[Federal Register Volume 61, Number 197 (Wednesday, October 9, 1996)]
[Notices]
[Pages 52962-52977]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: X96-11009]


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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 September 16, 1996, through September 27, 
1996. The last biweekly notice was published on September 25, 1996 (61 
FR 50338).

Notice Of Consideration Of Issuance Of Amendments To Facility 
Operating Licenses, Proposed No Significant Hazards Consideration 
Determination, And Opportunity For A Hearing

    The Commission has made a proposed determination that the following 
amendment requests involve no significant hazards consideration. Under 
the Commission's regulations in 10 CFR 50.92, this means that operation 
of the facility in accordance with the proposed amendment would not (1) 
involve a significant increase in the probability or consequences of an 
accident previously evaluated; or (2) create the possibility of a new 
or different kind of accident from any accident previously evaluated; 
or (3) involve a significant reduction in a margin of safety. The basis 
for this proposed determination for each amendment request is shown 
below.
    The Commission is seeking public comments on this proposed 
determination. Any comments received within 30 days after the date of 
publication of this notice will be considered in making any final 
determination.
    Normally, the Commission will not issue the amendment until the 
expiration of the 30-day notice period. However, should circumstances 
change during the notice period such that failure to act in a timely 
way would result, for example, in derating or shutdown of the facility, 
the Commission may issue the license amendment before the expiration of 
the 30-day notice period, provided that its final determination is that 
the amendment involves no significant hazards consideration. The final 
determination will consider all public and State comments received 
before action is taken. Should the Commission take this action, it will 
publish in the Federal Register a notice of issuance and provide for 
opportunity for a hearing after issuance. The Commission expects that 
the need to take this action will occur very infrequently.
    Written comments may be submitted by mail to the Chief, Rules 
Review and Directives Branch, Division of Freedom of Information and 
Publications Services, Office of Administration, U.S. Nuclear 
Regulatory Commission, Washington, DC 20555-0001, and should cite the 
publication date and page number of this Federal Register notice. 
Written comments may also be delivered to Room 6D22, Two White Flint 
North, 11545 Rockville Pike, Rockville, Maryland from 7:30 a.m. to 4:15 
p.m. Federal workdays. Copies of written comments received may be 
examined at the NRC Public Document Room, the Gelman Building, 2120 L 
Street, NW., Washington, DC. The filing of requests for a hearing and 
petitions for leave to intervene is discussed below.
    By November 8, 1996, the licensee may file a request for a hearing 
with respect to issuance of the amendment to the subject facility 
operating license and any person whose interest may be affected by this 
proceeding and who wishes to participate as a party in the proceeding 
must file a written request for a hearing and a petition for leave to 
intervene. Requests for a hearing and a petition for leave to intervene 
shall be filed in accordance with the Commission's ``Rules of Practice 
for Domestic Licensing Proceedings'' in 10 CFR Part 2. Interested 
persons should consult a current copy of 10 CFR 2.714 which is 
available at the Commission's Public Document Room, the Gelman 
Building, 2120 L Street, NW., Washington, DC and at the local public 
document room for the particular facility involved. If a request for a 
hearing or petition for leave to intervene is filed by the above date, 
the Commission or an Atomic Safety and Licensing Board, designated by 
the Commission or by the Chairman of the Atomic Safety and Licensing 
Board Panel, will rule on the request and/or petition; and the 
Secretary or the designated Atomic Safety and Licensing Board will 
issue a notice of a hearing or an appropriate order.
    As required by 10 CFR 2.714, a petition for leave to intervene 
shall set forth with particularity the interest of the petitioner in 
the proceeding, and how that interest may be affected by the results of 
the proceeding. The petition should specifically explain the reasons 
why intervention should be permitted with particular reference to the 
following factors: (1) the nature of the petitioner's right under the 
Act to be made a party to the proceeding; (2) the nature and extent of 
the petitioner's property, financial, or other interest in the 
proceeding; and (3) the possible effect of any order which may be 
entered in the proceeding on the petitioner's interest. The petition 
should also identify the specific aspect(s) of the subject matter of 
the proceeding as to which petitioner wishes to intervene. Any person 
who has filed a petition for leave to intervene or who has been 
admitted as a party may amend the petition without requesting leave of 
the Board up to 15 days prior to the first prehearing conference 
scheduled in the proceeding, but such an amended petition must satisfy 
the specificity requirements described above.
    Not later than 15 days prior to the first prehearing conference 
scheduled in the proceeding, a petitioner shall file a supplement to 
the petition to intervene which must include a list of the contentions 
which are sought to be litigated in the matter. Each contention must 
consist of a specific statement of the issue of law or fact to be 
raised or controverted. In addition, the petitioner shall provide a 
brief explanation of the bases of the contention and a concise 
statement of the alleged facts or expert opinion which support the 
contention

[[Page 52963]]

and on which the petitioner intends to rely in proving the contention 
at the hearing. The petitioner must also provide references to those 
specific sources and documents of which the petitioner is aware and on 
which the petitioner intends to rely to establish those facts or expert 
opinion. Petitioner must provide sufficient information to show that a 
genuine dispute exists with the applicant on a material issue of law or 
fact. Contentions shall be limited to matters within the scope of the 
amendment under consideration. The contention must be one which, if 
proven, would entitle the petitioner to relief. A petitioner who fails 
to file such a supplement which satisfies these requirements with 
respect to at least one contention will not be permitted to participate 
as a party.
    Those permitted to intervene become parties to the proceeding, 
subject to any limitations in the order granting leave to intervene, 
and have the opportunity to participate fully in the conduct of the 
hearing, including the opportunity to present evidence and cross-
examine witnesses.
    If a hearing is requested, the Commission will make a final 
determination on the issue of no significant hazards consideration. The 
final determination will serve to decide when the hearing is held.
    If the final determination is that the amendment request involves 
no significant hazards consideration, the Commission may issue the 
amendment and make it immediately effective, notwithstanding the 
request for a hearing. Any hearing held would take place after issuance 
of the amendment.
    If the final determination is that the amendment request involves a 
significant hazards consideration, any hearing held would take place 
before the issuance of any amendment.
    A request for a hearing or a petition for leave to intervene must 
be filed with the Secretary of the Commission, U.S. Nuclear Regulatory 
Commission, Washington, DC 20555-0001, Attention: Docketing and 
Services Branch, or may be delivered to the Commission's Public 
Document Room, the Gelman Building, 2120 L Street, NW., Washington DC, 
by the above date. 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 20555-0001, 
and to the attorney for the licensee.
    Nontimely filings of petitions for leave to intervene, amended 
petitions, supplemental petitions and/or requests for a hearing will 
not be entertained absent a determination by the Commission, the 
presiding officer or the Atomic Safety and Licensing Board that the 
petition and/or request should be granted based upon a balancing of 
factors specified in 10 CFR 2.714(a)(1)(i)-(v) and 2.714(d).
    For further details with respect to this action, see the 
application for amendment which is available for public inspection at 
the Commission's Public Document Room, the Gelman Building, 2120 L 
Street, NW., Washington, DC, and at the local public document room for 
the particular facility involved.

Baltimore Gas and Electric Company, Docket Nos. 50-317 and 50-318, 
Calvert Cliffs Nuclear Power Plant, Unit Nos. 1 and 2, Calvert 
County, Maryland

    Date of amendments request: September 10, 1996
    Description of amendments request: The proposed amendment would 
extend the Engineered Safety Features Actuation System (ESFAS) 
automatic actuation logic channel functional test surveillance interval 
from monthly to quarterly. The amendment request is based on analysis 
documented in Combustion Engineering Owners Group (CEOG) Topical 
Reports CEN-327 (Reference a), CEN 327, Supplement 1 (Reference b), and 
CEN-403, Revision 1-A, (Reference c). We have confirmed that the 
information presented in CEN-327 and CEN-403 is applicable to Calvert 
Cliffs, and agree with the methodology used to develop the topical 
reports. In a related matter, the licensee, also requests that the 
surveillance test interval for the containment sump isolation valves be 
extended from monthly to quarterly.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:
    1. Would not involve a significant increase in the probability 
or consequences of an accident previously evaluated.
    The Reactor Protective System and the Engineered Safety Features 
Actuation System (ESFAS) provide the actuation signals to safety 
equipment necessary to mitigate design basis accidents and 
transients. The proposed change would increase the surveillance test 
interval from monthly to quarterly for the ESFAS automatic actuation 
logic channel functional tests and associated actuation relays. The 
proposed change will also extend the containment sump isolation 
valve automatic opening verification surveillance interval from 
monthly to quarterly. The ESFAS instruments and containment sump 
isolation valves are not initiators in any previously evaluated 
accidents. Therefore, the proposed change does not involve an 
increase in the probability of an accident previously evaluated.
    The ESFAS automatic actuation logic circuitry and actuation 
relays are essentially digital devices, which are not subject to 
time-related instrument drift. Therefore, a plant-specific 
instrument drift analysis for these components is not required. 
However, in support of Calvert Cliffs License Amendment Request, 
dated May 27, 1994, a plant-specific setpoint drift analysis for 
each sensor loop demonstrated that the observed changes in 
instrument uncertainties for the extended surveillance test interval 
did not exceed the 30-day setpoint assumptions. This provides 
confidence that the 90-day test interval will not impact the ability 
to detect and monitor system degradation. A review of previous 
containment sump isolation valve surveillance test procedures 
revealed no valve or valve operator failures. Additionally, single 
failure criteria continues to be satisfied by two redundant and 
independent valves on each unit. Therefore, the proposed changes 
will not change the ability of the ESFAS instrumentation or 
associated engineered safety features equipment to respond to and 
mitigate the consequences of any previously evaluated accident.
    Therefore, the proposed change does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    2. Would not create the possibility of a new or different type 
of accident from any accident previously evaluated.
    The proposed extended surveillance test interval for the ESFAS 
instruments, actuation relays, and containment sump isolation valve 
automatic opening verification does not involve any changes in 
equipment or the function of these instruments. The proposed change 
does not represent a change in the configuration or operation of the 
plant. The ESFAS setpoints will not be affected since the ESFAS 
automatic actuation logic circuitry and actuation relays are not 
subject to time-related instrument drift. Therefore, the proposed 
change does not create the possibility of a new or different type of 
accident from any accident previously evaluated.
    3. Would not involve a significant reduction in a margin of 
safety.
    The proposed change will not affect the functions of the ESFAS 
instruments or associated equipment. Topical Reports CEN-327, ``RPS/
ESFAS Extended Test Interval Evaluation,'' and CEN-327, Supplement 
1, quantified the corresponding changes in core

[[Page 52964]]

melt frequency for the representative fault tree models that were 
developed for Calvert Cliffs. Additionally, the ESFAS actuation 
relay failure data presented in CEN-403, Revision 1-A, ``ESFAS 
Subgroup Relay Test Interval Extension,'' justifies extending the 
test interval for these relays. The proposed change has two 
principal effects with opposing impacts on core melt frequency. The 
first impact is a slight increase in core melt frequency that 
results from the increased possibility of an undetected 
instrumentation failure due to the extended surveillance interval. 
This assumed unavailability results from less frequent testing. The 
undetected ESFAS failure represents the potential for the failure of 
the appropriate engineered safety features to actuate when required. 
The opposing impact on core melt risk is the corresponding reduction 
in core melt frequency that would result due to the reduced exposure 
of the plant to test-induced transients. Topical Report CEN-327 
determined that the two changes are nearly equal, and the net result 
is no distinguishable effect on plant safety. The NRC issued a 
Safety Evaluation Report which found that the above evaluations were 
acceptable for justifying the extensions in the surveillance test 
intervals for the ESFAS automatic actuation logic channel functional 
tests from 30 days to 90 days. In addition to the evaluation of risk 
given in Topical Report CEN-327, we have evaluated the plant 
specific risk associated with these proposed changes and concluded 
that changing the surveillance intervals from monthly to quarterly 
results in a net decrease in the annual core melt frequency.
    The ESFAS setpoints will not be changed since ESFAS automatic 
actuation logic circuitry and actuation relays are not subject to 
time-related instrument drift. The conclusions of the accident 
analyses in the Calvert Cliffs Updated Final Safety Analysis Report 
remain valid and the safety limits continue to be met.
    Extending the containment sump isolation valve automatic opening 
surveillance interval from monthly to quarterly will not 
significantly reduce the margin of safety. Both Units 1 and 2 are 
provided with two containment sump isolation valves, which satisfy 
single failure criteria. Historical review of surveillance test 
procedures and Nuclear Plant Reliability Data System data revealed 
no failures of these valves or associated valve operators. We have 
also evaluated the plant specific risk associated with this proposed 
change to the surveillance interval and conclude that the risk is 
acceptable.
    Based on the generic and plant specific risk evaluations and the 
demonstrated low failure rate of the components, we conclude that 
these proposed changes do not involve a significant reduction in the 
margin of safety.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendments request involves no significant hazards consideration.
    Local Public Document Room location: Calvert County Library, Prince 
Frederick, Maryland 20678.
    Attorney for licensee: Jay E. Silbert, Esquire, Shaw, Pittman, 
Potts and Trowbridge, 2300 N Street, NW., Washington, DC 20037.
    NRC Project Director: Alexander W. Dromerick, Acting Director

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: August 29, 1996, as supplemented on 
September 20, 1996
    Description of amendment request: The proposed amendments would 
change the Technical Specifications to implement 10 CFR Part 50, 
Appendix J, Option B, by referring to Regulatory Guide 1.163, 
``Performance-Based Containment Leakage-Test Program,'' with certain 
exceptions detailed in the licensee's application.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:
    1. Involve a significant increase in the probability or 
consequences of an accident previously evaluated.
    The proposed amendment cannot affect the probability of an 
accident since it involves only changes in the containment leakage 
rate testing program. There is no credible accident which can be 
initiated by containment leakage rate testing.
    The proposed amendment will not affect the consequences of a[n] 
accident since the allowable containment leakage rates, which 
determine the offsite consequences of a[n] accident, are unchanged. 
Only the frequency of measuring the leakage rates may be changed.
    2. Create the possibility of a new or different kind of accident 
from any accident previously evaluated.
    The proposed amendment will not create the possibility of a new 
or different kind of accident since there are no changes to any 
systems, structures, or components, and no changes in the method of 
operation of any system, structure, or component.
    3. Involve a significant reduction in a margin of safety.
    The proposed amendment will not involve a significant reduction 
in the margin of safety. As documented in the 10 CFR 50, Appendix J, 
Option B Proposed Rule and Final Rule published in the Federal 
Register, the additional industry wide risk resulting from the 
proposed change is marginal and within acceptable limits.
    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 60603
    NRC Project Director: Robert A. Capra

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

    Date of amendment request: August 23, 1996
    Description of amendment request: The proposed amendment would 
change the technical specifications to allow fuel enrichments of up to 
5.0 weight percent uranium-235.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:
    1. Does Not Involve a Significant Increase in the Probability or 
Consequences of an Accident Previously Evaluated.
    The calculated k-effective including uncertainties, demonstrates 
substantial margin to criticality in the fuel assembly storage 
locations for both normal and accident conditions; therefore, the 
probability of a previously evaluated accident is not significantly 
increased. Since a criticality accident is demonstrated to not be 
feasible under the specified conditions, the consequences of a 
previously evaluated accident are not significantly increased. 
Administrative controls are utilized in order to assure that a fuel 
assembly is not placed in an unanalyzed configuration. Therefore, 
this change does not involve a significant increase in the 
probability or consequences of any accident previously evaluated.
    2. Does Not Create the Possibility of a New or Different Kind of 
Accident from any Previously Evaluated.
    The increase in fuel enrichment could be considered a change in 
plant equipment; however, it would only affect reactivity. The 
reactivity increase has been analyzed and shown that no new or 
different kinds of accidents from any previously evaluated exist. 
The proposed change does not involve the addition of any plant 
equipment, nor does it modify the method of operation of any plant 
equipment. Also, the proposed change would not alter the design or 
configuration of the plant beyond the standard functional 
capabilities of the equipment. Therefore, this change does not 
create the possibility of a new or different kind of accident from 
any previously evaluated.
    3. Does Not Involve a Significant Reduction in the Margin of 
Safety.

[[Page 52965]]

    The proposed change has been analyzed to maintain a k-effective 
of less than the criticality acceptance criteria of 0.95 including 
uncertainties at the 95/95 probability/confidence level for all 
storage configurations. Additionally, the optimum moderation 
condition for the new fuel storage racks has been analyzed and 
determined to meet the acceptance criteria of maintaining k-
effective of 0.98. The use of physical restraints for blocking the 
storage locations where fuel is prohibited to be stored in the spent 
fuel pools prevents misloading of fuel into these locations. A 
dropped assembly and/or the misplacement of a fuel assembly for each 
storage configuration has been analyzed. By crediting 1000 ppm boron 
(ANO-2 Technical Specifications require 1600 ppm), the 95/95 k-
effective is well below 0.95. Therefore, this change does not 
involve a significant reduction in the margin of safety.
    Based upon the reasoning presented above and the previous 
discussion of the amendment request, Entergy Operations proposes 
that the requested change does not involve a significant hazards 
consideration.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: Tomlinson Library, Arkansas 
Tech University, Russellville, AR 72801
    Attorney for licensee: Nicholas S. Reynolds, Esquire, Winston and 
Strawn, 1400 L Street, NW., Washington, DC 20005-3502
    NRC Project Director: William D. Beckner

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

    Date of amendment request: August 23, 1996
    Description of amendment request: The proposed amendment would 
revise Technical Specification (TS) Section 5.0, Design Features, and 
would for the most part, adopt NUREG-1432, Revision 1, improved 
``Standard Technical Specifications for Combustion Engineering Plants'' 
(ISTS), for this section of the TSs.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:
    1. Does Not Involve a Significant Increase in the Probability or 
Consequences of an Accident Previously Evaluated.
    The proposed amendment revises the Section 5.0, Design Features, 
and would for the most part, adopt NUREG-1432 Revision 1, ``Standard 
Technical Specifications Combustion Engineering Plants,'' for this 
section of the technical specifications. This proposed change will 
also allow the relocation of portions of the design features section 
of the technical specifications to other licensee controlled 
documents that are controlled under 10 CFR 50.59. This approach is 
consistent with the NRC final policy statement and the staff's 
Technical Specification line item improvement program. The 
relocation of information to licensee controlled documents will 
improve the usability and readability of technical specifications 
without changing any of the design requirements for the facility.
    This amendment request does not remove or modify any of the 
design requirements for the facility or affect any accident 
initiators, conditions or assumptions for any accident previously 
evaluated. Therefore, this change does not involve a significant 
increase in the probability of any accident previously evaluated.
    This amendment request is administrative in nature and does not 
affect any system or component functional requirements. This change 
does not affect the operation of the plant or affect any component 
that is used to mitigate the consequences of any accident. 
Therefore, this change does not involve a significant increase in 
the consequences of any accident previously evaluated.
    Therefore, this change does not involve a significant increase 
in the probability or consequences of any accident previously 
evaluated.
    2. Does Not Create the Possibility of a New or Different Kind of 
Accident from any Previously Evaluated.
    The relocation of existing requirements from the technical 
specifications to other licensee controlled documents and the 
reformatting of the design features section of the technical 
specifications to the NUREG-1432 format are changes that are 
administrative in nature. This change does not modify or remove any 
plant design requirement. The proposed change will not affect any 
plant system or structure, nor will it affect any system functional 
or operability requirements. Consequently, no new failure modes are 
introduced as a result of this change. Therefore, this change does 
not create the possibility of a new or different kind of accident 
from any previously evaluated.
    3. Does Not Involve a Significant Reduction in the Margin of 
Safety.
    The proposed amendment request represents a relocation of a 
portion of the information previously located in the technical 
specification design features section to other licensee controlled 
documents that are controlled under 10 CFR 50.59. The proposed 
change is administrative in nature because the design requirements 
for the facility remain the same. The proposed change does not 
represent a change in the configuration or operation of the plant. 
Therefore, this change does not involve a significant reduction in 
the margin of safety.
    Therefore, based upon the reasoning presented above and the 
previous discussion of the amendment request, Entergy Operations has 
determined that the requested change does not involve a significant 
hazards consideration.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: Tomlinson Library, Arkansas 
Tech University, Russellville, AR 72801
    Attorney for licensee: Nicholas S. Reynolds, Esquire, Winston and 
Strawn, 1400 L Street, NW., Washington, DC 20005-3502
    NRC Project Director: William D. Beckner

Niagara Mohawk Power Corporation, Docket No. 50-220, Nine Mile 
Point Nuclear Station Unit No. 1, Oswego County, New York

    Date of amendment request: July 16, 1996
    Description of amendment request: The amendment would change the 
Technical Specifications to permit the use of 10 CFR Part 50, Appendix 
J, Option B, Performance-Based Containment Leakage Rate Testing in 
accordance with the implementation guidance in NRC's Regulatory Guide 
(RG) 1.163 dated September 1995.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:
    The operation of Nine Mile Point Unit 1, in accordance with the 
proposed amendment, will not involve a significant increase in the 
probability or consequences of an accident previously evaluated.
    NMP1 [Nine Mile Point, Unit 1] is currently implementing Option 
A of Appendix J of 10 CFR 50 for Type, A, B, and C testing. The 
proposed change to the Technical Specifications and the Bases would 
implement Option B to Appendix J of 10 CFR 50 at NMP1 for Type A, B, 
and C testing. Option B would allow increased testing intervals 
after satisfying certain performance based criteria.
    Appendix J describes the requirements for leakage of the primary 
containment and its components penetrating the primary containment. 
The leakage testing interval of the primary containment and its 
components is not a precursor or initiator to an accident. The 
primary containment and its penetrations minimizes the leakage of 
radioactivity into the environment during an accident which 
pressurizes the primary containment.
    Therefore, the proposed change does not involve a significant 
increase in the probability of an accident previously evaluated.

[[Page 52966]]

    The proposed change to the Technical Specifications and the 
Bases would replace the detailed and prescriptive technical 
requirements contained in Option A of Appendix J with performance 
based requirements in accordance with supporting regulatory/industry 
documents referenced in Option B of Appendix J. This proposed change 
includes a description of the 10 CFR 50 Appendix J Testing Program 
Plan in Section 6.16 of the Technical Specifications.
    This program plan, with two exceptions, is consistent with RG 
1.163. Therefore, this program plan establishes leakage-rate test 
methods, procedures, acceptance criteria and analyses which comply 
with Option B of Appendix J to 10 CFR 50.
    The implementation of this program continues to provide adequate 
assurance that during a DBA-LOCA [design-basis accident/loss-of-
coolant accident], the primary containment and its components will 
continue to limit leakage rates to less than the allowable leakage 
rates described in the Technical Specifications and thereby limit 
leakage consistent with the assumptions of the accident analyses. 
Therefore, the increased test intervals permitted by Option B for 
the primary containment and its penetrations will continue to 
implement the safety objectives underlying the requirements of 
Appendix J.
    As discussed below relative to the margin of safety, the impact 
of the proposed change on the consequences of a release is 
negligible. The slight increase in the risk to the population is 
compensated by the corresponding risk reduction benefits associated 
with the reduction in component cycling, stress, and wear associated 
with increased test intervals.
    Accordingly, operation with the proposed change to the Technical 
Specifications and the Bases will not significantly increase the 
consequences of an accident previously evaluated.
    The operation of Nine Mile Point Unit 1, in accordance with the 
proposed amendment, will not create the possibility of a new or 
different kind of accident from any accident previously evaluated.
    The proposed change would implement Option B of Appendix J to 10 
CFR 50 for Type A, B, and C testing. Option B would allow increased 
testing intervals after satisfying certain performance based 
criteria.
    No new plant operating modes, system operating configurations 
nor failure modes are introduced by the proposed change. The primary 
containment and its penetrations will continue to perform their 
accident mitigating function.
    Accordingly, operation with the proposed change will not create 
the possibility of a new or different kind of accident from any 
accident previously evaluated.
    The operation of Nine Mile Point Unit 1, in accordance with the 
proposed amendment, will not involve a significant reduction in a 
margin of safety.
    A regulatory impact analysis of implementing performance-based 
requirements indicates that relaxing the frequency of Type A, B, and 
C testing leads to an increase in overall risk of approximately two 
percent. As indicated in the Staff's Regulatory Impact Analysis, 
this increase is considered to be marginal to safety.
    As indicated above, increasing test intervals can slightly 
increase the risk to the population associated with the consequences 
of a release; however, this is compensated by the corresponding risk 
reduction benefits associated with the reduction in component 
cycling, stress, and wear associated with increased test intervals. 
Therefore, when considering the total integrated risk, the risk 
associated with increased test intervals is negligible.
    The proposed change is consistent with current plant safety 
analyses. In addition, the proposed change does not require 
revisions to the design of NMP1. As such, the proposed TS changes 
will maintain the same level of reliability of the equipment 
associated with containment integrity, assumed to operate in the 
plant safety analysis, or provide continued assurance that specified 
parameters affecting leak rate integrity, will remain within their 
acceptance limits.
    The as-left leakage after performing a required leakage test 
continues to be less than 0.60 La for combined Type B and C leakage 
and less than or equal to 0.75 La for Type A leakage. These as-left 
acceptance criteria and the testing frequency as established by the 
10 CFR 50 Appendix J Testing Program Plan provide assurance that the 
measured leakage rate will not exceed the maximum allowable leakage 
of La during plant operation.
    Visual examination of accessible interior and exterior surfaces 
of the primary containment continues to be performed prior to 
initiating a Type A test. The total number of visual examinations 
performed will continue to be three times during a 10-year period. 
Therefore, visual examinations of the primary containment will 
continue to allow for the timely uncovering of evidence of 
structural deterioration and satisfy the requirements of RG 1.163.
    The leakage limits of LCO 3.3.3 will continue to be met prior to 
reactor coolant system temperature exceeding 215*F and anytime 
Primary Containment is required. Satisfying these leakage limits 
provides assurance that the measured leakage rate will not exceed 
the maximum allowable leakage rate of La during plant operation. 
Therefore, operation with the proposed change 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 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: Reference and Documents 
Department, Penfield Library, State University of New York, Oswego, New 
York 13126.
    Attorney for licensee: Mark J. Wetterhahn, Esquire, Winston & 
Strawn, 1400 L Street, NW., Washington, DC 20005-3502.
    NRC Project Director: Guy S. Vissing, Acting Director

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

    Date of amendment requests: May 31, 1996
    Description of amendment requests: The proposed amendments would 
revise the combined Technical Specifications (TS) for the Diablo Canyon 
Nuclear Power Plant (DCPP), Unit Nos. 1 and 2 to revise 23 TS 
surveillance frequencies from at least once every 18 months to at least 
once per refueling interval (nominally 24 months) and to make 
administrative changes for 6 other TS to maintain consistency for TS 
that are not proposed for surveillance extension. The specific TS 
changes proposed include those for 2 response time tests, 3 containment 
spray system tests, and 24 ventilation system tests.
    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 six administrative changes regarding laboratory carbon 
testing are administrative changes only and do not affect the 
probability or consequences of accidents.
    The 23 proposed TS surveillance interval increases from 18 to 24 
months do not alter the intent or method by which the inspections, 
tests, or verifications are conducted, do not alter the way any 
structure, system, or component functions, and do not change the 
manner in which the plant is operated. The surveillance, 
maintenance, and operating histories indicate that the equipment 
will continue to perform satisfactorily with longer surveillance 
intervals. No recurring surveillance or maintenance problems were 
identified for response time, containment spray system, or control 
room ventilation system testing.
    Recurring maintenance issues on the fuel handling building and 
auxiliary building ventilation systems regarding the system control 
panels and certain dampers have been addressed. These ventilation 
systems are in service during all modes of operation and experience 
normal wear. None of the problems are related to refueling frequency 
testing. The monthly surveillance tests provide assurance of system 
operability for the control panels. The preventative maintenance 
program for the dampers is independent of refueling shutdowns and 
provides assurance that degradation mechanisms such as corrosion and 
wear are adequately addressed.

[[Page 52967]]

    There are no known mechanisms that would significantly degrade 
the performance of the evaluated equipment during normal plant 
operation. All potential time-related degradation mechanisms have 
insignificant effects in the timeframe of interest (maximum of 30 
months). Based on the past performance of the equipment, the 
probability or consequences of accidents would not be significantly 
affected by the proposed surveillance interval increases.
    Therefore, the proposed changes do not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    2. The proposed change does not create the possibility of a new 
or different kind of accident from any accident previously 
evaluated.
    The six administrative changes regarding laboratory carbon 
testing are administrative changes only and do not affect the type 
of accidents possible.
    The containment spray system and control room, auxiliary 
building, and fuel handling building ventilation systems are not 
associated with the initiation of any accident. The reactor trip and 
engineered safety feature actuation system response times are 
assumed in the accident analysis. However, the proposed surveillance 
interval increases would not affect the type of accidents possible.
    For the 23 proposed TS changes involving surveillance interval 
increases from 18- to 24-months, the surveillance and maintenance 
histories indicate that the equipment will continue to effectively 
perform their respective design functions over the longer operating 
cycles. Additionally, the increased surveillance intervals do not 
result in any physical modifications, affect safety function 
performance or the manner in which the plant is operated, or alter 
the intent or method by which surveillance tests are performed. Only 
a few problems have been identified and generally have not recurred. 
All potential time-related degradations have insignificant effects 
in the timeframe of interest. Therefore, the proposed changes do not 
create the possibility of a new or different kind of accident from 
any accident previously evaluated.
    3. The proposed change does not involve a significant reduction 
in a margin of safety.
    The six administrative changes regarding laboratory carbon 
testing are administrative changes only and do not affect the margin 
of safety.
    For the 23 proposed TS changes involving 18- to 24-month 
surveillance interval increases, evaluation of historical 
surveillance and maintenance data indicates there have been only a 
few problems experienced with the evaluated equipment. There are no 
indications that potential problems would be cycle-length dependent 
or that potential degradation would be significant for the timeframe 
of interest; therefore, increasing the surveillance interval will 
have little, if any, impact on any margin of safety. There is no 
safety analysis impact since these changes will have no effect on 
any safety limit, protection system setpoint, or limiting condition 
for operation, and there are no hardware changes that would impact 
existing safety analysis acceptance criteria. Safety margins would 
not be significantly affected by the proposed surveillance interval 
increases.
    Therefore, the proposed changes do not involve a significant 
reduction in a margin of safety.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment requests involve no significant hazards consideration.
    Local Public Document Room location: California Polytechnic State 
University, Robert E. Kennedy Library, Government Documents and Maps 
Department, San Luis Obispo, California 93407
    Attorney for licensee: Christopher J. Warner, Esq., Pacific Gas and 
Electric Company, P.O. Box 7442, San Francisco, California 94120
    NRC Project Director: William H. Bateman

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

    Date of amendment request: September 13, 1996
    Description of amendment request: The proposed amendment changes 
the Administrative Controls Section 5.6.6 of the Ginna Station 
Technical Specifications which would allow referencing of Revision of 
the Ginna Station Pressure and Temperature Limits Report (PTLR) for the 
Reactor Coolant System (RCS) pressure and temperature (P/T) limits and 
low temperature overpressure protection (LTOP) limits.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:
    1. Operation of Ginna Station in accordance with the proposed 
changes does not involve a significant increase in the probability 
or consequences of an accident previously evaluated. The proposed 
changes only revise the reference to the PTLR in the Administrative 
Controls section of technical specifications and correct a 
typographical error. The changes complete implementation of Generic 
Letter 96-03 by referencing NRC approved methodology within the 
Administrative Controls. As such, these changes are administrative 
in nature and do not impact initiators or analyzed events or assumed 
mitigation of accident or transient events. Therefore, these changes 
do not involve a significant increase in the probability or 
consequences of an accident previously analyzed.
    2. Operation of Ginna Station in accordance with the proposed 
changes does not create the possibility of a new or different kind 
of accident from any accident previously evaluated. The proposed 
changes do not involve a physical alteration of the plant (i.e., no 
new or different type of equipment will be installed) or changes in 
the methods governing normal plant operation. The proposed changes 
will not impose any new or different requirements. Thus, this change 
does not create the possibility of a new or different kind of 
accident from any accident previously evaluated.
    3. Operation of Ginna Station in accordance with the proposed 
changes does not involve a significant reduction in a margin of 
safety. The proposed changes will not reduce a margin of plant 
safety because the changes have been shown to ensure that the P/T 
and LTOP limits in the revised PTLR continue to meet all necessary 
requirements for reactor vessel integrity. These changes are 
administrative in nature. As such, no question of safety is 
involved, and 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: Rochester Public Library, 115 
South Avenue, Rochester, New York 14610
    Attorney for licensee: Nicholas S. Reynolds, Winston & Strawn, 1400 
L Street, NW., Washington, DC 20005
    NRC Project Director: Alexander W. Dromerick, Acting Director

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

    Date of amendment request: August 21, 1996 (TS 96-03)
    Description of amendment request: The proposed change would result 
in an amendment to Licenses DPR-77 and DPR-79 to change the Technical 
Specifications (TS) for the Sequoyah Nuclear Plant, Units 1 and 2. The 
proposed change would revise TS 3.7.1.3, ``Condensate Storage Tank,'' 
to include a new mode of applicability that reads: ``Mode 4 when steam 
generator is relied upon for heat removal.'' In addition, other 
proposed changes to TS 3.7.1.3 and a Bases change to Bases Sections 3/
4.3.1 and 3/4.3.2, ``Protective and Engineered Safety Features (ESF) 
Instrumentation,'' are included to provide improvements and establish 
requirements that are consistent with Westinghouse Standard Technical

[[Page 52968]]

Specifications (NUREG-1431, Revision 1).
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:
    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 revises SQN's condensate storage tank 
(CST) Specification 3.7.1.3 to incorporate requirements from the 
Westinghouse Standard Technical Specification (STS) contained in 
NUREG-1431, Revision 1. The proposed change is consistent with the 
STS for ensuring that SQN's CST remains operable in Modes 1, 2, 3, 
and Mode 4 when steam generator (SG) is relied upon for heat 
removal. In addition, the proposed change provides a general TS 
improvement by incorporating STS phraseology within the action 
requirements. Included with this change is an increase in the 
completion time for achieving hot shutdown. The current completion 
time of 6 hours is increased to 12 hours. This change allows 
sufficient time, while in Mode 4, to transition from SGs to residual 
heat removal entry conditions. The 12-hour completion time is 
reasonable based on operating experience to reach the required plant 
condition in an orderly manner without challenging plant systems.
    [The Tennessee Valley Authority's] TVA's proposed change also 
includes deletion of Surveillance Requirement (SR) 4.7.1.3.2. This 
SR demonstrates operability of SQN's essential raw cooling water 
(ERCW) system every 12 hours whenever the ERCW system is used as a 
supply source for the auxiliary feedwater (AFW) system. Deletion of 
this SR is consistent with STS requirements and is justified based 
on: (1) current SQN TS 3.7.4 requirements ensure operability of 
SQN's ERCW in Modes 1, 2, 3, and 4, and (2) newly proposed Action 
(b) requirements ensure that SQN's ERCW system is verified operable 
every 12 hours whenever the CST is inoperable.
    The proposed changes provide TS requirements for SQN's CST that 
are conservative with respect to assumptions used in SQN's accident 
analysis as contained in the Final Safety Analysis Report (FSAR). 
This change does not involve a physical modification to the plant or 
affect any instrumentation setpoints. 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.
    The proposed changes provide TS requirements for SQN's CST that 
are conservative with respect to assumptions used in SQN's accident 
analysis as contained in the FSAR. No new event initiator has been 
created, nor has any hardware been changed. This change does not 
involve a physical change to SQN's CST or any other system. 
Therefore, the proposed change 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.
    TVA's proposed change replaces SQN's CST TS requirements with TS 
requirements from the Westinghouse STS (NUREG-1431, Revision 1). The 
proposed change to SQN's CST TS to add ``Mode 4 when steam generator 
is relied upon for heat removal,'' provides consistency with the 
mode requirements of SQN's AFW TS and resolves a disparity that 
currently exists between these TSs. The allowed outage time for an 
inoperable CST remains unchanged and is consistent with the allowed 
outage time in STS. The proposed change to delete a SR for verifying 
operability of the ERCW system is considered acceptable based on 
other existing TSs that verify operability of SQN's ERCW system. 
Overall, similarity exists between SQN's current CST specification 
and the STS version. Consequently, with the exception of format, the 
TS requirements remain essentially unchanged.
    The proposed changes provide a line-item improvement for SQN's 
CST TS that are conservative with respect to the assumptions used in 
SQN's accident analysis as contained in the FSAR. This changes does 
not involve a setpoint change or physical modification to the plant. 
Accordingly, the margin of safety has not been reduced.
    The NRC has reviewed the licensee's analysis and, based on 
thisreview, 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

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

    Date of amendment request: August 21, 1996 (TS 96-06)
    Description of amendment request: The proposed change would result 
in an amendment to Licenses DPR-77 and DPR-79 to change the Technical 
Specifications for the Sequoyah Nuclear Plant (SQN), Units 1 and 2. The 
proposed change would remove Surveillance Requirement 4.8.1.1.1.b that 
verifies the ability to transfer the unit power supply from the unit 
generator supported circuit to the preferred power circuit. The current 
SQN design and operating configurations do not require the use of this 
transfer feature making this surveillance unnecessary.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:
    Operation of 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 will delete a surveillance associated with a 
feature that does not provide a safety function based on the current 
SQN design and operating procedures. The transfer feature that is 
currently verified will either be achieved by the system alignment 
or covered by the applicable TS action requirements. This transfer 
feature provided automatic system alignment to preferred offsite 
power circuits for accident mitigation purposes. This feature, or 
the lack of, is not considered a source of any accident and the 
proposed change to delete the associated surveillance will not 
increase the possibility of an accident previously evaluated. Safety 
functions are maintained by the current offsite circuit alignment 
without the transfer feature or associated surveillance. Therefore, 
the consequences of an accident can not be increased and may be 
reduced by eliminating the use of active devices to satisfy safety 
functions.
    2. Create the possibility of a new or different kind of accident 
from any previously analyzed. The offsite circuit transfer feature 
is not considered to be a source of an accident and the deletion of 
a surveillance to verify the operability of this transfer will not 
impact this potential. Therefore, the deletion of Surveillance 
4.8.1.1.1.b will not create the possibility of a new or different 
kind of accident previously analyzed.
    3. Involve a significant reduction in a margin of safety.
    The current SQN alignment satisfies all required offsite circuit 
alignments necessary to support accident mitigation functions 
without the use of active devices. In addition, any time delays 
associated with the transfer actuation to realign the offsite 
circuits, that is tested by the surveillance proposed to be deleted, 
are eliminated by the current alignment. Therefore, the margin of 
safety associated with the affected safety function is not reduced 
and may be increased by the elimination of active devices and their 
associated time delays.
    The NRC has reviewed the licensee's analysis and, based on 
thisreview, 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

[[Page 52969]]

    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: August 22, 1996 (TS 96-08)
    Description of amendment request: The proposed change would result 
in an amendment to Licenses DPR-77 and DPR-79 to change the Technical 
Specifications (TSs) for the Sequoyah Nuclear Plant, Units 1 and 2. The 
proposed changes would delete TS Table 4.8-1, ``Diesel Generator 
Reliability,'' and revise TS Section 3.8.1 to allow a once per 18 
month, 7-day allowed outage time (AOT) for the emergency diesel 
generators (EDGs). The first change would remove the accelerated 
testing requirements for the EDGs in accordance with Generic Letter 96-
01, ``Removal of Accelerated Testing and Special Reporting Requirements 
for Emergency Diesel Generators from Technical Specifications.'' The 
second change would revise the Units 1 and 2 TS to allow a once per 18 
month 7-day AOT for planned maintenance activities, particularly an 
upcoming major 12-year overhaul of all four EDGs.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:
    Operation of 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.
    Deletion of Table 4.8-1, in accordance with Generic Letter (GL) 
94-01, is an administrative change that will not impact the plant 
design or operation. None of the assumptions used in evaluating the 
radiological consequences of an accident are changed. A new or 
altered release path is not created. Therefore, this change does not 
involve an increase in the probability of any accident previously 
evaluated.
    The emergency diesel generators (EDGs) supply backup power to 
the essential safety systems in the event of a loss-of-offsite 
(normal) power. The EDGs cannot initiate an accident. The requested 
change will not impact the plant design or operation. The increased 
out of service time does not invalidate assumptions used in 
evaluating the radiological consequences of an accident and does not 
provide a new or altered release path. Therefore, this change does 
not involve an increase in the probability of any accident 
previously evaluated.
    An increase in the allowed outage time (AOT) would not change 
the conditions, operating configuration, or minimum amount of 
operable equipment assumed in the plant Final Safety Analysis Report 
for accident mitigation. The longer AOT would provide a longer time 
window for maintenance, but would lessen the total EDG 
unavailability per year. Based on the small increase in plant risk 
during maintenance, and the decrease in overall plant risk as a 
result of this change, this change will not result in a significant 
increase in the consequences of an accident.
    2. Create the possibility of a new or different kind of accident 
from any previously analyzed.
    Deletion of Table 4.8-1, in accordance with GL 94-01, is an 
administrative change that will not impact plant the plant design or 
operation. Appropriate testing, in accordance with the Maintenance 
Rule, will continue. Therefore, this change does not create the 
possibility of a new or different kind of accident from any 
previously analyzed.
    The proposed change to extend the AOT for the EDGs does not 
alter the physical design, or configuration of the plant. The EDG 
operation remains unchanged, therefore, this change does 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.
    Deletion of Table 4.8-1, in accordance with GL 94-01, ensures 
that the requirements and provisions of 10 CFR 50.65 and the 
guidance of Regulatory Guide 1.160 are met. The program put in place 
by these documents will ensure that any degradation of the EDGs is 
identified and appropriate action is taken. Therefore, this change 
does not involve a significant reduction in the margin of safety.
    A change to the maintenance schedule was performed to conform 
with vendor recommendations. This change in schedule required an 
increase in the duration of the 18 month and longer maintenance 
activities. Due to the number of shared systems, three of the four 
EDGs are required to meet all of the safety functions for each unit. 
However, the TSs conservatively assume four EDGs are necessary for 
unit operation; therefore, loss of any one EDG causes entry into a 
LCO action statement on both units. Performing the required 
maintenance with a 72-hour AOT will result in more EDG 
unavailability per year than would be required if the AOT was 7 
days. Therefore, the 7-day AOT would not result in a significant 
reduction in the margin of safety.
    The NRC has reviewed the licensee's analysis and, based on 
thisreview, 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: August 28, 1996 (TS 96-07)
    Description of amendment request: The proposed change would result 
in an amendment to Licenses DPR-77 and DPR-79 to change the Technical 
Specifications (TS) for the Sequoyah Nuclear Plant, Units 1 and 2. The 
proposed change would revise the setpoint tolerance for the pressurizer 
safety valves (PSVs) and main steam safety valves (MSSVs) from plus or 
minus one percent to plus or minus three percent. An analysis performed 
by Framatome Technologies Incorporated to support this change is 
provided in the licensee's submittal. These parameters are contained in 
TS 3.4.2, TS 3.4.3.1, and Table 3.7-2. Additionally, the sentence 
``Following testing, lift settings shall be within plus or minus 1%.'' 
would be added to Surveillance Requirements (SR) 4.4.2, 4.4.3.1, and 
4.7.1.1.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:
    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 evaluation contained in Enclosure 5 [Framatome Report No. 
77-1257369-01, ``Safety Evaluation of Safety Valve Setpoint 
Tolerance Relaxation,'' dated August 1, 1996] discusses the 
consequences of this change as it pertains to the accidents 
previously analyzed. The positive increase of the setpoint 
tolerance, from one percent to three percent, of the PSV[s] and the 
MSSV[s] does not challenge the design limits of the installed 
systems. This conclusion is demonstrated by means of the reanalysis 
of the bounding overpressure events. The negative tolerance for the 
MSSVs, from minus one percent to minus three percent, will result in 
an increase in mass discharged through these valves. The increase 
was evaluated and the analysis indicated that the dose release 
remained within the limits required by 10 CFR 100. Based on the 
results of this review, there is no increase in the probability of a 
previously evaluated accident or a significant increase in the 
consequences of an accident previously evaluated.
    2. Create the possibility of a new or different kind of accident 
from any previously analyzed.
    The proposed TS change will increase the setpoint tolerance for 
the PSVs and the MSSVs. This change does not involve an

[[Page 52970]]

equipment addition or change in the method the plant is operated. 
Therefore, the possibility of a new or different kind of accident 
from any previously analyzed is not created.
    3. Involve a significant reduction in a margin of safety.
    The proposed change is a change in the lift setpoint tolerance 
of the existing valves. The analysis demonstrates that the design 
limits are not exceeded nor are the dose release limits exceeded due 
to this increase in setpoint tolerance. Additionally, the valves 
will be returned to the original tolerance of plus or minus one 
percent. This will ensure that the maximum margin is retained; 
therefore, the margin of safety has not been reduced by this change.
    The NRC has reviewed the licensee's analysis and, based on 
thisreview, 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

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

    Date of amendment request: August 7, 1996
    Description of amendment request: The proposed amendment would 
revise Technical Specification (TS) 1.0, ``Definitions,'' by defining a 
refueling interval to be [less than or equal to] 730 days; and would 
revise TS 3/4.0, ``Applicability,'' TS 3/4.6.2.1, ``Containment Systems 
- Depressurization and Cooling Systems - Containment Spray System,'' 
and TS 3/4.6.3.1, ``Containment Systems - Containment Isolation 
Valves,'' to reflect performing surveillance tests during a refueling 
interval rather than every 18 months.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensees have 
provided their analysis of the issue of no significant hazards 
consideration, which is presented below:
    Toledo Edison has reviewed the proposed changes and determined 
that a significant hazards consideration does not exist because 
operation of the Davis-Besse Nuclear Power Station (DBNPS), Unit No. 
1, in accordance with these changes would:
    1a. Not involve a significant increase in the probability of an 
accident previously evaluated because no such accidents are affected 
by the proposed revisions to increase the surveillance test 
intervals from 18 to 24 months for the containment spray system 
(Surveillance Requirements 4.6.2.1.b), or the containment isolation 
valves (Surveillance Requirements 4.6.3.1.2). The proposed change to 
TS 1.0, adding a definition for ``Refueling Interval,'' and the 
associated proposed change to TS Bases 3/4.0, are administrative 
changes associated with the 24 month cycle conversion. Initiating 
conditions and assumptions remain as previously analyzed for 
accidents in the DBNPS Updated Safety Analysis Report.
    These revisions do not involve any physical changes to systems 
or components, nor do they alter the typical manner in which the 
systems or components are operated.
    A review of historical 18 month surveillance data and 
maintenance records support an increase in the surveillance test 
intervals from 18 to 24 months (and up to 30 months on a non-routine 
basis) because no potential for a significant increase in a failure 
rate of an affected system or component was identified during these 
reviews.
    These proposed revisions are consistent with the NRC guidance on 
evaluating and proposing such revisions as provided in Generic 
Letter 91-04, ``Changes in Technical Specification Surveillance 
Intervals to Accommodate a 24-Month Fuel Cycle,'' dated April 2, 
1991.
    1b. Not involve a significant increase in the consequences of an 
accident previously evaluated because the source term, containment 
isolation or radiological releases are not being changed by these 
proposed revisions. Existing system and component redundancy is not 
being changed by these proposed changes. Existing system and 
component operation is not being changed by these proposed changes. 
The assumptions used in evaluating the radiological consequences in 
the DBNPS Updated Safety Analysis Report are not invalidated.
    A review of historical 18 month surveillance data and 
maintenance records support an increase in the surveillance test 
intervals from 18 to 24 months (and up to 30 months on a non-routine 
basis) because no potential for a significant increase in a failure 
rate of an affected system or component was identified during these 
reviews.
    2. Not create the possibility of a new or different kind of 
accident from any accident previously evaluated because these 
revisions do not involve any physical changes to systems or 
components, nor do they alter the typical manner in which the 
systems or components are operated. A review of historical 18 month 
surveillance data and maintenance records support an increase in the 
surveillance test intervals from 18 to 24 months (and up to 30 
months on a non-routine basis) because no potential for a 
significant increase in a failure rate of a system or component was 
identified during these reviews. No changes are being proposed to 
the type of testing currently being performed, only to the length of 
the surveillance test interval.
    3. Not involve a significant reduction in a margin of safety 
because a review of the historical 18 month surveillance data and 
maintenance records identified no potential for a significant 
increase in a failure rate of a system or component due to 
increasing the surveillance test interval to 24 months. Existing 
system and component redundancy is not being changed by these 
proposed changes.
    There are no new or significant changes to the initial 
conditions contributing to accident severity or consequences. 
Therefore, there are no significant reductions in a margin of 
safety.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: University of Toledo, William 
Carlson Library, Government Documents Collection, 2801 West Bancroft 
Avenue, Toledo, Ohio 43606
    Attorney for licensee: Jay E. Silberg, Esquire, Shaw, Pittman, 
Potts and Trowbridge, 2300 N Street, NW., Washington, DC 20037
    NRC Project Director: Gail H. Marcus
    Toledo Edison Company, Centerior Service Company, and The 
Cleveland Electric Illuminating Company, Docket No. 50-346, Davis-
Besse Nuclear Power Station, Unit No. 1, Ottawa County, Ohio
    Date of amendment request: September 4, 1996
    Description of amendment request: The proposed amendment would 
revise Technical Specification (TS) 6.2.3, ``Facility Staff Overtime,'' 
by removing specific overtime limits and working hours and by adding 
procedural controls to perform a monthly review of overtime 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:
    Toledo Edison has reviewed the proposed changes and determined 
that a significant hazards consideration does not exist because 
operation of the Davis-Besse Nuclear Power Station (DBNPS), Unit 
Number 1, in accordance with these changes would:1a. Not involve a 
significant increase in the probability of an accident previously 
evaluated because no change is being made to any accident initiator. 
No previously analyzed accident scenario is changed, and initiating 
conditions and assumptions remain as previously analyzed.
    The proposed change to TS 6.2.3, ``Facility Staff Overtime,'' to 
relocate specific overtime limits and working hours to the DBNPS 
Updated Safety Analysis Report (USAR) is consistent with the NRC 
Staff's determination previously provided on a

[[Page 52971]]

generic basis in the Safety Evaluation to License Amendment Number 
127 and 116 to the Operating Licenses (Number NPF-10 and NPF-15), 
for the San Onofre Generating Nuclear Station, Units 2 and 3, dated 
February 9, 1996. The appropriate relocation of TS requirements, 
such as portions of TS 6.2.3, to licensee-controlled documents is 
also addressed generically in the NRC's ``Final Policy Statement on 
Technical Specification Improvements for Nuclear Power Reactors'', 
dated July 23, 1993.
    The relocated overtime limits and working hours will be subject 
to review and evaluation under Section 50.59, ``Changes, Tests, and 
Experiments'', of Title 10 of the Code of Federal Regulations (10 
CFR) prior to any changes being made. The other changes to TS 6.2.3 
are editorial, with an exception being that a new requirement has 
been added for plant procedures to ensure that an individual's 
overtime is reviewed monthly by the Plant Manager or his designee(s) 
to ensure excessive hours have not been assigned.
    Overtime will remain controlled by plant administrative 
procedures and USAR requirements generally following the guidance of 
the NRC's Policy Statement on working hours contained within Generic 
Letter 82-12, ``Nuclear Power Plant Staff Working Hours.''
    1b. Not involve a significant increase in the consequences of an 
accident previously evaluated because the proposed changes do not 
affect accident conditions or assumptions used in evaluating the 
radiological consequences of an accident. The proposed changes do 
not alter the source term, containment isolation or allowable 
radiological releases.
    The proposed changes to TS 6.2.3 only alter the administrative 
location of and the regulatory controls applicable to plant staff 
specific overtime limits and working hours. Therefore, there is no 
significant increase in the consequences of an accident previously 
evaluated.
    2. Not create the possibility of a new or different kind of 
accident from any accident previously evaluated because the proposed 
changes do not change the way the plant is operated, and no new or 
different failure modes have been defined for any plant system or 
component important to safety, nor has any limiting single failure 
been identified as a result of the proposed changes. No new or 
different types of failures or accident initiators are introduced by 
the proposed changes.
    The proposed changes to TS 6.2.3 only alter the administrative 
location of and the regulatory controls applicable to plant staff 
specific overtime limits and working hours. Therefore, there is no 
possibility created for a new or different kind of accident.
    3. Not involve a significant reduction in a margin of safety 
because facility staff overtime is not an input in the calculation 
of any safety margin with regard to TS Safety Limits, Limiting 
Safety System Settings, other TS Limiting Conditions for Operation 
or other previously defined margins for any structure, system, or 
component important to safety.
    The proposed changes to TS 6.2.3 only alter the administrative 
location of and the regulatory controls applicable to plant specific 
overtime limits and working hours. Therefore, there is no 
significant reduction in a margin of safety.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: University of Toledo, William 
Carlson Library, Government Documents Collection, 2801 West Bancroft 
Avenue, Toledo, Ohio 43606
    Attorney for licensee: Jay E. Silberg, Esquire, Shaw, Pittman, 
Potts and Trowbridge, 2300 N Street, NW., Washington, DC 20037
    NRC Project Director: Gail H. Marcus
    Toledo Edison Company, Centerior Service Company, and The 
Cleveland Electric Illuminating Company, Docket No. 50-346, Davis-
Besse Nuclear Power Station, Unit No. 1, Ottawa County, Ohio
    Date of amendment request: September 12, 1996
    Description of amendment request: The proposed amendment would 
revise Technical Specifications (TS) 3/4.1.3.4, ``Reactivity Control 
Systems - Rod Drop Time,'' and TS 3/4.5.2, ``Emergency Core Cooling 
Systems - Tavg [greater than or equal to] 280 deg.F,'' to change 
surveillance test intervals from every 18 months to each refueling 
interval (nominally 24 months). Additionally, the proposed amendment 
would remove a footnote for TS 4.5.2.b that is no longer applicable.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensees have 
provided their analysis of the issue of no significant hazards 
consideration, which is presented below:
    Toledo Edison has reviewed the proposed changes and determined 
that a significant hazards consideration does not exist because 
operation of the Davis-Besse Nuclear Power Station (DBNPS), Unit No. 
1, in accordance with these changes would:
    1a. Not involve a significant increase in the probability of an 
accident previously evaluated because no such accidents are affected 
by the proposed revisions to increase the surveillance test 
intervalsfrom 18 months to 24 months for the reactivity control 
systems (Surveillance Requirement 4.1.3.4.c), or the emergency core 
cooling systems (Surveillance Requirement 4.5.2.b), or the proposed 
administrative change to Surveillance Requirement 4.5.2.b to remove 
a time-conditional footnote which has expired. Initiating conditions 
and assumptions remain as previously analyzed for all accidents in 
the DBNPS Updated Safety Analysis Report.
    These revisions do not involve any physical changes to systems 
or components, nor do they alter the typical manner in which the 
systems or components are operated.
    A review of historical 18 month surveillance data and 
maintenance records support an increase in the surveillance test 
intervals from 18 months to 24 months (and up to 30 months on a non-
routine basis) because no potential for a significant increase in a 
failure rate of an affected system or component was identified 
during these reviews.
    These proposed revisions are consistent with the NRC guidance on 
evaluating and proposing such revisions as provided in Generic 
Letter 91-04, ``Changes in Technical Specification Surveillance 
Intervals to Accommodate a 24-month Fuel Cycle,'' dated April 2, 
1991.
    1b. Not involve a significant increase in the consequences of an 
accident previously evaluated because the source term, containment 
isolation or radiological releases are not being changed by these 
proposed revisions. Existing system and component redundancy is not 
being changed by these proposed changes. Existing system and 
component operation is not being changed by these proposed changes. 
The assumptions used in evaluating the radiological consequences in 
the DBNPS Updated Safety Analysis Report are not invalidated.
    A review of historical 18 month surveillance data and 
maintenance records support an increase in the surveillance test 
intervals from 18 to 24 months (and up to 30 months on a non-routine 
basis) because no potential for a significant increase in a failure 
rate of an affected system or component was identified during these 
reviews.
    2. Not create the possibility of a new or different kind of 
accident from any accident previously evaluated because these 
revisions do not involve any physical changes to systems or 
components, nor do they alter the typical manner in which the 
systems or components are operated. A review of historical 18 month 
surveillance data and maintenance records support an increase in the 
surveillance test intervals from 18 months to 24 months (and up to 
30 months on a non-routine basis) because no potential for a 
significant increase in a failure rate of a system or component was 
identified during these reviews. No changes are being proposed to 
the type of testing currently being performed, only to the length of 
the surveillance test interval.
    3. Not involve a significant reduction in a margin of safety 
because a review of the historical 18 month surveillance data and 
maintenance records identified no potential for a significant 
increase in a failure rate of a system or component due to 
increasing the surveillance test interval to 24 months. Existing 
system and component redundancy and operation is not being changed 
by these proposed changes.
    There are no new or significant changes to the initial 
conditions contributing to accident severity or consequences. 
Therefore, there are no significant reductions in a margin of 
safety.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are

[[Page 52972]]

satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: University of Toledo, William 
Carlson Library, Government Documents Collection, 2801 West Bancroft 
Avenue, Toledo, Ohio 43606
    Attorney for licensee: Jay E. Silberg, Esquire, Shaw, Pittman, 
Potts and Trowbridge, 2300 N Street, NW., Washington, DC 20037
    NRC Project Director: Gail H. Marcus
    Toledo Edison Company, Centerior Service Company, and The 
Cleveland Electric Illuminating Company, Docket No. 50-346, Davis-
Besse Nuclear Power Station, Unit No. 1, Ottawa County, Ohio
    Date of amendment request: September 17, 1996
    Description of amendment request: The proposed amendment would 
revise the surveillance interval from 18 months to less than or equal 
to 730 days, nominally 24 months, for Technical Specification (TS) 3/
4.5.2, ``Emergency Core Cooling Systems - ECCS Subsystems - Tavg 
greater than or equal to 280 degrees F;'' TS 3/4.6.5.1, ``Containment 
Systems - Shield Building - Emergency Ventilation System;'' TS 3/
4.7.6.1, ``Plant Systems - Control Room Emergency Ventilation System;'' 
TS 3/4.7.7, ``Plant Systems - Snubbers;'' TS 3/4.9.12, ``Refueling 
Operations - Storage Pool Ventilation;'' and TS Bases 3/4.7.7 - 
``Snubbers.''
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensees have 
provided their analysis of the issue of no significant hazards 
consideration, which is presented below:
    Toledo Edison has reviewed the proposed changes and determined 
that a significant hazards consideration does not exist because 
operation of the Davis-Besse Nuclear Power Station (DBNPS), Unit No. 
1 in accordance with these changes would:
    1a. Not involve a significant increase in the probability of an 
accident previously evaluated because no such accidents are affected 
by the proposed revisions to increase the surveillance test 
intervals from 18 months to 24 months for the trisodium phosphate 
dodecahydrate (TSP) volume, (Surveillance Requirement 4.5.2.d.4), 
Shield Building and Storage Pool Emergency Ventilation Systems, 
(Surveillance Requirements 4.6.5.1.b, 4.6.5.1.d, 4.9.12.1, and 
4.9.12.2), and the Control Room Emergency Ventilation System, 
(Surveillance Requirements 4.7.6.1.c and 4.7.6.1.e) and Snubbers 
(Surveillance Requirement 4.7.7.2.b and associated Bases 3/4.7.7). 
Initiating conditions and assumptions remain as previously analyzed 
for all accidents in the DBNPS Updated Safety Analysis Report.
    These revisions do not involve any physical changes to systems 
or components, nor do they alter the typical manner in which the 
systems or components are operated.
    A review of historical 18 month surveillance data and 
maintenance records support an increase in the surveillance test 
intervals from 18 to 24 months (and up to 30 months on a non-routine 
basis) because no potential for a significant increase in a failure 
rate of a system or component was identified during these reviews.
    These proposed revisions are consistent with the NRC guidance on 
evaluating and proposing such revisions as provided in Generic 
Letter 91-04, ``Changes in Technical Specification Surveillance 
Intervals to Accommodate a 24-Month Fuel Cycle,'' dated April 2, 
1991.
    1b. Not involve a significant increase in the consequences of an 
accident previously evaluated because the source term, containment 
isolation or radiological releases are not being changed by these 
proposed revisions. Existing system and component redundancy is not 
being changed by these proposed changes. Existing system and 
component operation is not being changed by these proposed changes. 
The assumptions used in evaluating the radiological consequences in 
the DBNPS Updated Safety Analysis Report are not invalidated.
    2. Not create the possibility of a new or different kind of 
accident from any accident previously evaluated because these 
revisions do not involve any physical changes to systems or 
components, nor do they alter the typical manner in which the 
systems or components are operated. A review of historical 18 month 
surveillance data and maintenance records support an increase in the 
surveillance test intervals from 18 to 24 months (and up to 30 
months on a non-routine basis) because no potential for a 
significant increase in a failure rate of a system or component was 
identified during these reviews. No changes are being proposed to 
the type of testing currently being performed, only to the length of 
the surveillance test interval.
    3. Not involve a significant reduction in a margin of safety 
because the review results of the historical 18 month surveillance 
data and maintenance records identified no potential for a 
significant increase in a failure rate of a system or component due 
to increasing the surveillance test interval to 24 months. Existing 
system and component redundancy and operation is not being changed 
by these proposed changes.
    There are no new or significant changes to the initial 
conditions contributing to accident severity or consequences. 
Therefore, there are no significant reductions in a margin of 
safety.
    The NRC staff has reviewed the licensees' analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: University of Toledo, William 
Carlson Library, Government Documents Collection, 2801 West Bancroft 
Avenue, Toledo, Ohio 43606
    Attorney for licensee: Jay E. Silberg, Esquire, Shaw, Pittman, 
Potts and Trowbridge, 2300 N Street, NW., Washington, DC 20037
    NRC Project Director: Gail H. Marcus

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

    Date of amendment request: August 22, 1996
    Description of amendment request: The proposed amendment would 
revise the license for each unit and the bases for Technical 
Specification (TS) Section 15.3.1, ``Reactor Coolant System.'' The 
licensed power level would be changed from 1518 to 1518.5 megawatts 
thermal to agree with other sections of the TSs.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration which is presented below:
    1. Operation of this facility under the proposed Technical 
Specifications will not create a significant increase in the 
probability or consequences of an accident previously evaluated.
    There is no physical change to the facilities as a result of the 
proposed license amendment and all Limiting Conditions for 
Operation, Limiting Safety System Settings and Safety Limits 
specified in the Technical Specifications remain unchanged. The 
proposed change is administrative only and restores consistency 
within the PBNP license and licensing basis. Therefore, this 
amendment will not cause a significant increase in the probability 
or consequences of an accident previously evaluated.
    2. Operation of this facility under the proposed Technical 
Specifications change will not create the possibility of a new or 
different kind of accident from any accident previously evaluated.
    The proposed amendment has no effect on the physical 
configuration of the facilities or the manner in which they operate. 
The design and design basis of the plants remains the same. The 
current plant safety analysis therefore remains complete and 
accurate in addressing the design basis events and in analyzing 
plant response and consequences for the facilities. The Limiting 
Conditions for Operation, Limiting Safety System Settings and Safety 
Limits specified in the Technical Specifications for the facilities 
are not affected by the proposed license amendment. The plant 
conditions for which the design basis accident analysis have been 
performed remain valid. Therefore, the proposed license amendment 
cannot create the possibility of a new or different kind of accident 
from any accident previously evaluated.
    3. Operation of this facility under the proposed Technical 
Specifications change will not create a significant reduction in a 
margin of safety.
    Plant safety margins are established through the Limiting 
Conditions for

[[Page 52973]]

Operation, Limiting Safety System Settings and Safety Limits 
specified in the Technical Specifications. Since there is no change 
to the physical design or operation of the plant, there is no change 
to any of these margins. Thus, the proposed license amendment does 
not involve a reduction in any margin of safety.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: Joseph P. Mann Library, 1516 
Sixteenth Street, Two Rivers, Wisconsin 54241
    Attorney for licensee: Gerald Charnoff, Esq., Shaw, Pittman, Potts, 
and Trowbridge, 2300 N Street, NW., Washington, DC 20037
    NRC Project Director: Gail H. Marcus

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.
    Consumers Power Company, Palisades Plant, Van Buren County, 
Michigan
    Date of amendment request: December 11, 1995, supplemented by 
letters dated January 18, 1996, and September 3, 1996Brief
    Description of amendment request: The proposed amendment would 
revise the Palisades Technical Specifications (TS) Administrative 
Controls section (Section 6) and other TS associated with the 
administrative controls section to adopt the format of NUREG-1432, 
``Standard Technical Specifications, Combustion Engineering Plants.'' 
The amendment would also revise certain other surveillance intervals 
and administrative requirements.Date of individual notice in the 
Federal Register: September 20, 1996 (61 FR 49493)
    Expiration date of individual notice: October 21, 1996
    Local Public Document Room location: Van Wylen Library, Hope 
College, Holland, Michigan 49423

Consumers Power Company, Palisades Plant, Van Buren County, 
Michigan

    Date of amendment request: August 14, 1996 (also refer to related 
application dated January 18, 1996)
    Brief Description of amendment request: The proposed amendment 
would revise the Palisades Technical Specifications (TS) to extend the 
surveillance interval frequency for the primary coolant pump (PCP) 
flywheels by one operating cycle. By letter dated January 18, 1996, the 
licensee previously submitted a request to amend the TS to delete the 
requirement to perform PCP flywheel inspections. NRC review of the 
original request will not be completed in time for the upcoming 
refueling outage scheduled for November 1996; therefore, the licensee 
has submitted this separate request to extend the surveillance 
frequency by one operating cycle.Date of individual notice in the 
Federal Register: September 24, 1996 (61 FR 50054
    Expiration date of individual notice: October 24, 1996
    Local Public Document Room location: Van Wylen Library, Hope 
College, Holland, Michigan 49423

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

    Date of amendment request: May 9, 1996, as supplemented by letter 
dated August 27, 1996
    Brief description of amendment request: The amendment would revise 
the Technical Specifications to allow the surveillance of the relief 
mode of operation of each of the 20 safety/relief valves without 
physically lifting the disk off the seat at power.Date of individual 
notice in the Federal Register: September 11, 1996 (61 FR 47971)
    Expiration date of individual notice: October 11, 1996
    Local Public Document Room location: Judge George W. Armstrong 
Library, 220 S. Commerce Street, Natchez, MS 39120

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

    Date of amendment request: August 27, 1996
    Description of amendment request: The proposed amendment would 
clarify the Technical Specifications limiting condition for operation 
and surveillance requirements for the charging pumps and high pressure 
safety injection pumps when the unit is shut down (Modes 5 and 6). Date 
of publication of individual notice in Federal Register: September 20, 
1996 (61 FR 49498)
    Expiration date of individual notice: October 21, 1996
    Local Public Document Room location: Learning Resources Center, 
Three Rivers Community-Technical College, 574 New London Turnpike, 
Norwich, Connecticut, and the Waterford Library, ATTN: Vince Juliano, 
49 Rope Ferry Road, Waterford, Connecticut

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

    Date of amendment request: August 27, 1996
    Brief description of amendment request: This amendment proposes to 
delete License Condition 2.C.(24)(a) for Unit 2 which required 
establishment by June 3, 1981, of regularly scheduled 8-hour shifts 
without reliance on routine use of overtime. The proposed amendment 
also modifies Technical Specification 6.2.2 for both units to 
incorporate limits on overtime.
    Date of publication of individual notice in Federal Register: 
September 12, 1996 (61 FR 48175)
    Expiration date of individual notice: October 15, 1996
    Local Public Document Room location: Salem Free Public Library, 112 
West Broadway, Salem, NJ 08079

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

[[Page 52974]]

published in the Federal Register as indicated.
    Unless otherwise indicated, the Commission has determined that 
these amendments satisfy the criteria for categorical exclusion in 
accordance with 10 CFR 51.22. Therefore, pursuant to 10 CFR 51.22(b), 
no environmental impact statement or environmental assessment need be 
prepared for these amendments. If the Commission has prepared an 
environmental assessment under the special circumstances provision in 
10 CFR 51.12(b) and has made a determination based on that assessment, 
it is so indicated.
    For further details with respect to the action see (1) the 
applications for amendment, (2) the amendment, and (3) the Commission's 
related letter, Safety Evaluation and/or Environmental Assessment as 
indicated. All of these items are available for public inspection at 
the Commission's Public Document Room, the Gelman Building, 2120 L 
Street, NW., Washington, DC, and at the local public document rooms for 
the particular facilities involved.

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

    Date of application for amendments: August 16, 1996
    Brief description of amendments: The amendments would defer the 
implementation date as stated in Amendment No. 150 for Dresden, Unit 2, 
and Amendment No. 145 for Dresden, Unit 3, until January 15, 1997.
    Date of issuance: September 26, 1996
    Effective date: Immediately, to be implemented on or before January 
15, 1997.
    Amendment Nos.: 151 and 146
    Facility Operating License Nos. DPR-19 and DPR-25: The amendments 
revised the implementation date.
    Date of initial notice in Federal Register: August 22, 1996 (61 FR 
43391) The Commission's related evaluation of the amendments is 
contained in a Safety Evaluation dated September 26, 1996. No 
significant hazards consideration comments received: No
    Local Public Document Room location: Morris Area Public Library 
District, 604 Liberty Street, Morris, Illinois 60450.

Commonwealth Edison Company, Docket Nos. 50-373 and 50-374, LaSalle 
County Station, Units 1 and 2, LaSalle County, Illinois

    Date of application for amendments: April 9, 1996
    Brief description of amendments: The amendments revise the 
Technical Specifications to eliminate the main steamline radiation 
monitoring system high radiation trip function for initiating an (1) 
automatic reactor scram, (2) automatic closure of the main steamline 
isolation valves, and (3) automatic closure of the reactor 
recirculation water sample line isolation valves and main steam line 
drain isolation valves.
    Date of issuance: September 20, 1996
    Effective date: Immediately, to be implemented within 90 days.
    Amendment Nos.: 115 and 100
    Facility Operating License Nos. NPF-11 and NPF-18: The amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: May 22, 1996 (61 FR 
25701) The Commission's related evaluation of the amendments is 
contained in a Safety Evaluation dated September 20, 1996.No 
significant hazards consideration comments received: No
    Local Public Document Room location: Jacobs Memorial Library, 
Illinois Valley Community College, Oglesby, Illinois 61348.

Consumers Power Company, Docket No. 50-255, Palisades Plant, Van 
Buren County, Michigan

    Date of application for amendment: February 6, 1996
    Brief description of amendment: The amendment deletes the 
requirement to perform alternate train testing of redundant components 
when emergency core cooling system and containment cooling system 
components are found to be inoperable or are to be removed from service 
for maintenance.
    Date of issuance: September 26, 1996
    Effective date: September 26, 1996
    Amendment No.: 172
    Facility Operating License No. DPR-20. Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: June 5, 1996 (61 FR 
28611) The Commission's related evaluation of the amendment is 
contained in a Safety Evaluation dated September 26, 1996.No 
significant hazards consideration comments received: No.
    Local Public Document Room location: Van Wylen Library, Hope 
College, Holland, Michigan 49423

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

    Date of application for amendments: November 15, 1995
    Brief description of amendments: The amendments revise the 
Technical Specifications to modify Section 3/4.7.5, ``Standby Nuclear 
Service Water Pond,'' for the Catawba Nuclear Station, Units 1 and 2, 
raising the minimum water level by 1 foot (from elevation 570 to 571 
feet).
    Date of issuance: September 20, 1996
    Effective date: As of the date of issuance to be implemented within 
30 days
    Amendment Nos.: 152 and 144
    Facility Operating License Nos. NPF-35 and NPF-52: Amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: December 20, 1995 (60 
FR 65676) The Commission's related evaluation of the amendments is 
contained in a Safety Evaluation dated September 20, 1996.No 
significant hazards consideration comments received: No
    Local Public Document Room location: York County Library, 138 East 
Black Street, Rock Hill, South Carolina 29730

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

    Date of application for amendment: April 29, 1996, as supplemented 
September 12, 1996
    Brief description of amendment: This amendment revises TS 5.3.1 to 
allow the use of ZIRLO as an alternate zirconium-based fuel rod 
material and removes the word ``clad'' to be consistent with the text 
of the NRC's improved Standard Technical Specifications (NUREG-1431). 
Limited substitution of fuel rods by ZIRLO filler rods is permitted. 
The proposed revision to Note 2 on TS Table 3.9-1 to specify that the 
maximum burnup in the peak fuel rod in a fuel assembly stored in Region 
2 spent fuel racks should not exceed the NRC-approved limit for WCAP-
12610 was withdrawn by letter dated September 12, 1996.
    Date of issuance: September 13, 1996
    Effective date: As of date of issuance and shall be implemented 
within 60 days.
    Amendment No: 82
    Facility Operating License No. NPF-73. Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: May 22, 1996 (61 FR 
25703) The Commission's related evaluation of the amendment is 
contained in a Safety Evaluation dated September 13, 1996.No 
significant hazards consideration comments received: No
    Local Public Document Room location: B. F. Jones Memorial Library, 
663 Franklin Avenue, Aliquippa, PA 15001

[[Page 52975]]

Entergy Operations, Inc., Docket No. 50-313, Arkansas Nuclear One, 
Unit No. 1, Pope County, Arkansas

    Date of amendment request: May 19, 1995, as supplemented by letters 
dated July 21, 1995, and June 10, September 10 and 13, 1996
    Brief description of amendment: The amendment revises the technical 
specifications to permit the reactor building personnel airlock doors 
to remain open during fuel handling.
    Date of issuance: September 20, 1996
    Effective date: September 20, 1996
    Amendment No.: 184
    Facility Operating License No. DPR-51. Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: August 2, 1995 (60 FR 
39437) The additional information contained in the supplemental letters 
dated July 21, 1995, and June 10, September 10 and 13, 1996, were 
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 
amendment is contained in a Safety Evaluation dated September 20, 
1996.No significant hazards consideration comments received: No.
    Local Public Document Room location: Tomlinson Library, Arkansas 
Tech University, Russellville, AR 72801

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

    Date of application for amendment: April 4, 1995, as supplemented 
April 25 and August 19, 1996
    Brief description of amendment: The amendment revised operating 
criteria and requirements associated with containment personnel air 
locks.
    Date of issuance: September 26, 1996
    Effective date: September 26, 1996
    Amendment No.: 175
    Facility Operating License No. NPF-6. Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: August 2, 1995 (60 FR 
39438) The Commission's related evaluation of the amendment is 
contained in a Safety Evaluation dated September 26, 1996.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: December 18, 1995, as 
supplemented on May 3, June 11, July 1, July 3, and August 22, 1996.
    Brief description of amendments: The amendments increase the 
authorized rated thermal power from 2200 Megawatt-thermal (MWt) to 2300 
MWt. The amendment also approves changes to the Technical 
Specifications to implement uprated power operation.
    Date of issuance: September 26, 1996
    Effective date: September 26, 1996, to be implemented within 120 
days
    Amendment Nos. 191 and 185Facility Operating Licenses Nos. DPR-31 
and DPR-41: Amendments revised the Technical Specifications.
    Date of initial notice in Federal Register: July 3, 1996 (61 FR 
34889) The initial Federal Register notice included information from 
the licensee's May 3 and May 11, 1996 supplemental letters. The July 1, 
July 3, and August 22, 1996 letters provided clarifying information 
that did not change the initial proposed no significant hazards 
consideration determination. The Commission's related evaluation of the 
amendments is contained in an Environmental Assessment dated September 
12, 1996 and in a Safety Evaluation dated September 26, 1996. No 
significant hazards consideration comments received: No
    Local Public Document Room location: Florida International 
University, University Park, Miami, Florida 33199.

Niagara Mohawk Power Corporation, Docket No. 50-410, Nine Mile 
Point Nuclear Station, Unit 2, Oswego County, New York

    Date of application for amendment: May 15, 1996
    Brief description of amendment: The amendment revises Technical 
Specification 3/4.3.2, ``Isolation Actuation Instrumentation,'' to 
establish a range of allowable values and trip setpoints for high 
temperatures in the Main Steam Line Tunnel Lead Enclosure.
    Date of issuance: September 17, 1996
    Effective date: As of the date of issuance to be implemented within 
30 days.
    Amendment No.: 77
    Facility Operating License No. NPF-69: Amendment revises the 
Technical Specifications.
    Date of initial notice in Federal Register: July 3, 1996 (61 FR 
34893) The Commission's related evaluation of the amendment is 
contained in a Safety Evaluation dated September 17, 1996.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.

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

    Date of application for amendment: January 16, 1996
    Brief description of amendment: The amendment changes the Technical 
Specification (TS) Limiting Condition for Operation Section 3.6.1 and 
Surveillance Requirement Section 4.6.1, ``Primary Containment,'' and 
the corresponding Bases, as well as, adds Administrative Controls 
Section 6.19, ``Containment Leakage Rate Testing Program.'' These 
changes will allow the use of the performance-based containment leakage 
testing requirements described in 10 CFR Part 50, Appendix J, Option B, 
for Type B, for Type A, B, and C testing.
    Date of issuance: September 20, 1996
    Effective date: As of the date of issuance.
    Amendment No.: 203
    Facility Operating License No. DPR-65. Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: February 14, 1996 (61 
FR 5816) as corrected on February 29, 1996 (61 FR 7825)The Commission's 
related evaluation of the amendment is contained in a Safety Evaluation 
dated September 20, 1996.No significant hazards consideration comments 
received: No.
    Local Public Document Room location: Learning Resources Center, 
Three Rivers Community-Technical College, 574 New London Turnpike, 
Norwich, CT 06360, and the Waterford Library, ATTN: Vince Juliano, 49 
Rope Ferry Road, Waterford, CT 06385

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

    Date of application for amendment: June 22, 1995, as supplemented 
August 10, 1995, and March 26, 1996
    Brief description of amendment: The amendment modifies the 
Technical Specification requirements for avoidance and protection from 
thermal- hydraulic instabilities to be consistent with the previously 
approved Boiling

[[Page 52976]]

Water Reactor Owners Group long-term solution Option I-D described in 
the Licensing Topical Report, ``BWR Owners Group Long-Term Stability 
Solutions Licensing Methodology (NEDO-31960),'' dated June 1991, and 
Supplement 1 to NEDO-31960, dated March 1992. The amendment also adds 
the fuel cycle dependent stability power and flow limits in the Core 
Operating Limits Report.
    Date of issuance: September 17, 1996
    Effective date: September 17, 1996, with full implementation within 
60 days
    Amendment No.: 97
    Facility Operating License No. DPR-22. Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: August 30, 1995 (60 FR 
45181) The August 10, 1995, and March 26, 1996, letters provided a 
nonproprietary version of the topical report GENE-637-043-0295 and 
clarifying information, respectively. This information was within the 
scope of the original application and did not change the staff's 
initial proposed no significant hazards consideration determination. 
Therefore, renoticing was not warranted. The Commission's related 
evaluation of the amendment is contained in a Safety Evaluation dated 
September 17, 1996.No significant hazards consideration comments 
received: No.
    Local Public Document Room location: Minneapolis Public Library, 
Technology and Science Department, 300 Nicollet Mall, Minneapolis, 
Minnesota 55401

Omaha Public Power District, Docket No. 50-285, Fort Calhoun 
Station, Unit No. 1, Washington County, Nebraska

    Date of amendment request: May 31, 1995
    Brief description of amendment: The amendment revised the technical 
specifications to require additional restrictions on the component 
cooling water system heat exchangers.
    Date of issuance: September 19, 1996
    Effective date: September 19, 1996
    Amendment No.: 175
    Facility Operating License No. DPR-40. Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: July 5, 1995 (60 FR 
35083) The Commission's related evaluation of the amendment is 
contained in a Safety Evaluation dated September 19, 1996.No 
significant hazards consideration comments received: No.
    Local Public Document Room location: W. Dale Clark Library, 215 
South 15th Street, Omaha, Nebraska 68102

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

    Date of application for amendment: July 12, 1996, as 
supplementedAugust 19, 1996, and August 21, 1996.
    Brief description of amendment: The amendment extends the 
surveillance interval for certain instruments from 18 to 24 months.
    Date of issuance: September 24, 1996
    Effective date: As of the date of issuance to be implemented within 
30 days.
    Amendment No.: 169
    Facility Operating License No. DPR-64: Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: August 14, 1996 (61 FR 
42282) The Commission's related evaluation of the amendment is 
contained in a Safety Evaluation dated September 24, 1996.No 
significant hazards consideration comments received: No
    Local Public Document Room location: White Plains Public Library, 
100 Martine Avenue, White Plains, New York 10610.

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

    Date of application for amendments: May 3, 1996 (TS 352)
    Brief description of amendments: The amendments provide 
administrative changes to the technical specifications.
    Date of issuance: September 18, 1996
    Effective Date: September 18, 1996
    Amendment Nos.: 231, 246 and 206
    Facility Operating License Nos. DPR-33, DPR-52 and DPR-68: 
Amendments revised the Technical Specifications.
    Date of initial notice in Federal Register: August 14, 1996 (61 FR 
42284) The Commission's related evaluation of the amendment is 
contained in a Safety Evaluation dated September 18, 1996.No 
significant hazards consideration comments received: None
    Local Public Document Room location: Athens Public library, 405 E. 
South Street, Athens, Alabama 35611

Vermont Yankee Nuclear Power Corporation, Docket No. 50-271, 
Vermont Yankee Nuclear Power Station, Vernon, Vermont

    Date of application for amendment: June 28, 1996, as 
supplementedAugust 30, 1996.
    Brief description of amendment: The amendment revises the Technical 
Specifications (TSs) to increase the required shutdown margin. It also 
revises TSs associated with this shutdown margin increase to allow 
calculational determination of the highest worth control rod and to 
relax the action requirements in the event the required shutdown margin 
is not met. The amendment also makes appropriate editorial changes and 
minor editorial corrections to the affected TSs.
    Date of issuance: September 25, 1996
    Effective date: As of the date of issuance, to be implemented 
within 30 days.
    Amendment No.: 148
    Facility Operating License No. DPR-28. Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: July 31, 1996 (61 FR 
40031) The August 30, 1996, letter provided clarifying information that 
did not change the scope of the application or affect the initial 
determination. The Commission's related evaluation of the amendment is 
contained in aSafety Evaluation dated September 25, 1996.No significant 
hazards consideration comments received: No
    Local Public Document Room location: Brooks Memorial Library, 224 
Main Street, Brattleboro, VT 05301.

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

    Date of application for amendment: April 25, 1995
    Brief description of amendment: The amendment adds a reactor water 
cleanup system high blowdown containment isolation trip function and 
associated limiting condition for operation and surveillance 
requirements to the Technical Specifications.
    Date of issuance: September 19, 1996Effective date: September 19, 
1996, to be implemented within 30 days of issuance.
    Amendment No.: 147
    Facility Operating License No. NPF-21: The amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: June 28, 1996 (61 FR 
33777) The Commission's related evaluation of the amendment is 
contained in a Safety Evaluation dated September 19, 1996.No 
significant hazards consideration comments received: No.
    Local Public Document Room location:  Richland Public Library, 955 
Northgate Street, Richland, Washington 99352

[[Page 52977]]

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

    Date of application for amendment: June 10, 1996, as supplemented 
on August 27, and September 5, 1996.
    Brief description of amendment: The amendment revises Kewaunee 
Nuclear Power Plant Technical Specification 4.2.b, ``Steam Generator 
Tubes,'' and its associated basis, by allowing the use of Westinghouse 
laser-welded sleeves to repair defective steam generator tubes.
    Date of issuance: September 24, 1996
    Effective date: September 24, 1996, and is to be implemented within 
30 days of the date of issuance.
    Amendment No.: 127
    Facility Operating License No. DPR-43: Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: July 3, 1996 (61 FR 
34902) The August 27, 1996, submittal increased the TS required sample 
size for in-service inspection of repaired tubes in both SGs. The 
September 5, 1996, submittal incorporated the EPRI guidelines for SG 
inspection scope expansion for repaired SG tubes into the TS. These 
submittals provided clarifying information and 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 September 24, 1996. No significant hazards 
consideration comments received: No.
    Local Public Document Room location: University of Wisconsin, 
Cofrin Library, 2420 Nicolet Drive, Green Bay, Wisconsin 54311-7001

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

    Date of application for amendment: May 1, 1996 as supplemented on 
May 31, August 14, August 26, and September 11, 1996.
    Brief description of amendment: The amendment revises Kewaunee 
Nuclear Power Plant Technical Specification 4.2.b, ``Steam Generator 
Tubes,'' its associated bases, and Figure TS 4.2-1 by redefining the 
pressure boundary for Westinghouse mechanical hybrid expansion joint 
(HEJ) steam generator (SG) tube sleeves.
    Date of issuance: September 25, 1996
    Effective date: September 25, 1996, and is to be implemented within 
30 days of the date of issuance.
    Amendment No.: 128
    Facility Operating License No. DPR-43: Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: May 22, 1996 (61 FR 
25715) The May 31, August 14, August 26, and September 11, 1996, 
submittals 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 September 25, 1996.No significant hazards 
consideration comments received: No
    Local Public Document Room location: University of Wisconsin, 
Cofrin Library, 2420 Nicolet Drive, Green Bay, Wisconsin 54311-7001

    Dated at Rockville, Maryland, this 2nd day of October 1996.

    For the Nuclear Regulatory Commission
Steven A. Varga,
Director, Division of Reactor Projects - I/II, Office of Nuclear 
Reactor Regulation
[Doc. 96-25743 Filed 10-8-96; 8:45 am]
BILLING CODE 7590-01-F