[Federal Register Volume 62, Number 19 (Wednesday, January 29, 1997)]
[Notices]
[Pages 4341-4360]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: X97-10129]


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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 January 4, 1997, through January 16, 1997. 
The last biweekly notice was published on January 15, 1997 (62 FR 
2185).

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 February 28, 1997, the licensee may file a request for a hearing 
with respect to issuance of the amendment to the subject facility 
operating license and any person whose interest may be affected by this 
proceeding and who wishes to participate as a party in the proceeding 
must file a written request for a hearing and a petition for leave to 
intervene. Requests for a hearing and a petition for leave to intervene 
shall be filed in accordance with the Commission's ``Rules of Practice 
for Domestic Licensing Proceedings'' in 10 CFR Part 2. Interested 
persons should consult a current copy of 10 CFR 2.714 which is 
available at the Commission's Public Document Room, the Gelman 
Building, 2120 L Street, NW., Washington, DC and at the local public 
document room for the particular facility involved. If a request for a 
hearing or petition for leave to intervene is filed by the above date, 
the Commission or an Atomic Safety and Licensing Board, designated by 
the Commission or by the Chairman of the Atomic Safety and Licensing 
Board Panel, will rule on the request and/or petition; and the 
Secretary or the designated Atomic Safety and Licensing Board will 
issue a notice of a hearing or an appropriate order.
    As required by 10 CFR 2.714, a petition for leave to intervene 
shall set forth with particularity the interest of the petitioner in 
the proceeding, and how that interest may be affected by the results of 
the proceeding. The petition should specifically explain the reasons 
why intervention should be permitted with particular reference to the 
following factors: (1) the nature of the petitioner's right under the 
Act to be made a party to the proceeding; (2) the nature and extent of 
the petitioner's property, financial, or other interest in the 
proceeding; and (3) the possible effect of any order which may be 
entered in the proceeding on the petitioner's interest. The petition 
should also identify the specific aspect(s) of the subject matter of 
the proceeding as to which petitioner wishes to intervene. Any person 
who has filed a petition for leave to intervene or who has been 
admitted as a party may amend the petition without requesting leave of 
the Board up to 15 days prior to the first prehearing conference 
scheduled in the proceeding, but such an amended petition must satisfy 
the specificity requirements described above.

[[Page 4342]]

    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. 2
    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: December 30, 1996
    Description of amendment request: The amendment revises (1) 
chemistry data (nickel content) shown on Technical Specification (TS) 
Figures 3.4-2 and 3.4-3 for TS 3/4.4.9, ``Pressure/Temperature 
Limits,'' and (2) the associated Bases 3/4.4.9 to reflect changes to 
chemistry and material properties and changes to comply with recent 
U.S. Nuclear Regulatory Commission (NRC) rule changes to 10 CFR 50, 
Appendix G.
    Basis for proposed no significant hazards determination: As 
required by 10 CFR 50.91(a), the licensee has provided its analysis of 
the issue of no significant hazards consideration, which is presented 
below:
    This change does not involve a significant hazards consideration 
for the following reasons:
    1. The proposed amendment does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    There are no physical changes to any plant equipment created by 
the proposed changes. The chemistry and material property changes do 
not impact the ability of the reactor vessel to maintain [its] 
pressure boundary integrity as previously evaluated. The decrease in 
EOL USE [End-of-Life Upper Shelf Energy] for weld heat 5P6771 is 
relatively minor and remains above the required value that has been 
prescribed by the NRC to provide the necessary level of ductility 
assumed for reactor vessel integrity evaluations. Therefore, the 
accident initiating and mitigating aspects of the pressure vessel 
are not affected. In addition, neither the proposed change requiring 
the ISLH [In-Service Leak and Hydrostatic] test to be complete 
before the core is critical nor the proposed change allowing fuel in 
the reactor vessel during ISLH affects any accident initiating 
mechanisms. The proposed change requiring the ISLH test to be 
completed before the core is critical will not increase the 
consequences of previously evaluated accidents because it 
conservatively assures the core is subcritical. Although the 
proposed change allows fuel in the vessel during ISLH utilizing the 
ISLH Pressure-Temperature (P-T) limits, the consequences of a 
pressure boundary leak have not changed because ISLH testing is 
already allowed using the normal plant P-T limits. In addition, the 
ISLH will be required to be completed before the core is allowed to 
go critical. The consequences of a leak with fuel in the vessel 
during ISLH are the same using either the normal P-T limits or the 
ISLH limits.
    Therefore, there would be no increase in the probability or 
consequences of an accident previously evaluated.
    2. The proposed amendment does not create the possibility of a 
new or different kind of accident from any accident previously 
evaluated.
    There are no physical changes to any plant equipment or new 
components created by the proposed changes. The chemistry and 
material property changes do not impact the pressure boundary 
integrity of the reactor vessel. The decrease in EOL USE for weld 
heat 5P6771 is relatively minor and remains above the required value 
that has been prescribed by the NRC to provide the necessary level 
of ductility assumed for reactor vessel integrity evaluations. 
Therefore, the accident initiating aspects of the pressure vessel 
are not affected. In addition, neither the proposed change requiring 
the ISLH test to be complete before the core is critical nor the 
proposed change allowing fuel in the reactor vessel during ISLH 
creates any new accident initiating mechanisms.
    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.

[[Page 4343]]

    The changes in chemical and material properties do not adversely 
affect any reactor vessel integrity evaluations, such as PTS 
[Pressurized Thermal Shock] or P-T limits. The USE for weld heat 
5P6771 does decrease slightly as described in TS Bases Table B 3/
4.4-1. However, the predicted EOL USE remains above the value 
prescribed in 10 CFR 50, Appendix G and is not a significant 
reduction in the margin of safety. With regard to the proposed 
changes allowing fuel in the reactor vessel during ISLH, the 
existing TS Bases specifically state that fuel is not to be in the 
reactor vessel when the ISLH P-T curve is utilized. However, this 
change is consistent with the revised 10 CFR 50, Appendix G rule and 
as such, is not a significant reduction in the margin of safety.
    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: Mark Reinhart, Acting

Commonwealth Edison Company, Docket No. 50-010, Dresden Nuclear 
Generating Station, Unit 1, Grundy County, Illinois

    Date of amendment request: October 23, 1996
    Description of amendment request: The proposed change would amend 
the Dresden Unit 1 Appendix A Technical Specifications (TS). The 
proposed amendment is a complete revision of the TS to the same format 
as Dresden Unit 2/3 TS.
    Basis for proposed no significant hazards determination: As 
required by 10 CFR 50.91(a), the licensee has provided its analysis of 
the issue of no significant hazards consideration, which is presented 
below:
    1. Will operation of the facility according to this proposed 
change involve a significant increase in the probability or 
consequences of an accident previously evaluated?
    No. In general the proposed amendment represents the conversion 
of current requirements to a more generic format, or the addition of 
requirements which are based on the current safety analysis 
(Decommissioning Plan). Implementation of these changes will not 
reduce reliability of equipment assumed to operate in the current 
safety analysis (Decommissioning Plan), or will provide continued 
assurance that specified parameters remain within their acceptance 
limits, and as such, will not significantly increase the probability 
or consequences of a previously evaluated accident.
    Some of the proposed changes represent minor curtailments of the 
current requirements which are based on generic guidance or 
previously approved provisions for other stations. The proposed 
amendment for Dresden Station Unit 1's Technical Specifications in 
general is based on STS [Standard Technical Specifications] 
guidelines or NRC accepted changes to other facilities such as 
Trojan or San Onofre Unit 1. Any deviations from STS requirements do 
not significantly increase the probability or consequences of any 
previously evaluated accidents for Dresden Station Unit 1. The 
proposed amendment is consistent with the current safety analysis 
(Decommissioning Plan) and has been previously determined to 
represent sufficient requirements for the assurance and reliability 
of equipment assumed to operate in the safety analysis 
(Decommissioning Plan), or provide continued assurance that 
specified parameters remain within their acceptance limits. As such, 
these changes will not significantly increase the probability or 
consequences of a previously evaluated accident.
    2. Will operation of the facility according to this proposed 
change create the possibility of a new or different kind of accident 
from any accident previously evaluated.
    No. In general, the proposed amendment represents the conversion 
of current requirements to a more generic format, or the addition of 
requirements which are based on the current safety analysis 
(Decommissioning Plan). Others represent minor curtailments of the 
current requirements which are based on generic guidance or 
previously approved provisions for other stations. These changes do 
not involve revisions to the design of the station. Some of the 
changes may involve revision in the operation of the station; 
however, these provide additional restrictions which are in 
accordance with the current safety analysis (Decommissioning Plan).
    The proposed amendment for Dresden Station Unit 1's Technical 
Specifications in general is based on STS guidelines or NRC accepted 
changes to other facilities such as Trojan or San Onofre Unit 1. The 
proposed amendment has been reviewed for acceptability at the 
Dresden Nuclear Power Station considering similarity of system or 
component design versus the STS of later operating plants. Any 
deviations from STS requirements do not create the possibility of a 
new of different kind of accident previously evaluated for Dresden 
Station, Unit 1. No new modes of operation are introduced by the 
proposed changes. The proposed changes maintain at least the present 
level of operability. Therefore, the proposed changes do not create 
the possibility of a new or different kind of accident from any 
previously evaluated.
    3. Will operation of the facility according to this proposed 
change involve a significant reduction in a margin of safety?
    No. In general, the proposed amendment represents the conversion 
of current requirements to a more generic format, or the addition of 
requirements which are based on the current safety analysis 
(Decommissioning Plan). Others represent minor curtailments of the 
current requirements which are based on generic guidance or 
previously approved provisions for other stations. Some of the later 
individual items may introduce minor reductions in the margin of 
safety when compared to the current requirements. However, other 
individual changes are the adoption of new requirements which will 
provide significant enhancement of the reliability of human 
performance assumed in the safety analysis (Decommissioning Plan), 
or provide enhanced assurance that specified parameters remain 
within their acceptance limits. These enhancements compensate for 
the individual minor reductions, such that taken together, the 
proposed changes will not significantly reduce the margin of safety.
    The proposed amendment to Technical Specification Section 6.0 
implements present requirements, or the intent of present 
requirements in accordance with the guidelines set forth in the STS. 
Any deviations from STS requirements do not significantly reduce the 
margin of safety for Dresden Station. The proposed changes are 
intended to improve readability, usability, and the understanding of 
technical specification requirements while maintaining acceptable 
levels of safe operation. The proposed changes have been evaluated 
and found to be acceptable for use at Dresden based on system 
design, safety analysis requirements and operational performance. 
Since the proposed changes are based on NRC accepted provisions at 
other operating plants that are applicable at Dresden and maintain 
necessary levels of system or component reliability, the proposed 
changes do not involve a significant reduction in the margin of 
safety.
    The NRC staff has reviewed the analysis of the licensee and, based 
on this review, it appears that the three standards of 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: Morris Area Public Library 
District, 604 Liberty Street, Morris, Illinois 60450
    Attorney for licensee: Michael I. Miller, Esquire, Sidley and 
Austin, One First National Plaza, Chicago, Illinois 60603
    NRC Project Director: Seymour H. Weiss

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

    Date of amendment request: December 2, 1996
    Description of amendment request: The proposed amendments would

[[Page 4344]]

revise Technical Specification 3/4.4.2 to reduce the number of required 
Safety/Relief Valves (SRVs). This change will support a modification to 
remove five of the currently installed SRVs due to the current excess 
capacity, and to reduce maintenance costs and worker radiation dose. 
The current requirement for 17 of the 18 installed SRVs to be operable 
would be changed to require 12 of the 13 installed SRVs to be operable.
    Basis for proposed no significant hazards 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 because:
    The probability of an accident previously evaluated will not 
increase as a result of this change, because the change in valve 
configuration, and the accompanying piping modification does not 
alter any of the initiators of an accident or cause them to occur 
more frequently. The piping modifications will be performed 
consistent with the current piping classifications for the affected 
components. Removal of the SRVs will not impact the ability of the 
remaining SRVs to perform their functions, as described below.
    The consequences of an ASME Overpressurization Event are not 
significantly increased and do not exceed the previously accepted 
licensing criteria for this event. General Electric (GE) has 
calculated the revised peak vessel pressure for LaSalle Station to 
be 1341 psig, which is below the 1375 psig criterion of the ASME 
Code for upset conditions, referenced in Section 5.2.2, 
Overpressurization Protection, of the Updated Final Safety Analysis 
Report (UFSAR), and NUREG-0519 (Safety Evaluation Report related to 
the operation of LaSalle County Station, Units 1 and 2, March 1981), 
and Section 15.2-4, Closure of Main Steam Isolation Valves (BWR) of 
NUREG-0800 (Standard Review Plan). The consequences of this event 
will continue to be verified on a cycle-specific basis, beginning 
with LaSalle Unit 1 Cycle 9 (L1C9). These analysis results will be 
approved as part of the normal reload licensing 10 CFR 50.59 
processes.
    GE has also performed an analysis of the limiting Anticipated 
Transient Without Scram (ATWS) event, which is the MSIV Closure Event 
(MSIVC). This analysis calculated the peak vessel pressure to be 1378 
psig, which is well below the 1500 psig criterion of the ASME Code for 
emergency conditions. General Electric has verified that these results 
will not be impacted with the introduction of Siemens fuel.
    The conclusions given in the safety analyses with regards to 
primary containment dynamic loads, main steam piping loads, Loss-of-
Coolant Accident (LOCA) impact, Minimum Critical Power Ratio (MCPR) 
impact and SRV availability also show that current accident and 
transient analyses are not impacted by this change beyond those 
reanalyzed by GE and discussed above.
    There is no increase in the amount or types of radioactive 
release for any of the affected accidents or transients.
    Therefore, there is not a significant increase in the 
consequences of an accident previously evaluated.
    2) Create the possibility of a new or different kind of accident 
from any accident previously evaluated because:
    The as-left SRV piping configuration will continue to be 
consistent with the current classifications for these piping and 
supports, and have been evaluated by Sargent and Lundy analyses. 
This ensures no different types of events may be caused by piping 
failures at these locations. This is the only physical modification 
proposed by this submittal, and it will not create the possibility 
of a new or different kind of accident from those previously 
evaluated. Other systems are not modified with this change and have 
been shown in this submittal to continue to function as intended 
with the new system configuration, with the exception of the 
abandoned discharge line snubbers which may be replaced with struts, 
except where they will be retained as snubbers due to thermal 
expansion requirements. The changed supports are required to 
function only as struts with the revised piping. Consideration and 
evaluation of this function ensure no new or different accidents are 
created.
    3) Involve a significant reduction in the margin of safety 
because:
    While the calculated peak vessel pressures for the ASME 
Overpressurization Event and the MSIVC ATWS Event are increased due 
to the proposed SRV removals, the new peak pressures remain below 
the respective licensing acceptance limits associated with these 
events.
    The actual cycle-specific reload analysis of the ASME 
Overpressurization Event will be verified to be within the licensing 
acceptance limit for that event prior to each cycle startup, as 
required in the normal reload 10CFR50.59 process. These licensing 
acceptance limits have been previously evaluated as providing a 
sufficient margin of safety. For other accidents and transients, 
including suppression pool loadings, the SRV removals have a 
negligible, if any, effect on the results, so the margin of safety 
is preserved.
    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: Jacobs Memorial Library, 
Illinois Valley Community College, Oglesby, Illinois 61348
    Attorney for licensee: Michael I. Miller, Esquire; Sidley and 
Austin, One First National Plaza, Chicago, Illinois 60603
    NRC Project Director: Robert A. Capra

Consolidated Edison Company of New York, Docket No. 50-247, Indian 
Point Nuclear Generating Unit No. 2, Westchester County, New York

    Date of amendment request: August 14, 1996
    Description of amendment request: The proposed amendment would 
revise Technical Specification Sections 3.3 (Engineered Safety 
Features) and 6.9.1.9 (Core Operating Limits Report (COLR)); the basis 
of Section 3.3, 3.6 (Containment) and 3.10 (Control Rods). These 
changes would incorporate the best estimate approach into the licensing 
basis for the Indian Point Unit No. 2 large break loss-of-coolant 
accident (LOCA) analysis.
    Basis for proposed no significant hazards determination: As 
required by 10 CFR 50.91(a), the licensee has provided its analysis of 
the issue of no significant hazards consideration, which is presented 
below:
    1) Does the proposed license amendment involve a significant 
increase in the probability or consequences of an accident 
previously evaluated?
    Response:
    No physical changes are being made by this change. The plant 
conditions assumed in the analysis are bounded by the design 
conditions for all equipment in the plant. Therefore, there will be 
no increase in the probability of a loss-of-coolant accident. The 
consequences of a LOCA are not being increased. That is, it is shown 
that the emergency core cooling system is designed so that its 
calculated cooling performance conforms to the criteria contained in 
50.46 paragraph b, that is it meets the five criteria listed in 
Section II [see application dated August 14, 1996] of this 
evaluation. No other accident is potentially affected by this 
change. Therefore, neither the probability nor the consequences of 
an accident previously analyzed is increased due to the proposed 
change.
    2) Does the proposed license amendment create the possibility of 
a new or different kind of accident from any previously analyzed?
    Response:
    There are no physical changes being made to the plant. No new 
modes of plant operation are being introduced. The parameters 
assumed in the analysis are within the design limits of existing 
plant equipment. All plant systems will perform equally during the 
response to a potential accident. Therefore, the possibility of a 
new or different kind of accident than previously analyzed will not 
be increased.
    3. Does the proposed amendment involve a significant reduction 
in the margin of safety?
    Response:
    It has been shown that the analytic technique used in the 
analysis realistically describes the expected behavior of the Indian 
Point Unit No. 2 reactor system during a

[[Page 4345]]

postulated loss of coolant accident. Uncertainties have been 
accounted for as required by 10 CFR 50.46. A sufficient number of 
loss of coolant accidents with different break sizes, different 
locations and other variations in properties have been analyzed to 
provide assurance that the most severe postulated loss of coolant 
accidents were calculated. It has been shown by the analysis that 
there is a high level of probability that all criteria contained in 
10 CFR 50.46 paragraph b) are met. Therefore the proposed amendment 
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 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: White Plains Public Library, 
100 Martine Avenue, White Plains, New York 10610.
    Attorney for licensee: Brent L. Brandenburg, Esq., 4 Irving Place, 
New York, New York 10003.
    NRC Project Director: S. Singh Bajwa, Acting Director

Consolidated Edison Company of New York, Docket No. 50-247, Indian 
Point Nuclear Generating Unit No. 2, Westchester County, New York

    Date of amendment request: August 21, 1996
    Description of amendment request: The proposed amendment would 
change the licensee's Technical Specifications (TSs) Section 3.3.G 
(Hydrogen Recombiner System and Post-Accident Containment Venting 
System), the basis for Section 3.3.G, and Section 4.4, Table 4.4-1 
(Containment Isolation Valves). The change would remove the existing 
flame-type hydrogen recombiners, its supporting equipment, and replace 
it with passive autocatalytic recombiners (PARs). In addition, the 
design basis analysis of post-accident hydrogen generation would be 
recalculated.
    Basis for proposed no significant hazards determination: As 
required by 10 CFR 50.91(a), the licensee has provided its analysis of 
the issue of no significant hazards consideration, which is presented 
below:
    1) Does the proposed license amendment involve a significant 
increase in the probability or consequences of an accident 
previously evaluated?
    Neither the probability nor the consequences of a post-LOCA 
[loss-of-coolant accident] combustible gas accident are increased by 
the change in recombiners or in the change to hydrogen generation 
analysis. The probability of a 10 CFR 59.44 type LOCA is not 
affected. The consequences of such an accident are not significantly 
changed.
    Accidents associated with failure of the flame-recombiner flue 
(hydrogen/oxygen) system as well as with failure of the flame-
recombiner containment isolation valves have been eliminated.
    No other accident is potentially affected by this change.
    2) Does the proposed license amendment create the possibility of 
a new or different kind of accident from any previously evaluated?
    No new modes of plant operation are being introduced other than 
elimination of operation of the flame-type recombiners and 
associated support equipment. Recombiner failure is believed to be 
far less likely with the PAR design but in any event, the 
containment vent system is being maintained in its current role as 
backup to recombiner systems. All other plant systems will perform 
equally during the response to a potential accident. Therefore, the 
possibility of a new or different kind of accident than previously 
analyzed will not be increased.
    3) Does the proposed amendment involve a significant reduction 
in the margin of safety?
    The proposed amendment involves margin in the hydrogen 
flammability limit, in the hydrogen generation assumptions and in 
the number of PAR devices assumed. Furthermore, sensitivity analysis 
on PAR effectiveness indicates that additional margin exists for 
success even with degraded PAR performance. It has been shown by the 
analysis that the criteria of 10 CFR 50.44(d) can be met with 
margin. Therefore, the proposed amendment 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 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: White Plains Public Library, 
100 Martine Avenue, White Plains, New York 10610.
    Attorney for licensee: Brent L. Brandenburg, Esq., 4 Irving Place, 
New York, New York 10003.
    NRC Project Director: S. Singh Bajwa, Acting Director

Consolidated Edison Company of New York, Docket No. 50-247, Indian 
Point Nuclear Generating Unit No. 2, Westchester County, New York

    Date of amendment request: August 22, 1996
    Description of amendment request: The proposed amendment would 
revise the licensee's Technical Specification Sections 3.3 and 4.5 
(Engineered Safety Features). The proposed revision would delete the 
requirement to utilize sodium hydroxide (NaOH) as an additive in the 
posted-accident containment spray system.
    Basis for proposed no significant hazards 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:
    ...consistent with the Commission's criteria in 10 CFR 50.92, we 
have determined that the proposed change does not involve a 
significant hazards consideration because the operation of Indian 
Point Unit No. 2 in accordance with this change would not:
    1) involve a significant increase in the probability or 
consequences of an accident previously evaluated. The proposed 
revisions are based on conservative analyses utilizing new, approved 
methodologies. The analysis shows the sodium hydroxide spray 
additive can be removed without significantly affecting the 
radiological consequences of a postulated LOCA [loss-of-coolant 
accident] and that the calculated off-site doses would remain within 
the 10 CFR 100 guidelines. In order to maintain acceptable pH levels 
in the recirculating ECC [emergency core cooling] solution, baskets 
of trisodium phosphate will be stored in strategic locations in 
containment.
    2) create the probability of a new or different kind of accident 
from any accident previously evaluated. The proposed change allows 
the containment safeguards to mitigate the consequences of a design 
basis LOCA in a manner equivalent to that previously approved.
    3) involve a significant reduction in a margin of safety. With 
the proposed change, all safety criteria previously evaluated are 
still met and remain conservative.
    Therefore, based on the above, we conclude that the proposed 
changes do not constitute a significant hazards consideration.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: White Plains Public Library, 
100 Martine Avenue, White Plains, New York 10610.
    Attorney for licensee: Brent L. Brandenburg, Esq., 4 Irving Place, 
New York, New York 10003.
    NRC Project Director: S. Singh Bajwa, Acting Director

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

    Date of amendment request: January 3, 1997
    Description of amendment request: The proposed amendments would 
eliminate from various parts of the Technical Specifications any 
requirement for the low steam pressure

[[Page 4346]]

signal as an initiator of safety injection. The licensee stated that 
the function of the signal is adequately performed by other signals 
(such as the low pressurizer pressure signal).
    Basis for proposed no significant hazards determination: As 
required by 10 CFR 50.91(a), the licensee has provided its analysis of 
the issue of no significant hazards consideration, which is presented 
below:
    1. The proposed change will not involve a significant increase 
in the probability or consequences of an accident previously 
evaluated.
    The proposed change, to delete the SI [safety injection] signal 
on low steam line pressure, will only prevent an unnecessary SI 
actuation as an event occurs which involves secondary system 
depressurization. No consequences will significantly increase, 
because for each event previously analyzed it has been shown that 
either SI on low steam pressure is not demanded, or that another SI 
signal (e.g., low pressurizer pressure) is generated in sufficient 
time to meet applicable acceptance criteria. The probability of an 
accident will not increase.
    2. The proposed change will not create the possibility of any 
new accident not previously evaluated.
    The initiation of SI on a low steam line pressure signal may 
occur during events which involve a depressurization of the 
secondary side, including excessive auxiliary feedwater addition. 
There are other SI initiation signals which will accomplish this 
same function if needed. Removing this actuation signal will not 
create any new failure modes or necessitate any new hardware 
configurations (other than the deletion of the signal itself). No 
new accident scenarios are created.
    3. There is no significant reduction in a margin of safety.
    Analysis has shown that for any transient for which SI would 
have occurred on low steam line pressure, transient response is 
maintained within acceptable limits. Steam line break mass and 
energy releases inside containment do not violate the existing 
environmental qualification envelope. Steam line breaks outside 
containment are not adversely affected by this change.
    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 
proposed amendments involve no significant hazards consideration.
    Local Public Document Room location: York County Library, 138 East 
Black Street, Rock Hill, South Carolina 29730
    Attorney for licensee: Mr. Albert Carr, Duke Power Company, 422 
South Church Street, Charlotte, North Carolina 28242
    NRC Project Director: Herbert N. Berkow

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

    Date of amendment request: November 26, 1996
    Description of amendment request: Change Reactor Coolant System 
Pressure and Temperature Curvers
    Basis for proposed no significant hazards determination: As 
required by 10 CFR 50.91(a), the licensee has provided its analysis of 
the issue of no significant hazards consideration, which is presented 
below:
    Criterion 1 - Does not Involve a Significant Increase in the 
Probability or Consequences of an Accident Previously Evaluated.
    The proposed change revises the pressure/temperature limits in 
accordance with the 10 CFR 50.60 requirements or in accordance with 
Code Case N-514. This approach utilizes the latest NRC guidelines 
relative to estimating neutron irradiation damage of the reactor 
vessel, as well as maintaining conservative limits with respect to 
the low temperature overpressure protection (LTOP) system. 
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 proposed change will not create the possibility of a new or 
different kind of accident from any previously evaluated since it 
does not introduce new systems, failure modes or plant 
perturbations. 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 Margin 
of Safety.
    The proposed change will not involve a significant reduction in 
the margin of safety since the proposed pressure/temperature 
limitations have been developed consistent with the requirements of 
10 CFR 50.60. The operational limits have been developed to maintain 
the necessary margins of safety through 32 effective full power 
years using methodologies previously reviewed and approved by the 
NRC. The objective of these limits is to prevent non-ductile failure 
during any normal operating condition, including anticipated 
operational occurrences and system hydrostatic tests.
    The LTOP safety factors are based on reanalyzed conditions for 
32 effective full power years of operation utilizing methodology 
contained in ASME Code Case N-514. The LTOP evaluation under Code 
Case N-514 for low temperature transients is considered more 
appropriate than the ASME Section XI. The code case establishes a 
factor of 110% of the pressure determined to satisfy Appendix G, 
paragraph G-2215 of ASME Section XI, Division 1 as a design limit, 
instead of 100% required by Section XI. This proposed alternative is 
acceptable because the Code Case recognizes the conservatism of the 
ASME Appendix G curves and allows establishing a LTOP setpoint which 
retains an acceptable margin of safety while maintaining operational 
margins for reactor coolant pump operation at low temperatures and 
pressures. The Code Case provides an acceptable margin of safety 
against flaw initiation and reactor vessel failure, and reduces the 
potential for an undesired LTOP actuation. The application of Code 
Case N-514 for ANO-1 will ensure an acceptable level of safety. 
Therefore, this change does not involve a significant reduction in 
the margin of safety.
    Therefore, based upon the reasoning presented above and the 
previous discussion of the amendment request, Entergy Operations has 
determined that the requested change does not involve significant 
hazards consideration.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: Tomlinson Library, Arkansas 
Tech University, Russellville, AR 72801
    Attorney for licensee: Nicholas S. Reynolds, Esquire, Winston and 
Strawn, 1400 L Street, N.W., Washington, DC 20005-3502
    NRC Project Director: William D. Beckner

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

    Date of amendment request: October 7, 1996
    Description of amendment request: Modify Plant Protection System 
Test Interval to 123 days.
    Basis for proposed no significant hazards determination: As 
required by 10 CFR 50.91(a), the licensee has provided its analysis of 
the issue of no significant hazards consideration, which is presented 
below:
    Criterion 1 - Does Not Involve a Significant Increase in the 
Probability or Consequences of an Accident Previously Evaluated.
    The proposed changes included in this amendment request are 
being made to surveillance intervals, allowances to use CISAM 
elements and various administrative changes. These changes do not 
alter the functional characteristics of any plant component and do 
not allow any new modes of operation of any components. These 
changes do not involve a significant increase in the probability of 
any event initiator to occur. Therefore, this amendment request does 
not involve a significant increase in the probability of any 
accident previously evaluated.
    Increasing the surveillance interval for the RPS and ESFAS 
instrumentation has two principal effects with opposing impacts on 
risk. The first impact is a slight increase in core damage frequency 
that results from the increased unavailability of the

[[Page 4347]]

instrumentation in question from the extended testing interval. The 
unavailability of the tested instrumentation components is 
translated to result in a failure of the reactor to trip, an 
anticipated transient without a scram, or a failure of the 
appropriate engineered safety feature to actuate when required. The 
opposing impact on risk is the corresponding reduction in core 
damage frequency that would result due to the reduced exposure of 
the plant to test induced transients.
    Representative fault tree models were developed for ANO-2 and 
the corresponding core damage frequency increases and decreases were 
quantified in CEN-327 and CEN-327 Supplement 1. The NRC staff found 
that changes in the RPS unavailabilities that result from extending 
the surveillance test interval (STI) from 30 days to 90 days were 
not considered to be significant. Estimates of the reduction in 
scram frequency from the reduction in test induced scrams and the 
corresponding reduction in core damage frequency were found 
acceptable. Sequential testing intervals of 90 days were found to 
result in a net reduction in risk.
    CE NPSD-576 employed the same methodology used in CEN-327 and 
its supplement to evaluate the impact of extending the surveillance 
intervals from monthly sequential testing to every four months 
(triannual) on a staggered test basis. The corresponding changes in 
RPS and ESFAS unavailabilities are quantified in CE NPSD-576 and are 
shown to be less than their counterparts in CEN-327 and its 
supplement. Thus, triannual staggered testing should be acceptable 
as it results in lower RPS and ESFAS unavailabilities than for a 90 
day test interval with sequential testing which has been found to be 
acceptable to the NRC.
    The TS amendment request provided the option to use cycle 
independent shape annealing matrix (CISAM) elements. The CISAM 
elements will be validated during startup testing and will be 
required to meet additional acceptance criteria as well as that used 
for the cycle specific shape annealing matrix (SAM) elements. If the 
CISAM is determined to be no longer valid, a cycle specific SAM will 
be calculated and used in the CPCs. Therefore, the CPCs will operate 
as designed and this change will not affect the consequences of any 
accident previously evaluated.
    The CPC addressable constant surveillance requirements and the 
various administrative changes affected by this TS change do not 
affect the consequences of any accident previously evaluated.
    Therefore, these changes do 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.
    This amendment request does not involve any changes in equipment 
and will not alter the manner in which the plant will be operated.
    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 RPS/ESFAS extended testing interval yields no significant 
reduction in the margin to safety. The instrument drift occurring 
over the proposed STI will not cause the setpoint values to exceed 
those assumed in the safety analysis and specified in the TS. There 
are no changes to equipment or plant operations that will result 
from this change. The implementation of these proposed changes is 
expected to result in an overall improvement in safety due to the 
fact that reduced testing will result in fewer inadvertent trips, 
less frequent actuation of EFAS components, and less frequent 
distraction of the operations personnel.
    The CPC addressable constant surveillance interval extension 
included in this amendment request is consistent with the 
methodology found in NUREG-1432, ``Standard Technical Specifications 
Combustion Engineering Plants'' (ISTS). Requiring the addressable 
constant verification to be performed as part of the CPC channel 
functional test should detect an error in these constants prior to 
restoring the channel to operable status instead of allowing the 
error to go undetected until the next surveillance period. Although 
the surveillance interval is extended by this TS change, this change 
does not involve a significant reduction in the margin of safety.
    The CPC CISAM elements and the various administrative changes 
included in this TS change do not involve a significant reduction in 
the margin of safety.
    Therefore, these changes do not involve a significant reduction 
in the margin of safety.
    Therefore, based upon the reasoning presented above and the 
previous discussion of the amendment request, Entergy Operations has 
determined that the requested change does not involve significant 
hazards consideration.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: Tomlinson Library, Arkansas 
Tech University, Russellville, AR 72801
    Attorney for licensee: Nicholas S. Reynolds, Esquire, Winston and 
Strawn, 1400 L Street, N.W., Washington, DC 20005-3502
    NRC Project Director: William D. Beckner

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

    Date of amendment request: December 19, 1996
    Description of amendment request: Change Request Concerning 
Addition to the Core Operating Limit Report References
    Basis for proposed no significant hazards determination: As 
required by 10 CFR 50.91(a), the licensee has provided its analysis of 
the issue of no significant hazards consideration, which is presented 
below:
    Criterion 1 - Does not Involve a Significant Increase in the 
Probability or Consequences of an Accident Previously Evaluated.
    The proposed change to add the technical manual for the 
Combustion Engineering Nuclear Transient Simulation (CENTS) code to 
the Core Operating Limits Report (COLR) references is administrative 
in nature. The CENTS code has been reviewed and approved by the NRC. 
The physical design or operation of the plant is not impacted by 
this proposed change. The proposed change does not adversely impact 
transient analysis assumptions or results. The COLR-related safety 
analyses will continue to be performed utilizing NRC-approved 
methodologies, and specific reload changes will be evaluated under 
the provisions of 10CFR50.59. 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 proposed change to reference the NRC-approved CENTS code is 
administrative in nature. No physical alterations of plant 
configuration, changes to plant operating procedures, or operating 
parameters are proposed. No new equipment is being introduced, and 
no equipment is being operated in a manner inconsistent with its 
design. The COLR-related safety analyses will continue to be 
performed utilizing NRC-approved methodologies. A 10CFR50.59 safety 
review will continue to be performed to evaluate specific reload 
changes. 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 proposed change to reference the CENTS code is 
administrative in nature. Existing technical specification 
operability and surveillance requirements are not reduced by the 
proposed change. The cycle- specific COLR limits for future reloads 
will continue to be developed based on NRC-approved methodologies. 
Technical specifications will continue to require that the core be 
operated within these limits and specify appropriate actions to be 
taken if the limits are violated. The COLR-related safety analyses 
will continue to be performed utilizing NRC-approved methodologies, 
and specific reload changes will be evaluated per 10CFR50.59. 
Therefore, this change does not involve a significant reduction in 
the margin of safety.
    Therefore, based upon the reasoning presented above and the 
previous discussion of the amendment request, Entergy Operations has 
determined that the requested change does not involve a significant 
hazards consideration.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff

[[Page 4348]]

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 No. 50-368, Arkansas Nuclear One, 
Unit No. 2, Pope County, Arkansas

    Date of amendment request: December 19, 1996
    Description of amendment request: Change Request Concerning Power 
Calibration Requirements
    Basis for proposed no significant hazards determination: As 
required by 10 CFR 50.91(a), the licensee has provided its analysis of 
the issue of no significant hazards consideration, which is presented 
below:
    Criterion 1 - Does not Involve a Significant Increase in the 
Probability or Consequences of an Accident Previously Evaluated.
    The proposed change will redefine the tolerance band allowed for 
linear power level, the Core Protection Calculator (CPC) delta T 
Power, and CPC nuclear power signals. Changing the tolerance range 
from [plus or minus] 2% to between -0.5% and 10% between 15% and 80% 
rated thermal power, will require more conservative tolerances than 
are currently allowed. This change will ensure that the power 
indications are more conservative relative to the existing safety 
analyses. 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 proposed change to Technical Specification power calibration 
tolerance limits are conservative relative to the current 
requirements. This amendment request does not change the design or 
operation of any plant systems or components. 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 Margin 
of Safety.
    The allowed tolerance band for the linear power level, CPC delta 
T power, and CPC nuclear power signals between 15 and 80% power has 
been redefined. The new requirements are more conservative than the 
tolerances that currently exist in the Technical Specifications. 
This change will ensure that the power indications are more 
conservative relative to the existing safety analyses. Therefore, 
this change does not involve a significant reduction in the margin 
of safety.
    Therefore, based upon the reasoning presented above and the 
previous discussion of the amendment request, Entergy Operations has 
determined that the requested change does not involve significant 
hazards consideration.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: Tomlinson Library, Arkansas 
Tech University, Russellville, AR 72801
    Attorney for licensee: Nicholas S. Reynolds, Esquire, Winston and 
Strawn, 1400 L Street, N.W., Washington, DC 20005-3502
    NRC Project Director: William D. Beckner

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

    Date of amendment request: December 19, 1996
    Description of amendment request: Change Request Concerning Reactor 
Coolant System Volume
    Basis for proposed no significant hazards 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.
    This proposed change allows the relocation of the reactor 
coolant system volume in the design features section of technical 
specifications to the safety analysis report. Future changes will be 
controlled under 10CFR50.59. This change is considered 
administrative in nature. Appropriate values of reactor coolant 
system volume are used in the safety analyses. This change does not 
affect any system or component functional requirements. The 
operation of the plant is not affected by this change.
    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 relocation of existing requirements from the technical 
specifications to another licensee controlled document is 
administrative in nature. This change does not modify or remove any 
plant design requirement. The proposed change will not affect any 
plant system or structure, nor will it affect any system functional 
or operability requirements. Therefore, no new failure modes are 
introduced as a result of this change.
    Therefore, this change does not create the possibility of a new 
or different kind of accident from any previously evaluated.
    Criterion 3 - Does Not Involve a Significant Reduction in the 
Margin of Safety.
    The proposed amendment request relocates the coolant system 
volume located in the technical specifications design feature 
section to another licensee controlled document, the ANO-2 Safety 
Analysis Report, which is controlled under 10CFR50.59. The proposed 
change is administrative in nature because the design requirements 
for the facility remain the same. The proposed change does not 
represent a change in the configuration or operation of the plant.
    Therefore, this change does not involve a significant reduction 
in the margin of safety.
    Therefore, based upon the reasoning presented above and the 
previous discussion of the amendment request, Entergy Operations has 
determined that the requested change does not involve a significant 
hazards consideration.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: Tomlinson Library, Arkansas 
Tech University, Russellville, AR 72801
    Attorney for licensee: Nicholas S. Reynolds, Esquire, Winston and 
Strawn, 1400 L Street, 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: December 19, 1996
    Description of amendment request: Change Control Room Ventillation 
System Requirements
    Basis for proposed no significant hazards 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 control room emergency ventilation and air conditioning 
systems are not initiators of an accident previously evaluated. 
Extension of the allowable outage time for one inoperable control 
room emergency air conditioning system from 7 days to 30 days is 
acceptable based on the low probability of an event occurring that 
would require control room isolation and a concurrent or subsequent 
failure of the remaining operable control room emergency air 
conditioning system. An evaluation using probabilistic safety 
assessment techniques has shown the frequency of this event to be an 
acceptably low level (4.67E-6/yr). The ANO-1 surveillance 
requirements for the control room emergency ventilation and air

[[Page 4349]]

conditioning system have been updated for consistency with the ANO-2 
requirements and are consistent with RG 1.52, March 1978, Revision 2 
and ASTM D3803-1989. The change in the ANO-2 Mode of Applicability 
for the control room radiation monitoring instrumentation is 
acceptable because the only identified accident scenario requiring 
control room isolation on high radiation while in Modes 5 and 6 is 
the fuel handling accident and this analysis shows that the dose 
consequences to the control room operators are acceptable in the 
event of a fuel handling accident, assuming that the normal control 
room ventilation system is properly isolated. The remainder of the 
changes have been made for consistency between the ANO-1 and ANO-2 
TS and are considered to be more restrictive or administrative in 
nature.
    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 control room emergency ventilation and air conditioning 
systems are not accident initiators. The proposed changes introduce 
no new mode of plant operation and no new possibility for an 
accident is introduced by modifying the ANO-1 surveillance testing 
requirements for the control room emergency ventilation and air 
conditioning systems.
    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.
    With the exception of the AOT extension and the relaxation of 
the ANO-2 Mode of Applicability for the control room radiation 
monitoring instrumentation, all the ANO-1 and ANO-2 changes are 
considered administrative or more restrictive and are intended to 
clarify and make consistent the requirements of the control room 
emergency habitability equipment. Although the AOT extension does 
involve an incremental reduction in the margin of safety due to 
slight increase in the frequency of an event requiring control room 
isolation, followed by failure of the operable emergency control 
room chiller, a probabilistic safety assessment has shown this 
slight increase in frequency (approximately 3.58E-6/yr) to be 
acceptably low. The change in the ANO-2 Mode of Applicability for 
the control room radiation monitoring instrumentation is acceptable 
because the only identified accident scenario requiring control room 
isolation on high radiation while in Modes 5 and 6 is the fuel 
handling accident and this analysis shows that the dose consequences 
to the control room operators are acceptable in the event of a fuel 
handling accident, assuming that the normal control room ventilation 
system is properly isolated.
    Therefore, this change does not involve a significant reduction 
in the margin of safety.
    Therefore, based upon the reasoning presented above and the 
previous discussion of the amendment request, Entergy Operations has 
determined that the requested change does not involve significant 
hazards consideration.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: Tomlinson Library, Arkansas 
Tech University, Russellville, AR 72801
    Attorney for licensee: Nicholas S. Reynolds, Esquire, Winston and 
Strawn, 1400 L Street, N.W., Washington, DC 20005-3502
    NRC Project Director: William D. Beckner

Georgia Power Company, Oglethorpe Power Corporation, Municipal 
Electric Authority of Georgia, City of Dalton, Georgia, Docket No. 
50-366, Edwin I. Hatch Nuclear Plant, Unit 2, Appling County, 
Georgia

    Date of amendment request: December 3, 1996
    Description of amendment request: The two proposed changes would 
revise Technical Specification (TS) 2.1.1.2 for Hatch Nuclear Plant, 
Unit 2, Safety Limit Minimum Critical Power Ratio (SLMCPR) values. The 
revision is based upon unique plant evaluations for the current Cycle 
13 and the use of General Electric (GE) GE-13 fuel, a 9 x 9 fuel 
design, in the next Cycle 14. The proposed SLMCPRs for Hatch Unit 2 are 
1.08 and 1.09 (single-loop operation) for the current Cycle 13, and 
1.12 and 1.14 (single-loop operation) for Cycle 14.
    The new SLMCPRs were calculated using NRC-approved methods and 
interim implementing procedures. The SLMCPRs are set high enough to 
ensure that greater than 99.9% of all fuel rods in the core avoid 
transition boiling if the limit is not violated. The SLMCPRs 
incorporate a margin for uncertainty in the core operating state for 
uncertainties that are fuel-type dependent, including fuel bundle 
nuclear characteristics, critical power correlation, and manufacturing 
tolerances. These interim procedures were revised to incorporate the 
following cycle-specific parameters: (1) Actual core loading, (2) 
Conservative variations of projected control blade patterns, (3) Actual 
bundle parameters (e.g., local peaking), and (4) Full cycle exposure 
range.
    Basis for proposed no significant hazards determination: As 
required by 10 CFR 50.91(a), the licensee has provided its analysis of 
the issue of no significant hazards consideration which is presented 
below:
    1. Does the change involve a significant increase in the 
probability of consequences of an accident previously evaluated?
    The derivation of the revised SLMCPRs for Plant Hatch Unit 2 for 
incorporation into the Technical Specifications, and its use to 
determine cycle-specific thermal limits, were performed using NRC-
approved methods. Additionally, interim implementing procedures 
incorporating cycle-specific parameters were used. Based upon the 
use of these calculations, revised SLMCPRs cannot increase the 
probability or severity of an accident. The basis of the SLMCPR 
calculation is to ensure that <ls-thn-eq> 99.9% of all fuel rods in 
the core avoid transition boiling if the limit is not violated. The 
new SLMCPRs preserve the existing margin to transition boiling and 
fuel damage in the event of a postulated accident. Thus, it can be 
concluded that the probability of fuel damage is not increased and 
the proposed Technical Specifications changes do not involve an 
increase in the probability or consequences of an accident 
evaluation.
    2. Do the proposed changes create the possibility of a new or 
different type of accident from any previously evaluated?
    The SLMCPR is a Technical Specifications numerical value 
designed to ensure that fuel damage from transition boiling does not 
occur as a result of the limiting postulated accident. The SLMCPRs 
were calculated using NRC-approved methods. Additionally, interim 
procedures incorporating cycle-specific parameters were used in the 
analysis. Therefore, the proposed changes do not create the 
possibility of a new or different kind of accident from any accident 
previously evaluated.
    3. Do the proposed changes involve a significant reduction in 
the margin of safety?
    The margin of safety as defined in the Bases will remain the 
same. The new SLMCPRs were calculated using NRC-approved methods 
which are in accordance with the current fuel design and licensing 
criteria. Additionally, interim implementing procedures, which 
incorporate cycle-specific parameters were used. The SLMCPR remains 
high enough to ensure that <ls-thn-eq> 99.9% of all fuel rods in the 
core will avoid transition boiling if the limit is not violated, 
thereby preserving the fuel cladding integrity. Therefore, the 
proposed changes do not involve a significant reduction in a margin 
of safety.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 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

[[Page 4350]]

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: January 7, 1997
    Description of amendment request: The proposed amendments would 
change the Technical Specifications (TS) for Plant Hatch, Units 1 and 
2, associated with Surveillance Requirement (SR) testing that requires 
manually actuating every safety/relief valve (S/RV) during each unit 
startup from a refueling outage. The proposed changes would provide an 
alternate method of testing the S/RVs during shutdown conditions rather 
than during unit startup as is currently done. This approach would 
reduce valve leakage, thereby reducing the possibility of inadvertent 
valve actuation and resultant plant transients. Additionally, deletion 
of testing for the safety mode of the S/RVs is proposed since other 
testing provides operability verification.
    Furthermore, the licensee proposes relief from the applicable 
requirements of the ASME OM Code (1995), Appendix I, paragraph I 
3.4.1(d), which also requires manual actuating of S/RVs during unit 
startup.
    Current Unit 1 and Unit 2 SRs 3.5.1.12 and 3.6.1.6.1 require that 
each S/RV be manually actuated at pressure conditions. Georgia Power 
Company (GPC) proposes to revise SRs 3.5.1.12 and 3.6.1.6.1 that would 
require the S/RVs to be manually actuated in the relief mode during a 
plant outage before steam is generated. The solenoid valve would be 
energized, the actuator would stroke, and the pilot rod lift would be 
measured. This in-situ test would verify that, given a signal to the 
solenoid, the pilot disc rod would lift. If steam were present, the 
pilot disc would open and initiate opening of the main stage.
    The licensee also proposes to delete current Units 1 and 2 SR 
3.4.3.2, which also requires that each S/RV be manually actuated 
because this test is not necessary to assure S/RV operability in the 
safety mode since other tests, taken together, confirm the entire S/RV 
assembly functions adequately.
    Basis for proposed no significant hazards 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:
    Georgia Power Company [GPC] has reviewed the proposed license 
amendment request and determined its adoption does not involve a 
significant hazards consideration. In support of this determination, 
an evaluation of each of the three 10 CFR 50.92 standards follows.
    1. The proposed changes do not involve a significant increase in 
the probability or consequences of an accident previously evaluated.
    Since the proposed Technical Specifications changes and ASME 
Code relief do not impose any physical changes to the S/RVs, their 
design function is unaffected. The submittal only proposes changes 
to the manner in which the S/RVs are tested. As discussed in 
Enclosure 1 [of the licensee's submittal], the combination of 
current S/RV testing and the proposed alternate
    testing will continue to adequately demonstrate the operability 
of the S/RVs for both the safety and relief modes. Under the 
proposed testing requirements, it is expected that S/RV leakage will 
decrease; thus, the probability of occurrence of an inadvertent S/RV 
actuation is actually reduced.
    FSAR [Final Safety Analysis Report] analyzed events, such as 
MSIV [main steam isolation valve] closure, generator load reject, 
turbine trip with failure of switchyard breakers to open, and 
pressure regulator failure, take credit for the S/RVs mitigating the 
consequences of these events. These proposed changes will not 
increase the consequences of these events, since a series of S/RV 
tests (on the bench and installed) will ensure all S/RV components 
necessary to ensure valve opening will function. The S/RVs will 
therefore be capable of performing their design functions.
    Furthermore, reducing the number of manual actuations of the S/
RVs decreases the likelihood of a stuck open S/RV, which is an 
analyzed event in the Hatch FSAR.
    Therefore, the probability of occurrence and the consequences of 
previously analyzed events are not increased.
    2. The proposed changes do not create the possibility of [a new 
or different kind of accident from any accident] previously 
evaluated.
    The proposed changes affect the manner in which S/RV operability 
is verified in that one Technical Specifications SR [surveillance 
requirement] is being deleted and two are being revised; however, 
they do not affect the way the S/RVs are operated. The S/RVs will 
not be operated or tested in a manner contrary to their design. As a 
result, no new mode of operation is introduced. That is, the 
proposed changes do not create the possibility of a new or different 
kind of accident from any previously evaluated.
    3. The proposed changes do not involve a significant reduction 
in the margin of safety.
    The present method of S/RV testing unnecessarily challenges the 
valves, and is linked to S/RV degradation through pilot valve and/or 
main valve leakage. This Technical Specifications change should 
decrease S/RV leakage and improve S/RV reliability by reducing the 
potential for spurious valve actuation at full power. In this sense, 
the margin of safety is actually increased; e.g., the likelihood for 
spurious S/RV actuation is reduced.
    Deleting the test of installed S/RVs at rated temperature and 
pressure will not significantly reduce the margin of safety for 
events in which S/RV actuation is assumed, since each S/RV receives 
a series of tests which insure each component necessary for 
successful opening of the S/RV functions properly. Thus, the S/RV is 
assured of opening in either the safety or the relief mode. For 
example, at Wyle Labs, the valves undergo testing at operating steam 
pressure. This test ensures operability of the pilot and main discs 
and also verifies set pressure, reseat pressure, and main steam 
stroke time. As noted previously, upon successful completion of 
these tests, including verification of zero seat leakage, the valves 
receive a written certification from the lab and are returned to 
Plant Hatch for installation.
    GPC further proposes that, upon installation, but before steam 
is generated, the valves receive a test requiring the solenoid to be 
energized. This test provides additional verification that the pilot 
disc opens. The remaining segments of the S/RV tests verify the 
ability of ADS and LLS logic to energize the solenoid.
    In summary, this amendment does not involve a significant 
reduction in the margin of safety, because of the reduction in S/RV 
degradation, and because remaining tests confirm the valves will 
function properly when required.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: Appling County Public Library, 
301 City Hall Drive, Baxley, Georgia 31513
    Attorney for licensee: Ernest L. Blake, Jr., Esquire, Shaw, 
Pittman, Potts and Trowbridge, 2300 N Street, NW., Washington, DC 20037
    NRC Project Director: Herbert N. Berkow

GPU Nuclear Corporation, et al., Docket No. 50-289, Three Mile 
Island Nuclear Station, Unit No. 1, Dauphin County, Pennsylvania

    Date of amendment request: December 16, 1996
    Description of amendment request: The amendment request, if 
approved, would reflect the change in the legal name of the operator of 
TMI-1 from GPU Nuclear Corporation to GPU Nuclear Inc. and reflect in 
the TMI-1 license and the Technical Specifications the registered trade 
name of GPU Energy.
    Basis for proposed no significant hazards determination: As 
required by 10 CFR 50.91(a), the licensee has provided its analysis of 
the issue of no significant hazards consideration (SHC), which is 
presented below:
    GPU Nuclear Inc. has determined that the proposed TMI-1 license 
amendment and technical specification change request

[[Page 4351]]

involve no significant hazards consideration as defined in 10 CFR 
50.92 because:
    Operation of the facility in accordance with the proposed 
amendment would not involve a significant increase in the 
probability of occurrence or the consequences of an accident 
previously evaluated. The proposed amendment adds to the license and 
the technical specifications the trade name of the Owners of TMI-1. 
The change in the legal name of the operator of TMI-1 is a cosmetic 
change made to reflect the name changes made throughout the GPU 
family of companies. The name change has no impact on plant design 
or operation.
    Operation of the facility in accordance with the proposed 
amendment would not create the possibility of a new or different 
kind of accident from any accident previously evaluated because no 
new failure modes are created by the proposed changes. The use of a 
common trade name for the Owners of TMI-1 and the change in the 
legal name of the operator of TMI-1 has no impact on plant design or 
operation. Thus, there is no creation of the possibility of a new or 
different kind of accident from those previously evaluated.
    Operation of the facility in accordance with the proposed 
amendment will not involve a signficant reduction in a margin of 
safety. The proposed amendment does not change any operating limits 
for reactor operation.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: Law/Government Publications 
Section, State Library of Pennsylvania, (REGIONAL DEPOSITORY) Walnut 
Street and Commonwealth Avenue, Box 1601, Harrisburg, PA 17105.
    Attorney for licensee: Ernest L. Blake, Jr., Esquire, Shaw, 
Pittman, Potts & Trowbridge, 2300 N Street, NW., Washington, DC 20037. 
NRC Acting Project Director: Patrick D. Milano

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

    Date of amendment request: August 4, 1995 as supplemented December 
20, 1996 [AEP:NRC:1129E and 1129M]
    Description of amendment request: The proposed amendment would 
modify the technical specifications (T/S) to allow for repair of hybrid 
expansion joint (HEJ) sleeves under redefined repair boundary limits. 
This alternate plugging criterion would assess the integrity of parent 
tube indications based on the degraded joint geometry, with reference 
to the specific location of the flaw. The continued operability of the 
HEJ sleeved tube would be based on the measured diameter difference, or 
diameter delta (delta D), between the sleeve peak hardroll diameter and 
the diameter of the sleeve adjacent to the parent tube flaw in the 
upper joint.
    Basis for proposed no significant hazards 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:
    Conformance of the proposed amendments to the standards for a 
determination of no significant hazard as defined in 10 CFR 50.92 
(three factor test) is shown in the following.
    (1) Operation of Cook Nuclear Plant unit 1 in accordance with 
the proposed license amendment does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    The HEJ sleeved tube structural integrity limits defined by this 
amendment provide for structural integrity consistent with the 
guidance of RG 1.121. Tube structural integrity consistent with the 
most limiting RG 1.121 loading is inherently provided by a measured 
[delta D] of less than 1 mil, although the criterion specifies a 
minimum of 3 mils must be verified. The structural integrity 
characteristics of a postulated degraded parent tube with a 3 mil 
[delta D] provides for axial restraint capability of more than 
double the most limiting RG 1.121 loading, which indicates that the 
postulated separated tube would not become axially displaced 
relative to the sleeve during any plant condition.
    Based on tube pull data from Cook Nuclear Plant and other plants 
it is expected that TSP intersections would provide a substantial 
axial restraint capability. This interaction is neglected in the 
analysis of the criterion, and provides for extra safety margin.
    Based on the destructive examination results for sections of HEJ 
sleeved tubes removed in 1994 from another plant, the parent tube 
flaw morphology is described as circumferentially oriented with 
multiple initiation sites. This segmented morphology indicates that 
the previously performed structural capability testing is 
conservative. Additional axial load bearing capability is provided 
by the segmented morphology since end cap loading would be 
transmitted through the tube by the non-degraded ligaments of the 
segmented crack network, and tube separation therefore, is not 
likely or credible.
    The consequences of any postulated failure of a sleeved tube to 
which the criteria has been applied would be bounded by the current 
steam generator tube rupture event discussed in the Cook Nuclear 
Plant Final Safety Analysis Report (FSAR). Axial displacement of any 
tube, sleeved or unsleeved, is bounded by approximately 1.1 inch. A 
tube which experiences axial displacement by this amount would be 
expected to exhibit a release rate well below the normal makeup 
capacity. In order for a HEJ sleeved tube to exhibit reactor coolant 
system release rates approaching the release rates assumed in the 
FSAR the tube must be displaced by approximately 3 inches. In order 
for the postulated separated tube to experience axial displacement 
of any magnitude, it must be assumed that the HEJ hardroll provides 
no structural benefit and that the tube-to-TSP interaction is 
frictionless.
    Postulated primary to secondary leakage during a main steam line 
break event will be assessed against the limit of 8.4 gpm in the 
faulted loop, calculated as part of the voltage based plugging limit 
for tube support plate intersections. The total of all leakage 
sources must be shown to be less than this value.
    Application of the 3 mil [delta D] criterion (excluding eddy 
current uncertainty) does not change existing reactor coolant system 
flow conditions, therefore, existing LOCA analysis results will be 
unaffected. Plant response to design basis accidents for the current 
tube plugging and flow conditions are not affected by the repair 
process; no new tube diameter restriction is introduced.
    (2) The proposed license amendment does not create the 
possibility of a new or different kind of accident from any accident 
previously evaluated.
    Application of the proposed 3 mil [delta D] HEJ sleeved tube 
structural integrity criterion will not introduce significant or 
adverse changes to the plant design basis. The 3 mil [delta D] 
criteria provides for structural integrity of the HEJ sleeved tube 
assembly which significantly exceeds the limiting RG 1.121 loading 
condition. Under these conditions neither a single nor a multiple 
tube rupture event is considered credible.
    The general outline of the HEJ sleeve is unaffected, and the 
application of the proposed criterion does not change the sleeve 
configuration or size/shape. The application of the criterion also 
does not represent a potential to affect other plant components.
    (3) The proposed license amendment does not involve a 
significant reduction in a margin of safety.
    The proposed criterion has been shown to provide structural 
integrity of the tube bundle consistent with the most limiting RG 
1.121 tube integrity recommendations. In order for tube rupture to 
occur, the degraded parent tube must experience a complete 
circumferential separation and be subsequently axially displaced by 
approximately 3 inches. The inherent structural integrity provided 
by the interference fit of the HEJ in addition to the axial 
restraint provided by tube support plate intersections above the HEJ 
provides for structural integrity far exceeding the RG 1.121 loading 
of 2264 lb. Even in the event that a degraded HEJ sleeved parent 
tube were to experience axial displacement, the maximum amount of 
displacement the tube could experience is bounded by 1.11 inch. 
Postulating that the tube were to become displaced by this amount, 
primary to secondary leakage would be limited to well less than the 
normal makeup capacity due to the proximity between the 
hydraulically expanded sleeve OD and tube ID.
    Pulled HEJ sleeved tube samples from another plant with HEJ 
sleeved tubes indicate that the crack morphology is described as 
circumferentially oriented cracking with multiple initiation sites. 
This segmented

[[Page 4352]]

morphology provides for additional structural margin not modeled in 
the testing program.
    Existing flow equivalency calculations for the HEJ sleeved tubes 
will be unaffected by the application of the criterion.
    Based on the preceding analysis it is concluded that operation 
of Cook Nuclear Plant unit 1 following the application of the 3 mil 
[delta D] HEJ sleeved tube structural integrity limit does not 
increase the probability of an accident previously evaluated, create 
the possibility of a new or different kind of accident from any 
accident previously evaluated, or reduce any margins to plant 
safety. Therefore, the license amendment does not involve a 
significant hazards consideration as defined in 10 CFR 50.92.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: Maud Preston Palenske Memorial 
Library, 500 Market Street, St. Joseph, Michigan 49085
    Attorney for licensee: Gerald Charnoff, Esq., Shaw, Pittman, Potts 
and Trowbridge, 2300 N Street, NW, Washington, DC 20037
    NRC Project Director: Gail H. Marcus

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

    Date of application for amendment: October 30, 1996
    Description of amendment request: These amendments revise the 
safety limit minimum critical power ratios (SLMCPRs) at Peach Bottom 
Atomic Power Station, Unit 3.
    Basis for proposed no significant hazards 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 [technical specification] changes do not 
involve a significant increase in the probability or consequences of 
an accident previously evaluated.
    The derivation of the cycle-specific SLMCPRs for incorporation 
into the TS, and its use to determine cycle-specific thermal limits, 
have been performed using USNRC [U.S. Nuclear Regulatory 
Commission]-approved methods as discussed in ``General Electric 
Standard Application for Reactor Fuel,'' NEDE-24011-P-A-11, and U.S. 
Supplement, NEDE-24011-P-A-11-US, November 17, 1995 and interim 
(reconfirmation) implementing procedures. This change in SLMCPRs 
cannot increase the probability or severity of an accident.
    The basis of the SLMCPR calculation is to ensure that greater 
than 99.9% of all fuel rods in the core avoid transition boiling if 
the limit is not violated. The new SLMCPRs preserve the existing 
margin to transition boiling and fuel damage in the event of a 
postulated accident. The fuel licensing acceptance criteria for the 
SLMCPR calculation apply to PBAPS [Peach Bottom Atomic Power 
Station], Unit 3, Cycle 11 in the same manner as they have applied 
previously. The probability of fuel damage is not increased. 
Therefore, the proposed TS changes do not involve an increase in the 
probability or consequences of an accident previously evaluated.
    2) The proposed TS changes do not create the possibility of a 
new or different kind of accident from any accident previously 
evaluated.
    The SLMCPR is a TS numerical value, designed to ensure that 
transition boiling does not occur in 99.9% of all fuel rods in the 
core during the limiting postulated accident. It cannot create the 
possibility of any new type of accident. The new SLMCPRs are 
calculated using USNRC-approved methods (<gr-thn-eq>General Electric 
Standard Application for Reactor Fuel,'' NEDE-24011-P-A-11, and U.S. 
Supplement, NEDE-24011-P-A-11-US, November 17, 1995) and interim 
(reconfirmation) implementing procedures.
    Therefore, the proposed TS changes do not create the possibility 
of a new or different kind of accident, from any accident previously 
evaluated.
    3) The proposed TS changes do not involve a significant 
reduction in a margin of safety.
    The margin of safety as defined in the TS Bases will remain the 
same. The new SLMCPRs are calculated using USNRC-approved methods 
(<gr-thn-eq>General Electric Standard Application for Reactor 
Fuel,'' NEDE-24011-P-A-11, and U.S. Supplement, NEDE-24011-P-A-11-
US, November 17, 1995) and interim (reconfirmation) implementing 
procedures which are in accordance with the current fuel licensing 
criteria. The SLMCPRs ensure that greater than 99.9% of all fuel 
rods in the core will avoid transition boiling if the limit is not 
violated, thereby preserving the fuel cladding integrity. Therefore, 
the proposed TS changes do not involve a reduction in a margin of 
safety.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: Government Publications 
Section, State Library of Pennsylvania, (REGIONAL DEPOSITORY) Education 
Building, Walnut Street and Commonwealth Avenue, Box 1601, Harrisburg, 
PA 17105.
    Attorney for licensee: J. W. Durham, Sr., Esquire, Sr. V. P. and 
General Counsel, PECO Energy Company, 2301 Market Street, Philadelphia, 
PA 19101
    NRC Project Director: John F. Stolz

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

    Date of amendment request: January 15, 1997
    Description of amendment request: The amendment proposes to 
relocate the snubber operability, surveillance, and record requirements 
for components (snubbers) in the Technical Specifications (TS) to plant 
controlled documents.
    Basis for proposed no significant hazards determination: As 
required by 10 CFR 50.91(a), the licensee has provided its analysis of 
the issue of no significant hazards consideration, which is presented 
below:
    Operation of the FitzPatrick plant in accordance with the 
proposed Amendment would not involve a significant hazards 
consideration as defined in 10 CFR 50.92, based on the following:
    1. These changes do not involve a significant increase in the 
probability or consequences of an accident previously evaluated 
because:
    The changes relocate operability, surveillance, and record 
requirements for components (snubbers) which do not meet the 
criteria for inclusion in the Technical Specifications (TS). The 
affected components are not assumed to be initiators of analyzed 
events and are not assumed to mitigate accident or transient events. 
The snubber requirements will be relocated from the TS to plant 
controlled documents. These requirements will be maintained pursuant 
to 10 CFR 50.59. Therefore, the changes do not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    2. The changes do not create the possibility of a new or 
different type of accident previously evaluated because:
    The changes do not necessitate a physical alteration of the 
plant (no new or different type of equipment will be installed) or 
affect parameters governing normal plant operation. Adequate control 
of future changes to snubber requirements will be maintained. Thus, 
these changes do not create the possibility of a new or different 
kind of accident from any accident previously evaluated for the 
plant.
    3. The proposed changes do not involve a reduction in a margin 
of safety because:
    The changes do not involve a change to the operability, 
surveillance, and record requirements for the snubber program as 
they currently exist in the TS, nor do they impact on any safety 
analysis assumptions. The proposed changes relocate snubber 
requirements from the TS to plant controlled documents. Changes to 
the requirements in these documents are subject to the requirements 
of 10 CFR 50.59. In addition, exceptions to code requirements for 
testing will require NRC approval. Regulations and

[[Page 4353]]

FitzPatrick commitments to the NRC contain the necessary 
programmatic requirements for the plant controlled documents. 
Operating limitations will continue to be imposed, and required 
surveillances will continue to be performed in accordance with 
regulations, FitzPatrick commitments to the NRC, and written 
procedures and instructions that are auditable by the NRC. If 
snubber inoperability causes a TS system or component to be 
inoperable, then the affected system or component Limiting Condition 
for Operation (LCO) will be entered. Based on the above, the 
proposed changes do not involve a significant reduction in a margin 
of safety.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: Reference and Documents 
Department, Penfield Library, State University of New York, Oswego, New 
York 13126.
    Attorney for licensee: Mr. Charles M. Pratt, 1633 Broadway, New 
York, New York 10019.
    NRC Project Director: S. Singh Bajwa, Acting Director

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

    Date of amendment request: January 7, 1997
    Description of amendment request: The proposed amendment would 
revise Technical Specification (TS) 3/4.2.5 to incorporate an exception 
to the provisions of TS 4.0.4 and to clarify the time at which the 
surveillance can be performed by adding that the surveillance is to be 
performed within 24 hours after attaining steady state conditions at or 
above 90% rated thermal power. The revised surveillance would also 
contain editorial enhancements that do not change the intent of the 
current surveillance. TS Table 3.2-1 for Salem Unit 1 would be revised 
to delete reference to three loop operation (which is not permitted at 
Salem Unit 1) in order to eliminate potential confusion when applying 
this table.
    Basis for proposed no significant hazards 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 proposed on the RCS [Reactor Coolant System] flow 
measurement and exemption to Specification 4.0.4 do not affect the 
operation of the equipment during conditions when they are required 
to perform their safety function. No physical changes to the plant 
result from the proposed changes made to the surveillance 
requirements. The measurement of RCS flow does not impact the 
probability of an accident.
    Testing is being performed with the plant in the condition in 
which the automatic initiation signals for low RCS flow would result 
in a time consistent with the TS requirements.
    Protection System in providing a reactor trip upon a loss of RCS 
flow. Degradations in flow will occur over a long duration; however, 
testing will continue to be performed within twenty-four hours upon 
achieving steady state greater than or equal to 90% RTP [Rated 
Thermal Power] after refueling which is a sufficiently short 
duration after startup to identify flow degradations.
    Changes proposed to refer to Table 3.2-1 for the DNB [Departure 
from Nucleate Boiling] parameters and to delete the Unit 1 three 
loop operation parameters, and the inclusion of the type of test 
performed are editorial in nature.
    Therefore, the consequences of an accident previously evaluated 
are not significantly increased by the proposed changes.
    2. The proposed change does not create the possibility of a new 
or different kind of accident from any accident previously 
evaluated.
    The proposed changes do not involve any modifications to 
existing plant equipment, do not alter the function of any plant 
systems, do not introduce any new operating configurations or new 
modes of plant operation, nor change the safety analyses. The point 
at which RCS flow is measured using a heat balance will not impact 
the ability to maintain or monitor Reactor Coolant flows. The 
proposed changes will, therefore, not create the possibility of a 
new or different kind of accident from any accident previously 
evaluated.
    Changes proposed to refer to Table 3.2-1 for the DNB parameters 
and to delete the Unit 1 three loop operation parameters, and the 
inclusion of the type of test performed are editorial in nature.
    [The proposed changes will, therefore, 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 changes to the RCS flow surveillance do not decrease the 
scope of the existing testing, but will clarify the point at which 
the testing is performed.
    The time in which testing is performed, after achieving steady 
state conditions after reaching greater than or equal to 90% RTP 
ensures that testing is performed in a timely manner. Flow margins 
established as a result of previous testing will not be 
significantly reduced in light of recent outage activities. Future 
changes that might impact margins established by the testing will be 
reviewed in accordance with the requirements of 10 CFR 50.59.
    Changes proposed to refer to Table 3.2-1 for the DNB parameters 
and to delete the Unit 1 three loop operation parameters, and the 
inclusion of the type of test performed are editorial in nature.
    All changes are consistent with the intent of Salem's current TS 
[Technical Specification] and with the 18 month surveillances 
specified in NUREG-1431, Revision 1.
    The proposed change, therefore, 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: Salem Free Public library, 112 
West Broadway, Salem, NJ 08079
    Attorney for licensee: Mark J. Wetterhahn, Esquire, Winston and 
Strawn, 1400 L Street, NW, Washington, DC 20005-3502
    NRC Project Director: John F. Stolz

Southern Nuclear Operating Company, Inc., Docket No. 50-348, Joseph 
M. Farley Nuclear Plant, Unit 1, Houston County, Alabama

    Date of amendments request: December 26, 1996
    Description of amendments request: The proposed amendment would 
revise Technical Specification 3/4.4.6 ``Steam Generators'' and its 
associated Bases. Specifically, the steam generator repair limit would 
be modified to clarify that the appropriate method for determining 
serviceability for tubes with outside diameter stress corrosion 
cracking at the tube support plate is by a methodology that more 
reliably assesses structural integrity. This amendment request is in 
accordance with NRC's Generic Letter 95-05, ``Voltage-Based Repair 
Criteria for Westinghouse Steam Generator Tubes Affected by Outside 
Diameter Stress Corrosion Cracking.''
    Basis for proposed no significant hazards 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 Farley units in accordance with the proposed 
license amendment does not involve a significant increase in the 
probability or consequences of an accident previously evaluated.
    Testing of model boiler specimens for free standing tubes at 
room temperature conditions shows burst pressures as high as 
approximately 5000 psi for indications of outer diameter stress 
corrosion cracking with voltage measurements as high as 26.5 volts. 
Burst testing performed on pulled tubes,

[[Page 4354]]

including tubes pulled from Farley Unit 1, with up to 7.5 volt 
indications show burst pressures in excess of 5300 psi at room 
temperature. ... [T]ube burst criteria are inherently satisfied 
during normal operating conditions by the presence of the tube 
support plate. Furthermore, correcting for the effects of 
temperature on material properties and minimum strength levels (as 
the burst testing was done at room temperature), tube burst 
capability significantly exceeds the R.G. [Regulatory Guide] 1.121 
criterion requiring the maintenance of a margin of 1.43 times the 
steam line break pressure differential on tube burst if through-wall 
cracks are present without regard to the presence of the tube 
support plate. Considering the existing data base, this criterion is 
satisfied with bobbin coil indications with signal amplitudes over 
twice the 2.0 volt voltage-based repair criteria, regardless of the 
indicated depth measurement. This structural limit is based on a 
lower 95% confidence level limit of the data at operating 
temperatures. The 2.0 volt criterion provides an extremely 
conservative margin of safety to the structural limit considering 
expected growth rates of outside diameter stress corrosion cracking 
at Farley. Alternate crack morphologies can correspond to a voltage 
so that a unique crack length is not defined by a burst pressure to 
voltage correlation. However, relative to expected leakage during 
normal operating conditions, no field leakage has been reported from 
tubes with indications with a voltage level of under 7.7 volts for a 
3/4 inch tube with a 10 volt correlation to 7/8 inch tubing (as 
compared to the 2.0 volt proposed voltage-based tube repair limit). 
Thus, the proposed amendment does not involve a significant increase 
in the probability or consequences of an accident.
    Relative to the expected leakage during accident condition 
loadings, the accidents that are affected by primary-to-secondary 
leakage and steam release to the environment are Loss of External 
Electrical Load and/or Turbine Trip, Loss of All AC Power to Station 
Auxiliaries, Major Secondary System Pipe Failure, Steam Generator 
Tube Rupture, Reactor Coolant Pump Locked Rotor, and Rupture of a 
Control Rod Drive Mechanism Housing. Of these, the Major Secondary 
System Pipe Failure is the most limiting for Farley in considering 
the potential for off-site doses. The offsite dose analyses for the 
other events which model primary-to-secondary leakage and steam 
releases from the secondary side to the environment assume that the 
secondary side remains intact. The steam generator tubes are not 
subjected to a sustained increase in differential pressure, as is 
the case following a steam line break event. This increase in 
differential pressure is responsible for the postulated increase in 
leakage and associated offsite doses following a steam line break 
event. In addition, the steam line break event results in a bypass 
of containment for steam generator leakage. Upon implementation of 
the voltage-based repair criteria, it must be verified that the 
expected distributions of cracking indications at the tube support 
plate intersections are such that primary-to-secondary leakage would 
result in site boundary dose within the current licensing basis. 
Data indicate that a threshold voltage of 2.8 volts could result in 
through-wall cracks long enough to leak at steam line break 
conditions. Application of the proposed repair criteria requires 
that the current distribution of a number of indications versus 
voltage be obtained during the refueling outages. The current 
voltage is then combined with the rate of change in voltage 
measurement and a voltage measurement uncertainty to establish an 
end of cycle voltage distribution and, thus, leak rate during steam 
line break pressure differential. The leak rate during a steam line 
break is further increased by a factor related to the probability of 
detection of the flaws. If it is found that the potential steam line 
break leakage for degraded intersections planned to be left in 
service coupled with the reduced allowable specific activity levels 
result in radiological consequences outside the current licensing 
basis, then additional tubes will be plugged or repaired to reduce 
steam line break leakage potential to within the acceptance limit. 
Thus, the consequences of the most limiting design basis accident 
are constrained to present licensing basis limits, and therefore 
there is no change to the probability or consequences of an accident 
previously evaluated.
    2) The proposed license amendment does not create the 
possibility of a new or different kind of accident from any accident 
previously evaluated.
    Implementation of the proposed voltage-based tube repair 
criteria does not introduce any significant changes to the plant 
design basis. Use of the criteria does not provide a mechanism that 
could result in an accident outside of the region of the tube 
support plate elevations. Neither a single or multiple tube rupture 
event would be expected in a steam generator in which the repair 
criteria have been applied during all plant conditions. The bobbin 
probe signal amplitude repair criteria are established such that 
operational leakage or excessive leakage during a postulated steam 
line break condition is not anticipated. Southern Nuclear has 
previously implemented a maximum leakage limit of 140 gpd per steam 
generator. The R.G. 1.121 criterion for establishing operational 
leakage limits that require plant shutdown are based upon leak-
before-break considerations to detect a free span crack before 
potential tube rupture. The 140 gpd limit provides for leakage 
detection and plant shutdown in the event of the occurrence of an 
unexpected single crack resulting in leakage that is associated with 
the longest permissible crack length. R.G. 1.121 acceptance criteria 
for establishing operating leakage limits are based on leak-before-
break considerations such that plant shutdown is initiated if the 
leakage associated with the longest permissible crack is exceeded. 
The longest permissible crack is the length that provides a factor 
of safety of 1.43 against bursting at steam line break pressure 
differential. A voltage amplitude of approximately 9 volts for 
typical outside diameter stress corrosion cracking corresponds to 
meeting this tube burst requirement at the 95% prediction interval 
on the burst correlation. Alternate crack morphologies can 
correspond to a voltage so that a unique crack length is not defined 
by the burst pressure versus voltage correlation. Consequently, a 
typical burst pressure versus through-wall crack length correlation 
is used below to define the ``longest permissible crack'' for 
evaluating operating leakage limits.
    The single through-wall crack lengths that result in tube burst 
at 1.43 times steam line break pressure differential and steam line 
break conditions are about 0.54 inch and 0.84 inch, respectively. 
Normal leakage for these crack lengths would range from about 0.4 
gallons per minute to 4.5 gallons per minute, respectively, while 
lower 95% confidence level leak rates would range from about 0.06 
gallons per minute to 0.6 gallons per minute, respectively.
    An operating leak rate of 140 gpd per steam generator has been 
implemented. This leakage limit provides for detection of 0.4 inch 
long cracks at nominal leak rates and 0.6 inch long cracks at the 
lower 95% confidence level leak rates. Thus, the 140 gpd limit 
provides for plant shutdown prior to reaching critical crack lengths 
for steam line break conditions at leak rates less than a lower 95% 
confidence level and for three times normal operating pressure 
differential at less than nominal leak rates.
    Considering the above, the implementation of voltage-based 
repair criteria will not create the possibility of a new or 
different kind of accident from any previously evaluated.
    3) The proposed license amendment does not involve a significant 
reduction in margin of safety.
    The use of the voltage-based repair criteria is demonstrated to 
maintain steam generator tube integrity commensurate with the 
requirements of Generic Letter 95-05 and R.G. 1.121. R.G. 1.121 
describes a method acceptable to the NRC staff for meeting GDC 
[General Design Criteria] 2, 14, 15, 31, and 32 by reducing the 
probability of the consequences of steam generator tube rupture. 
This is accomplished by determining the limiting conditions of 
degradation of steam generator tubing, as established by inservice 
inspection, for which tubes with unacceptable cracking should be 
removed from service. Upon implementation of the criteria, even 
under the worst case conditions, the occurrence of outside diameter 
stress corrosion cracking at the tube support plate elevations is 
not expected to lead to a steam generator tube rupture event during 
normal or faulted plant conditions. The most limiting effect would 
be a possible increase in leakage during a steam line break event. 
Excessive leakage during a steam line break event, however, is 
precluded by verifying that, once the criteria are applied, the 
expected end of cycle distribution of crack indications at the tube 
support plate elevations would result in minimal, and acceptable 
primary to secondary leakage during the event and, hence, help to 
demonstrate radiological conditions are less than an appropriate 
fraction of the 10 CFR [Part] 100 guideline.
    The margin to burst for the tubes using the voltage-based repair 
criteria is comparable to that currently provided by existing 
technical specifications.
    In addressing the combined effects of LOCA [loss-of-coolant 
accident] + SSE [safe shutdown earthquake] on the steam generator

[[Page 4355]]

component (as required by GDC 2), it has been determined that tube 
collapse may occur in the steam generators at some plants. This is 
the case as the tube support plates may become deformed as a result 
of lateral loads at the wedge supports at the periphery of the plate 
due to either the LOCA rarefaction wave and/or SSE loadings. Then, 
the resulting pressure differential on the deformed tubes may cause 
some of the tubes to collapse.
    There are two issues associated with steam generator tube 
collapse. First, the collapse of steam generator tubing reduces the 
RCS [reactor coolant system] flow area through the tubes. The 
reduction in flow area increases the resistance to flow of steam 
from the core during a LOCA which, in turn, may potentially increase 
Peak Clad Temperature (PCT). Second, there is a potential the 
partial through-wall cracks in tubes could progress to through-wall 
cracks during tube deformation or collapse or that short through-
wall indications would leak at significantly higher leak rates than 
included in the leak rate assessments.
    Consequently, a detailed leak-before-break analysis was 
performed and it was concluded that the leak-before-break 
methodology (as permitted by GDC 4) is applicable to the Farley 
reactor coolant system primary loops and, thus, the probability of 
breaks in the primary loop piping is sufficiently low that they need 
not be considered in the structural design basis of the plant. 
Excluding breaks in the RCS primary loops, the LOCA loads from the 
large branch line breaks were analyzed at Farley and were found to 
be of insufficient magnitude to result in steam generator tube 
collapse or significant deformation.
    Regardless of whether or not leak-before-break is applied to the 
primary loop piping at Farley, any flow area reduction is expected 
to be minimal (much less than 1%) and PCT margin is available to 
account for this potential effect. Based on analyses' results, no 
tubes near wedge locations are expected to collapse or deform to the 
degree that secondary to primary in-leakage would be increased over 
current expected levels. For all other steam generator tubes, the 
possibility of secondary-to-primary leakage in the event of a LOCA + 
SSE event is not significant. In actuality, the amount of secondary-
to-primary leakage in the event of a LOCA + SSE is expected to be 
less than that originally allowed, i.e., 500 gpd per steam 
generator. Furthermore, secondary-to-primary in-leakage would be 
less than primary-to-secondary leakage for the same pressure 
differential since the cracks would tend to tighten under a 
secondary-to-primary pressure differential. Also, the presence of 
the tube support plate is expected to reduce the amount of in-
leakage.
    Addressing the R.G. 1.83 considerations, implementation of the 
tube repair criteria is supplemented by 100% inspection requirements 
at the tube support plate elevations having outside diameter stress 
corrosion cracking indications, reduced operating leakage limits, 
eddy current inspection guidelines to provide consistency in voltage 
normalization, and rotating probe inspection requirements for the 
larger indications left in service to characterize the principle 
degradation mechanism as outside diameter stress corrosion cracking.
    As noted previously, implementation of the voltage-based repair 
criteria will decrease the number of tubes that must be taken out of 
service with tube plugs or repaired. The installation of steam 
generator tube plugs or tube sleeves would reduce the RCS flow 
margin, thus implementation of the voltage-based repair criteria 
will maintain the margin of flow that would otherwise be reduced 
through increased tube plugging or sleeving.
    Considering the above, it is concluded that the proposed change 
does not result in a significant reduction in margin with respect to 
plant safety as defined in the Final Safety Analysis Report or any 
bases of the plant Technical Specifications.
    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: Houston-Love Memorial Library, 
212 W. Burdeshaw Street, Post Office Box 1369, Dothan, Alabama 36302
    Attorney for licensee: M. Stanford Blanton, Esq., Balch and 
Bingham, Post Office Box 306, 1710 Sixth Avenue North, Birmingham, 
Alabama 35201
    NRC Project Director: Herbert N. Berkow

Southern Nuclear Operating Company, Inc., Docket No. 50-348, Joseph 
M. Farley Nuclear Plant, Unit 1, Houston County, Alabama

    Date of amendments request: January 10, 1997
    Description of amendments request: The proposed amendments would 
implement repair of tubes using laser welded tube sleeves for the steam 
generators at Farley Units 1 and 2 as described in WCAP-13088, Revision 
4, and WCAP-14740. In addition, for Unit 2, references to a one-cycle 
limited implementation of L* are being removed. The approval for the 
limited implementation of L* expired at the last Unit 2 outage in the 
fall of 1996.
    Basis for proposed no significant hazards 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 Farley Units 1 and 2 in accordance with the 
proposed license amendment does not involve a significant increase 
in the probability or consequences of an accident previously 
evaluated.
    The laser welded sleeve configurations as described within WCAP-
13088, Revision 4 and WCAP-14740 have been designed and analyzed in 
accordance with the requirements of the ASME Code [American Society 
of Mechanical Engineers Boiler and Pressure Vessel Code]. Fatigue 
and stress analyses of the sleeved tube assemblies produced 
acceptable results. Mechanical testing has shown that the structural 
strength of Alloy 690 sleeves under normal, faulted and upset 
conditions is within acceptable limits. Leakage testing for 7/8 inch 
tube sleeves has demonstrated that significant primary-to-secondary 
leakage is notexpected during all plant conditions, including the 
case where the seal weld is not produced in the lower joint of the 
tubesheet sleeve.
    Initial acceptance of welded joints uses ultrasonic inspection 
to verify that all weld thicknesses meet the minimum specified 
conditions over the entire circumference. A plugging limit of 24% 
allowable depth of penetration of the sleeve tube wall thickness 
applies for each type of laser welded sleeve that may be installed 
in the Farley Nuclear Plant steam generators and is determined for 
uprated conditions with a limiting steam pressure for reduced 
T<INF>hot and 20% steam generator tube plugging conditions. These 
conditions represent the limiting primary-to-secondary operating 
pressure differential, which is bounding for the sleeve plugging 
limit and structural analysis inputs. However, the state-of-the-art 
in eddy current inspection capability is such that no probes are 
qualified to size the depth of penetration of stress corrosion 
cracking. It is generally believed that the detection threshold of 
these probes is well below 40% throughwall. Southern Nuclear 
Operating Company will plug on detection any crack-like indications 
that may occur in the sleeve using the sleeve inspection probe of 
record until an inspection process is qualified to size depth of 
penetration of stress corrosion cracking into the tube wall.
    The hypothetical consequences of failure of the sleeve would be 
bounded by the current steam generator tube rupture analysis 
included in the Farley Nuclear Plant FSAR [Final Safety Analysis 
Report]. Due to the slight reduction in diameter caused by the 
sleeve wall thickness, it is expected that primary coolant release 
rates would be slightly less than assumed for the steam generator 
tube rupture analysis (depending on the break location), and 
therefore, would result in lower total primary fluid mass release to 
the secondary system. Combinations of tubesheet sleeves and tube 
support plate sleeves would reduce the primary fluid flow through 
the sleeved tube assembly due to the series of diameter reductions 
the fluid would have to pass on its way to the break area. The 
overall effect would be reduced steam generator tube rupture release 
rates.
    As addressed previously, the proposed Technical Specification 
change to support the installation of full length tubesheet, 
elevated tubesheet, or tube support plate elevation Alloy 690 laser 
welded sleeves as described in WCAP-13088, Revision 4 and WCAP-14740 
does not adversely impact any other previously evaluated design 
basis accident or the results of LOCA [loss-
    of-coolant accident] and non-LOCA accident analyses for the 
current Technical Specification minimum reactor coolant

[[Page 4356]]

system flow rate. The results of the analyses and testing, as well 
as plant operating experience, demonstrate that the sleeve assembly 
is an acceptable means of restoring tube integrity to a condition 
consistent with its original design basis. Also, per Regulatory 
Guide 1.83, Revision 1 recommendations, the condition of the sleeved 
tube can be monitored through periodic inspections with present eddy 
current techniques.
    Conformance of the sleeve design with the applicable sections of 
the ASME Code and results of the leakage and mechanical tests 
support the conclusion that the installation of laser welded tube 
sleeves will not increase the probability or consequences of an 
accident previously evaluated. Depending upon the break location for 
a postulated steam generator tube rupture event, implementation of 
tube sleeving could act to reduce the radiological consequences to 
the public due to reduced primary to secondary flow rate through a 
sleeved tube compared to a non-sleeved tube based on the restriction 
afforded by the sleeve wall thickness.
    Removal of the references to the interim use of an L* repair 
criteria will not involve a significant increase in the probability 
or consequences of an accident previously evaluated.
    2. The proposed license amendment does not create the 
possibility of a new or different kind of accident from any accident 
previously evaluated.
    Implementation of laser welded sleeving will not introduce 
significant or adverse changes to the plant design basis. Sleeving 
also does not represent a potential to affect any other plant 
component. Stress and fatigue analysis of the repair has shown the 
ASME Code minimum stress values are not exceeded. Implementation of 
laser welded sleeving maintains overall tube bundle structural and 
leakage integrity at a level consistent to that of the originally 
supplied tubing during all plant conditions. Leak and mechanical 
testing of sleeves support the conclusions of the calculations that 
each sleeve joint retains both structural and leakage integrity 
during all conditions. Sleeving of tubes does not provide a 
mechanism resulting in an accident outside of the area affected by 
the sleeves. Any hypothetical accident as a result of potential tube 
or sleeve degradation in the repaired portion of the tube is bounded 
by the existing tube rupture accident analysis. Since the sleeve 
design does not affect any other component or location of the tube 
outside of the immediate area repaired, in addition to the fact that 
the installation of sleeves and the impact on current plugging level 
analyses is accounted for, the possibility that laser welded 
sleeving creates a new or different type of accident is not 
credible.
    Removal of the references to the interim use of an L* repair 
criteria will not create the possibility of a new or different kind 
of accident from any accident previously evaluated.
    3. The proposed license amendment does not involve a significant 
reduction in margin of safety.
    The laser welded sleeving repair of degraded steam generator 
tubes as identified in WCAP-13088, Revision 4, has been shown by 
analysis to restore the integrity of the tube bundle consistent with 
its original design basis condition as the requirements of the ASME 
Code are satisfied. The safety factors used in the design of sleeves 
for the repair of degraded tubes are consistent with the safety 
factors in the ASME Boiler and Pressure Vessel Code used in steam 
generator design. The design of the tubesheet sleeve lower joints 
for the 7/8 inch sleeves (for both the full length and elevated 
tubesheet sleeve) have been verified by testing to preclude 
realistic leakage during normal and postulated accident conditions.
    The portions of the installed sleeve assembly which represent 
the reactor coolant pressure boundary can be monitored for the 
initiation and progression of sleeve/tube wall degradation, thus 
satisfying the recommendations of Regulatory Guide 1.83, Revision 1 
and the surveillance requirements included in Specification 4.4.6.0. 
The portion of the tube bridged by the sleeve joints is effectively 
removed from the pressure boundary, and the sleeve then forms the 
new pressure boundary. The areas of the sleeved tube assembly which 
require inspection are defined in WCAP-13088, Revision 4.
    The effect of sleeving on the design transients and accident 
analyses have been reviewed based on the installation of sleeves up 
to the level of steam generator tube plugging coincident with the 
minimum reactor flow rate. The installation of sleeves is to be 
evaluated as the equivalent of some level of steam generator tube 
plugging. Evaluation of the installation of sleeves is based on the 
determination that LOCA evaluations for the licensed minimum reactor 
coolant flow bound the effect of a combination of tube plugging and 
sleeving up to an equivalent of the actual steam generator tube 
plugging limit. Information provided in WCAP-13088, Revision 4, 
describes the method to determine the flow equivalency for all 
combinations of tubesheet and tube support plate sleeves in order 
that the minimum flow requirements are met.
    Implementation of laser welded sleeving will reduce the 
potential for primary-to-secondary leakage during a postulated steam 
line break while maintaining available primary coolant flow area in 
the event of a LOCA. By effectively isolating degraded areas of the 
tube through repair, primary pressure boundary integrity is restored 
and the potential for primary-to-secondary leakage during all plant 
conditions is minimized. These degraded tubes are returned to a 
condition consistent with the design basis. While the installation 
of a sleeve causes a reduction in primary coolant flow, the 
reduction is significantly below the reduction incurred by plugging. 
Therefore, greater primary coolant flow area is maintained through 
sleeving.
    Removal of the references to the interim use of an L* repair 
criteria will 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: Houston-Love Memorial Library, 
212 W. Burdeshaw Street, Post Office Box 1369, Dothan, Alabama 36302
    Attorney for licensee: M. Stanford Blanton, Esq., Balch and 
Bingham, Post Office Box 306, 1710 Sixth Avenue North, Birmingham, 
Alabama 35201
    NRC Project Director: Herbert N. Berkow

Tennessee Valley Authority, Docket No. 50-390 Watts Bar Nuclear 
Plant, Unit 1, Rhea County, Tennessee

    Date of amendment request: January 10, 1997
    Description of amendment request: The proposed amendment would 
modify the Watts Bar Nuclear Plant (WBN) Unit 1 Technical 
Specifications (TS) in order to implement the 1995 rule change to 10 
CFR Part 50, Appendix J. The revised Appendix J provided an Option B 
which allows performance based testing for containment leakage rate 
testing. The TS in Section 3.6 and associated Bases, TS Section 3.0.2 
and TS Section 5.7 would be changed. Also, the schedular exemption for 
containment airlock testing now specified in the facility license in 
Section 2.D(1) would no longer be required and would be deleted.
    Basis for proposed no significant hazards 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 
amendment would not involve a significant increase in the 
probability or consequences of an accident previously evaluated.
    The proposed amendment to WBN TSs is in accordance with Option B 
to 10 CFR 50, Appendix J. The proposed amendment adds a voluntary 
performance-based option for containment leak-rate testing. The 
changes being proposed do not affect the precursor for an accident 
or transient analyzed in Chapter 15 of WBN Final Safety Analysis 
Report. The proposed change does not increase the total allowable 
primary containment leakage rate. The proposed change does not 
reflect a revision to the physical design and/or operation of the 
plant. [T]herefore, operation of the facility, in accordance with 
the proposed change, does not significantly affect the probability 
or consequences of an accident previously evaluated.
    (2) Operation of the facility in accordance with the proposed 
amendment would not create the possibility of a new or different 
kind of accident from any accident previously evaluated.
    The proposed amendment to WBN TSs is in accordance with the new 
performance-based option (Option B) to 10 CFR 50,

[[Page 4357]]

Appendix J. The changes being proposed will not change the physical 
plant or the modes of operation defined in the facility license. The 
proposed changes do not increase the total allowable primary 
containment leakage rate. The changes do not involve the addition or 
modification of equipment, nor do they alter the design or operation 
of plant systems. Therefore, operation of the facility in accordance 
with the propsoed change does not create the possibility of a new or 
diferent kind of accident from any previously evaluated.
    (3) Operation of the facility in accordance with the proposed 
amendment would not involve a significant reduction in margin of 
safety.
    The proposed change to WBN TSs is in accordance with the new 
option to 10 CFR 50, Appendix J. The proposed option is formulated 
to adopt performance-based approaches. This option removes the 
current prescriptive details from the TS. The proposed changes do 
not affect plant safety analyses or change the physical design or 
operation of the plant. The proposed change does not increase the 
total allowable primary containment leakage rate. Therefore, 
operation of the facility, in accordance with 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: Chattanooga-Hamilton County 
Library, 1001 Broad Street, Chattanooga, TN 37402
    Attorney for licensee: General Counsel, Tennessee Valley Authority, 
400 West Summit Hill Drive, ET llH, Knoxville, Tennessee 37902
    NRC Project Director: Frederick J. Hebdon

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

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

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

    Date of amendment request: November 26, 1996, as supplemented 
December 17, 1996
    Description of amendment request: The proposed amendments would 
allow a one-time only change necessary to replace the existing 125-volt 
dc battery cells with new cells. Date of publication of individual 
notice in Federal Register: December 13, 1996 (61 FR 65605)
    Expiration date of individual notice: January 13, 1997
    Local Public Document Room location: J. Murrey Atkins Library, 
University of North Carolina at Charlotte, 9201 University City 
Boulevard, North Carolina 28223-0001

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.

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

    Date of application for amendment: August 23, 1996
    Brief description of amendment: This amendment makes Technical 
Specifications changes allowing fuel enrichment of up to 5.0 weight 
percent Uranium-235. The previous limit was 4.1 weight percent. This 
change allows Arkansas Nuclear One, Unit-2, to receive, store, and use 
nuclear fuel of 5.0 weight percent Uraninum-235.
    Date of issuance: January 14, 1997
    Effective date: January 14, 1997
    Amendment No.: 178
    Facility Operating License No. NPF-6. Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: October 9, 1996 (61 FR 
52964) The Commission's related evaluation of the amendment is 
contained in a Safety Evaluation dated January 14, 1997.No significant 
hazards consideration comments received: No.
    Local Public Document Room location: Tomlinson Library, Arkansas 
Tech University, Russellville, AR 72801

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

    Date of application for amendment: November 24, 1996, as 
supplemented on December 2, 1996.
    Brief description of amendment: This amendment adds small break-
loss-of coolant accident methodology CENPD-137, Supplement 1-P and its 
approval letter by the NRC as a reference to Section 6.9.5.1. This code 
previously approved by the NRC increases the steam generator tube 
plugging limit to 30% with an associated reduction of 10% in RCS flow. 
This amendment also corrects a typographical error in Specification 
6.9.5.1.8, and Specifications 6.9.5.1.10 through 6.9.5.1.14 are 
numbered to accommodate these changes.
    Date of issuance: January 14, 1997
    Effective date: January 14, 1997
    Amendment No.: 179
    Facility Operating License No. NPF-6. Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: December 3, 1996 (61 FR 
64173) However, on December 9, 1996,

[[Page 4358]]

the licensee verified that the number of plugged tubes would not exceed 
their current 10% limit established by the old code. This determination 
removed the basis for considering this request as exigent. Since the 
potential does exist for the plugging to exceed the 10% in the future, 
the technical specification amendment request is therefore, a valid 
request on a normal schedule. This change did not alter the staff's 
initial proposed no safety hazard condition determination, therefore 
noticing was not warranted. The Commission's related evaluation of the 
amendment is contained in a Safety Evaluation dated January 14, 1997.
    Local Public Document Room location: Tomlinson Library, Arkansas 
Tech University, Russellville, AR 72801

Entergy Gulf States, Inc., Cajun Electric Power Cooperative, and 
Entergy Operations, Inc., Docket No. 50-458, River Bend Station, 
Unit 1, West Feliciana Parish, Louisiana

    Date of amendment request: October 24, 1996
    Brief description of amendment: The amendment revises the technical 
specifications to delete the accelerated testing requirements for the 
standby diesel generators. This action is consistent with the 
provisions of Generic Letter 94-01, ``Removal of Accelerated Testing 
and Special Reporting Requiremets for Emergency Diesel Generators,'' 
dated May 31, 1994.
    Date of issuance: January 14, 1997
    Effective date: January 14, 1997
    Amendment No.: 90
    Facility Operating License No. NPF-47. The amendment revised the 
Technical Specifications/operating license.
    Date of initial notice in Federal Register: December 4, 1996 (61 FR 
64384) The Commission's related evaluation of the amendment is 
contained in a Safety Evaluation dated January 14, 1997.No significant 
hazards consideration comments received. No.
    Local Public Document Room location: Government Documents 
Department, Louisiana State University, Baton Rouge, LA 70803

Entergy Gulf States, Inc., Cajun Electric Power Cooperative, and 
Entergy Operations, Inc., Docket No. 50-458, River Bend Station, 
Unit 1, West Feliciana Parish, Louisiana

    Date of amendment request: May 30, 1996
    Brief description of amendment: The amendment revises the technical 
specification surveillance requirement 3.8.3.4 to specify a 5-start 
pressure for the air recievers associated with the Division III, High 
Pressure Core Spray emergency diesel generator.
    Date of issuance: January 16, 1997
    Effective date: January 16, 1997
    Amendment No.: 91
    Facility Operating License No. NPF-47. The amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: July 3, 1996 (61 FR 
34892) The Commission's related evaluation of the amendment is 
contained in a Safety Evaluation dated January 16, 1997.No significant 
hazards consideration comments received. No.
    Local Public Document Room location: Government Documents 
Department, Louisiana State University, Baton Rouge, LA 70803

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

    Date of application for amendment: September 23, 1996
    Brief description of amendment: Changes to Technical Specification 
(TS) to delete a note for the Surveillance Requirement 3.3.7.1 for the 
Engineered Safeguard Actuation System Logic.Date of issuance: January 
6, 1997
    Effective date: January 6, 1997
    Amendment No.: 155
    Facility Operating License No. DPR-72. Amendment revised the TS.
    Date of initial notice in Federal Register: October 23, 1966 (61 FR 
55034) The Commission's related evaluation of the amendment is 
contained in a Safety Evaluation dated January 6, 1997.No significant 
hazards consideration comments received: No.
    Local Public Document Room location: Coastal Region Library, 8619 
W. Crystal Street, Crystal River, Florida 32629

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: October 23, 1996, as supplemented by 
letter dated November 6, 1996.
    Brief description of amendments: The amendments revised Technical 
Specification 3.4.6.1, regarding reactor coolant system leakage 
detection instrumentation, to adopt the requirements found in NUREG-
1431, ``Standard Technical Specifications Westinghouse Plants,'' for 
the reactor coolant system leakage detection instrumentation.
    Date of issuance: January 8, 1997
    Effective date: January 8, 1997
    Amendment Nos.: Unit 1 - Amendment No. 86; Unit 2 - Amendment No. 
73
    Facility Operating License Nos. NPF-76 and NPF-80. The amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: December 4, 1996 (61 FR 
64387) The Commission's related evaluation of the amendments is 
contained in a Safety Evaluation dated January 8, 1997.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

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

    Date of application for amendment: April 19, 1996, and as 
supplemented on August 15, 1996
    Brief description of amendment: The amendment introduces new 
Technical Specification (TS) 3.10.10, ``Single Control Rod Withdrawal - 
Refueling,'' under TS 3.10, ``SPECIAL OPERATIONS.'' The purpose of this 
Special Operations LCO is to permit the withdrawal of a single control 
rod for testing in MODE 5 without imposing the requirements for 
establishing the secondary containment and main control room boundaries 
as normally required during Core Alterations.
    Date of issuance: January 13, 1997
    Effective date: January 13, 1997
    Amendment No.: 112
    Facility Operating License No. NPF-62: The amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: May 22, 1996 (61 FR 
25707) and September 25, 1996 (61 FR 50344). The August 15, 1996, 
submittal changed the focus of the original amendment request, 
therefore, it was re-noticed in the FEDERAL REGISTER. The Commission's 
related evaluation of the amendment is contained in a Safety Evaluation 
dated January 13, 1997.No significant hazards consideration comments 
received: No
    Local Public Document Room location: The Vespasian Warner Public 
Library, 120 West Johnson Street, Clinton, Illinois 61727

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

    Date of application for amendment: September 12, 1995

[[Page 4359]]

    Brief description of amendment: The amendment revises Technical 
Specification 6.3.1 to add a requirement that the Assistant Operations 
Manager hold a senior reactor operator (SRO) license if the Operations 
Manager does not hold an SRO license for Millstone Unit 3.
    Date of issuance: January 7, 1997
    Effective date: As of the date of issuance, to be implemented 
within 60 days.
    Amendment No.: 132
    Facility Operating License No. NPF-49. Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: March 27, 1996 (61 FR 
13530) The Commission's related evaluation of the amendment is 
contained in a Safety Evaluation dated January 7, 1997.No significant 
hazards consideration comments received: No.
    Local Public Document Room location: Learning Resources Center, 
Three Rivers Community-Technical College, 574 New London Turnpike, 
Norwich, CT 06360, and the Waterford Library, ATTN: Vince Juliano, 49 
Rope Ferry Road, Waterford, CT 06385

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

    Date of application for amendment: April 26, 1996, as supplemented 
August 23, 1996.
    Brief description of amendment: The amendment changes requirements 
regarding reactor coolant system leakage testing following refueling 
outage and other sytem pressure testing of reactor coolant system 
following repairs, replacements, or modifications.
    Date of issuance: January 7, 1997
    Effective date: As of the date of issuance to be implemented within 
60 days.
    Amendment No.: 171
    Facility Operating License No. DPR-64: Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: June 5, 1996 (61 FR 
28602) The August 23, 1996, letter provided clarifying information that 
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 January 7, 1997No significant 
hazards consideration comments received: No
    Local Public Document Room location: White Plains Public Library, 
100 Martine Avenue, White Plains, New York 10610.

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

    Date of application for amendment: October 29, 1996
    Brief description of amendment: The amendment corrects an error 
with respect to Table 3.3.2-1, Function 6c of the Technical 
Specifications (TSs) which references the incorrect Required Action for 
inoperable channels of the auxiliary feedwater pump actuation on Steam 
Generator Level - Low Low logic. The TSs are revised to correct the 
Required Action to place the inoperable channel in ``trip'' within 6 
hours or initiate a plant shutdown to Mode 4.
    Date of issuance: January 9, 1997
    Effective date: January 9, 1997
    Amendment No.: 66
    Facility Operating License No. DPR-18: Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: December 4, 1996 (61 FR 
64395) The Commission's related evaluation of the amendment is 
contained in a Safety Evaluation dated January 9, 1997No significant 
hazards consideration comments received: No
    Local Public Document Room location: Rochester Public Library, 115 
South Avenue, Rochester, New York 14610.

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

    Date of application for amendment: October 29, 1996
    Brief description of amendment: This amendment revises the MODE of 
applicability for the motor-driven auxiliary feedwater pump actuation 
on opening of the main feedwater pump breakers to correct an error 
introduced during Amendment No. 61.
    Date of issuance: January 9, 1997
    Effective date: January 9, 1997
    Amendment No.: 67
    Facility Operating License No. DPR-18: Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: December 4, 1996 (61 FR 
64395) The Commission's related evaluation of the amendment is 
contained in a Safety Evaluation dated January 9, 1997No significant 
hazards consideration comments received: No
    Local Public Document Room location: Rochester Public Library, 115 
South Avenue, Rochester, New York 14610.
    Wisconsin Electric Power Company, Docket Nos. 50-266 and 50-301, 
Point Beach Nuclear Plant, Unit Nos. 1 and 2, Town of Two Creeks, 
Manitowoc County, Wisconsin
    Date of application for amendments: April 29, 1996, as supplemented 
October 21, December 2, and December 16, 1996
    Brief description of amendments: These amendments revise Technical 
Specification (TS) Section 15.3.14, ``Fire Protection System,'' and 
Section 15.4.15, ``Fire Protection System,'' and relocate the 
requirements of the fire protection program from the TS and 
incorporate, by reference, the NRC-approved fire protection program 
into the Final Safety Analysis Report. In addition, the amendments 
revise the operating licenses to include the NRC's standard fire 
protection condition. The amendments also approve administrative 
changes consistent with the relocation as well as corrections to 
several typographical errors.
    Date of issuance: January 8, 1997
    Effective date: January 8, 1997, and implementation within 90 days 
from the date of issuance. Implementation shall include the relocation 
of Technical Specification requirements to the appropriate licensee-
controlled document as identified in the licensee's application dated 
April 29, 1996, as supplemented October 21, December 2, and December 
16, 1996, and reviewed in the staff's safety evaluation dated January 
8, 1997.
    Amendment Nos.:  Unit 1 - 170, Unit 2 - 174
    Facility Operating License Nos. DPR-24 and DPR-27: Amendments 
revise the Technical Specifications and the operating licenses.
    Date of initial notice in Federal Register: June 5, 1996 (61 FR 
28621) The October 21, December 2, and December 16, 1996, supplements 
provided corrected license and TS pages and a 90-day implementation 
schedule. These supplements were within the scope of the original 
application and did not change the staff's initial proposed no 
significant hazards considerations determination. The Commission's 
related evaluation of the amendments is contained in a Safety 
Evaluation dated January 8, 1997.No significant hazards consideration 
comments received: No
    Local Public Document Room location: Joseph P. Mann Library, 1516 
Sixteenth Street, Two Rivers, Wisconsin 54241

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

    Date of application for amendment: October 31, 1996

[[Page 4360]]

    Brief description of amendment: The amendment revises Kewaunee 
Nuclear Power Plant Technical Specification 6.9, ``Reporting 
Requirements,'' by deleting the annual requirement to submit a 
description of changes made pursuant to 10 CFR 50.59. Administrative 
changes are also made to correct inconsistencies in the TS Table of 
Contents and in a footnote for Table TS 3.5-1.
    Date of issuance: January 6, 1997
    Effective date: January 6, 1997
    Amendment No.: 131
    Facility Operating License No. DPR-43: Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: December 4, 1996 (61 FR 
64397) The Commission's related evaluation of the amendment is 
contained in a Safety Evaluation dated January 6, 1997.No significant 
hazards consideration comments received: No.
    Local Public Document Room location: University of Wisconsin, 
Cofrin Library, 2420 Nicolet Drive, Green Bay, Wisconsin 54311-7001
    Dated at Rockville, Maryland, this 22nd day of January 1997.
    For The Nuclear Regulatory Commission
Elinor G. Adensam,
Deputy Director, Division of Reactor Projects - III/IV, Office of 
Nuclear Reactor Regulation
[Doc. 97-1994 Filed 1-28-97; 8:45 am]
BILLING CODE 7590-01-F