provide TS improvements that ensure the plant operates 
within the bounds of SQN's accident analysis as contained in the 
Final Safety Analysis Report (FSAR) and only affects the definitions 
and does not have any affect on any work performed. The change to 
core alteration is to clarify those components that may result in 
reactivity changes. The change will not effect movement of fuel or 
components that effect reactivity, therefore, a fuel handling 
accident will not be effected. The change in the definition of 
quadrant power tilt ratio (QPTR) allows the alternate method of 
determining QPTR to be utilized. The current TS surveillance 
requirement (SR) and bases allow alternate means for determining 
QPTR, therefore, revising the definition will have no effect on any 
accident. The revision to the mode parameters is administrative in 
nature, therefore it will have no effect on any accident. This 
change has no impact on accident initiators and does not involve a 
physical modification to the plant. Accordingly, the proposed 
changes do not involve an increase in the probability or 
consequences of an accident previously evaluated.
    2. Create the possibility of a new or different kind of accident 
from any previously analyzed.
    This revision will not change any plant equipment, system 
configurations, or accident assumptions. This change will better 
define the associated parameters and will eliminate potential 
ambiguity and confusion. The change in the definition of core 
alteration allows components that do not affect reactivity to be 
moved within the reactor vessel. The change in the definition will 
not effect the monitoring of QPTR with one channel inoperable. The 
core will be monitored in accordance with the SRs. Therefore, this 
change will not affect the safety function of any components and 
will not create the possibility of a new or different kind of 
accident.
    3. Involve a significant reduction in a margin of safety.
    The proposed changes provide improvements for SQN's TS. This 
change does not involve a physical modification to the plant nor 
change the methods of monitoring plant parameters. Accordingly, the 
margin of safety has not been reduced.

    [[Page 20532]] The NRC has reviewed the licensee's analysis and, 
based on this review, it appears that the three standards of 10 CFR 
50.92(c) are satisfied. Therefore, the NRC staff proposes to determine 
that the amendment request involves no significant hazards 
consideration.
    Local Public Document Room location: Chattanooga-Hamilton County 
Library, 1101 Broad Street, Chattanooga, Tennessee 37402.
    Attorney for licensee: General Counsel, Tennessee Valley Authority, 
400 West Summit Hill Drive, ET 11H, Knoxville, Tennessee 37902.
    NRC Project Director: Frederick J. Hebdon.

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

    Date of amendment request: March 24, 1995.
    Description of amendment request: This amendment request proposes 
to revise Technical Specification (TS) 1.7, ``Containment Integrity,'' 
TS 3/4.6.1, ``Containment Integrity,'' TS 3/4.6.3, ``Containment 
Isolation Valves,'' and the associated Bases. These proposed changes 
will relocate TS Table 3.6-1, ``Containment Isolation Valves,'' to Wolf 
Creek Generating Station procedures. This proposed change is in 
accordance with the guidance provided in Generic Letter 91-08, 
``Removal of Component Lists from Technical Specifications,'' dated May 
6, 1991.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    (1) The proposed change does not involve a significant increase 
in the probability or consequences of an accident previously 
evaluated.
    The proposed changes simplify the technical specifications, meet 
the regulatory requirements for control of containment isolation, 
and are consistent with the guidelines of GL 91-08. The procedural 
details of Technical Specification Table 3.6-1 have not been 
changed, but only relocated to a different controlling document. The 
proposed changes are administrative in nature, should result in 
improved administrative practices, and do not affect plant 
operations.
    The probability of occurrence of a previously evaluated accident 
is not increased because this change does not introduce any new 
potential accident initiating conditions. The consequences of an 
accident previously evaluated is not increased because the ability 
of [the] containment to restrict the release of any fission product 
radioactivity to the environment will not be degraded by this 
change.
    (2) The proposed change does not create the possibility of a new 
or different kind of accident from any previously evaluated.
    The proposed changes are administrative in nature, do not result 
in physical alterations or changes to the operation of the plant, 
and cause no change in the method by which any safety-related system 
performs its function. Therefore, this proposed change will not 
create the possibility of a new or different kind of accident from 
any previously evaluated.
    (3) The proposed change does not involve a significant reduction 
in the margin of safety.
    The administrative change to relocate Technical Specification 
Table 3.6-1 to appropriate plant procedures does not alter the basic 
regulatory requirements for containment isolation and will not 
adversely affect containment isolation capability for credible 
accident scenarios. Adequate control of the content of the table is 
assured by existing plant procedures.
    The proposed relocation of Technical Specification Table 3.6-1 
does not alter the requirements for containment isolation valve 
operability currently in the technical specifications. The LCO 
[limiting condition for operation] and Surveillance Requirements 
would be retained in the revised technical specifications. 
Therefore, the proposed change will not affect the meaning, 
application, and function of the current technical specification 
requirements for the valves in Table 3.6-1.

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

Previously Published Notices of Consideration of Issuance of Amendments 
to Facility Operating Licenses, Proposed No Significant Hazards 
Consideration Determination, and Opportunity for a Hearing

    The following notices were previously published as separate 
individual notices. The notice content was the same as above. They were 
published as individual notices either because time did not allow the 
Commission to wait for this biweekly notice or because the action 
involved exigent circumstances. They are repeated here because the 
biweekly notice lists all amendments issued or proposed to be issued 
involving no significant hazards consideration.
    For details, see the individual notice in the Federal Register on 
the day and page cited. This notice does not extend the notice period 
of the original notice.

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

    Date of amendment request: April 4, 1995, as supplemented by letter 
dated April 5, 1995.
    Description of amendment request: The proposed amendment would 
change the technical specifications on moderator temperature 
coefficient. The proposed change constitutes a one time deviation not 
to perform the two-thirds end-of-cycle moderator temperature 
coefficient test for Cycle 7.
    Date of individual notice in the Federal Register: April 11, 1995 
(60 FR 18432).
    Expiration date of individual notice: May 11, 1995.
    Local Public Document Room location: University of New Orleans 
Library, Louisiana Collection, Lakefront, New Orleans, Louisiana 70122.

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 
[[Page 20533]] prepared an environmental assessment under the special 
circumstances provision in 10 CFR 51.12(b) and has made a determination 
based on that assessment, it is so indicated.
    For further details with respect to the action see (1) the 
applications for amendment, (2) the amendment, and (3) the Commission's 
related letter, Safety Evaluation and/or Environmental Assessment as 
indicated. All of these items are available for public inspection at 
the Commission's Public Document Room, the Gelman Building, 2120 L 
Street, NW., Washington, DC, and at the local public document rooms for 
the particular facilities involved.

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

    Date of application for amendment: February 24, 1995.
    Brief description of amendment: The amendment revises Technical 
Specification Section 4.6.1.2.a, Primary Containment/Containment 
Leakage, to reference 10 CFR Part 50, Appendix J, as modified by 
approved exemptions.
    Date of issuance: April 10, 1995.
    Effective date: As of the date of issuance to be implemented within 
30 days.
    Amendment No.: 183.
    Facility Operating License No. DPR-69: Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: March 8, 1995 (60 FR 
12789).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated April 10, 1995.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: Calvert County Library, Prince 
Frederick, Maryland 20678.

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

    Date of application for amendments: September 30, 1994, as 
supplemented on March 24, 1995.
    Brief Description of amendments: The proposed change will revise 
Technical Specification requirements to eliminate the reactor scram and 
isolation functions of the Main Steam Line Radiation Monitors. The 
March 24, 1995, supplement provided clarifying information only and did 
not affect the NRC's determination of no significant hazards 
considerations.
    Date of issuance: March 31, 1995.
    Effective date: March 31, 1995.
    Amendment No.: 160.
    Facility Operating License Nos. DPR-71 and DPR-62. Amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: November 9, 1994 (59 FR 
55867). The March 24, 1995, submittal provided clarifying information 
only and did not affect the no significant hazards consideration as 
published in the Federal Register.
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated March 31, 1995.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: University of North Carolina 
at Wilmington, William Madison Randall Library, 601 S. College Road, 
Wilmington, North Carolina 28403-3297.

Carolina Power & Light Company, Docket No. 50-261, H. B. Robinson Steam 
Electric Plant, Unit No. 2, Darlington County, South Carolina

    Date of application for amendment: July 22, 1994, as supplemented 
March 6, 1995.
    Brief description of amendment: The amendments change the Technical 
Specifications to implement a performance based assessment program, 
including corresponding organizational and functional changes. 
Specifically, the changes affect the independent assessment of plant 
activity and the independent review function, the independent 
assessment of plant activity and the Independent Safety Engineering 
Group. These functions will be performed by the Nuclear Assessment 
Section (NAS). The NAS's fundamental role will be to: (1) Assist plant 
management in the early identification of issues that may prevent the 
plant from achieving quality, and (2) ensure effective correction of 
deficiencies.
    Date of issuance: March 31, 1995.
    Effective date: March 31, 1995.
    Amendment No.: 160.
    Facility Operating License No. DPR-23. Amendment revises the 
Technical Specifications.
    Date of initial notice in Federal Register: August 31, 1994 (59 FR 
45017). The March 6, 1995, submittal added Radiation Protection to the 
list of assessments in TS 6.5.5.2 and reworded Section 6.5.4.4, but did 
not change the no significant hazards consideration determination as 
published in the Federal Register.
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated March 31, 1995.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: Hartsville Memorial Library, 
147 West College Avenue, Hartsville, South Carolina 29550.

Carolina Power & Light Company, Docket No. 50-261, H. B. Robinson Steam 
Electric Plant, Unit No. 2, Darlington County, South Carolina

    Date of application for amendment: August 23, 1994, as supplemented 
March 2, 1995.
    Brief description of amendment: The amendment increases the trip 
voltage settings of the degraded grid voltage relays which are shown in 
TS Table 3.5-1, Engineering Safety Feature System Initiation Instrument 
Setting Limits, Item 6b.
    Date of issuance: April 14, 1995.
    Effective date: April 14, 1995.
    Amendment No.: 161.
    Facility Operating License No. DPR-23. Amendment revises the 
Technical Specifications.
    Date of initial notice in Federal Register: March 2, 1995 (60 FR 
11692). The licensee's March 2, 1995, submittal provided clarifying 
information that did not affect the no significant hazards 
consideration determination.
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated April 14, 1995.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: Hartsville Memorial Library, 
147 West College Avenue, Hartsville, South Carolina 29550.

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: June 9, 1994.
    Brief description of amendments: The amendments implement a partial 
application of the Generic Electric ARTS (Average Power Range Monitor 
(APRM)/Rod Block Monitor (RBM)/Technical Specification) Improvement 
Program. Four new ARTS thermal limits replace the existing flow-
referenced APRM trip setpoint setdown requirements and the Minimum 
Critical Power Ratio (MCPR) Kf factor.
    Date of issuance: April 13, 1995.
    Effective date: Immediately to be implemented within 60 days.
    Amendment Nos.: 103 and 88.
    Facility Operating License Nos. NPF-11 and NPF-18. The amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: August 3, 1994 (59 FR 
39582). [[Page 20534]] 
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated April 13, 1995.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: Jacobs Memorial Library, 
Illinois Valley Community College, Oglesby, Illinois 61348.

Detroit Edison Company, Docket No. 50-341, Fermi-2, Monroe County, 
Michigan

    Date of application for amendment: July 29, 1993, as supplemented 
October 8, 1993.
    Brief description of amendment: The amendment revises the Technical 
Specification ACTION STATEMENTS related to operability of the control 
room emergency filtration system. A portion of the amendment request 
was denied. A separate Notice of Denial of Amendment has been sent to 
the Federal Register for publication.
    Date of issuance: March 31, 1995.
    Effective date: March 31, 1995, with full implementation within 45 
days.
    Amendment No.: 103.
    Facility Operating License No. NPF-43. Amendment revises the 
Technical Specifications.
    Date of initial notice in Federal Register: August 18, 1993 (58 FR 
43925). The October 8, 1993, letter provided clarifying information 
within the scope of the original submittal and did not change the 
staff's initial no significant hazards consideration determination.
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated March 31, 1995.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: Monroe County Library System, 
3700 South Custer Road, Monroe, Michigan 48161.

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

    Date of application for amendments: January 18, 1995.
    Brief description of amendments: The amendments revise TS Table 
4.3-3 to allow the analog channel operational test interval for 
radiation monitoring instrumentation to be increased from monthly to 
quarterly, and are consistent with the guidance in Generic Letter 93-
05, ``Line-Item Technical Specifications Improvements to Reduce 
Surveillance Requirements for Testing During Power Operation.''
    Date of issuance: April 3, 1995.
    Effective date: As of the date of issuance to be implemented within 
30 days.
    Amendment Nos.: 154 and 136.
    Facility Operating License Nos. NPF-9 and NPF-17. Amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: March 1, 1995 (60 FR 
11132).
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated April 3, 1995.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: Atkins Library, University of 
North Carolina, Charlotte (UNCC Station), North Carolina 28223.

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

    Date of application for amendment: July 22, 1993.
    Brief description of amendment: The amendment revised operability 
requirements for the Reactor Protection System and the Engineered 
Safety Features Acturation System.
    Date of issuance: April 3, 1995.
    Effective date: April 3, 1995.
    Amendment No.: 159.
    Facility Operating License No. NPF-6. Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: September 1, 1993 (58 
FR 46229).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated April 3, 1995.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: Tomlinson Library, Arkansas 
Tech University, Russellville, AR 72801.

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

    Date of application for amendments: July 19, 1994, resubmitted 
October 20, 1994, and supplemented February 20, 1995.
    Brief description of amendments: These amendments relate to the 
maximum allowable reactor thermal power operation with inoperable main 
steam safety valves.
    Date of issuance: April 11, 1995.
    Effective date: April 11, 1995.
    Amendment Nos.: 172 and 166.
    Facility Operating Licenses Nos. DPR-31 and DPR-41. Amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: November 23, 1994 (59 
FR 60380).
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated April 11, 1995. The February 20, 1995 
submittal provided additional information that did not change the 
staff's proposed no significant hazards consideration determination.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: Florida International 
University, University Park, Miami, Florida 33199.

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

    Date of application for amendment: October 28, 1994.
    Brief description of amendment: The amendment revised the license 
by deleting the ``Plan for the Integrated Scheduling of Plant 
Modifications for the Duane Arnold Energy Center,'' as a condition of 
the license.
    Date of issuance: April 3, 1995.
    Effective date: April 3, 1995.
    Amendment No.: 208.
    Facility Operating License No. DPR-49. Amendment revised the 
license.
    Date of initial notice in Federal Register: March 1, 1995 (60 FR 
11134).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated April 3, 1995.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: Cedar Rapids Public Library, 
500 First Street SE., Cedar Rapids, Iowa 52401.

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

    Date of application for amendment: December 6, 1994, as 
supplemented on February 27, and April 4, 1995.
    Brief description of amendment: The amendment revises the Technical 
Specifications to allow the use of the Combustion Engineering sleeving 
process for repairing steam generator tubes. (The current requirement 
specifies that degraded steam generator tubes be repaired by plugging.)
    Date of issuance: April 14, 1995.
    Effective date: As of the date of issuance, to be implemented 
within 30 days.
    Amendment No.: 149.
    Facility Operating License No. DPR-36. Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: January 18, 1995 (60 FR 
3673). The February 27, and April 4, 1995, [[Page 20535]] letters 
provided clarifying information that did not change the initial 
proposed no significant hazards consideration determination.
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated April 14, 1995.
    No significant hazards consideration comments received: Yes.
    Comments were provided by letter dated February 17, 1995, from the 
Maine State Nuclear Safety Inspector, Office of Nuclear Safety, 
Division of Health Engineering, Department of Human Services. The NRC 
staff responded to his comments in its letter dated March 15, 1995.
    Local Public Document Room location: Wiscasset Public Library, High 
Street, P.O. Box 367, Wiscasset, ME 04578.

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

    Date of application for amendment: September 30, 1994, as 
supplemented February 13, 1995.
    Brief description of amendment: The amendment revises the Technical 
Specifications to (1) clarify the definition of core alterations, (2) 
change the verbiage in the Limiting Condition For Operation (LCO) 
addressing the refueling operations, (3) make changes of surveillance 
requirements involving source range instrumentation, and (4) change the 
LCO regarding the Residual Heat Removal and coolant circulation water 
levels to be consistent with the guidance provided in NUREG-1431, 
Standard Technical Specifications for Westinghouse plants.
    Date of issuance: April 12, 1995.
    Effective date: As of the date of issuance to be implemented within 
30 days.
    Amendment No.: 107.
    Facility Operating License No. NPF-49. Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: November 9, 1994 (59 FR 
55877). The February 13, 1995, letter provided clarifying information 
that did not change the initial proposed no significant hazards 
consideration determination.
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated April 12, 1995.
    No significant hazards consideration comments received: No.
    Local Public Document Room Location: Learning Resources Center, 
Three Rivers Community-Technical College, Thames Valley Campus, 574 New 
London Turnpike, Norwich, CT 06360.

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

    Date of application for amendment: January 26, 1993, as 
supplemented August 4, 1993.
    Brief description of amendment: The amendment changes the Technical 
Specifications governing electrical power systems, AC and DC power 
sources, and onsite power distribution for shutdown conditions (modes 5 
and 6).
    Date of issuance: April 12, 1995.
    Effective date: As of the date of issuance to be implemented within 
30 days.
    Amendment No.: 108.
    Facility Operating License No. NPF-49. Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: May 12, 1993 (58 FR 
28058). The August 4, 1993, submittal provided clarifying information 
that did not change the initial proposed no significant hazards 
consideration determination.
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated April 12, 1995.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: Learning Resources Center, 
Three Rivers Community-Technical College, Thames Valley Campus, 574 New 
London Turnpike, Norwich, CT 06360.

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

    Date of amendment request: April 23, 1993.
    Description of amendment request: The amendment revises the 
Appendix A Technical Specifications to allow longer surveillance test 
intervals and allowed outage times for the reactor protection system 
and the engineered safety features actuation system. Also, the 
amendment removes the requirement to perform the reactor trip system 
analog channel operational test on a staggered basis.
    Date of issuance: April 10, 1995.
    Effective date: As of the date of issuance to be implemented within 
30 days.
    Amendment No.: 36.
    Facility Operating License No. NPF-86. Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: August 4, 1993 (58 FR 
41507).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated April 10, 1995.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: Exeter Public Library, 47 
Front Street, Exeter, NH 03833.

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

    Date of application for amendments: February 16, 1994 (Reference 
LAR 94-03)
    Brief description of amendments: The amendments revise TS 4.6.1.2, 
``Containment Integrity,'' to allow a more flexible schedule for 
testing the primary containment integrated leakage rate.
    Date of issuance: April 11, 1995.
    Effective date: April 11, 1995, to be implemented within 30 days of 
issuance.
    Amendment Nos.: Unit 1--Amendment No. 99; Unit 2--Amendment No. 98
    Facility Operating License Nos. DPR-80 and DPR-82. The amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: March 30, 1994 (59 FR 
14893).
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated April 11, 1995.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: California Polytechnic State 
University, Robert E. Kennedy Library, Government Documents and Maps 
Department, San Luis Obispo, California 93407.

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

    Date of application for amendments: August 17, 1994.
    Brief description of amendments: The amendments revise the combined 
technical specifications (TS) to change TS 3/4.4.9.1, ``Reactor Coolant 
System--Pressure/Temperature Limits,'' Figures 3.4-2, ``Reactor Coolant 
System Heatup Limitations--Applicable Up to 8 EFPY,'' and 3.4-3, 
``Reactor Coolant System Cooldown Limitations--Applicable Up to 8 
EFPY,'' to extend the applicability up to 12 effective full-power years 
(EFPYs). TS 3/4/4/9/3, ``Overpressure Protection Systems,'' is revised 
to [[Page 20536]] specify a new low-temperature overprotection (LTOP) 
system actuation pressure setpoint. The associated Bases were also 
appropriately revised. Additionally, TS 3/4.1.2.2, ``Flow Paths--
Operating;'' TS 3/4.1.2.4, ``Charging Pumps--Operating;'' TS 3/4.4.1.3, 
``Hot Shutdown;'' TS 3/4.4.1.4.1, ``Cold Shutdown--Loops Filled;'' TS 
3/4.4.9.3, ``Overpressure Protection Systems;'' and TS 3/4.5.3, ``ECCS 
Subsystems--Tavg Less than 350 Degrees F,'' are revised to specify 
a new LTOP system enable temperature.
    Date of issuance: April 13, 1995.
    Effective date: April 13, 1995.
    Amendment Nos.: Unit 1--Amendment No. 100; Unit 2--Amendment No. 
99.
    Facility Operating License Nos. DPR-80 and DPR-82. The amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: October 12, 1994 (59 FR 
51622).
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated April 13, 1995.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: California Polytechnic State 
University, Robert E. Kennedy Library, Government Documents and Maps 
Department, San Luis Obispo, California 93407.

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

    Date of application for amendment: June 13, 1994.
    Brief description of amendment: The amendment removes License 
Condition 2.E from the Facility Operating License. License Condition 
2.E incorporated the requirements of U.S. Department of Interior 
publication ``Environmental Criteria for Electric Transmission 
Systems''--1970, which applies to the construction cleanup, 
restoration, and maintenance of transmission lines. The NRC staff has 
determined that removing this condition from the Facility Operating 
License has no bearing on plant safety or the health and safety of the 
public, and is therefore acceptable.
    Date of issuance: March 31, 1995.
    Effective date: As of the date of issuance to be implemented within 
30 days.
    Amendment No.: 224.
    Facility Operating License No. DPR-59. Amendment revised the 
Facility Operating License.
    Date of initial notice in Federal Register: March 1, 1995 (60 FR 
11140).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated March 31, 1995.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: Reference and Documents 
Department, Penfield Library, State University of New York, Oswego, New 
York 13126.

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

    Date of amendments request: June 10, 1995.
    Brief Description of amendments: The amendments allow modifications 
to be made for both units to relocate the lower level steam generator 
water level taps during the upcoming refueling outages. These 
modifications affect the Technical Specifications associated with the 
reactor trip system and engineered safety feature actuation system 
setpoints.
    Date of issuance: April 7, 1995.
    Effective date: As of the date of issuance to be implemented within 
30 days
    Amendment Nos.: 114 and 105.
    Facility Operating License Nos. NPF-2 and NPF-8. Amendments revise 
the Technical Specifications.
    Date of initial notice in Federal Register: March 6, 1995 (60 FR 
12253).
    Commission's related evaluation of the amendments is contained in a 
Safety Evaluation dated April 7, 1995.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: Houston-Love Memorial Library, 
212 W. Burdeshaw Street, Post Office Box 1369, Dothan, Alabama 36302.

Southern Nuclear Operating Company, Inc., Joseph M. Farley Nuclear 
Plant, Unit 2, Houston County, Alabama

    Date of application for amendment: December 7, 1994, as 
supplemented February 14 and March 20, 1995.
    Brief description of amendment: The December 7, 1994, submittal 
requested a permanent change to the Technical Specifications for both 
units related to steam generator tube support plate voltage-based 
repair criteria in accordance with the draft Generic Letter on this 
issue.
    Date of issuance: April 7, 1995.
    Effective date: As of the date of issuance to be implemented within 
30 days.
    Amendment No.: 106.
    Facility Operating License No. NPF-8. Amendment revises the 
Technical Specifications.
    Date of initial notice in Federal Register: February 15, 1995 (60 
FR 8754). The February 14 and March 20, 1995, letters provided 
clarifying information that did not change the scope of the original 
December 7, 1994, application and the initial proposed no significant 
hazards consideration determination.
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated April 7, 1995.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: Houston-Love Memorial Library, 
212 W. Burdeshaw Street, Post Office Box 1369, Dothan, Alabama 36302.

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

    Date of application for amendments: December 16, 1994; supplemented 
February 10, 1995 (TS 94-07).
    Brief description of amendments: The amendments revise the 
technical specifications to reduce the high reactor power level 
setpoints when one or more main steam safety valves are inoperable and 
incorporate related changes.
    Date of issuance: April 4, 1995.
    Effective date: April 4, 1995.
    Amendment Nos.: 196 and 187.
    Facility Operating License Nos. DPR-77 and DPR-79. Amendments 
revise the technical specifications.
    Date of initial notice in Federal Register: March 1, 1995 (60 FR 
11140).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated April 4, 1995.
    No significant hazards consideration comments received: None.
    Local Public Document Room location: Chattanooga-Hamilton County 
Library, 1101 Broad Street, Chattanooga, Tennessee 37402.

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

    Date of amendment request: February 14, 1994 (Reference LAR 94-002, 
TXX-94008), as supplemented by letters dated May 17, 1994 (Reference 
TXX-94142), and April 3, 1995 (TXX-95098).
    Brief description of amendments: The amendments revised Technical 
Specification (TS) 3/4.2.4, ``Quadrant Power Tilt Ratio,'' by replacing 
the existing TS and associated Bases concerning the quadrant power tilt 
ratio with a TS consistent with the improved Standard Technical 
Specifications (NUREG-1431). [[Page 20537]] 
    Date of issuance: April 4, 1995.
    Effective date: April 4, 1995.
    Amendment Nos.: Unit 1--Amendment No. 36; Unit 2--Amendment No. 22.
    Facility Operating License Nos. NPF-87 and NPF-89. The amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: July 20, 1994 (59 FR 
37087).
    The additional information contained in the supplemental letter 
dated April 3, 1995, was clarifying in nature and thus, within the 
scope of the initial notice and did not affect the staff's proposed no 
significant hazards consideration determination.
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated April 4, 1995.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: University of Texas at 
Arlington Library, Government Publications/Maps, 702 College, P.O. Box 
19497, Arlington, TX 76019.

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

    Date of amendment request: February 28, 1995 (Reference LAR 95-01).
    Brief description of amendments: These amendments replace 
Surveillance Requirement (SR) 4.6.2.1b of Technical Specification (TS) 
3/4.6.2, ``Depressurization and Cooling Systems--Containment Spray 
System,'' with the corresponding SR from NUREG-1431. Bases Section 3/
4.6.2.1 ``Containment Spray System'' has also been revised to expand 
the detail consistent with the corresponding Bases from NUREG-1431. The 
SR, and its associated Bases, for confirming the performance of the 
containment spray pumps are changed by replacing the specific pump head 
and flow values with the general requirement that the pumps provide the 
required head at the flow test point while the specific required values 
are moved to the Comanche Peak Steam Electric Station Technical 
Requirements Manual.
    Date of issuance: April 6, 1995.
    Effective date: April 6, 1995.
    Amendment Nos.: Unit 1--Amendment No. 37; Unit 2--Amendment No. 23.
    Facility Operating License Nos. NPF-87 and NPF-89. The amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: March 6, 1995 (60 FR 
12255).
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated April 6, 1995.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: University of Texas at 
Arlington Library, Government Publications/Maps, 702 College, P.O. Box 
19497, Arlington, TX 76019.

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

    Date of amendment request: April 22, 1994 (LAR 94-011, TXX-94116).
    Brief description of amendments: These amendments changed 
Surveillance Requirement 4.7.1.2 of Technical Specification 3/4.7.1.2 
``Auxiliary Feedwater System,'' for the operational test frequency of 
the motor driven and turbine driven pumps from ``at least once per 31 
days on a STAGGERED TEST BASIS'' to ``at least once per 92 days on a 
STAGGERED TEST BASIS.'' This change is consistent with ASME Section XI 
requirements and Generic Letter 93-05, ``Line-Item Technical 
Specifications Improvements to Reduce Surveillance Requirements for 
Testing During Power Operations.''
    Date of issuance: April 7, 1995.
    Effective date: April 7, 1995.
    Amendment Nos.: Unit 1--Amendment No. 38; Unit 2--Amendment No. 24.
    Facility Operating License Nos. NPF-87 and NPF-89. The amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: August 3, 1994 (59 FR 
39598).
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated April 7, 1995.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: University of Texas at 
Arlington Library, Government Publications/Maps, 702 College, P.O. Box 
19497, Arlington, TX 76019.

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

    Date of application for amendment: December 9, 1994, as 
supplemented on January 27, 1995.
    Brief description of amendment: The amendment revises the Technical 
Specification Surveillance Requirement 4.6.1.2.a and its associated 
Bases. The change defers the requirement to perform the Type A 
Containment Integrated Leak Rate Test until Refuel 8 (October 1996), in 
conjunction with the exemption to 10 CFR Part 50, Appendix J.
    Date of issuance: April 5, 1995.
    Effective date: April 5, 1995.
    Amendment No.: 98.
    Facility Operating License No. NPF-30. Amendment revises the 
Technical Specifications.
    Date of initial notice in Federal Register: March 1, 1995 (60 FR 
11141).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated April 5, 1995.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: Callaway County Public 
Library, 710 Court Street, Fulton, Missouri 65251.

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

    Date of amendment request: December 12, 1994.
    Brief description of amendment: The amendment clarifies the 
surveillance requirements for verifying the correct required position 
for the valves in the auxiliary feedwater system.
    Date of issuance: April 3, 1995.
    Effective date: April 3, 1995, to be implemented within 30 days of 
issuance.
    Amendment No.: 85.
    Facility Operating License No. NPF-42. The amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: January 18, 1995 (60 FR 
3677).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated April 3, 1995.
    No significant hazards consideration comments received: No.
    Local Public Document Room locations: Emporia State University, 
William Allen White Library, 1200 Commercial Street, Emporia, Kansas 
66801 and Washburn University School of Law Library, Topeka, Kansas 
66621.

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

    Date of application for amendment: April 11, 1994, as supplemented 
on November 30, and December 22, 1994, and March 3, 1995.
    Brief description of amendment: The amendment revises Kewaunee 
Nuclear Power Plant (KNPP) Technical Specification (TS) 3.1.f, 
``Minimum Conditions for Criticality,'' and its associated basis, by 
specifying that the moderator temperature coefficient (MTC) shall be no 
greater than 5.0 pcm/ deg.F when at or below 60% rated thermal power 
and shall be zero or negative when above 60% rated thermal power. 
Additionally, the MTC shall be no less [[Page 20538]] negative than -8 
pcm/ deg.F for 95% of the cycle time at full power. The amendment also 
incorporates required actions to be implemented, if the MTC 
specification is not met.
    Date of issuance: April 3, 1995.
    Effective date: April 3, 1995.
    Amendment No.: 117.
    Facility Operating License No. DPR-43. Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: September 28, 1994 (59 
FR 49442).
    The November 30, and December 22, 1994, and March 3, 1995, 
submittals, provided clarifying information and expanded the basis 
portion of the TS, but 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 April 3, 1995.
    No significant hazards consideration comments received: None.
    Local Public Document Room location: University of Wisconsin 
Library Learning Center, 2420 Nicolet Drive, Green Bay, Wisconsin 
54301.

Notice of Issuance of Amendments to Facility Operating Licenses and 
Final Determination of No Significant Hazards Consideration and 
Opportunity for a Hearing (Exigent Public Announcement or Emergency 
Circumstances)

    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 for the 
amendment 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.
    Because of exigent or emergency circumstances associated with the 
date the amendment was needed, there was not time for the Commission to 
publish, for public comment before issuance, its usual 30-day Notice of 
Consideration of Issuance of Amendment, Proposed No Significant Hazards 
Consideration Determination, and Opportunity for a Hearing.
    For exigent circumstances, the Commission has either issued a 
Federal Register notice providing opportunity for public comment or has 
used local media to provide notice to the public in the area 
surrounding a licensee's facility of the licensee's application and of 
the Commission's proposed determination of no significant hazards 
consideration. The Commission has provided a reasonable opportunity for 
the public to comment, using its best efforts to make available to the 
public means of communication for the public to respond quickly, and in 
the case of telephone comments, the comments have been recorded or 
transcribed as appropriate and the licensee has been informed of the 
public comments.
    In circumstances where failure to act in a timely way would have 
resulted, for example, in derating or shutdown of a nuclear power plant 
or in prevention of either resumption of operation or of increase in 
power output up to the plant's licensed power level, the Commission may 
not have had an opportunity to provide for public comment on its no 
significant hazards consideration determination. In such case, the 
license amendment has been issued without opportunity for comment. If 
there has been some time for public comment but less than 30 days, the 
Commission may provide an opportunity for public comment. If comments 
have been requested, it is so stated. In either event, the State has 
been consulted by telephone whenever possible.
    Under its regulations, the Commission may issue and make an 
amendment immediately effective, notwithstanding the pendency before it 
of a request for a hearing from any person, in advance of the holding 
and completion of any required hearing, where it has determined that no 
significant hazards consideration is involved.
    The Commission has applied the standards of 10 CFR 50.92 and has 
made a final determination that the amendment involves no significant 
hazards consideration. The basis for this determination is contained in 
the documents related to this action. Accordingly, the amendments have 
been issued and made effective 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 
application for amendment, (2) the amendment to Facility Operating 
License, 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 room for the particular facility involved.
    The Commission is also offering an opportunity for a hearing with 
respect to the issuance of the amendment. By May 26, 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. 
[[Page 20539]] Any person who has filed a petition for leave to 
intervene or who has been admitted as a party may amend the petition 
without requesting leave of the Board up to 15 days prior to the first 
prehearing conference scheduled in the proceeding, but such an amended 
petition must satisfy the specificity requirements described above.
    Not later than 15 days prior to the first prehearing conference 
scheduled in the proceeding, a petitioner shall file a supplement to 
the petition to intervene which must include a list of the contentions 
which are sought to be litigated in the matter. Each contention must 
consist of a specific statement of the issue of law or fact to be 
raised or controverted. In addition, the petitioner shall provide a 
brief explanation of the bases of the contention and a concise 
statement of the alleged facts or expert opinion which support the 
contention and on which the petitioner intends to rely in proving the 
contention at the hearing. The petitioner must also provide references 
to those specific sources and documents of which the petitioner is 
aware and on which the petitioner intends to rely to establish those 
facts or expert opinion. Petitioner must provide sufficient information 
to show that a genuine dispute exists with the applicant on a material 
issue of law or fact. Contentions shall be limited to matters within 
the scope of the amendment under consideration. The contention must be 
one which, if proven, would entitle the petitioner to relief. A 
petitioner who fails to file such a supplement which satisfies these 
requirements with respect to at least one contention will not be 
permitted to participate as a party.
    Those permitted to intervene become parties to the proceeding, 
subject to any limitations in the order granting leave to intervene, 
and have the opportunity to participate fully in the conduct of the 
hearing, including the opportunity to present evidence and cross-
examine witnesses. Since the Commission has made a final determination 
that the amendment involves no significant hazards consideration, if a 
hearing is requested, it will not stay the effectiveness of the 
amendment. Any hearing held would take place while the amendment is in 
effect.
    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 the 
factors specified in 10 CFR 2.714(a)(1)(i)-(v) and 2.714(d).

Commonwealth Edison Company, Docket Nos. 50-295 and 50-304, Zion 
Nuclear Power Station, Units 1 and 2, Lake County, Illinois

    Date of application for amendments: March 29, 1995.
    Brief description of amendments: The amendments change the 
Technical Specifications by revising the periodicity of the channel 
functional test of the turbine driven auxiliary feedwater pump from 
quarterly to each refueling outage.
    Date of issuance: April 14, 1995.
    Effective date: April 14, 1995.
    Amendment Nos.: 161 and 149.
    Facility Operating License Nos. DPR-39 and DPR-48. The amendments 
revised the Technical Specifications.
    Public comments requested as to proposed no significant hazards 
consideration: No.
    The Commission's related evaluation of the amendments, finding of 
emergency circumstances, and final determination of no significant 
hazards consideration are contained in a Safety Evaluation dated April 
14, 1995.
    Attorney for licensee: Michael I. Miller, Esquire; Sidley and 
Austin, One First National Plaza, Chicago, Illinois 60690.
    Local Public Document Room location: Waukegan Public Library, 128 
N. County Street, Waukegan, Illinois 60085.
    NRC Project Director: Robert A. Capra.

    Dated at Rockville, Maryland, this 19th day of April 1995.

    For the Nuclear Regulatory Commission.
Elinor G. Adensam,
Acting Director, Division of Reactor Projects--III/IV, Office of 
Nuclear Reactor Regulation.
[FR Doc. 95-10127 Filed 4-25-95; 8:45 am]
BILLING CODE 7590-01-P