[Federal Register Volume 61, Number 168 (Wednesday, August 28, 1996)]
[Notices]
[Pages 44353-44368]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: X96-30828]


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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 August 3, 1996, through August 16, 1996. The 
last biweekly notice was published on August 14, 1996 (61 FR 42274).

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 September 27, 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

[[Page 44354]]

Domestic Licensing Proceedings'' in 10 CFR Part 2. Interested persons 
should consult a current copy of 10 CFR 2.714 which is available at the 
Commission's Public Document Room, the Gelman Building, 2120 L Street, 
NW., Washington, DC and at the local public document room for the 
particular facility involved. If a request for a hearing or petition 
for leave to intervene is filed by the above date, the Commission or an 
Atomic Safety and Licensing Board, designated by the Commission or by 
the Chairman of the Atomic Safety and Licensing Board Panel, will rule 
on the request and/or petition; and the Secretary or the designated 
Atomic Safety and Licensing Board will issue a notice of a hearing or 
an appropriate order.
    As required by 10 CFR 2.714, a petition for leave to intervene 
shall set forth with particularity the interest of the petitioner in 
the proceeding, and how that interest may be affected by the results of 
the proceeding. The petition should specifically explain the reasons 
why intervention should be permitted with particular reference to the 
following factors: (1) the nature of the petitioner's right under the 
Act to be made a party to the proceeding; (2) the nature and extent of 
the petitioner's property, financial, or other interest in the 
proceeding; and (3) the possible effect of any order which may be 
entered in the proceeding on the petitioner's interest. The petition 
should also identify the specific aspect(s) of the subject matter of 
the proceeding as to which petitioner wishes to intervene. Any person 
who has filed a petition for leave to intervene or who has been 
admitted as a party may amend the petition without requesting leave of 
the Board up to 15 days prior to the first prehearing conference 
scheduled in the proceeding, but such an amended petition must satisfy 
the specificity requirements described above.
    Not later than 15 days prior to the first prehearing conference 
scheduled in the proceeding, a petitioner shall file a supplement to 
the petition to intervene which must include a list of the contentions 
which are sought to be litigated in the matter. Each contention must 
consist of a specific statement of the issue of law or fact to be 
raised or controverted. In addition, the petitioner shall provide a 
brief explanation of the bases of the contention and a concise 
statement of the alleged facts or expert opinion which support the 
contention and on which the petitioner intends to rely in proving the 
contention at the hearing. The petitioner must also provide references 
to those specific sources and documents of which the petitioner is 
aware and on which the petitioner intends to rely to establish those 
facts or expert opinion. Petitioner must provide sufficient information 
to show that a genuine dispute exists with the applicant on a material 
issue of law or fact. Contentions shall be limited to matters within 
the scope of the amendment under consideration. The contention must be 
one which, if proven, would entitle the petitioner to relief. A 
petitioner who fails to file such a supplement which satisfies these 
requirements with respect to at least one contention will not be 
permitted to participate as a party.
    Those permitted to intervene become parties to the proceeding, 
subject to any limitations in the order granting leave to intervene, 
and have the opportunity to participate fully in the conduct of the 
hearing, including the opportunity to present evidence and cross-
examine witnesses.
    If a hearing is requested, the Commission will make a final 
determination on the issue of no significant hazards consideration. The 
final determination will serve to decide when the hearing is held.
    If the final determination is that the amendment request involves 
no significant hazards consideration, the Commission may issue the 
amendment and make it immediately effective, notwithstanding the 
request for a hearing. Any hearing held would take place after issuance 
of the amendment.
    If the final determination is that the amendment request involves a 
significant hazards consideration, any hearing held would take place 
before the issuance of any amendment.
    A request for a hearing or a petition for leave to intervene must 
be filed with the Secretary of the Commission, U.S. Nuclear Regulatory 
Commission, Washington, DC 20555-0001, 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.

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

    Date of amendment request: July 19, 1996
    Description of amendment request: The proposed amendment would 
revise Technical Specification (TS) Section 3/4.6.2, Containment Spray 
System, to extend the surveillance interval for performance of an air 
or smoke flow test through containment spray nozzles from once per 5 
years to once per 10 years. This change is consistent with the guidance 
in NRC Generic Letter 93-05, ``Line Item Technical Specifications 
Improvements to Reduce Surveillance Requirements for Testing During 
Power Operations,'' and NUREG-1366, ``Improvements To Technical 
Specifications Surveillance Requirements.''
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:
    1. The proposed amendment does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    The proposed extended testing frequency of containment spray 
nozzles will not affect any initiators of any previously evaluated 
accidents or change the manner of operation for any system or 
component. The containment spray system serves a mitigating function 
by removing heat and fission products from a post accident 
containment atmosphere. Increasing the surveillance test interval 
will not affect the system's ability to provide this function. 
Therefore, there would be no increase in the probability or 
consequences of an accident previously evaluated.

[[Page 44355]]

    2. The proposed amendment does not create the possibility of a 
new or different kind of accident from any accident previously 
evaluated.
    Since the proposed change affects only a surveillance frequency, 
it will not involve any physical alterations to plant equipment or 
alter the manner in which any safety-related system performs its 
function. Therefore, the proposed change does not create the 
possibility of a new or different kind of accident from any accident 
previously evaluated.
    3. The proposed amendment does not involve a significant 
reduction in the margin of safety.
    The proposed change does not affect any Final Safety Analysis 
Report (FSAR) Chapter 15 accident analyses or impact the margin of 
safety for the containment spray system as defined in the Bases to 
the Technical Specifications. Therefore, the proposed change does 
not involve a significant reduction in the margin of safety.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: Cameron Village Regional 
Library, 1930 Clark Avenue, Raleigh, North Carolina 27605
    Attorney for licensee: William D. Johnson, Vice President and 
Senior Counsel, Carolina Power & Light Company, Post Office Box 1551, 
Raleigh, North Carolina 27602
    NRC Project Director: Eugene V. Imbro

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

    Date of amendment request: June 10, 1996
    Description of amendment request: To change the technical 
specifications to reflect the transition from General Electric Company 
(GE) to Siemens Power Corporation (SPC) as the fuel supplier for the 
Quad Cities Nuclear Power Station, Units 1 and 2.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:
    1) Involve a significant increase in the probability or 
consequences of an accident previously evaluated.
    The probability of an evaluated accident is derived from the 
probabilities of the individual precursors to that accident. The 
consequences of an evaluated accident are determined by the 
operability of plant systems designed to mitigate those 
consequences. Limits will be established consistent with NRC 
approved methods to ensure that fuel performance during normal, 
transient, and accident conditions is acceptable. The proposed 
Technical Specifications amendment reflects previously approved SPC 
methodology used to analyze normal operations, including anticipated 
operational occurrences (AOOs), and to determine the potential 
consequences of accidents.
    Licensing Methods and Models
    The proposed amendment is to support operation with NRC approved 
fuel and licensing methods supplied from Siemens Power Corporation. 
In accordance with FSAR Chapter 15, the same accidents and 
transients will be analyzed with the new fuel and methods as were 
analyzed by GE for GE fuel. The analysis methods and models are NRC 
approved. These approved methods and models are used to determine 
the fuel thermal limits (e.g., LHGR, APLHGR, MCPR). The SPC core 
monitoring code enables the site to monitor keff as well as rod 
density to perform the reactivity anomaly surveillance. This is 
consistent with GE methodology. The support systems for minimizing 
the consequences of transients and accidents are not affected by the 
proposed amendment. Therefore, the change in licensing analysis 
methods and models does not significantly increase the probability 
of an accident or the consequences of an accident previously 
identified.
    New Fuel Design
    The use of ATRIUM 9B fuel at Quad Cities does not involve a 
significant increase in the probability or consequences of any 
accident previously evaluated in the FSAR. The ATRIUM-9B fuel is 
generically approved for use as a reload BWR fuel type (Reference: 
ANF-89-014(P)(A) Rev. 1 Supplement 1, General Mechanical Design for 
Advanced Nuclear Fuels 9X9-IX and 9X9-9X BWR Reload Fuel). Limiting 
postulated occurrences and normal operation have been analyzed using 
NRC-approved methods for the ATRIUM 9B fuel design to ensure that 
safety limits are protected and that acceptable transient and 
accident performance is maintained.
    The reload fuel has no adverse impact on the performance of in-
core neutron flux instrumentation or CRD response. The ATRIUM-9B 
fuel design will not adversely affect performance of neutron 
instrumentation nor will it adversely affect the movement of control 
blades relative to the GE fuel. The exterior dimensions of the 
ATRIUM-9B fuel have been evaluated by ComEd; the SPC fuel provides 
adequate clearances relative to the GE10 fuel installed at Quad 
Cities. Thus, no increased interactions with the adjacent control 
blade and nuclear instrumentation are created. Additionally, given 
the above mentioned overall envelope similarities, no problems are 
anticipated with other station equipment such as the fuel storage 
racks, the new fuel inspection stand and the spent fuel pool fuel 
preparation machine. Therefore, the probability of adverse 
interactions between the Siemens fuel and components in the core and 
fuel handling equipment is not significantly increased.
    The ATRIUM 9B design is neutronically compatible with the 
existing fuel types and core components in the Quad Cities core. SPC 
tests have demonstrated that the ATRIUM-9B fuel design is 
hydraulically compatible with the GE9/GE10 fuel. The bundle pressure 
drop characteristics of the ATRIUM 9B bundle are similar to those of 
the GE9/GE10 fuel design, hence core thermal-hydraulic stability 
characteristics are not adversely affected by the ATRIUM 9B design. 
Cycle stability calculations are performed by SPC. Therefore, the 
probability of thermal hydraulic instability is not significantly 
increased.
    An evaluation of the Emergency Procedures is being performed to 
ensure that the use of the ATRIUM-9B fuel at Quad Cities does not 
alter any assumptions previously made in evaluating the radiological 
consequences of an accident at Quad Cities Station. Therefore, the 
radiological consequences of accidents are not significantly 
increased.
    Methods approved by the NRC are being used in the evaluation of 
fuel performance during normal and abnormal operating conditions. 
The ComEd and SPC methods to be used for the cycle specific 
transient analyses have been previously NRC approved. The proposed 
methodologies are administrative in nature and do not significantly 
affect any accident precursors or accident results; as such, the 
proposed incorporation of the SPC methodologies for Quad Cities does 
not significantly increase the probability or consequences of any 
previously evaluated accidents. The description of the fuel is 
modified to include the water box design of the NRC approved ATRIUM-
9B fuel. This change is administrative.
    Review of the above concludes that the probability of occurrence 
and the consequences of an accident previously evaluated in the 
safety analysis report have not been significantly increased.
    * * * * *
    2) Create the possibility of a new or different kind of accident 
from any accident previously evaluated:
    Creation of the possibility of a new or different kind of 
accident would require the creation of one or more new precursors of 
that accident. New accident precursors may be created by 
modifications of the plant configuration, including changes in 
allowable modes of operation.
    Licensing Methods and Models
    The proposed Technical Specification amendment reflects 
previously approved SPC methodology used to analyze normal 
operations, including AOOs, and to determine the potential 
consequences of accidents. In accordance with FSAR Chapter 15, the 
same accidents and transients will be analyzed with the new fuel and 
methods as were analyzed by GE for GE fuel. As stated above, the 
proposed changes do not permit modes of operation which differ from 
those currently permitted; therefore, the possibility of a new or 
different kind of accident is not created. Plant support equipment 
is not affected by the proposed changes; therefore, no new failure 
modes are created.
    New Fuel Design
    The basic design concept of a 9x9 fuel pin array with an 
internal water box has been used in various lead assembly programs 
and in reload quantities in Europe since 1986.

[[Page 44356]]

WNP-2 has loaded reload quantities since 1991. Approximately 650 
water box assemblies have been irradiated in the United States 
through 1995, with a substantially higher number being irradiated 
overseas. The NRC has reviewed and approved the ATRIUM-9B fuel 
design (Reference: ANF-89-014(P)(A) Rev. 1 Supplement 1, Generic 
Mechanical Design for Advanced Nuclear Fuels 9X9-IX and 9X9-9X BWR 
Reload Fuel). The similarities in fuel design and operation between 
GE and SPC, and the previous Boiling Water Reactor experience with 
both vendors' fuel indicate there would be no new or different types 
of accidents for Quad Cities than have been considered for the 
existing fuel. Therefore, the use of ATRIUM-9B fuel at Quad Cities 
does not create the possibility of a new or different kind of 
accident from any accident previously evaluated.
    * * * * *
    3) Involve a significant reduction in the margin of safety for 
the following reasons:
    The existing margin to safety is provided by the existing 
acceptance criteria (e.g., 10CFR50.46 limits). The proposed 
Technical Specification amendment reflects previously approved SPC 
methodology used to demonstrate that the existing acceptance 
criteria are satisfied. The revised methodology has been previously 
reviewed and approved by the USNRC for application to reload cores 
of GE BWRs. References for the Licensing Topical Reports which 
document this methodology, and include the Safety Evaluation Reports 
prepared by the USNRC, are added to the Reference section of the 
Technical Specifications as part of this amendment.
    Licensing Methods and Models
    The proposed amendment does not involve changes to the existing 
operability criteria. NRC approved methods and established limits 
(implemented in the COLR) ensure acceptable margin is maintained. 
The ComEd and SPC reload methodologies for the ATRIUM-9B reload 
design are consistent with the Technical Specification Bases. The 
Limiting Conditions for Operation are taken into consideration while 
performing the cycle specific and generic reload safety analyses. 
NRC approved methods are listed in Section 6 of the Technical 
Specifications.
    Analyses performed with NRC-approved methodology have 
demonstrated that fuel design and licensing criteria will be met 
during normal and abnormal operating conditions. The same margins of 
safety are utilized by SPC as GE (e.g., limits on peak cladding 
temperature, cladding oxidation, plastic strain). Therefore, there 
is not a significant reduction in the margin of safety.
    New Fuel Design
    The exterior dimensions of the ATRIUM-9B fuel assembly result in 
equivalent clearances relative to the GE10B. Thus, no increased 
interactions with the adjacent control blade and nuclear 
instrumentation are created. The change does not adversely impact 
equipment important to safety; therefore,the margin of safety is not 
significantly reduced.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
requested amendments involve no significant hazards consideration.
    Local Public Document Room location: Dixon Public Library, 221 
Hennepin Avenue, Dixon, Illinois 61021
    Attorney for licensee: Michael I. Miller, Esquire; Sidley and 
Austin, One First National Plaza, Chicago, Illinois 60603
    NRC Project Director: Robert A. Capra

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

    Date of amendment request: August 12, 1996
    Description of amendment request: The proposed change would 
implement the performance-based containment leak rate testing 
provisions of Option B to 10 CFR Part 50 Appendix J for the Type A 
(containment) testing program.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:
    The following analysis is presented, pursuant to 10 CFR 50.91, 
to demonstrate that the proposed change will not create a 
Significant Hazard Consideration.
    1. The proposed change will not involve a significant increase 
in the probability or consequences of an accident previously 
evaluated.
    Containment leak rate testing is not an initiator of any 
accident; the proposed change does not affect reactor operations or 
accident analysis, and has no significant radiological consequences. 
Therefore, this proposed change will not involve an increase in the 
probability or consequences of any previously-evaluated accident.
    2. The proposed change will not create the possibility of any 
new accident not previously evaluated.
    The proposed change does not affect normal plant operations or 
configuration, or change any design basis. The proposed changes will 
not affect the response of [the] containment during a design basis 
accident.
    3. There is no significant reduction in a margin of safety.
    The proposed changes are based on NRC-accepted provisions, and 
maintain necessary levels of reliability of containment integrity. 
The performance-based approach to leakage rate testing recognizes 
that historically good results of containment testing provide 
appropriate assurance of future containment integrity; this supports 
the conclusion that the impact on the health and safety of the 
public as a result of extended test intervals is negligible.
    Based on the above, no significant hazards consideration is 
created by the proposed 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: Oconee County Library, 501 
West South Broad Street, Walhalla, South Carolina 29691
    Attorney for licensee: J. Michael McGarry, III, Winston and Strawn, 
1200 17th Street, NW., Washington, DC 20036
    NRC Project Director: Herbert N. Berkow

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

    Date of amendment request: May 31, 1996
    Description of amendment request: The proposed amendment revises 
the surveillance test interval for the reactor protection system 
reactor trip breakers, reactor trip modules, and electronic trip relays 
from 1 month to 6 months. In addition to requesting a change to the 
Arkansas Nuclear One, Unit 1 Technical Specifications, the request also 
proposes the same changes to NUREG-1430, Standard Technical 
Specifications - Babcock and Wilcox Plants.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:
    Criterion 1 - Does Not Involve a Significant Increase in the 
Probability or Consequences of an Accident Previously Evaluated.
    The accident mitigation features of the plant are not affected 
by the proposed test interval extension. The results of the B&W 
Owners Group Topical Report BAW-10167, Supplement 3, ``Justification 
for increasing The Reactor Trip System On-Line Test Intervals,'' 
show that the test interval extension of the reactor protection 
system trip devices is not a significant contributor to trip system 
unavailability or the risk of core damage.
    Therefore, this change does not involve a significant increase 
in the probability or consequences of any accident previously 
evaluated.
    Criterion 2. Does Not Create the Possibility of a New or 
Different Kind of Accident from any Previously Evaluated.
    The reactor trip device surveillance test interval is not, in 
and of itself, considered to be an accident initiator. Failure of a 
trip device to function is an analyzed condition and does not 
constitute a new or different kind of accident.
    Therefore, this change does not create the possibility of a new 
or different kind of accident from any previously evaluated.
    Criterion 3. Does Not Involve a Significant Reduction in the 
Margin of Safety.
    The results of the B&W Owners Group Topical Report BAW-10167, 
Supplement 3, ``Justification for Increasing The Reactor Trip

[[Page 44357]]

System On-Line Test Intervals,'' show that the test interval 
extension of the reactor protection system trip devices is not a 
significant contributor to trip system unavailability or the risk of 
core damage. In addition, the uncertainty analysis contained in BAW-
10167 confirms the robustness of the results by demonstrating that 
even with an order of magnitude change in the failure data, the 
incremental increase due to an increased test interval is 
insignificant. Entergy Operations has reviewed BAW-10167 and found 
it applicable to ANO-1.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: Tomlinson Library, Arkansas 
Tech University, Russellville, AR 72801
    Attorney for licensee: Nicholas S. Reynolds, Esquire, Winston and 
Strawn, 1400 L Street, N.W., Washington, DC 20005-3502
    NRC Project Director: William D. Beckner

Entergy Operations, Inc., Docket Nos. 50-313 and 50-368, Arkansas 
Nuclear One, Unit Nos. 1 and 2 (ANO-1&2), Pope County, Arkansas

    Date of amendment request: May 9, 1996
    Description of amendment request: The proposed amendment changes 
the name of Arkansas Power and Light Company (AP&L) to Entergy 
Arkansas, Inc. in both the Operating License and the Technical 
Specifications. AP&L is licensed to own and possess Arkansas Nuclear 
One (ANO). The company licensed to operate ANO, Entergy Operations, 
Inc. is unaffected by this change.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:
    1. Does Not Involve a Significant Increase in the Probability or 
Consequences of an Accident Previously Evaluated.
    The proposed change documents changing the legal name of the 
company. The proposed change will not affect any other obligations. 
The company will continue to own all of the same assets, will 
continue to serve the same customers, and will continue to honor all 
existing obligations and commitments. 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 administrative changes in the operating license requirements 
do not involve any change in the design of the plant. 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 change is administrative in nature and does not 
reduce the margin of safety imposed by any current requirements. 
Therefore, this change does not involve a significant reduction in 
the margin of safety.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: Tomlinson Library, Arkansas 
Tech University, Russellville, AR 72801
    Attorney for licensee: Nicholas S. Reynolds, Esquire, Winston and 
Strawn, 1400 L Street, N.W., Washington, DC 20005-3502
    NRC Project Director: William D. Beckner

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

    Dates of amendment request: July 17, 1996
    Description of amendment request: The licensee proposed to change 
the Turkey Point Units 3 and 4 Technical Specifications (TS) to 
implement 10 CFR 50, Appendix J, Option B, for containment leakage 
testing. Changes include relocating the details for containment testing 
to the ``containment leakage rate testing program'' and adding the 
requirements of the containment leakage rate testing program to TS 
6.8.4, which describes facility programs.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below.
    (1) Operation of the facility in accordance with the proposed 
amendments would not involve a significant increase in the 
probability or consequences of an accident previously evaluated.
    The proposed amendments do not involve a significant increase in 
the probability or consequences of an accident previously evaluated 
because:
    a) These proposed changes are all consistent with NRC 
requirements and guidance for implementation of 10 CFR 50, Appendix 
J, Option B.
    b) Based on industry and NRC evaluations performed in support of 
developing Option B, these changes potentially result in a minor 
increase in the consequences of an accident previously evaluated due 
to the expanded testing intervals. However, the proposed changes do 
not result in an increase in the core damage frequency since the 
containment system is used for mitigation purposes only.
    c) These changes are expected to result in increased attention 
to components with poor leakage test history as part of the 
performance-based nature of Option B, such that the marginally 
increased consequences from the expanded testing intervals may be 
further reduced or negated.
    Therefore, these changes do not involve a significant increase 
in the probability or consequences of an accident previously 
evaluated.
    (2) Operation of the facility in accordance with the proposed 
amendments would not create the possibility of a new or different 
kind of accident from any accident previously evaluated.
    The use of the modified specifications can not create the 
possibility of a new or different kind of accident from any 
previously evaluated since the proposed amendments will not change 
the physical plant or the modes of plant operation defined in the 
facility operating license. No new failure mode is introduced due to 
the implementation of a performance-based program for containment 
leakage rate testing, since the proposed changes do not involve the 
addition or modification of equipment, nor do they alter the design 
or operation of affected plant systems, structures, or components.
    (3) Operation of the facility in accordance with the proposed 
amendments would not involve a significant reduction in a margin of 
safety.
    The operating limits and functional capabilities of the affected 
systems, structures, and components are basically unchanged by the 
proposed amendments due to the following reasons:
    a) The acceptance criteria for total integrated containment 
leakage of 1.0 La is consistent with the current technical 
specifications and is within the design basis accident assumptions, 
and therefore does not reduce the margin of safety.
    b) The increase in intervals between leak-test surveillances 
will not significantly reduce the margin of safety as shown by 
findings in NUREG 1493, ``Performance-Based Containment Leak-Test 
Program'', which was based on implementation of the performance-
based testing of Option B.
    Therefore these changes do not involve a significant reduction 
in the margin of safety.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 50.92(c) are 
satisfied.Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: Florida International 
University, University Park, Miami, Florida 33199
    Attorney for licensee: J. R. Newman, Esquire, Morgan, Lewis & 
Bockius, 1800 M Street, NW., Washington, DC 20036
    NRC Project Director: Frederick J. Hebdon

[[Page 44358]]

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

    Date of amendment request: May 21, 1996
    Description of amendment request: The proposed change to the 
condensate storage tank (CST) level indication would ensure that the 
water level is sufficient to provide 50,000 gallons of water for core 
spray makeup to the reactor pressure vessel.
    Technical Specification (TS) Surveillance Requirement (SR) 
3.5.2.2.b for ECCS - Shutdown states: ``Condensate storage tank (CST) 
water level is [greater than or equal to] 12 feet.'' The corresponding 
Bases state: ''... the CST contains [greater than or equal to] 150,000 
gallons of water, equivalent to 12 feet, ensures that the CS System can 
supply at least 50,000 gallons of makeup water to the RPV.''
    Subsequent licensee analyses confirmed that Plant Hatch Units 1 and 
2 CST configurations are different; that is, for both CSTs, a water 
level of 12 feet is not equivalent to the required capacity of 150,000 
gallons of water. Based on these calculations, the correct level for 
the Unit 1 CST is 13 feet, and the correct level for the Unit 2 CST is 
15 feet.
    The proposed change would revise Unit 1 and Unit 2 SR 3.5.2.2.b to 
require a CST water level of greater than or equal to 13 feet and 
greater than or equal to 15 feet, respectively, to ensure at least 
50,000 gallons of water are available for core spray (CS) makeup to the 
reactor pressure vessel (RPV).
    The associated Bases for each unit will be revised accordingly.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration which is presented below:
    1. The proposed TS change does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated, because this administrative change to the CST 
water level does not alter the operation of any plant system or 
component. The proposed change does not involve a physical 
modification to any structure, system, or component. The minimum CST 
water level for each unit is being increased to account for the 
height of the CS suction standpipe within each CST and the 
differences in the Unit 1 and
    Unit 2 CST diameters (gallons/ft of water) as follows:
    a. Unit 1 - The proposed minimum water level is calculated as: 
CS suction standpipe height of 9 ft + (50,000 gallons divided by 
12,704 gallons/ft) = 12.93 ft or 13 ft.
    b. Unit 2 - The proposed minimum water level is calculated as: 
CS suction standpipe height of 10 ft + (50,000 gallons divided by 
11,343 gallons/ft) = 14.4 ft or 15 ft.
    The revised minimum levels ensure at least 50,000 gallons of 
water are provided above the top of the standpipe in each unit's CST 
and are available for CS makeup to the RPV, as stated in the 
applicable Bases. The TS Limiting Conditions for Operation (LCO) 
remain unaffected by the proposed change.
    2. The proposed TS change does not create the possibility of a 
new or different kind of accident from any accident previously 
evaluated. Revising Surveillance Requirement acceptance criteria 
does not result in any physical modification to the plant or 
operation of any existing equipment.
    3. The proposed TS change does not involve a significant 
reduction in a margin of safety, since this administrative change 
only ensures the existing TS Bases are satisfied by increasing the 
minimum CST water level requirement to ensure at least 50,000 
gallons of water are available for CS injection to the RPV. The 
proposed change does not involve a physical modification to any 
structure, system or component, and does not modify the operation of 
any existing equipment.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: Appling County Public Library, 
301 City Hall Drive, Baxley, Georgia 31513
    Attorney for licensee: Ernest L. Blake, Jr., Esquire, Shaw, 
Pittman, Potts and Trowbridge, 2300 N Street, NW., Washington, DC 20037
    NRC Project Director: Herbert N. Berkow

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

    Date of amendment request: July 8, 1996
    Description of amendment request: The proposed amendment would 
clarify that the component cooling water system surge tank level 
instrumentation can be demonstrated operable by performing a channel 
calibration test during any plant mode of operation.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:
    1. The proposed change does not involve a significant increase 
in the probability or consequences of an accident previously 
evaluated.
    The proposed change to Technical Specification Surveillance 
Requirement 4.7.3.b.3 will not effect any accident initiators or 
precursors and will not alter the design assumptions for the systems 
or components used to mitigate the consequences of an accident. 
Calibration is performed on level instrumentation of Component 
Cooling Water System trains that are out of service for scheduled 
maintenance. Isolation redundancy is provided by instrumentation 
associated with the trains that are in service during the 
calibration. Since the surveillance will continue to be performed at 
the specified interval, this proposed change will not increase the 
probability of occurrence of an accident previously evaluated. The 
surveillance does not differ from those previously performed; 
therefore, there is no impact on the consequences of an accident 
previously evaluated.
    2. The proposed change does not create the possibility of a new 
or different kind of accident from any accident previously 
evaluated.
    Clarifying the surveillance interval for surge tank level 
instrumentation does not involve installation or operation of new or 
different kinds of equipment. There is no change in the procedures 
as described in the Technical Specifications. The change only 
clarifies the interval at which the subject calibration will be 
performed. Therefore, the proposed change does not create the 
possibility of a new or different kind of accident from any accident 
previously evaluated.
    3. The proposed change does not involve a significant reduction 
in a margin of safety.
    The specified surveillance will remain as stated in the 
Technical Specifications. Consequently, there is no reduction in the 
effectiveness of the surveillance in ensuring equipment operability. 
Calibration is performed on level instrumentation of Component 
Cooling Water System trains that are out of service for scheduled 
maintenance. Isolation redundancy is provided by instrumentation 
associated with the trains that are in service during the 
calibration. Consequently, clarifying the interval at which the 
calibration is performed will have no significant impact on the 
margin of safety.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
request for amendments involves no significant hazards consideration.
    Local Public Document Room location: Wharton County Junior College, 
J. M. Hodges Learning Center, 911 Boling Highway, Wharton, TX 77488
    Attorney for licensee: Jack R. Newman, Esq., Morgan, Lewis & 
Bockius, 1800 M Street, N.W., Washington, DC 20036-5869
    NRC Project Director: William D. Beckner

[[Page 44359]]

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

    Date of amendment request: August 8, 1996
    Description of amendment request: The proposed amendment would 
allow the transition from Mode 4 to Mode 3 with the turbine-driven 
auxiliary feedwater pump inoperable and allow a 72-hour period after 
the entry into Mode 3 to complete all necessary operability testing.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:
    1. The proposed change does not involve a significant increase 
in the probability or consequences of an accident previously 
evaluated.
    The proposed change will allow entry into Mode 3 with an 
inoperable Turbine Driven Auxiliary Feedwater pump. Since the 
operability test on the Turbine Driven Auxiliary Feedwater pump can 
only be performed once steam pressure is greater than or equal to 
1000 psig, this change will allow the plant to reach the Mode where 
steam pressure greater than or equal to 1000 psig is available to 
perform the operability testing on the Turbine Driven Auxiliary 
Feedwater pump. The allowance of 72 hours to complete the 
surveillance testing will make the surveillance requirements 
consistent with the allowed outage time already established in the 
Action Statements. The proposed change does not affect the 
probability of an accident. The Turbine Driven Auxiliary Feedwater 
pump is not assumed to be an initiator of any analyzed event. The 
consequences of an accident previously evaluated remain unchanged by 
allowing the pump to be inoperable until suitable conditions exist 
to perform the operability testing. The operability testing will 
continue to demonstrate that the Turbine Driven Auxiliary Feedwater 
pump will perform as required prior to entry into Mode 2. This 
change will not alter assumptions relative to the mitigation of an 
accident or transient event. Therefore, this change will 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 previously evaluated.
    This change will not physically alter the plant (no new or 
different type of equipment will be installed). The changes in 
methods governing normal plant operation are consistent with current 
safety analysis assumptions. The proposed change will allow entry 
into Mode 3 with the Turbine Driven Auxiliary Feedwater pump 
inoperable in order to perform the pump Operability Test on the 
turbine driven AFW [Auxiliary Feedwater] pump once steam pressure is 
greater than or equal to 1000 psig. Therefore, this change does not 
create the possibility of a new or different kind of accident from 
any accident previously evaluated.
    3. The proposed change does not involve a significant reduction 
in a margin of safety.
    The proposed change will allow entry into Mode 3 with the 
Turbine Driven AFW pump inoperable in order to perform the pump 
Operability Test on the turbine driven AFW pump once steam pressure 
is greater than or equal to 1000 psig. This will allow time for the 
plant to obtain suitable test conditions with steam pressure greater 
than or equal to 1000 psig. The margin of safety is not affected by 
this change. The operability testing will continue to maintain 
assurance that the AFW Pumps will perform as required prior to entry 
into Mode 2. The safety analysis assumptions will still be 
maintained, thus, no question of safety exists. Therefore, this 
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 standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
request for amendments involves no significant hazards consideration.
    Local Public Document Room location: Wharton County Junior College, 
J. M. Hodges Learning Center, 911 Boling Highway, Wharton, TX 77488
    Attorney for licensee: Jack R. Newman, Esq., Morgan, Lewis & 
Bockius, 1800 M Street, N.W., Washington, DC 20036-5869
    NRC Project Director: William D. Beckner

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

    Date of amendment request: June 4, 1996
    Description of amendment request: The proposed amendment would 
modify the Seabrook Station, Unit No. 1 Technical Specifications to 
implement Option B to 10 CFR Part 50, Appendix J by referring to 
Regulatory Guide 1.163, ``Performance-Based Containment Leak-Test 
Program. The following Technical Specifications would be affected by 
the proposed amendment:
    1. Definitions: Definition 1.7, Containment Integrity (Item d.) 
would be revised to reflect that leakage rates would be in accordance 
with the Containment Leakage Rate Testing Program.
    2. Limiting Conditions for Operation and Surveillance Requirements:
    a. Containment Integrity: Surveillance Requirement 4.6.1.1.c would 
be deleted because the specific guidance would be contained in the 
Containment Leakage Rate Testing Program.
    b. Containment Leakage: Limiting Condition for Operation 3.6.1.2.a 
through 3.6.1.2.c and Surveillance Requirements 4.6.1.2.a through 
4.6.1.2.h would be revised to replace specific guidance with a 
reference to the Containment Leakage Rate Testing Program.
    c. Containment Leakage: The Action for Limiting Condition for 
Operation 3.6.1.2 would be revised to include the equivalent Action as 
required for Limiting Condition for Operation 3.6.1.1 when the overall 
integrated containment leak rate exceeds 1.0 La.
    d. Containment Air Locks: Limiting Conditions for Operation 
3.6.1.3.a and 3.6.1.3.b would be deleted and Surveillance Requirements 
4.6.1.3.a and 4.6.1.3.b would be revised to replace specific guidance 
with a reference to the Containment Leakage Rate Testing Program. The 
footnote addressing the exemption to Appendix J regarding testing the 
air locks prior to establishing containment integrity would be 
maintained in the Containment Leakage Rate Testing Program.
    e. Containment Vessel Structural Integrity: Surveillance 
Requirement 4.6.1.6 would be revised to replace specific guidance with 
a reference to the Containment Leakage Rate Testing Program.
    f. Containment Ventilation System: Limiting Condition for Operation 
3.6.1.7, Action b. would be revised to replace specific guidance with a 
reference to the Containment Leakage Rate Testing Program. Surveillance 
Requirement 4.6.1.7.1 would be revised to replace specific guidance 
with a reference to the Containment Leakage Rate Testing Program.
    g. Containment Enclosure Building: Limiting Condition for Operation 
3.6.5.3 and Surveillance Requirement 4.6.5.3 would be revised to 
include a reference to the requirements in the Containment Leakage Rate 
Testing Program.
    3. Bases: Sections 3/4.6.1.2, Containment Leakage; 3/4.6.1.7, 
Containment Ventilation System; and 3/4.6.5.3, Containment Enclosure 
Building Structural Integrity, would be revised to reflect the above 
changes including a reference to the Containment Leakage Rate Testing 
Program. In addition, a statement would be added to Section 3/4.6.1.2 
to clarify the operability of containment regarding allowable leakage 
rates.
    4. Administrative Controls: Section 6.15 would be added to 
establish a Containment Leakage Rate Testing Program, as specified in 
Regulatory Guide 1.163, dated September 1995.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the

[[Page 44360]]

licensee has provided its analysis of the issue of no significant 
hazards consideration. The NRC staff has reviewed the licensee's 
analysis against the standards of 10 CFR 50.92(c). The NRC staff's 
review is presented below.
    A. The changes do not involve a significant increase in the 
probability or consequences of an accident previously evaluated (10 
CFR 50.92(c)(1)) because the proposed changes merely revise the 
testing criteria for containment penetrations. The revised criteria 
will be based on the guidance in Regulatory Guide 1.163, 
``Performance-Based Containment Leak-Test Program.''
    This guidance allows for the use of relaxed testing frequencies 
for containment penetrations that have performed satisfactorily on a 
historical basis.
    To support consideration of Option B to Appendix J, the NRC 
staff reviewed the potential impact of performance-based testing 
frequencies for containment penetrations. The NRC staff review is 
documented in NUREG-1493 ``Performance-Based Containment Leak-Test 
Program.'' One of the staff's conclusions was that reducing the 
frequency of Type A tests (Integrated Leak Rate Tests) from three 
per 10 years to one per 10 years leads to a marginal increase in 
risk. For Type B and C testing (Local Leak Rate Tests), the change 
in testing frequency will not have significant impact since, under 
existing requirements, leakage contributes less than 0.1 percent of 
the overall accident risk. The use of a performance-based testing 
program will continue to provide assurance that the accident 
analysis assumptions remain bounding.
    B. The changes do not create the possibility of a new or 
different kind of accident from any accident previously evaluated 
(10 CFR 50.92(c)(2)) because they do not affect the manner by which 
the facility is operated or involve changes to structures, systems, 
or components that affect the operational characteristics of the 
facility. The changes merely revise the testing criteria for the 
containment penetrations, and establish a Containment Leakage Rate 
Testing Program to ensure that the performance history of each 
penetration is satisfactory prior to changing any test frequency. 
Since there is no change to the facility or the way in which the 
facility is operated, there is no possibility of creating a new or 
different kind of accident than previously analyzed.
    C. The changes do not involve a significant reduction in a 
margin of safety (10 CFR 50.92(c)(3)). During the development of 10 
CFR Part 50, Appendix J, Option B, the NRC staff determined the 
reduction in safety associated with the implementation of the 
performance-based testing program. The staff concluded that reducing 
the frequency of Type A tests (Integrated Leak Rate Tests) from 
three per 10 years to one per 10 years would have an imperceptible 
impact upon risk. For Type B and C testing (Local Leak Rate Tests), 
the change in testing frequency will not have significant impact 
since this leakage contributes less than 0.1 percent of the overall 
risk based on the existing regulations. The use of Option B will 
have minimal impact on the radiological release rates since most 
penetration leakage is well below the specified limits. The staff 
noted that the accident risk is relatively insensitive to 
containment leakage rate because accident risk is dominated by 
accident sequences that result in failure of or bypass of the 
containment. The use of a performance-based testing program will 
continue to provide assurance that the accident analysis assumptions 
remain bounding.
    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: Exeter Public Library, 
Founders Park, Exeter, NH 03833
    Attorney for licensee: Lillian M. Cuoco, Esquire, Northeast 
Utilities Service Company, Post Office Box 270, Hartford CT 06141-0270
    NRC Project Director: Phillip F. McKee

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

    Date of amendment request: May 17, 1996
    Description of amendment request: The proposed amendment would 
revise the technical specifications (TS) to relocate the operability 
requirements for shock suppressors (snubbers) from the TS to the 
Updated Safety Analysis Report (USAR) and incorporate snubber 
examination and testing requirements into TS 3.3.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:
    1. The proposed change does not involve a significant increase 
in the probability or consequences of an accident previously 
evaluated.
    The proposed change will relocate operability requirements for 
shock suppressors (snubbers) from the Technical Specifications (TS) 
to the Updated Safety Analysis Report (USAR) and/or plant 
procedures. On July 16, 1993, the NRC issued a Final Policy 
Statement on Technical Specification Improvements for Nuclear Power 
Reactors. The Final Policy Statement contains four criteria which 
can be used to determine which constraints on the design and 
operation of nuclear power plants are appropriate for inclusion in 
TS. The NRC has incorporated these criteria into 10 CFR 50.36, 
``Technical specifications.'' Snubbers do not meet any of the four 
criteria for inclusion as a Limiting Condition for Operations within 
the TS, and therefore it is proposed that these requirements be 
relocated from the TS. The proposed change would not reduce or 
revise any of the current requirements for snubber operability, only 
relocate the requirements. Any changes to the requirements contained 
in the USAR and/or plant procedures can be made without NRC approval 
only when the changes meet the criteria of 10 CFR 50.59. Changes to 
the snubber operability requirements that do not meet the criteria 
of 10 CFR 50.59 must be approved by the NRC by license amendment. 
Therefore, the relocation of the requirements on snubber operability 
from the TS to the USAR does not increase the probability or 
consequences of any accident previously analyzed.
    The proposed change also deletes sections of the TS which are 
redundant or in conflict with the American Society of Mechanical 
Engineers (ASME) Boiler and Pressure Vessel Code. Snubbers are 
required to be examined and tested in accordance with ASME Section 
XI by 10 CFR 50.55a. The proposed change will ensure that the TS 
implement ASME Section XI examination and testing requirements for 
snubbers in accordance with 10 CFR 50.55a. Where differences between 
the deleted sections of the TS and ASME Section XI requirements 
exist, the Section XI requirements are similar or more conservative 
than the TS. For example, although the functional test sample size 
differs between the methodologies, both ensure that a very high 
percentage of the snubbers in the plant are operable within 
acceptance limits. Therefore, the proposed revision does not reduce 
the effectiveness of snubber examination and testing.
    The proposed change would not reduce the operability 
requirements, acceptance criteria, or examination and testing of 
snubbers. Therefore, the proposed change would not increase 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.
    There will be no physical alterations to the plant 
configuration, changes to setpoint values, or changes to the 
implementation of setpoints or limits as a result of this proposed 
change.
    The proposed change deletes duplicate or conflicting 
requirements between the TS and the ASME Section XI. In these areas, 
the proposed deletions would remove the TS requirements and testing 
would be conducted in accordance with ASME Section XI as directed by 
10 CFR 50.55a. Although the requirements of ASME Section XI differ 
from the TS in some cases, the differences do not decrease the 
effectiveness of testing and examination as compared to the TS 
requirements. Other areas, such as snubber operability requirements 
and service life monitoring, which are presently addressed by TS, 
but are not covered under ASME Section XI, will be maintained in the 
USAR so that these requirements cannot be deleted without NRC 
approval.
    Therefore, the proposed change does not create the possibility 
of a new or different kind of accident from any previously 
evaluated.
    3. The proposed change does not involve a significant reduction 
in a margin of safety.
    The proposed change does not reduce the operability, 
examination, or testing requirements for snubbers. Snubbers will 
still be required to meet the requirements of ASME Section XI and 10 
CFR 50.55a except where specific written relief has been granted by 
the NRC. Therefore, the proposed change does not involve a 
significant reduction in a margin of safety.

[[Page 44361]]

    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: W. Dale Clark Library, 215 
South 15th Street, Omaha, Nebraska 68102
    Attorney for licensee: Perry D. Robinson, Winston & Strawn, 1400 L 
Street, N.W., Washington, DC 20005-3502
    NRC Project Director: William H. Bateman

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

    Date of amendment request: May 20, 1996
    Description of amendment request: The proposed amendment would 
revise the technical specifications (TS) to clarify surveillance test 
requirements of TS 3.1, Tables 3-1, 3-2, 3-3, 3-3A, and 3-5.
    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 changes to the Table of Contents are administrative in 
nature to reflect the removal of incore instrumentation 
(Specification 2.10.3) from the TS by Amendment 167 and for 
consistency. Amendment 169 inadvertently reinserted incore 
instrumentation back into the Table of Contents.
    The change to Specification 2.1.7(1)b is necessary because the 
requirement to test the signal to alarm meter relay located in 
Specification 3.1, Table 3-3, Item 6 is being deleted. The test, 
which verifies the high and low pressurizer level alarm settings and 
the pressurizer heater cutout function is unnecessary. Operating 
experience has shown that a shiftly pressurizer level verification 
as proposed for Specification 3.1, Table 3-3, Item 6.a is sufficient 
to detect any level deviation and verify that operation is within 
safety analyses assumptions. The level alarms serve as early warning 
devices but do not provide an accident mitigation function. 
Replacing the monthly test with a channel check is in accordance 
with NUREG-1432, Combustion Engineering (CE), Standard Technical 
Specifications (STS), Surveillance Requirement (SR) 3.3.11.1 (post 
accident monitoring instrumentation). The monthly channel check 
supplements the shiftly level verification.
    The Basis of Specification 3.1 is revised to clarify 
expectations regarding a channel check of channels that are normally 
off scale when the surveillance is required. In this situation, the 
channel check only verifies that they are off scale in the same 
direction. Off scale low current loop channels are verified to be 
reading at the bottom of the range and not failed downscale. These 
statements are taken from the Bases of CE STS SR 3.3.4.1 Engineered 
Safety Features Actuation System (ESFAS) Instrumentation (Analog).
    In addition, the Basis of Specification 3.1 is revised to 
clarify that power operated relief valve (PORV) actuation is not 
required during the channel functional test of the PORV low 
temperature setpoint (Table 3-3, Item 18.a). PORV actuation is not 
required because it could depressurize the reactor coolant system. 
This clarification is modeled after a similar statement from the 
Bases of SR 3.4.12.6 (Low Temperature Overpressure Protection (LTOP) 
System) of the CE STS.
    Changing Specification 3.1, Tables 3-1, 3-2, 3-3, and 3-3A by 
using defined terms to enable the Surveillance Method to match the 
Surveillance Function is an administrative change designed to 
simplify the tables. Removal of the extraneous text does not alter 
the surveillance because the defined terms are equivalent in meaning 
to the deleted text.
    The reordering of several items in the tables into a Check-Test-
Calibrate sequence adds consistency to the tables. Text revisions in 
the Channel Description or Surveillance Function columns of Tables 
3-1 and 3-2 add clarity and/or consistency. Footnote No. 1 in Table 
3-1 concerning the bistable trip tester was deleted because it is 
unnecessary.
    The Surveillance Function of Table 3-1, Item 1.c (Power Range 
Safety Channels) is being changed to ``Test'' from ``Calibrate and 
Test.'' It is not necessary for Item 1.c to require both because 
Item 1.b already requires the power range safety channel adjustment 
(calibration) to be performed daily. As stated in the Basis of 
Specification 3.1, ``The minimum calibration frequencies of once-
per-day for the power range safety channels, ...are considered 
adequate.'' To further clarify the issue, the Basis of Specification 
3.1 is being revised to note that the daily calibration is a heat 
balance adjustment only.
    Changing Table 3-1, Item 4 (Thermal Margin/Low Pressure (TM/LP)) 
to use the defined term CHANNEL CALIBRATION will allow OPPD to relax 
the current TM/LP calibration requirements with a negligible impact 
on safety. Calibration of the temperature input and pressure input 
will still require calibration to known standards (i.e., resistance 
and pressure), but will allow the calibrations to be done separately 
instead of coincidently. The channel functional test that follows 
the channel calibration verifies proper function of the TM/LP 
circuitry.
    Removing the word ``Instruments'' from the Channel Description 
of Table 3-2, Item 14 makes the Channel Description consistent with 
the Surveillance Method. Table 3-2, Item 14 is not intended to 
verify safety injection tank (SIT) instrumentation operability but 
rather that the parameters level and pressure are within limits. 
Generic Letter (GL) 93-05, Item 7.4, states that the operability of 
SIT instrumentation is not directly related to the capability of a 
SIT to perform its safety function. GL 93-05 concludes that the 
surveillance should only confirm that the parameters defining SIT 
operability are within their specified limits.
    Items 22 & 24 are being added to Table 3-2 to clearly state the 
requirement for testing manual actuation of the Engineered Safety 
Features (ESF) channels for Off-site Power Low Signal (OPLS) and 
Auxiliary Feedwater. Although testing manual actuation of these 
channels is done via the existing Specifications, the requirement to 
do so is not clearly stated. Reordering Table 3-2, Item 23 into a 
Check-Test-Calibrate Surveillance Frequency sequence adds clarity 
and consistency.
    The addition of Footnote No. 7 to Table 3-2 clarifies that the 
refueling frequency ESF channel functional test pertains to the 
backup channels such as derived circuits and equipment that cannot 
be tested when the plant is at power. Operating certain relays 
during power operation could cause plant transients or equipment 
damage.
    The revisions to Table 3-3, Item 6, clarify that pressurizer 
level is the parameter to be verified and not the pressurizer level 
instruments. The revision to Item 6.a is consistent with CE STS SR 
3.4.9.1 (pressurizer water level). Reordering Item 6 into a Check-
Test-Calibrate Surveillance Function sequence makes Item 6 
consistent with the ordering of the other items in Table 3-3. The 
requirement to test the signal to alarm meter relay currently 
located in Specification 3.1, Table 3-3, Item 6.c is unnecessary. 
Operating experience has shown that a shiftly pressurizer level 
verification as proposed for Specification 3.1, Table 3-3, Item 6.a 
is sufficient to detect any level deviation and verify that 
operation is within safety analyses assumptions. Thus, the monthly 
``Test'' requirement will be replaced with a ``Check'' to supplement 
the less formal but more frequent shiftly level verification of Item 
6.a.
    Table 3-3, Items 21 (PORV Operation & Acoustic Position 
Indication Channel) and 23 (Safety Valve Acoustic Position 
Indication Channel) should be revised to a channel functional test 
from a channel/circuit check. An oscillator and installed impactors 
are used to generate noise signals and therefore, this surveillance 
is more accurately described as a channel functional test rather 
than a channel check.
    Table 3-3, Items 21 and 22 (PORV Block Valve Operation & 
Position Indication) should have the requirement to verify operation 
on the emergency power supply deleted. Permanent Class 1E power 
supplies the PORV and PORV Block Valve. Therefore, verification of 
PORV or PORV Block Valve operability while powered from the 
emergency power supply system provides no additional benefit. 
(Operability of the emergency power supply system is tested in 
accordance with Specification 3.7.) The proposed revision is in 
accordance with the exception for plants with a permanent Class 1E 
power supply to these valves as stated in CE STS, SR 3.4.11.4.
    Deletion of the requirement of TS 3.2, Table 3-5, Item 15, to 
test spent fuel pool surveillance coupons for a change in hardness 
corrects an oversight in the Application for Amendment dated 
December 7, 1992.
    As stated in the Safety Evaluation Report enclosed with 
Amendment 155, ``Each

[[Page 44362]]

coupon, upon its removal from the mounting jacket, will be analyzed 
according to the following tests:
    visual observation and photography
    neutron attenuation
    dimensional measurements (length, width, and thickness)
    weight and specific gravity.''
    The tests listed above are sufficient to detect degradation of 
the Boral material and do not require that the surveillance coupons 
be tested for hardness.
    Based on the above discussion, the proposed changes clarify and 
standardize existing surveillance requirements, remove redundant 
requirements, correct minor oversights from previous amendment 
requests or are in accordance with CE STS. Thus, none of the 
requested changes involve a significant increase in the probability 
or consequences of an accident previously evaluated.
    2. The proposed change does not create the possibility of a new 
or different kind of accident from any accident previously 
evaluated.
    The proposed revisions will not result in any physical 
alterations to the plant configuration, changes to setpoint values, 
or changes to the application of setpoints or limits. Therefore, the 
proposed changes do not create the possibility of a new or different 
kind of accident from any accident previously evaluated.
    3. The proposed change does not involve a significant reduction 
in a margin of safety.
    The proposed changes clarify existing surveillance requirements, 
remove redundant requirements, correct minor oversights from 
previous amendment requests or are in accordance with CE STS. Thus, 
none of the requested changes involves 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: W. Dale Clark Library, 215 
South 15th Street, Omaha, Nebraska 68102
    Attorney for licensee: Perry D. Robinson, Winston & Strawn, 1400 L 
Street, N.W., Washington, DC 20005-3502
    NRC Project Director: William H. Bateman

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

    Date of amendment request: May 20, 1996, as supplemented by letter 
dated July 25, 1996
    Description of amendment request: This amendment request would 
modify the Technical Specifications for the unit by: changing the 
Minimum Critical Power Ratio safety limit values, adding a reference to 
reflect the use of the ANF-B Critical Power Correlation, and modifying 
the associated Bases.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:
    1. The proposed change does not involve a significant increase 
in the probability or consequences of an accident previously 
evaluated.
    The change to the ANFB correlation and corresponding MCPR Safety 
Limits does not physically change the plant systems, structures, or 
components. Thus, the probability of an event evaluated in the SAR 
is not increased. The acceptance criterion for the MCPR Safety Limit 
(i.e., 99.9% of the fuel rods expected to avoid boiling transition) 
is not changed. Only the methodology used to demonstrate compliance 
is changed.
    Therefore, the consequences of anticipated operational 
occurrences (which must show the Safety Limit is not violated) are 
not changed. Results of incorporating this change will not 
significantly increase 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.
    As stated above, this methodology change does not impact the 
acceptance criteria for the MCPR Safety Limits and does not 
physically change the plant systems, structures, or components. 
Since no changes to the physical plant are being made, this change 
does not create the possibility of a new or different kind of 
accident from any accident previously evaluated.
    3. The proposed change does not involve a significant reduction 
in a margin of safety.
    A cycle specific MCPR Safety Limit analysis was performed by SPC 
[Siemens Power Corporation]. This analysis used NRC approved methods 
described in the SPC report: ANF-524(P)(A), Revision 2 and 
Supplement 1, Revision 2. The MCPR Safety Limit value is calculated 
such that at least 99.9% of the fuel rods are expected to avoid 
boiling transition during normal operation or anticipated operation 
occurrences. Both the existing analysis using XN-3 and the new 
analysis using ANFB utilize NRC approved methods to accomplish this 
same objective. Therefore, the change to an ANFB based Safety Limit 
does not involve a significant reduction in a margin of safety.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: Osterhout Free Library, 
Reference Department, 71 South Franklin Street, Wilkes-Barre, PA 18701
    Attorney for licensee: Jay Silberg, Esquire, Shaw, Pittman, Potts 
and Trowbridge, 2300 N Street NW., Washington, DC 20037
    NRC Project Director: John F. Stolz
    TU Electric Company, Docket Nos. 50-445 and 50-446, Comanche 
Peak Steam Electric Station (CPSES), Units 1 and 2, Somervell 
County, Texas
    Date of amendment request: July 31, 1996
    Brief description of amendments: Based on analyses of the core 
configuration and expected operation for CPSES Unit 1, Cycle 6, the 
proposed amendments would revise core safety limit curves and 
Overtemperature N-16 reactor trip setpoints. In addition, the TU 
Electric Small Break LOCA Topical Report on the Core Operating Limits 
Report Technical Specification is incorporated. The topical report 
change is applicable to both Units.
    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.a. Revision to the Unit 1 Core Safety Limits
    Analyses of reactor core safety limits are required as part of 
reload calculations for each cycle. TU Electric has performed the 
analyses of the Unit 1, Cycle 6 core configuration to determine the 
reactor core safety limits. The methodologies and safety analysis 
values result in new operating curves which, in general, permit 
plant operation over a similar range of acceptable conditions. This 
change means that if a transient were to occur with the plant 
operating at the limits of the new curve, a different temperature 
and power level might be attained than if the plant were operating 
within the bounds of the old curves. However, since the new curves 
were developed using NRC approved methodologies which are wholly 
consistent with and do not represent a change in the Technical 
Specification BASES for safety limits, all applicable postulated 
transients will continue to be properly mitigated. As a result, 
there will be no significant increase in the consequences, as 
determined by accident analyses, of any accident previously 
evaluated.
    1.b. Revision to Unit 1 Overtemperature N-16 Reactor Trip 
Setpoints, Parameters and Coefficients
    As a result of changes discussed, the Overtemperature N-16 
reactor trip setpoint has been recalculated. These trip setpoints 
help ensure that the core safety limits are maintained and that all 
applicable limits of the safety analysis are met.
    Based on the calculations performed, the safety analysis value 
for Overtemperature N-16 reactor trip setpoint has changed. This 
essentially means if a transient were to occur, the actual 
temperature and power level achievable prior to initiating a reactor 
trip could be slightly higher. However, the analyses performed show 
that, using the TU Electric methodologies, all applicable limits of 
the safety analysis are met. This setpoint

[[Page 44363]]

provides a trip function which allows the mitigation of postulated 
accidents and has no impact on accident initiation. Therefore, the 
changes in safety analysis values do not involve an increase in the 
probability of an accident and, based on satisfying all applicable 
safety analysis limits, there is no significant increase in the 
consequences of any accident previously evaluated.
    In addition, sufficient operating margin has been maintained in 
the overtemperature setpoint such that the risk of turbine runbacks 
or reactor trips due to upper plenum flow anomalies or other 
operational transients will be minimized, thus reducing potential 
challenges to the plant safety systems.
    1.c. Incorporation of TU Electric Small Break LOCA Topical 
Report, RXE-95-0001-P.
    TU Electric has submitted the topical report ``Small Break Loss 
of Coolant Accident Analysis Methodology,'' RXE-95-001-P and plans 
to use the report to support Unit 1 Cycle 6. In order to accomplish 
this activity, it is necessary to include the topical report in the 
list of NRC-approved methodologies in Technical Specification 
6.9.1.6b. Use of this topical report is contingent upon NRC 
approval; therefore, inclusion of this report in Section 6 of the 
Technical Specifications is administrative in nature and does not 
change the probability or consequences of an accident.
    2. The proposed changes involve the use of revised safety 
analysis values and the calculation of new reactor core safety 
limits and reactor trip setpoints. As such, the changes play an 
important role in the analysis of postulated accidents but none of 
the changes effect plant hardware or the operation of plant systems 
in a way that could initiate an accident. Therefore, the proposed 
changes do not create the possibility of a new or different kind of 
accident from any accident previously evaluated.
    3. In reviewing and approving the methods used for safety 
analyses and calculations, the NRC has approved the safety analysis 
limits which establish the margin of safety to be maintained. While 
the actual impact on safety is discussed in response to question 1, 
the impact on margin of safety is discussed below:
    3.a.
    Revision to the Unit 1 Reactor Core Safety Limits
    The TU Electric reload analysis methods have been used to 
determine new reactor core safety limits. All applicable safety 
analysis limits have been met. The methods used are wholly 
consistent with Technical Specification BASES 2.1 which is the bases 
for the safety limits. In particular, the curves assure that for 
Unit 1, Cycle 6, the calculated DNBR is no less than the safety 
analysis limit and the average enthalpy at the vessel exit is less 
than the enthalpy of saturated liquid. The acceptance criteria 
remains valid and continues to be satisfied; therefore, no change in 
a margin of safety occurs.
    3.b. Revision to Unit 1 Overtemperature N-16 Reactor Trip 
Setpoints, Parameters and Coefficients
    Because the reactor core safety limits for CPSES Unit 1, Cycle 6 
are recalculated, the Reactor Trip System instrumentation setpoint 
values for the Overtemperature N-16 reactor trip setpoint which 
protect the reactor core safety limits must also be recalculated. 
The Overtemperature N-16 reactor trip setpoint helps prevent the 
core and Reactor Coolant System from exceeding their safety limits 
during normal operation and design basis anticipated operational 
occurrences. The most relevant design basis analysis in Chapter 15 
of the CPSES Final Safety Analysis Report (FSAR) which is affected 
by the change in the safety analysis value for the CPSES Unit 1 
Overtemperature N-16 reactor trip setpoint is the Uncontrolled Rod 
Cluster Control Assembly Bank Withdrawal at Power (FSAR Section 
15.4.2). This event has been re-analyzed with the revised safety 
analysis value for the Overtemperature N-16 reactor trip setpoint to 
demonstrate compliance with event specific acceptance criteria. 
Because all event acceptance criteria are satisfied, there is no 
degradation in a margin of safety.
    The nominal Reactor Trip System instrumentation setpoints values 
for the Overtemperature N-16 reactor trip setpoint (Technical 
Specification Table 2.2-1) are determined based on a statistical 
combination of all of the uncertainties in the channels to arrive at 
a total uncertainty. The total uncertainty plus additional margin is 
applied in a conservative direction to the safety analysis trip 
setpoint value to arrive at the nominal and allowable values 
presented in Technical Specification Table 2.2-1. Meeting the 
requirements of Technical Specification Table 2.2-1 assures that the 
Overtemperature N-16 reactor trip setpoint assumed in the safety 
analyses remains valid. The CPSES Unit 1, Cycle 6 Overtemperature N-
16 reactor trip setpoint is different from previous cycles which 
provides more operational flexibility to withstand mild transients 
without initiating automatic protective actions. Although the 
setpoint is different, the Reactor Trip System instrumentation 
setpoint values for the Overtemperature N-16 reactor trip setpoint 
are consistent with the safety analysis assumption which has been 
analytically demonstrated to be adequate to meet the applicable 
event acceptance criteria. Thus, there is no reduction in a margin 
of safety.
    3.c. Revise 6.9.1.6b to include Topical Report RXE-95-001-P, 
``Small Break Loss of Coolant Accident Methodology''
    TU Electric has submitted the topical report ``Small Break Loss 
of Coolant Accident Analysis Methodology,'' RXE-95-001-P and plans 
to use the report to support Unit 1 Cycle 6. In order to accomplish 
this activity, it is necessary to include the topical report in the 
list of NRC-approved methodologies in Technical Specification 
6.9.1.6b. Use of this topical report is contingent upon NRC 
approval; therefore, inclusion of this report in Section 6 of the 
Technical Specifications is administrative in nature and does not 
reduce the margin of safety.
    Using the NRC approved TU Electric methods, the reactor core 
safety limits are determined such that all applicable limits of the 
safety analyses are met. Because the applicable event acceptance 
criteria continue to be met, there is no significant reduction in 
the margin of safety.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: University of Texas at 
Arlington Library, Government Publications/Maps, 702 College, P.O. Box 
19497, Arlington, TX 76019
    Attorney for licensee: George L. Edgar, Esq., Morgan, Lewis and 
Bockius, 1800 M Street, N.W., Washington, DC 20036
    NRC Project Director: William D. Beckner

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

    Date of amendment request: July 31, 1996
    Brief description of amendments: The proposed amendments would 
revise the Technical Specifications by (1) changing the battery charger 
ratings; (2) by clarifying the meaning of the term ``associated 
inverter''; and by (3) deleting the protection channel and the vital 
bus ratings for the instrument busses identified for Mode 1 through 4. 
These changes are associated with a plant modification in which the 
inverters and battery chargers are being replaced and an installed 
spare inverter is being added for each safety train. These changes are 
equally applicable to CPSES Units 1 and 2.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:
    1. DO THE PROPOSED CHANGES INVOLVE A SIGNIFICANT INCREASE IN THE 
PROBABILITY OR CONSEQUENCES OF AN ACCIDENT PREVIOUSLY EVALUATED?
    CHANGE TO IDENTIFY BATTERY CHARGER RATINGS
    The first proposed change replaces the test amperes with the 
design value for the replacement battery charger and allows a 
voltage range (greater than or equal to 130 volts) instead of a 
single value. The intent of the surveillance requirement or the 
surveillance frequency is not changed. The replacement inverters and 
battery chargers will continue to provide the capacity needed to 
perform the required safety functions. The revised surveillance will 
continue to assure that the battery chargers are capable of 
performing as designed. Therefore this change does not impact the 
probability or the consequences of an accident previously evaluated.
    CLARIFICATION TO DEFINE ASSOCIATED INVERTER
    The second proposed change adds a foot note to clarify the term 
``associated inverter'' by describing it as, ''... the dedicated 
inverter or installed spare inverter.'' Also the Bases

[[Page 44364]]

for this specification is revised to reflect the basis for this 
change. This change allows use of an installed spare inverter (for 
each train) having the capability to energize the Instrument Bus for 
the protection channel or the vital bus. Procedural controls and 
interlocks ensure that the spare is available to feed only one of 
the protection channel or vital bus Instrument Bus at a time, in the 
event the dedicated inverter is not available. Procedural controls 
and interlocks also ensure that the installed spare inverter is fed 
from the same power source as that of the dedicated inverter not in 
service and whose loads are being fed by the spare inverter. This 
proposed design only allows the spare inverter for a safety train to 
be manually aligned to replace only one of the four inverters in 
that train at a time.
    The installation of a spare inverter for each train and the 
associated design configuration increases the availability of 
energized Instrument Bus for the protection channel and vital bus. 
These changes do not involve an increase in the probability or 
consequences of an accident previously evaluated.
    DELETION OF THE PROTECTION CHANNEL AND VITAL BUS RATINGS FOR 
INSTRUMENT BUS
    The third proposed change deletes specifying of the protection 
channel and vital bus KVA ratings for the Instrument Bus. The 
ratings of inverter that feeds these instrument buses are being 
described in other Licensing Bases Documents or Design Basis 
Documents. There is no change proposed to the intent of the action 
statements.
    This is considered an administrative change and does not impact 
the probability or consequences of an accident previously evaluated.
    2. DO THE PROPOSED CHANGES CREATE THE POSSIBILITY OF A NEW OR 
DIFFERENT KIND OF ACCIDENT FROM ANY ACCIDENT PREVIOUSLY EVALUATED?
    CHANGE TO IDENTIFY BATTERY CHARGER RATINGS
    Replacing the inverters and battery chargers and changing the 
parameters of the battery charger surveillance test to match the 
replacement chargers does not alter the functional modes of this 
portion of the design and does not result in any new failure modes. 
As such, it does not create the possibility of a new or different 
accident from any previously evaluated.
    CLARIFICATION TO DEFINE ASSOCIATED INVERTER
    The second proposed change allows use of an installed spare 
inverter for each train to energize the one of the Instrument Bus 
for the protection channel and vital bus at a time for the 
respective safety train while its dedicated inverter is not 
available. The spare inverter is such that it has the capability to 
support the maximum load for the protection channel or vital bus. 
Manually aligning the installed inverter to replace on[e] of the 
dedicated inverters is essentially equivalent to a repair activity 
which replaces a faulted inverter with a new inverter. In addition, 
procedural controls and interlocks are provided to ensure the proper 
alignment of the installed spare when it is used. The proposed 
changes do not create the possibility of a new or different accident 
from any previously evaluated.
    DELETION OF THE PROTECTION CHANNEL AND VITAL BUS RATINGS FOR 
INSTRUMENT BUS
    The third proposed change as discussed earlier does not change 
intent of the Technical Specifications action statements. This is an 
administrative change which does not introduce new failure modes and 
has no new or different accidents from any previously evaluated are 
created.
    3. DO THE PROPOSED CHANGES INVOLVE A SIGNIFICANT REDUCTION IN 
MARGIN OF SAFETY?
    The relevant Technical Specification sections proposed for 
changes: (1) ensure that the battery charger is capable of charging 
the battery by performing the surveillance at 18 month frequency; 
(2) establish operability requirements of the Instrument Bus for the 
protection channel and vital bus in MODES 1 through 6; and (3) 
identify the actions required for not meeting item 2.
    These proposed changes do not alter the intent of the above 
requirements; however replacement of the currently installed 
inverters with inverters which are expected to be more reliable and 
available and the addition of a spare inverter per safety train to 
energize Instrument Bus for protection channel and vital bus does 
increase the reliability of the instrument busses for the train. 
Allowing credit for this spare inverter in meeting the operability 
requirements of Instrument Bus for the protection channel and vital 
bus, minimize potential plant shutdowns due to non-energized 
instrument from its dedicated inverter. These changes do not involve 
a significant reduction in margin of safety.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: University of Texas at 
Arlington Library, Government Publications/Maps, 702 College, P.O. Box 
19497, Arlington, TX 76019
    Attorney for licensee: George L. Edgar, Esq., Morgan, Lewis and 
Bockius, 1800 M Street, N.W., Washington, DC 20036
    NRC Project Director: William D. Beckner

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

    Date of amendment request: August 9, 1996
    Description of amendment request: The proposed amendment would 
revise the Safety Limits for Minimum Critical Power Ratio (MCPR) based 
upon a Vermont Yankee plant and cycle specific analysis, performed by 
General Electric. The revised MCPR Safety Limits are needed to 
accommodate Vermont Yankee's core design for upcoming refueling cycle 
number 19. Specifically, the MCPR Safety Limits of 1.07 and 1.08 in the 
Vermont Yankee Technical Specifications (TS) section 1.1.A are proposed 
to be increased to 1.10 and 1.12 for two loop and single loop 
operation, respectively.
    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 Safety Limit Minimum Critical Power Ratio (MCPR) is 
defined to ensure that during normal operation and Anticipated 
Operational Transients (AOTs), at least 99.9% of the fuel rods in 
the core do not experience transition boiling. Core MCPR operating 
limits are developed to ensure these Safety Limits are maintained in 
the event of the worst case transient. Since the Safety Limit MCPR 
will be maintained at all times, operation under the proposed 
changes will ensure at least 99.9% of the fuel rods in the core do 
not experience transition boiling and no significant radiological 
release will result. Therefore, this Safety Limit MCPR change does 
not affect the probability or consequences of a previously evaluated 
accident.
    (2) The proposed changes do not involve any new modes of 
operation or any plant modifications. Establishment and monitoring 
of the operating limits will continue as per established procedure. 
The proposed changes to these limits do not result in the creation 
of any new precursors to an accident. Therefore, the proposed change 
does not create the possibility of a new or a different kind of 
accident from any previously analyzed.
    (3) The Safety Limit MCPR values were evaluated by General 
Electric based upon a cycle specific Vermont Yankee analysis, using 
NRC approved methods. The resulting limits are more conservative 
than the previous generic limits and will continue to assure that at 
least 99.9% of the fuel rods in the core do not experience 
transition boiling during analyzed transients. This acceptance 
criteria ensures the safety design limit of ``no damage to a nuclear 
system process barrier shall result from forces associated with 
AOTs.'' Therefore, the implementation of the proposed change does 
not involve a significant reduction in [a] margin of safety.
    The NRC staff has reviewed the licensee's analysis. The staff notes 
that, although the proposed change does not involve a plant 
modification, the reason for the proposed higher safety limit MCPRs is 
the cycle-specific core design and the local power distribution in the 
slightly higher enriched fresh GE-9B fuel bundles. This new fuel will 
be loaded during the September/October 1996 refueling outage. In 
conjunction with the proposed safety limit MCPRs and the core operating 
limits determined in accordance with Vermont Yankee TS 6.7.A.4, the new 
fuel load will not involve a significant increase in the probability or 
consequences of an

[[Page 44365]]

accident previously evaluated nor a significant reduction in a margin 
of safety. In addition, the new fuel load does not create the 
possibility of a new or different kind of accident from any accident 
previously evaluated. 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: Brooks Memorial Library, 224 
Main Street, Brattleboro, VT 05301
    Attorney for licensee: R. K. Gad, III, Ropes and Gray, One 
International Place, Boston, MA 02110-2624
    NRC Project Director: Jocelyn A. Mitchell, Acting Director

Notice Of Issuance Of Amendments To Facility Operating Licenses

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

Boston Edison Company, Docket No. 50-293, Pilgrim Nuclear Power 
Station, Plymouth County, Massachusetts

    Date of application for amendment: May 1, 1996
    Brief description of amendment: The proposed amendment will modify 
the definition of ``Core Alteration,'' and the limiting condition for 
operation, Surveillance conditions and Bases section associated with 
Technical Specification 3.7.C, ``Secondary Containment.''
    Date of issuance: August 12, 1996
    Effective date: August 12, 1996
    Amendment No.: 166
    Facility Operating License No. DPR-35: Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: June 5, 1996 (61 FR 
28606) The Commission's related evaluation of the amendment is 
contained in a Safety Evaluation dated August 12, 1996. No significant 
hazards consideration comments received: No
    Local Public Document Room location:  Plymouth Public Library, 11 
North Street, Plymouth, Massachusetts 02360.

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

    Date of application for amendment: July 17, 1995, as supplemented 
May 2, 1996, and July 1, 1996.
    Brief description of amendment: The change revises technical 
specification (TS) section 3.8 to specify that the spent fuel building 
refueling filter fan and at least one containment purge fan shall be 
shown to operate within plus or minus 10 percent of the design flow.
    Date of issuance: August 6, 1996
    Effective date: August 6, 1996
    Amendment No. 172
    Facility Operating License No. DPR-23. Amendment revises the 
Technical Specifications.
    Date of initial notice in Federal Register: September 13, 1995 (60 
FR 47615). The May 2, and July 1, 1996, letters provided clarifying 
information that did not affect the proposed no significant hazards 
consideration. The Commission's related evaluation of the amendment is 
contained in a Safety Evaluation dated August 6, 1996. No significant 
hazards consideration comments received: No
    Local Public Document Room location: Hartsville Memorial Library, 
147 West College Avenue, Hartsville, South Carolina 29550

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

    Date of application for amendment: June 6, 1996
    Brief description of amendment: The amendment revises technical 
specifications (TS) Section 4.2.3 to allow the licensee to defer the 
ultrasonic inspection of the reactor coolant pump flywheel for one 
operating cycle.
    Date of issuance: August 9, 1996
    Effective date: August 9, 1996
    Amendment No. 173
    Facility Operating License No. DPR-23. Amendment revises the 
Technical Specifications.
    Date of initial notice in Federal Register: July 3, 1996 (61 FR 
34888) The Commission's related evaluation of the amendment is 
contained in a Safety Evaluation dated August 9, 1996. No significant 
hazards consideration comments received: No
    Local Public Document Room location: Hartsville Memorial Library, 
147 West College Avenue, Hartsville, South Carolina 29550

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

    Date of application for amendment: May 31, 1996
    Brief description of amendment: The amendment revises Technical 
Specifications (TS) Table 3.3-7, Seismic Monitoring Instrumentation, 
and TS Table 4.3-4, Seismic Monitoring Instrumentation Surveillance 
Requirements, to correct the location described for one of the three 
Triaxial Peak Accelerograph recorders.
    Date of issuance: August 7, 1996
    Effective date: August 7, 1996
    Amendment No. 66
    Facility Operating License No. NPF-63. Amendment revises the 
Technical Specifications
    Date of initial notice in Federal Register: July 3, 1996 (61 FR 
34888) The Commission's related evaluation of the amendment is 
contained in a Safety Evaluation dated August 7, 1996. No significant 
hazards consideration comments received: No
    Local Public Document Room location: Cameron Village Regional 
Library, 1930 Clark Avenue, Raleigh, North Carolina 27605.

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 16, 1996
    Brief description of amendments: The amendments revise the 
Technical

[[Page 44366]]

Specifications (TSs) to eliminate selected response time testing 
requirements based on analyses performed by the Boiling Water Reactor 
Owners' Group as documented in NEDO-32291. The affected TS sections are 
3/4.3.1, ``Reactor Protection System Instrumentation;'' 3/4.3.2, 
``Isolation Actuation Instrumentation;'' and 3/4.3.3, ``Emergency Core 
Cooling System Actuation Instrumentation.''
    Date of issuance: August 14, 1996
    Effective date: Immediately, to be implemented within 60 days.
    Amendment Nos.: 114 and 99
    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 
25702) The Commission's related evaluation of the amendments is 
contained in a Safety Evaluation dated August 14, 1996. No significant 
hazards consideration comments received: No
    Local Public Document Room location: Jacobs Memorial Library, 
Illinois Valley Community College, Oglesby, Illinois 61348.

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

    Date of application for amendment: December 21, 1995
    Brief description of amendment: The amendment revises the Technical 
Specifications (TS) to implement 10 CFR Part 50, Appendix J - Option B, 
by referring to Regulatory Guide 1.163, ``Performance-Based Containment 
Leak-Test Program.'' Specifically, changes have been made to TS Section 
3/4.6.1.2, ``Primary Containment Leakage,'' TS 3/4.6.1.3, ``Primary 
Containment Air Locks,'' TS 3/4.6.1.5, ``Primary Containment Structural 
Integrity,'' TS 6.0, ``Administrative Controls,'' and their associated 
Bases.
    Date of issuance: August 8, 1996
    Effective date: August 8, 1996, with full implementation within 45 
days.
    Amendment No.: 108
    Facility Operating License No. NPF-43. Amendment revises the 
Technical Specifications.
    Date of initial notice in Federal Register: February 28, 1996 (61 
FR 7551) The Commission's related evaluation of the amendment is 
contained in a Safety Evaluation dated August 8, 1996. No significant 
hazards consideration comments received: No.
    Local Public Document Room location: Monroe County Library System, 
3700 South Custer Road, Monroe, Michigan 48161

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: April 19, 1996, and supplements 
dated May 10 and May 28, 1996.
    Brief description of amendments: The amendment changes the 
Technical Specifications to address frequency extension on a periodic 
basis, deletes separate notification requirements for an inoperable 
startup transformer, and allows the operating residual heat removal 
loop to be removed from operation, under certain conditions, during 
refueling.
    Date of Issuance: August 6, 1996
    Effective Date: August 6, 1996
    Amendment Nos.: 189 and 183Facility 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 
34892) The Commission's related evaluation of the amendments is 
contained in a Safety Evaluation dated August 6, 1996. No significant 
hazards consideration comments received: No.
    Local Public Document Room location: Florida International 
University, University Park, Miami, Florida 33199.

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

    Date of application for amendments: July 26, 1995, and supplemented 
March 13, May 3, and May 9, 1996.
    Brief description of amendments: Change TS 6.9.1.7, Core Operating 
Limits Report, resulting from a reanalysis of the small break loss-of-
coolant accident for the Turkey Point Units using the NOTRUMP code 
including the COSI safety injection (SI) condensation model.
    Date of issuance: August 13, 1996
    Effective date: August 13, 1996
    Amendment Nos. 190 and 184Facility Operating Licenses Nos. DPR-31 
and DPR-41: Amendments revised the Technical Specifications.
    Date of initial notice in Federal Register: September 13, 1995 (60 
FR 47618). The supplements dated March 13, May 3, and May 9, 1996 
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 a 
Safety Evaluation dated August 13, 1996. No significant hazards 
consideration comments received: No
    Local Public Document Room location: Florida International 
University, University Park, Miami, Florida 33199.

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

    Date of amendment request: May 1, 1996
    Brief description of amendments: The amendments changed the 
technical specifications to implement 10 CFR Part 50, Appendix J, 
Option B, by referring to Regulatory Guide 1.163, ``Performance-Based 
Containment Leak-Test Program.'' Part of the requested change, that 
regarding the frequency of leakage rate testing the normal containment 
purge valves and the supplementary containment purge valves, was 
denied.
    Date of issuance: August 13, 1996
    Effective date: August 13, 1996
    Amendment Nos.: 84 and 71
    Facility Operating License Nos. NPF-76 and NPF-80. The amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: June 5, 1996 (61 FR 
28616) The Commission's related evaluation of the amendments is 
contained in a Safety Evaluation dated August 13, 1996. No significant 
hazards consideration comments received: No
    Local Public Document Room location: Wharton County Junior College, 
J. M. Hodges Learning Center, 911 Boling Highway, Wharton, TX 77488

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

    Date of application for amendment: November 30, 1995
    Brief description of amendment: The amendment implements the Option 
I-D long-term stability solution and removes the existing SIL-380 Rev. 
1-based specifications. In addition, the amendment requires a plant 
scram be initiated should the plant enter natural circulation 
conditions and prohibits restarting a recirculation pump while in 
natural circulation. Finally, this amendment deletes Technical 
Specification (TS) actions and surveillance requirements related to 
core plate differential pressure noise while in single recirculation 
pump operation (SLO).
    Date of issuance: August 7, 1996
    Effective date: August 7, 1996
    Amendment No.: 215
    Facility Operating License No. DPR-49: Amendment revised the 
Technical Specifications.

[[Page 44367]]

    Date of initial notice in Federal Register: March 13, 1996 (61 FR 
10394) The Commission's related evaluation of the amendment is 
contained in a Safety Evaluation dated August 7, 1996. No significant 
hazards consideration comments received: No.
    Local Public Document Room location: Cedar Rapids Public Library, 
500 First Street, S. E., Cedar Rapids, Iowa 52401

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

    Date of application for amendment: November 15, 1995, as 
supplemented April 9, 1996
    Brief description of amendment: The amendment revises the 
requirements for the End of Cycle Recirculation Pump Trip logic to 
match more closely the assumptions applicable to the turbine trip 
events for which it was installed. The surveillance requirements are 
also revised, based on those same assumptions.
    Date of issuance: August 8, 1996
    Effective date: August 8, 1996
    Amendment No.: 216
    Facility Operating License No. DPR-49: Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: January 22, 1996 (61 FR 
1629) The April 9, 1996, submittal was clarifying in nature and did not 
affect the no significant hazards determination. The Commission's 
related evaluation of the amendment is contained in a Safety Evaluation 
dated August 8, 1996. No significant hazards consideration comments 
received: No.
    Local Public Document Room location:  Cedar Rapids Public Library, 
500 First Street, S. E., Cedar Rapids, Iowa 52401

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

    Date of application for amendment: January 18, 1996
    Brief description of amendment: The amendment revises the setpoint 
at which the Reactor Water Cleanup (RWCU) system isolates, based on 
reactor vessel water level. In particular, the amendment changes the 
Group 5 isolation from isolating on ``reactor water level low'' to 
``reactor water level low-low.''
    Date of issuance: August 8, 1996
    Effective date: August 8, 1996, and shall be implemented prior to 
startup from RFO 14.
    Amendment No.: 217
    Facility Operating License No. DPR-49: Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: February 14, 1996 (61 
FR 5814) The Commission's related evaluation of the amendment is 
contained in a Safety Evaluation dated August 8, 1996. No significant 
hazards consideration comments received: No.
    Local Public Document Room location: Cedar Rapids Public Library, 
500 First Street, S. E., Cedar Rapids, Iowa 52401

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

    Date of application for amendments: January 12, 1996 (AEP:NRC:1233)
    Brief description of amendments: The amendments modify the 
Technical Specifications to delete the surveillance requirement 
demonstrating operability of the emergency power supply for the 
pressurizer power operated relief valves and block valves.
    Date of issuance: August 15, 1996
    Effective date: August 15, 1996, with full implementation within 45 
days
    Amendment Nos.: 211 and 196
    Facility Operating License Nos. DPR-58 and DPR-74. Amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: February 28, 1996 (61 
FR 7554) The Commission's related evaluation of the amendments is 
contained in a Safety Evaluation dated August 15, 1996. No significant 
hazards consideration comments received: No.
    Local Public Document Room location:  Maud Preston Palenske 
Memorial Library, 500 Market Street, St. Joseph, Michigan 49085

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

    Date of application for amendment: February 7, 1996, as 
supplemented July 26, 1996.
    Brief description of amendment: The amendment revises the operating 
license, TSs and associated Bases to implement Option B ``Performance-
Based Requirements'' of Appendix J to 10 CFR Part 50 for Type A, B, and 
C leakage rate testing.
    Date of issuance: August 13, 1996
    Effective date: As of the date of issuance to be implemented within 
30 days.
    Amendment No.: 74
    Facility Operating License No. NPF-69: Amendment revises the 
Technical Specifications and operating license.
    Date of initial notice in Federal Register: May 8, 1996 (61 FR 
20849) The Commission's related evaluation of the amendment is 
contained in a Safety Evaluation dated August 13, 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: July 3, 1996
    Brief description of amendment: The amendment removes, on a one-
time basis during the cycle 13 mid-cycle offload/reload activities, the 
Technical Specification (TS) requirement that the boron concentration 
in all filled portions of the reactor coolant system be ``uniform.'' 
The requested change also adds a footnote indicating that it is 
acceptable for the boron concentration of the water volumes in the 
steam generators and the connecting piping to be as low as 1300 parts 
per million. The TS Bases are also updated to reflect the one-time TS 
change.
    Date of issuance: August 12, 1996
    Effective date: As of the date of issuance, to be implemented 
within 30 days.
    Amendment No.: 201
    Facility Operating License No. DPR-65. Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: July 11, 1996 (61 FR 
36583) The Commission's related evaluation of the amendment is 
contained in a Safety Evaluation dated August 12, 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 Waterford Library, ATTN: Vince Juliano, 49 Rope 
Ferry Road, Waterford, CT 06385

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

    Date of application for amendments: February 29, 1996
    Brief description of amendments: These amendments relocate 
Specification 3/4.9.6, ``Refueling Platform,'' to the Susquehanna Steam 
Electric Station Technical Requirements Manual, a document which is 
controlled under the requirements of 10 CFR 50.59.
    Date of issuance: August 13, 1996
    Effective date: August 13, 1996
    Amendment Nos.: 159 and 130

[[Page 44368]]

    Facility Operating License Nos. NPF-14 and NPF-22. The amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: April 10, 1996 (61 FR 
15992) The Commission's related evaluation of the amendments is 
contained in a Safety Evaluation dated August 13, 1996. No significant 
hazards consideration comments received: No
    Local Public Document Room location: Osterhout Free Library, 
Reference Department, 71 South Franklin Street, Wilkes-Barre, 
Pennsylvania 18701.

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 20, 1996 (TS 373)
    Brief description of amendment: The amendments incorpore the 
guidance of Generic Letter 87-09 in the technical specifications, 
allowing a 24-hour delay in implementing action requirements due to a 
missed surveillance requirement.
    Date of issuance: August 5, 1996
    Effective Date: August 5, 1996
    Amendment Nos.: 230, 245 and 205
    Facility Operating License Nos. DPR-33, DPR-52 and DPR-68: 
Amendments revised the Technical Specifications.
    Date of initial notice in Federal Register: June 19, 1996 (61 FR 
31185) The Commission's related evaluation of the amendment is 
contained in a Safety Evaluation dated August 5, 1996. No significant 
hazards consideration comments received: None
    Local Public Document Room location: Athens Public library, South 
Street, Athens, Alabama 35611

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

    Date of application for amendment: May 29, 1996
    Brief description of amendment: The amendment authorizes revision 
of the Final Safety Analysis Report (FSAR) to incorporate a 
modification to the facility that will reduce the single failure trip 
potential for the main feedwater and bypass valves.
    Date of issuance: August 13, 1996
    Effective date: August 13, 1996
    Amendment No.: 115
    Facility Operating License No. NPF-30: The amendment revised the 
Final Safety Analysis Report.
    Date of initial notice in Federal Register: July 3, 1996 (61 FR 
34900) The Commission's related evaluation of the amendment is 
contained in a Safety Evaluation dated August 13, 1996. No significant 
hazards consideration comments received: No.
    Local Public Document Room location: Callaway County Public 
Library, 710 Court Street, Fulton, Missouri 65251.

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

    Date of application for amendment: June 4, 1996
    Brief description of amendment: The amendment revises the Technical 
Specifications by reducing the surveillance test frequencies for the 
radiation monitoring system (Table TS 4.1-1) and the control rods 
(Table TS 4.1-3) in accordance with the guidance of Generic Letter 93-
05, ``Line-Item Technical Specifications Improvements to Reduce 
Surveillance Requirements for Testing During Power Operation,'' dated 
September 27, 1993.
    Date of issuance: August 7, 1996
    Effective date: August 7, 1996
    Amendment No.: 125
    Facility Operating License No. DPR-43: Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: July 3, 1996 (61 FR 
34901) The Commission's related evaluation of the amendment is 
contained in a Safety Evaluation dated August 7, 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

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

    Date of amendment request: July 29, 1994, as superseded by letter 
dated September 15, 1995, and subsequently supplemented by letters 
dated March 8, 1996, April 18, 1996, June 14, 1996, and July 12, 1996.
    Brief description of amendment: The amendment revises TS 3/4.8.1, 
``Electric Power Systems - A.C. Sources,'' and its associated Bases to 
achieve an overall improvement in emergency diesel generator 
reliability and availability.
    Date of issuance: August 9, 1996
    Effective date: August 9, 1996, to be implemented within 90 days of 
the date of issuance.
    Amendment No.: 101
    Facility Operating License No. NPF-42. The amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: May 22, 1996 (61 FR 
25716) The June 14, 1996, and July 12, 1996, supplemental letters 
provided Bases page changes and did not change the initial no 
significant hazards consideration determination. The Commission's 
related evaluation of the amendment is contained in a Safety Evaluation 
dated August 9, 1996. No significant hazards consideration comments 
received: No.
    Local Public Document Room locations: Emporia State University, 
William Allen White Library, 1200 Commercial Street, Emporia, Kansas 
66801 and Washburn University School of Law Library, Topeka, Kansas 
66621
    Dated at Rockville, Maryland, this 21st day of August 1996.
    For the Nuclear Regulatory Commission
Steven A. Varga,
Director, Division of Reactor Projects - I/II Office of Nuclear Reactor 
Regulation
[Doc. 96-21813 Filed 8-27-96; 8:45 am]
BILLING CODE 7590-01-F