[Federal Register Volume 61, Number 244 (Wednesday, December 18, 1996)]
[Notices]
[Pages 66702-66721]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: 96-31944]



[[Page 66702]]

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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 November 22, 1996, through December 6, 1996. 
The last biweekly notice was published on December 4, 1996.

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 January 17, 1996, the licensee may file a request for a hearing 
with respect to issuance of the amendment to the subject facility 
operating license and any person whose interest may be affected by this 
proceeding and who wishes to participate as a party in the proceeding 
must file a written request for a hearing and a petition for leave to 
intervene. Requests for a hearing and a petition for leave to intervene 
shall be filed in accordance with the Commission's ``Rules of Practice 
for Domestic Licensing Proceedings'' in 10 CFR Part 2. Interested 
persons should consult a current copy of 10 CFR 2.714 which is 
available at the Commission's Public Document Room, the Gelman 
Building, 2120 L Street, NW., Washington, DC and at the local public 
document room for the particular facility involved. If a request for a 
hearing or petition for leave to intervene is filed by the above date, 
the Commission or an Atomic Safety and Licensing Board, designated by 
the Commission or by the Chairman of the Atomic Safety and Licensing 
Board Panel, will rule on the request and/or petition; and the 
Secretary or the designated Atomic Safety and Licensing Board will 
issue a notice of a hearing or an appropriate order.
    As required by 10 CFR 2.714, a petition for leave to intervene 
shall set forth with particularity the interest of the petitioner in 
the proceeding, and how that interest may be affected by the results of 
the proceeding. The petition should specifically explain the reasons 
why intervention should be permitted with particular reference to the 
following factors: (1) The nature of the petitioner's right under the 
Act to be made a party to the proceeding; (2) the nature and extent of 
the petitioner's property, financial, or other interest in the 
proceeding; and (3) the possible effect of any order which may be 
entered in the proceeding on the petitioner's interest. The petition 
should also identify the specific aspect(s) of the subject matter of 
the proceeding as to which petitioner wishes to intervene. Any person 
who has filed a petition for leave to intervene or who has been 
admitted as a party may amend the petition without requesting leave of 
the Board up to 15 days prior to the first prehearing conference 
scheduled in the proceeding, but such an amended petition must satisfy 
the specificity requirements described above.
    Not later than 15 days prior to the first prehearing conference 
scheduled in the proceeding, a petitioner shall file a supplement to 
the petition to intervene which must include a list of the contentions 
which are sought to be litigated in the matter. Each contention must 
consist of a specific statement of the issue of law or fact to be 
raised or controverted. In addition, the petitioner shall provide a 
brief explanation of the bases of the contention and a concise 
statement of the alleged facts or expert opinion which support the 
contention and on which the petitioner intends to rely in proving the 
contention at the hearing. The petitioner must also provide references 
to those specific sources and documents of which the petitioner is 
aware and on which the petitioner intends to rely to establish those 
facts or expert opinion. Petitioner must provide sufficient information 
to show that a genuine dispute exists with the applicant on a material 
issue of law or fact. Contentions shall be limited to matters within 
the scope of the amendment under consideration. The contention must be 
one which, if proven, would entitle the petitioner to relief. A 
petitioner who fails to file such a supplement which satisfies these 
requirements with respect to at least one

[[Page 66703]]

contention will not be permitted to participate as a party.
    Those permitted to intervene become parties to the proceeding, 
subject to any limitations in the order granting leave to intervene, 
and have the opportunity to participate fully in the conduct of the 
hearing, including the opportunity to present evidence and cross-
examine witnesses.
    If a hearing is requested, the Commission will make a final 
determination on the issue of no significant hazards consideration. The 
final determination will serve to decide when the hearing is held.
    If the final determination is that the amendment request involves 
no significant hazards consideration, the Commission may issue the 
amendment and make it immediately effective, notwithstanding the 
request for a hearing. Any hearing held would take place after issuance 
of the amendment.
    If the final determination is that the amendment request involves a 
significant hazards consideration, any hearing held would take place 
before the issuance of any amendment.
    A request for a hearing or a petition for leave to intervene must 
be filed with the Secretary of the Commission, U.S. Nuclear Regulatory 
Commission, Washington, DC 20555-0001, Attention: Docketing and 
Services Branch, or may be delivered to the Commission's Public 
Document Room, the Gelman Building, 2120 L Street, NW., Washington DC, 
by the above date. Where petitions are filed during the last 10 days of 
the notice period, it is requested that the petitioner promptly so 
inform the Commission by a toll-free telephone call to Western Union at 
1-(800) 248-5100 (in Missouri 1-(800) 342-6700). The Western Union 
operator should be given Datagram Identification Number N1023 and the 
following message addressed to (Project Director): petitioner's name 
and telephone number, date petition was mailed, plant name, and 
publication date and page number of this Federal Register notice. A 
copy of the petition should also be sent to the Office of the General 
Counsel, U.S. Nuclear Regulatory Commission, Washington, DC 20555-0001, 
and to the attorney for the licensee.
    Nontimely filings of petitions for leave to intervene, amended 
petitions, supplemental petitions and/or requests for a hearing will 
not be entertained absent a determination by the Commission, the 
presiding officer or the Atomic Safety and Licensing Board that the 
petition and/or request should be granted based upon a balancing of 
factors specified in 10 CFR 2.714(a)(1) (i)-(v) and 2.714(d).
    For further details with respect to this action, see the 
application for amendment which is available for public inspection at 
the Commission's Public Document Room, the Gelman Building, 2120 L 
Street, NW., Washington, DC, and at the local public document room for 
the particular facility involved.

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

    Date of application for amendment request: September 20, 1996.
    Description of amendment request: The proposed amendments would 
update the Pressure Temperature (P-T) curves contained in the Technical 
Specifications to 22 Effective Full Power Years (EFPYs).
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Involve a significant increase in the probability or 
consequences of an accident previously evaluated because of the 
following:
    The proposed changes merely adjust the reference temperature for 
the limiting beltline material to account for irradiation effects 
and provide the same level of protection as previously evaluated. 
The adjusted reference temperature calculations were performed 
utilizing the guidance contained in Regulatory Guide 1.99, Revision 
2. The change is administrative in nature to reflect the extension 
of the operating limits to 22 EFPY. As such, these changes will not 
significantly increase the probability or consequences of a 
previously evaluated accident.
    2. Create the possibility of a new or different kind of accident 
from any accident previously evaluated because:
    The proposed changes do not create the possibility of a new or 
different kind of accident previously evaluated for Dresden or Quad 
Cities Stations. No new modes of operation are introduced by the 
proposed changes. The revised operating limits are merely an update 
of the old limits by taking into account the effects of irradiation 
on the limiting reactor vessel material. Use of the revised P-T 
curves will continue to provide the same level of protection as was 
previously reviewed and approved. Therefore, the proposed changes do 
not create the possibility of a new or different kind of accident 
from any previously evaluated.
    The associated change to the P-T curves related to this proposed 
amendment does not affect any activities or equipment and are not 
assumed in any safety analysis to initiate any accident sequence for 
Dresden or Quad Cities Stations; therefore, the proposed changes do 
not create the possibility of a new or different kind of accident 
from any previously evaluated.
    3. Involve a significant reduction in the margin of safety 
because:
    The proposed amendment reflects an update of the P-T curves to 
extend the operating limit to 22 EFPY. The revised curves are based 
on the latest NRC guidance along with actual data for the units. The 
new limits retain the margin of safety to the level expected for a 
new vessel, adjusted for irradiation effects as required by 10 CFR, 
Appendix G, thereby maintaining a conservative margin of safety.
    Therefore, the proposed changes do not involve a significant 
reduction in the margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 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: for Dresden, Morris Area 
Public Library District, 604 Liberty Street, Morris, Illinois 60450; 
for Quad Cities, Dixon Public Library, 221 Hennepin Avenue, Dixon, 
Illinois 61021.
    Attorney for licensee: Michael I. Miller, Esquire; Sidley and 
Austin, One First National Plaza, Chicago, Illinois 60603.
    NRC Project Director: Robert A. Capra.

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

    Date of amendment request: October 31, 1996.
    Description of amendment request: The proposed amendments would 
relocate the requirements for seismic monitoring instrumentation from 
the Technical Specifications to licensee controlled documents. The 
Technical Specifications affected are 3/4.3.7.2, ``Seismic Monitoring 
Instrumentation,'' Table 3.3.7.2-1, ``Seismic Monitoring 
Instrumentation,'' Table 4.3.7.2-1, ``Seismic Monitoring 
Instrumentation Surveillance Requirements,'' and Bases Section 3/
4.3.7.2, ``Seismic Monitoring Instrumentation.''
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:


[[Page 66704]]


    (1) Involve a significant increase in the probability or 
consequences of an accident previously evaluated because:
    The function of the seismic monitoring instrumentation is to 
monitor seismic activity above the Operating-Basis Earthquake (OBE) 
threshold, and to record observed seismic data for comparison to 
design basis response spectra. The seismic monitoring 
instrumentation does not provide any function to mitigate an 
accident or the consequences of an accident. The replacement seismic 
monitoring instrumentation will remain in place. The proposed 
Amendment is not a result of any changes to system function, alarm 
setpoints, or main control room annunciators. Rather, the Technical 
Specification requirements (as revised for the replacement 
instrumentation) are being relocated to licensee-controlled 
documents in accordance with NRC Generic Letter 95-10.
    The proposed change relocates requirements and surveillances for 
structures, systems, components or variables that do not meet the 
criteria for inclusion in Technical Specifications as identified in 
the Application of Selection Criteria to the LaSalle Technical 
Specifications. The affected structures, systems, components or 
variables are not assumed to be initiators of analyzed events and 
are not assumed to mitigate accident or transient events. The 
requirements and surveillances for these affected structures, 
systems, components or variables will be relocated from the 
Technical Specifications to an appropriate administratively 
controlled document which will be maintained pursuant to 10 CFR 
50.59. In addition, the affected structures, systems, components or 
variables are addressed in existing surveillance procedures which 
are also controlled by 10 CFR 50.59 and subject to the change 
control provisions imposed by plant administrative procedures, which 
endorse applicable regulations and standards. Therefore, this change 
does not involve a significant increase in the probability or 
consequences of an accident previously evaluated.
    (2) Create the possibility of a new or different kind of 
accident from any accident previously evaluated because:
    The seismic monitoring instrumentation does not provide any 
function to mitigate an accident or the consequences of an accident. 
The replacement seismic monitoring instrumentation will remain in 
place and will provide the same basic function as the existing 
instrumentation. The replacement instrumentation will provide 
enhanced system reliability and will not result in any changes to 
system function, alarm setpoints, or main control room annunciators. 
The Technical Specification requirements (as revised for the 
replacement instrumentation) are being relocated to licensee-
controlled documents in accordance with NRC Generic Letter 95-10.
    The proposed change does not involve any change in the methods 
governing normal plant operation. The proposed change will not 
impose or eliminate any requirements and adequate control of 
existing requirements will be maintained. Thus, this change does not 
create the possibility of a new or different kind of accident from 
any accident previously evaluated.
    (3) Involve a significant reduction in the margin of safety 
because:
    The replacement seismic monitoring instrumentation will have no 
impact on margin of safety. The intended function of the seismic 
monitoring instrumentation, i.e. to record observed seismic data for 
analysis to determine the impact on plant components, will be made 
more reliable by this modification. The Technical Specification 
requirements (as revised for the replacement instrumentation) are 
being relocated to licensee-controlled documents in accordance with 
NRC Generic Letter 95-10.
    The proposed change will not reduce a margin of safety because 
it has no impact on any safety analysis assumptions. In addition, 
the relocated requirements and surveillances for the affected 
structure, system, component or variable continue to meet the same 
requirements as the existing Technical Specifications. However, the 
LCO requirement specified in Section 3.3.7.2.a (to prepare and 
submit a Special Report to the NRC within 10 days of the seismic 
monitoring instrumentation being inoperable for more than 30 days) 
will not be included in the ATR [Administrative Technical 
Requirements] since the Technical Specification Special Report 
requirements are only applicable to the LCOs. Since any future 
changes to these requirements or the surveillance procedures will be 
evaluated per the requirements of 10 CFR 50.59, no reduction in a 
margin of safety will be permitted.
    The existing requirement for NRC review and approval of 
revisions, in accordance with 10 CFR 50.92, to these details 
proposed for relocation does not have a specific margin of safety 
upon which to evaluate. However, since the proposed change is 
consistent with the BWR Standard Technical Specification, NUREG-
1434, Rev. 1 approved by the NRC Staff, revising the Technical 
Specifications to reflect the approved level of detail ensures no 
significant reduction in the margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
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.

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

    Date of amendment request: November 7, 1996.
    Description of amendment request: The proposed amendments would 
change Specification 4.3.1.A.4.b from verifying greater than or equal 
to 17 percent steam generator secondary side wide range water level to 
greater than or equal to 17 percent steam generator secondary side 
narrow range water level.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. The proposed changes do not involve a significant increase in 
the probability or consequences of occurrence of any accident 
previously evaluated.
    Maintaining secondary side steam generator water level greater 
than or equal to 17 percent by wide range level indication is the 
current requirement by the technical specifications. By revising the 
requirement to require using the narrow range water level, no change 
in operating practices or plant configuration is made. The minimum 
requirement of 17 percent by narrow range level indication is more 
restrictive and conservative than 17 percent by wide range 
indication. The requirement to maintain secondary side steam 
generator water level greater than or equal to 17 percent by narrow 
range indication is currently required by operations procedure PT-O, 
Appendix F-1 and will be maintained. This change ensures that the 
requirements for natural circulation cooldown are maintained in Mode 
4. Therefore, changing this surveillance requirement does not 
involve a significant increase in the probability or consequences of 
an accident previously evaluated.
    2. The proposed changes do not create the possibility of a new 
or different kind of accident from any accident previously 
evaluated.
    The proposed changes do not require a physical alteration of the 
plant (no new or different equipment will be installed). The 
Technical Specifications will continue to require OPERABLE steam 
generator(s) for heat removal functions. The Technical 
Specifications will continue to require the performance of SR 
4.3.1.A.4.b. Changing the SR to use narrow level indication 
correctly states the steam generator water level required to support 
heat removal function. Thus, this change does not create the 
possibility of a new or different kind of accident from any accident 
previously evaluated.
    3. The proposed changes do not involve a significant reduction 
in a margin of safety.
    The proposed changes do not result in a significant reduction in 
a margin of safety because it has no impact on any safety analysis 
assumptions. The requirement to have OPERABLE steam generator(s) in 
MODE 4 for heat removal function is maintained. The requirement to 
perform SR 4.3.1.A.4.b is not changed. Changing the SR to use narrow 
level indication correctly states the steam generator water level 
required to support heat

[[Page 66705]]

removal function. Therefore, this change does not involve a 
significant reduction in a margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
requested amendments involve no significant hazards consideration.
    Local Public Document Room location: Waukegan Public Library, 128 
N. County Street, Waukegan, Illinois 60085.
    Attorney for licensee: Michael I. Miller, Esquire; Sidley and 
Austin, One First National Plaza, Chicago, Illinois 60603.
    NRC Project Director: Robert A. Capra.

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

    Date of amendment request: November 7, 1996.
    Description of amendment request: The proposed amendments would 
change the values for the reduced power range neutron flux high 
setpoint trip that are specified when one or more code main steam 
safety valves are inoperable.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. The proposed changes do not involve a significant increase in 
the probability or consequences of occurrence of any accident 
previously evaluated.
    The requirement to change the Power Range Neutron Flux High Trip 
setpoints to the Reduced Setpoint Values of Table 3.7-1 for the most 
restrictive loop if one or more code MSSVs are inoperable is not 
changed by this amendment. As such, no change in operating practices 
or plant configuration is being made.
    The amendment provides new reduced setpoint values for the Power 
Range Neutron flux High Trip to ensure that for the limiting 
transient (Loss of Load/Turbine Trip [LOL/TT]), a secondary side 
overpressurization condition does not occur. The new values were the 
result of calculation using an algorithm provided by Westinghouse in 
Westinghouse Nuclear Safety Advisory Letter NSAL-94-001, ``Operation 
at Reduced Power Levels with Inoperable MSSVs,'' January 25, 1994. 
The new values are much more restrictive than the previous values 
and ensure that the probability or consequences of an accident 
previously evaluated is not increased. Therefore, the new reduced 
setpoint values do not involve a significant increase in the 
probability or consequences of an accident previously evaluated.
    2. The proposed changes do not create the possibility of a new 
or different kind of accident from any accident previously 
evaluated.
    The proposed change does not require a physical alteration of 
the plant (no new or different equipment will be installed to 
implement this change). The Reduced Neutron Flux High Trip setpoints 
ensure that a secondary side overpressurization transient does not 
occur for the most limiting transient. In addition, no new modes of 
operations will be introduced by this change. Thus, this change does 
not create the possibility of a new or different kind of accident 
from any accident previously evaluated.
    3. The proposed changes do not involve a significant reduction 
in a margin of safety.
    This amendment provides new Reduced Power Range Neutron Flux 
High Trip setpoints.The Specification that requires the Power Range 
Neutron Flux High Trip setpoints be changed to the reduced values 
for one or more inoperable MSSVs is not changed. The reduced Trip 
setpoints are the result of new calculations using an algorithm 
provided by Westinghouse in Westinghouse Nuclear Safety Advisory 
Letter NSAL-94-001, ``Operation at Reduced Power levels with 
Inoperable MSSVs,'' January 25, 1994, and ensure the LOL/TT 
transient does not result in a secondary overpressurization. 
Therefore, this change does not involve a significant reduction in a 
margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
requested amendments involve no significant hazards consideration.
    Local Public Document Room location: Waukegan Public Library, 128 
N. County Street, Waukegan, Illinois 60085.
    Attorney for licensee: Michael I. Miller, Esquire; Sidley and 
Austin, One First National Plaza, Chicago, Illinois 60603.
    NRC Project Director: Robert A. Capra.

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

    Date of amendment request: November 7, 1996.
    Description of amendment request: The proposed amendments would 
clarify the operability requirements for the residual heat removal 
(RHR) loops during core alteration operations.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. The proposed changes do not involve a significant increase in 
the probability or consequences of occurrence of any accident 
previously evaluated.
    The ability to remove an RHR loop from operation for up to one 
hour per eight-hour period is currently allowed by technical 
specification 3.13.9.B.b. By adding a reference to LCO [Limiting 
Condition for Operation] 3.13.1.A.4. and adding the requirement to 
suspend CORE ALTERATIONS to Action 3.13.9.B.a. to be consistent with 
3.13.9.B.b., no change in operating practices or plant configuration 
is made. By maintaining the requirement to have an RHR loop in 
operation during MODE 6, and by requiring CORE ALTERATIONS to be 
suspended if an RHR loop is not back in operation after one hour, 
adequate corrective actions are implemented until the RHR loop is 
restored to operating status. Therefore, operation of the system is 
consistent with current technical specifications and this change 
does not involve a significant increase in the probability or 
consequences of an accident previously evaluated.
    2. The proposed changes do not create the possibility of a new 
or different kind of accident from any accident previously 
evaluated.
    The proposed changes do not require a physical alteration of the 
plant (no new or different equipment will be installed to implement 
this change). The Technical Specifications will continue to require 
an RHR loop to be in operation during MODE 6, and will only permit 
the loop to be not in operation for up to one hour in an eight-hour 
period. The Technical Specifications will continue to require 
compliance with these limitations and suspension of CORE ALTERATIONS 
if an RHR loop is not in operation for more than one hour. Thus, 
this change does not create the possibility of a new or different 
kind of accident from any accident previously evaluated.
    3. The proposed changes do not involve a significant reduction 
in a margin of safety.
    The proposed changes do not result in a significant reduction in 
a margin of safety because it has no impact on any safety analysis 
assumptions. The requirement to have an RHR loop in operation during 
MODE 6 is maintained, along with the ability to remove RHR from 
operation for up to one hour per eight-hour period. If an RHR loop 
is not in service beyond 1 hour per TS 3.13.9.B, CORE ALTERATIONS 
will be suspended. Therefore, this change does not involve a 
significant reduction in a margin of safety.

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

[[Page 66706]]

First National Plaza, Chicago, Illinois 60603.
    NRC Project Director: Robert A. Capra.

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

    Date of amendment request: November 7, 1996.
    Description of amendment request: The proposed amendment would 
revise Technical Specification 4.2.9, Service and Instrument Air 
System, to add an additional air compressor.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:
    The proposed change does not:

    (1) Involve a significant increase in the probability or 
consequences of an accident previously evaluated.
    Utilizing the existing piping configuration, both the new and 
the existing air compressors are capable of supporting either 
portion of the Service and Instrument Air System. The addition of 
the fourth air compressor will decrease the probability of an 
accident previously evaluated, because capacity is being added to 
the system. The consequences of an accident previously evaluated 
will not be affected by the addition of a fourth air compressor. The 
Service and Instrument Air System performs the non-safety related 
function of providing compressed air for service use and moisture 
free compressed instrument air for control air demands. The 
instrument air portion is designed so that its operation is required 
for plant reliability, not plant nuclear safety. Safety-related 
equipment supplied by instrument air is designed to fail in its safe 
condition upon loss of instrument air or, safety-related equipment 
(and nonsafety-related equipment determined to be important to 
safety) required to operate subsequent to instrument air failure is 
supplied by backup nitrogen accumulators.
    (2) Create the possibility of a new or different kind of 
accident from any accident previously evaluated.
    The operation of the equipment in the Service and Instrument Air 
System is essentially unchanged. The new air compressor is a similar 
design (nonlubricated), providing additional air volume at a quality 
comparable to the three existing air compressors. Therefore, the 
possibility of an accident of a different kind than any previously 
evaluated has not been created.
    (3) Involve a significant reduction in a margin of safety
    The Technical Specification does not specify a margin of safety 
for the operation of the Service and Instrument Air System, other 
than specifying that [``Instrument and service] air shall be 
supplied by three, nonlubricated air compressors, each rated at 70 
scfm [standard cubic feet per minute]. Instrument air shall also 
pass through a dryer.'' Addition of a fourth air compressor will 
increase the available capacity, thus increasing the margin of 
safety. Therefore, adding the statement ``and one, nonlubricated air 
compressor rated at 100 scfm'' to Technical Specification 4.2.9. 
will not reduce 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: North Central Michigan 
College, 1515 Howard Street, Petoskey, Michigan 49770.
    Attorney for licensee: Judd L. Bacon, Esquire, Consumers Power 
Company, 212 West Michigan Avenue, Jackson, Michigan 49201.
    NRC Project Director: John N. Hannon.

Duquesne Light Company, et al., Docket Nos. 50-334 and 50-412, Beaver 
Valley Power Station, Unit Nos. 1 and 2, Shippingport, Pennsylvania

    Date of amendment request: October 4, 1996.
    Description of amendment request: The proposed amendment would 
revise the surveillance requirements in Technical Specifications (TSs) 
4.1.2.3.1, 4.1.2.4.1, 4.5.2.b, and 4.6.2.1.b and associated Bases. The 
subject surveillance requirements are applicable to the charging/high 
head safety injection pumps, low head safety injection pump, and the 
containment quench spray pumps. The proposed changes would replace the 
current specific test acceptance criteria contained in these 
surveillance requirements with requirements to verify pump performance 
in accordance with the Inservice Testing Program, the Emergency Core 
Cooling System Flow Analysis, or the Containment Integrity Safety 
Analysis, as applicable. The proposed changes would also make minor 
editorial changes in these TSs and make conforming changes in the TS 
Index pages.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Does the change involve a significant increase in the 
probability or consequences of an accident previously evaluated?
    The change does not result in a modification to plant equipment 
nor does it affect the manner in which the plant is operated. Since 
the physical plant equipment and operating practices are not 
changed, as noted above, there is no change in the probability of an 
accident previously evaluated.
    The proposed change will not lower the pump performance 
operability criteria for the charging/high head safety injection, 
low head safety injection and quench spray pumps, as assumed in the 
safety analysis. The required values for developed pump head and 
flow will continue to satisfy accident mitigation requirements and 
will be maintained and controlled in the Inservice Testing (IST) 
Programs(s).
    Since the proposed change does not lower the pump's performance 
acceptance criteria, as assumed in the safety analysis, the 
containment depressurization system will continue to meet its design 
basis requirements. The proposed change will not impose additional 
challenges to the containment structure in terms of peak pressure. 
The calculated offsite dose consequences of a design basis accident 
(DBA) will remain unchanged since the one hour release duration and 
source term remain unchanged. The ability of the emergency core 
cooling system (ECCS) subsystems to provide sufficient emergency 
core cooling capability in the event of a loss of coolant accident 
(LOCA) remains unchanged. Therefore, peak cladding temperatures 
during a LOCA will continue to remain within acceptable limits. The 
ability of the ECCS subsystems to provide sufficient long term core 
cooling capability in the recirculation mode during the accident 
recovery period remains unchanged. The charging pumps, as part of 
the boron injection system, will continue to provide sufficient flow 
to ensure negative reactivity control during each mode of facility 
operation. Future changes to the pump head and flow requirements 
will be made under the 10 CFR 50.59 process to ensure that the 
system performance requirements continue to be met.
    The proposed change to the Bases section will ensure that safety 
analyses assumptions for assumed pump performance continue to be 
met. The words ``required developed head'' will be clearly defined 
to reflect that they refer to the value(s) assumed in the safety 
analysis for the pump's developed head at a specific or a given 
point. The proposed changes to the Index pages and the footnote in 
LCO 3.1.2.4 are administrative in nature and do not affect plant 
safety.
    Based on the above discussion, it is concluded that this change 
does not involve a significant increase in the probability or 
consequences of an accident previously evaluated.
    2. Does the change create the possibility of a new or different 
kind of accident from any accident previously evaluated?
    The proposed change does not alter the method of operating the 
plant. The charging pumps will continue to be in service during 
plant operation and be available to perform their function as high 
head safety injection pumps. This proposed change does not pose 
additional challenges to the design or function of the charging 
pumps. The low head safety injection and quench spray

[[Page 66707]]

systems are accident mitigation systems and are normally in standby. 
System operation would be initiated as required to mitigate the 
consequences of a DBA. The charging/high head safety injection, low 
head safety injection and quench pumps will continue to provide 
sufficient flow to mitigate the consequences of a DBA. These 
systems' operation continues [sic] [continue] to fulfill the safety 
functions for which they were designed and no changes to plant 
equipment will occur. As a result, an accident which is new or 
different than any already evaluated in the Updated Final Safety 
Analysis Report will not be created due to this change.
    Therefore, the proposed change does not create the possibility 
of a new or different kind of accident from any accident previously 
evaluated.
    3. Does the change involve a significant reduction in a margin 
of safety? The surveillance requirements for demonstrating that the 
pumps are operable will continue to assure the ability of the system 
to satisfy its design function. Therefore, the proposed change will 
not affect the ability of these systems to perform their safety 
function.
    The containment systems' design requirements to restore the 
containment to subatmospheric condition within one hour will 
continue to be satisfied. This proposed change does not have an 
effect on the containment peak pressure since the charging/high head 
safety injection, low head safety injection and quench spray pumps' 
performance requirements are not being lowered. The ability of the 
ECCS subsystems to provide sufficient emergency core cooling 
capability in the event of a LOCA remains unchanged. Therefore, peak 
cladding temperatures during a LOCA will continue to remain within 
acceptable limits. The ability of the ECCS subsystems to provide 
sufficient long term core cooling capability in the recirculation 
mode during the accident recovery period remains unchanged. The 
charging pumps, as part of the boron injection system, will continue 
to provide sufficient flow to ensure negative reactivity control 
during each mode of facility operation. There is no resultant change 
in dose consequences since source term remains unchanged and the 
containment will continue to reach a subatmospheric pressure within 
the first hour following a DBA.
    Each pump's performance requirements will continue to be 
controlled in a manner to ensure safety analysis assumptions are 
met.
    Therefore, based on the above discussions, it can be concluded 
that the proposed change does not involve a significant reduction in 
a margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: B. F. Jones Memorial Library, 
663 Franklin Avenue, Aliquippa, PA 15001.
    Attorney for licensee: Jay E. Silberg, Esquire, Shaw, Pittman, 
Potts & Trowbridge, 2300 N Street, NW., Washington, DC 20037.
    NRC Project Director: John F. Stolz.

Florida Power and Light Company, et al., Docket Nos. 50-335 and 50-389, 
St. Lucie Plant Units 1 and 2, St. Lucie County, Florida

    Date of amendment request: October 31, 1996.
    Description of amendment request: The proposed amendments will 
revise administrative controls Technical Specification (TS) 6.5.1, 
``Facility Review Group (FRG),'' and TS 6.8, ``Procedures and 
Programs.'' The revisions to TS 6.5.1 reduce the scope of procedures 
and procedure changes which require review by the FRG, transfer 
approval of certain procedures from the Plant Manager to the FRG, and 
require copies of FRG meeting minutes be provided to the Plant Manager. 
The changes to TS 6.8 reflect the corresponding changes in TS 6.5.1, 
and expand the scope of the section on temporary changes to procedures.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below.

    (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 amendments revise certain administrative controls 
involved with the on-site programmatic process for review and 
approval of plant procedures. Specifications that are in place to 
provide assurance that the unit operating staff qualifications are 
acceptable, and that written procedures are established, implemented 
and maintained for safety related activities are not being changed. 
The revisions are consistent with industry standards established 
pursuant to 10 CFR Part 50, Appendix B, and do not alter any 
parameter or equipment performance assumptions that are contained in 
plant safety analyses to evaluate the initiation or consequences of 
an accident. Therefore, operation of either facility in accordance 
with its proposed amendment would not involve a significant increase 
in the probability or consequences of an accident previously 
evaluated.
    (2) Operation of the facility in accordance with the proposed 
amendment would not create the possibility of a new or different 
kind of accident from any accident previously evaluated.
    The proposed amendments will not change the physical plant or 
the modes of plant operation defined in the Facility License for 
either St. Lucie unit. Changes proposed for the administrative 
controls do not involve the addition or modification of equipment 
nor do they alter the design or operation of plant systems. 
Therefore, operation of either facility in accordance with its 
proposed amendment would not create the possibility of a new or 
different kind of accident from any accident previously evaluated.
    (3) Operation of the facility in accordance with the proposed 
amendment would not involve a significant reduction in a margin of 
safety.
    The proposed amendments revise certain administrative controls 
involving the on-site programmatic process for review and approval 
of plant procedures. The scope, or the requirement to establish, 
maintain, and implement procedures for activities that could affect 
nuclear safety are not being changed. The proposed changes are 
consistent with approved industry standards and do not alter the 
basis for any technical specification that is related to the 
establishment of, or the maintenance of, a nuclear safety margin. 
Therefore, operation of either facility in accordance with its 
proposed amendment would not involve a significant reduction in a 
margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: Indian River Junior College 
Library, 3209 Virginia Avenue, Fort Pierce, Florida 34954-9003.
    Attorney for licensee: M. S. Ross, Attorney, Florida Power & Light, 
11770 US Highway 1, North Palm Beach, Fl 33408.
    NRC Project Director: Frederick J. Hebdon.

GPU Nuclear Corporation, et al., Docket No. 50-219, Oyster Creek 
Nuclear Generating Station, Ocean County, New Jersey

    Date of amendment request: October 31, 1996 (TSCR 205).
    Description of amendment request: The proposed change requests 
deletion of Technical Specification Table 3.5.2 which lists automatic 
primary containment isolation valves. In addition, this change request 
clarifies the applicability of an action statement which applies to 
several limiting conditions for operation in Section 3.5 and deletes 
closure time requirements for several automatic isolation valves in 
Section 4.5.F.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:


[[Page 66708]]


    1. The proposed deletion of the automatic primary containment 
isolation valve Table 3.5.2 and closure times for several valves in 
Specification 4.5.F.1 are administrative in nature and do not affect 
the purpose, function, operability and testing requirements of the 
automatic primary containment isolation valves or the isolation 
condenser isolation valves. The required action contained in 
Specification 3.5.A.7 has been moved to the associated 
specifications and has not changed. Capitalizing definitions and 
deleting unneeded pages are also administrative changes which 
enhance the usability of the Technical Specifications. Therefore, 
the proposed changes do not increase the probability of occurrence 
or consequence of an accident previously evaluated.
    2. The proposed changes are administrative and do not involve a 
physical change to plant configuration nor do they affect the 
performance of any equipment. Existing limiting conditions for 
operation and surveillance requirements are retained. Therefore, the 
possibility of a new or different kind of accident from any accident 
previously evaluated is not created.
    3. Deleting the list of valves in Table 3.5.2 and valve closure 
times in Specification 4.5.F.1 are administrative changes which do 
not affect the purpose or function of the automatic primary 
containment isolation valves. The listing of the automatic primary 
containment isolation valves and stroke time requirements will be in 
controlled plant procedures. Changes to the list or closure times 
can be made in accordance with review procedures required by Section 
6.5 of the Technical Specifications and 10 CFR 50.59. Similarly, 
inserting the statement of required action in Specification 3.5.A.7 
into the Specifications to which it applies does not modify the 
condition or the action to be taken and is an administrative change 
which clarifies the Technical Specifications. Therefore, the margin 
of safety is not reduced.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: Ocean County Library, 
Reference Department, 101 Washington Street, Toms River, NJ 08753.
    Attorney for licensee: Ernest L. Blake, Jr., Esquire. Shaw, 
Pittman, Potts & Trowbridge, 2300 N Street, NW., Washington, DC 20037.
    NRC Project Director: John F. Stolz.

GPU Nuclear Corporation, et al., Docket No. 50-219, Oyster Creek 
Nuclear Generating Station, Ocean County, New Jersey

    Date of amendment request: November 12, 1996, as supplemented 
November 27, 1996 (TSCR 224).
    Description of amendment request: The proposed technical 
specification change will reflect the implementation of the revised 10 
CFR Part 20, ``Standards for Protection Against Radiation.''
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Operation of the facility in accordance with the proposed 
amendment would not involve a significant increase in the 
probability of occurrence or the consequences of an accident 
previously evaluated.
    The proposed revisions to the liquid release rate limits and 
bases and gaseous effluent bases will not result in a change in the 
types or amounts of effluents released nor will there be an increase 
in individual or cumulative radiation exposures. In addition, these 
changes do not impact the operation or design of any plant 
structures, systems, or components. These changes ensure compliance 
with 10 CFR 50.36a and 10 CFR 50 Appendix I and result in levels of 
radioactive materials in effluents being maintained ALARA [as low as 
is reasonably achievable]. The revision to the high radiation area 
controls and dose measurement distance will ensure areas are 
conservatively posted as high radiation areas in compliance with 10 
CFR 20.1601(a)(1) and provide controls to ensure individuals are not 
overexposed. Other proposed changes consist of revisions to 10 CFR 
20 references to recognize the new section numbers, and 
administrative controls for record keeping to maintain compliance 
with the new Part 20.
    These changes will not result in a change to plant design or 
operation. Therefore, it can be concluded that the proposed changes 
do not involve an increase in 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 changes do not affect the plant design or operation nor do 
they result in a change to the configuration of any equipment. There 
will be no change in the types or increase in the amount of 
effluents released offsite.
    Therefore, this proposed change cannot create the possibility of 
a new or different 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 a margin of 
safety.
    The proposed revisions do not involve any change in the types or 
increase in the amount of effluents released offsite. The proposed 
changes do not involve any actual change in the methodology used in 
the control of radioactive wastes or radiological environmental 
monitoring. The methodology that will be used in the control of 
radioactive effluents and calculation of effluent monitor setpoints 
will result in the same effluent release rate as the current 
methodology now being used. The operational flexibility needed for 
releases allows the use of limits as proposed. In addition, the 
changes in measurement distances for determination of high radiation 
areas will not result in an increase in individual or cumulative 
occupational radiation exposures since it will result in a more 
conservative identification of high radiation areas. Compliance with 
the limits of the new 10 CFR 20.1301 will be demonstrated by 
operating within the limits of 10 CFR 50 Appendix I and 40 CFR 190. 
Thus, operation of the facility in accordance with the proposed 
amendment 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: Ocean County Library, 
Reference Department, 101 Washington Street, Toms River, NJ 08753.
    Attorney for licensee: Ernest L. Blake, Jr., Esquire. Shaw, 
Pittman, Potts & Trowbridge, 2300 N Street, NW., Washington, DC 20037.
    NRC Project Director: John F. Stolz.

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

    Date of amendment request: August 29, 1996, as supplemented October 
3, 1996. The October 3, 1996, submittal contained editorial changes 
only and did not change the initial no significant hazards 
consideration evaluation.
    Description of amendment request: The purpose of this amendment 
request is to incorporate certain improvements from the Standard 
Technical Specifications for B&W Plants, NUREG-1430.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration (SHC), which is presented below:
    GPU Nuclear has determined that this Technical Specification Change 
Request involves no significant hazards consideration as defined in 10 
CFR 50.92 because:

    1. Operation of the facility in accordance with the proposed 
amendment would not involve a significant increase in the 
probability of occurrence or the

[[Page 66709]]

consequences of an accident previously evaluated. The proposed 
amendment deletes limiting conditions for operation (LCOs) from the 
TMI-1 Technical Specifications that are no longer required to be 
addressed in Technical Specifications per 10 CFR 50.36(c)(2)(ii). 
The proposed amendment deletes Surveillance Requirements from the 
TMI-1 Technical Specifications that are related to the LCOs to be 
deleted. These items are addressed in licensee controlled documents. 
Certain design feature specifications are also to be deleted 
consistent with the RSTS [Revised Standard Technical Specifications] 
for B&W plants. The proposed changes do not modify the operation, 
limits or controls of systems, structures or components relied upon 
to prevent or mitigate the consequences or accidents previously 
evaluated.
    Also, the reliability of systems and components relied upon to 
prevent or mitigate the consequences of accidents previously 
evaluated is not degraded by the proposed changes. Therefore, this 
change does not involve a significant increase in the probability of 
occurrence or the 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 because no 
new failure modes are created by the proposed changes.
    3. Operation of the facility in accordance with the proposed 
amendment would not involve a significant reduction in a margin of 
safety because 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 Project Director: John F. Stolz.

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

    Date of amendment request: September 20, 1996.
    Description of amendment request: The proposed amendment would 
revise the Nine Mile Point Unit 1 (NMP1) Technical Specifications that 
involve the frequencies of surveillance requirements stated in Tables 
4.6.2a, 4.6.2b, 4.6.2g, and 4.6.11, and Sections 4.2.5b(1), 4.3.2b, 
4.3.6b(1), 4.3.6b(2), 4.3.6b(3), 4.3.6b(4), 4.3.6c(2), 4.6.13b.1, and 
4.6.13b.2. The surveillances associated with these tables and sections 
are currently satisfied during NMP1 refueling outages prior to restart 
of the unit. The proposed changes would permit surveillance testing 
either while the reactor is operating or during outage periods not 
associated with refueling. The requirements of the surveillance 
sections and tables addressed by this request that are not changed to 
be performed at power are being changed to allow surveillance credit to 
be taken for performance of the associated surveillances while the 
plant is in the Cold Shutdown, Refueling, or Major Maintenance modes. 
In addition to these proposed changes, typographical errors are 
corrected.
    Basis for proposed no significant hazards consideration 
determination: The licensee states that: ``The periods between 
surveillances will not be inappropriately lengthened. For the affected 
surveillances, NMP1 administrative controls will require that the 
interval between surveillance testing not exceed a period equal to 1.25 
times the nominal 24 months frequency (no longer than 30 months). The 
NMP1 plant preventive maintenance and surveillance database will be 
revised accordingly.''
    The licensee groups the systems affected by this request into four 
categories:

    Category 1: The associated system will remain operable and able 
to automatically perform its safety function during performance of 
surveillances that satisfy the proposed surveillance requirement.
    Category 2: The system is required for monitoring purposes only 
and provides no automatic safety actuation function and redundant, 
or redundant and alternate channels are available for required 
monitoring.
    Category 3: There is no change in the system configuration or 
plant operating conditions during the performance of associated 
surveillances whether the plant is shutdown for refueling or 
shutdown for maintenance. The surveillances performed to meet the 
requirements of NMP1 Technical Specifications Tables 4.6.2a 
Parameter 8 and 4.6.2g Parameter 6 are included in this category and 
may also be completed in concurrence with a unit shutdown. The only 
difference between the proposed changes and the normal unit shutdown 
sequence is that the mode switch may be taken to ``Shutdown'' in 
order to scram the plant. The response of the plant is the same as 
it is under the current plant shutdown procedures. There are no 
other differences in testing techniques or testing criteria from 
those previously required by the NMP1 Technical Specifications.
    Category 4: The system or equipment is isolated or out of 
service during the performance of the required surveillances. The 
associated surveillance may be performed concurrently with quarterly 
valve stroking, at which time the system or equipment is already out 
of service.

    As required by 10 CFR 50.91(a), the licensee has provided its 
analysis of the issue of no significant hazards consideration, which is 
presented below:
    The operation of Nine Mile Point Unit 1, in accordance with the 
proposed amendment, will not involve a significant increase in the 
probability or consequence of an accident previously evaluated.
    Each of the four categories [* * *] are evaluated separately below:

    Category 1: The associated systems will remain operable and able 
to automatically fulfill as designed any required safety functions 
that may become necessary during performance of required 
surveillances. No physical change to the plant design, materials, or 
standards is involved. No change to instrumentation operating 
characteristics outside current tolerances will be made. No plant 
transients will be initiated as a result of the proposed changes. No 
initiator of any accident previously evaluated is adversely 
affected. No system required to actuate to respond to any accident 
previously evaluated in the UFSAR [Updated Final Safety Analysis 
Report] is adversely affected by the proposed change.
    Category 2: The associated systems will be required for 
monitoring purposes only and provide no automatic safety actuation 
function and redundant, or redundant and alternate channels are 
available for required monitoring. Since redundant monitoring 
instrumentation will still be available as required by the technical 
specifications, the associated systems' functions in accident 
mitigation are not affected. No physical change to the plant design, 
materials, or standards is involved. No change to instrumentation 
operating characteristics outside current tolerances will be made. 
No plant transients will be initiated as a result of the proposed 
changes. No initiator of any accident previously evaluated is 
adversely affected. No system required to actuate to respond to any 
accident previously evaluated in the UFSAR is adversely affected by 
the proposed changes.
    Category 3: There will be no change in the system configuration 
or plant operating conditions during the performance of associated 
surveillances. The associated system's ability to perform required 
safety functions will not be affected, whether the plant is shutdown 
for refueling or shutdown for maintenance. The surveillances 
performed to meet the requirements of NMP1 Technical Specifications 
Tables 4.6.2a Parameter B and 4.6.2g Parameter 6 are included in 
this category and may also be performed in concurrence with a unit 
shutdown. The only difference between the proposed changes and the 
normal unit shutdown sequence is that the mode switch

[[Page 66710]]

may be taken to ``Shutdown'' in order to scram the plant. The 
response of the plant is the same as it is under the current plant 
shutdown procedures. There are no other differences in testing 
techniques or testing criteria from those previously required by the 
NMP1 Technical Specifications. No physical change to the plant 
design, materials, or standards is involved. No change to 
instrumentation operating characteristics outside current tolerances 
will be made. No unexpected plant transients will be initiated as a 
result of the proposed changes. No initiator of any accident 
previously evaluated is adversely affected. No system required to 
actuate to respond to any accident previously evaluated in the UFSAR 
is adversely affected by the proposed changes.
    Category 4: The associated system or equipment will be isolated 
or out of service during the performance of the required 
surveillances. The associated surveillances will be performed during 
quarterly valve stroking, at which time the system or equipment is 
already out of service. No physical change to the plant design, 
materials, or standards is involved. No change to instrumentation 
operating characteristics outside current tolerances will be made. 
No plant transients will be initiated as a result of the proposed 
changes. No initiator of any accident previously evaluated is 
adversely affected. No system required to actuate to respond to any 
accident previously evaluated in the UFSAR is adversely affected by 
the proposed changes.

    The correction of the typographical errors is administrative only 
and has no affect on plant systems or procedures. In all cases, 
equipment used for accident mitigation is not adversely affected. The 
ability of the operators to safely shut down NMP1 is not impaired. The 
changes will not adversely affect any accident precursor or initiator 
of any accident. For these reasons, the proposed changes will not 
involve a significant increase in the probability or consequences of an 
accident previously evaluated.
    The operation of Nine Mile Point Unit 1, in accordance with the 
proposed amendment, will not create the possibility of a new or 
different kind of accident from any accident previously evaluated.
    Each of the four categories [* * *] are evaluated separately below.

    Category 1: The associated systems will remain operable and able 
to automatically perform as designed any required safety functions 
that may become necessary during performance of required 
surveillances. No physical change to the plant design, materials, or 
standards is involved. No change to instrumentation operating 
characteristics outside current tolerances will be made. No accident 
initiator or failure of a different type than previously identified 
in the UFSAR is introduced. No different or new plant transients may 
result from those previously evaluated in the UFSAR.
    Category 2: The associated systems will be required for 
monitoring purposes only and provide no automatic safety actuation 
function. Since redundant, or redundant and alternate monitoring 
instrumentation will still be available as required by the technical 
specifications, the associated systems' functions in accident 
mitigation are not affected. No physical change to the plant design, 
materials, or standards is involved. No change to instrumentation 
operating characteristics outside current tolerances will be made. 
No accident initiator or failure of a different type than previously 
identified in the UFSAR is introduced. No different or new plant 
transients may result from those previously evaluated in the UFSAR.
    Category 3: There will be no change in the system configuration 
or plant operating conditions during the performance of associated 
surveillances. The associated system's ability to perform required 
safety functions will not be affected, whether the plant is shutdown 
for refueling or shutdown for maintenance. The surveillances 
performed to meet the requirements of NMP1 Technical Specifications 
Tables 4.6.2a Parameter 8 and 4.6.2g Parameter 6 are included in 
this category and may also be performed in concurrence with a unit 
shutdown. The only difference between the proposed changes and the 
normal unit shutdown sequence is that the mode switch may be taken 
to ``Shutdown'' in order to scram the plant. The response of the 
plant is the same as it is under the current plant shutdown 
procedures. There are no other differences in testing techniques or 
testing criteria from those previously required by the NMP1 
Technical Specifications. No physical change to the plant design, 
materials, or standards is involved. No change to instrumentation 
operating characteristics outside current tolerances will be made. 
No unexpected plant transients will be initiated as a result of the 
proposed changes. No accident initiator or failure of a different 
type than previously identified in the UFSAR is introduced. No 
different or new plant transients may result from those previously 
evaluated in the UFSAR.
    Category 4: The associated system or equipment will be isolated 
or out of service during the performance of the required 
surveillance. The associated surveillances will be performed during 
quarterly valve stroking, at which time the system or equipment is 
already out of service. No physical change to the plant design, 
materials, or standards is involved. No change to instrumentation 
operating characteristics outside current tolerances will be made. 
No plant transients will be initiated as a result of the proposed 
changes. No accident initiator or failure of a different type than 
previously identified in the UFSAR is introduced. No different or 
new plant transients may result from those previously evaluated in 
the UFSAR.

    The correction of the typographical errors is administrative only 
and has no affect on plant systems or procedures. In all cases, the 
changes will not adversely affect any accident precursor or initiator 
of any accident and, therefore, the changes do not introduce any new 
failure modes or conditions that may create a new or different 
accident. For these reasons, the proposed changes do not create the 
possibility of a new or different kind of accident from any accident 
previously evaluated in the UFSAR.
    The operation of Nine Mile Point Unit 1, in accordance with the 
proposed amendment, will not involve a significant reduction in a 
margin of safety.
    Each of the four categories [* * *] are evaluated separately below.

    Category 1: The associated systems will remain operable and able 
to automatically perform required safety functions during 
performance of surveillances that satisfy the surveillance 
requirement. There will be no effective change in the interval of 
the affected surveillances. The probability of instrument drift or 
the ability to detect a failed or drifted instrument remains 
unchanged. No physical change to the plant design, materials, or 
standards is involved. No change to instrumentation operating 
characteristics outside current tolerances will be made. No system 
required to actuate to respond to any accident is adversely affected 
by the proposed changes. Since each system's operability is not 
affected, the margin of safety associated with these systems will 
not be significantly reduced.
    Category 2: The associated systems will be required for 
monitoring purposes only and provide no automatic safety actuation 
function. Redundant, or redundant and alternate monitoring 
instrumentation will still be available as required by the technical 
specifications during the performance of the associated 
surveillances. No physical change to the plant design, materials, or 
standards is involved. No change to instrumentation operating 
characteristics outside current tolerances will be made. There will 
be no effective change in the intervals of the affected 
surveillances. The probability of instrument drift or the ability to 
detect a failed or drifted instrument remains unchanged. No plant 
transients will be initiated as a result of the proposed changes. No 
initiator of any accident previously evaluated is adversely 
affected. No system required to actuate to respond to any accident 
is adversely affected by the proposed changes. Therefore, the 
associated systems' functions in accident mitigation are not 
affected, and no margin of safety will be significantly reduced.
    Category 3: There will be no change in the system configuration 
or plant operating conditions during the performance of associated 
surveillances, the associated system's ability to perform required 
safety functions will not be affected, whether the plant is shutdown 
for refueling or shutdown for maintenance. The surveillances 
performed to meet the requirements of NMP1 Technical Specifications 
Tables 4.6.2a Parameter 8 and 4.6.2g Parameter 6 may also be 
completed in concurrence with a unit shutdown. The only difference 
between the proposed changes and the normal unit shutdown sequence 
is that the mode switch may be taken to ``Shutdown'' in order to

[[Page 66711]]

scram the plant. The response of the plant is the same as it is 
under the current plant shutdown procedures. There are no other 
differences in testing techniques or testing criteria from those 
previously required by the NMP1 Technical Specifications. No 
physical change to the plant design, materials, or standards is 
involved. No change to instrumentation operating characteristics 
outside current tolerances will be made. There will be no effective 
change in the intervals of the affected surveillances. The 
probability of instrument drift or the ability to detect a failed or 
drifted instrument remains unchanged. No unexpected plant transients 
will be initiated as a result of the proposed changes. No initiator 
of any accident if adversely affected. No system required to actuate 
to respond to any accident previously evaluated is adversely 
affected by the proposed changes. Therefore, no margin of safety 
will be significantly reduced.
    Category 4: The associated system or equipment will be isolated 
or out of service during the performance of the required 
surveillances. The associated surveillances will be performed during 
quarterly valve stroking, at which time the system or equipment will 
already be out of service. No physical change to the plant design, 
materials, or standards is involved. No change to instrumentation 
operating characteristics outside current tolerances will be made. 
There will be no effective change in the intervals of the affected 
surveillances. The probability of instrument drift or the ability to 
detect a failed or drifted instrument remains unchanged. No plant 
transients will be initiated as a result of the proposed changes. No 
accident initiator or failure of a different type than identified in 
the UFSAR is introduced. Therefore, no margin of safety will be 
significantly reduced.

    The correction of the typographical errors is administrative only 
and has no affect on plant systems or procedures. In all cases, the 
changes will not adversely affect any accident precursor or initiator 
of any accident and, therefore, the changes do not introduce any new 
failure modes or conditions that may create a new or different 
accident. None of the proposed changes involve physical modification of 
the plant or alterations to any accident or transient analysis. 
Therefore, for this and the above reasons, these proposed changes do 
not involve any 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: Reference and Documents 
Department, Penfield Library, State University of New York, Oswego, New 
York 13126.
    Attorney for licensee: Mark J. Wetterhahn, Esquire, Winston & 
Strawn, 1400 L Street, NW., Washington, DC 20005-3502.
    NRC Project Director: S. Singh Bajwa, Acting Director.

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

    Date of amendment request: October 16, 1996.
    Description of amendment request: The proposed amendment would 
change certain requirements stated in Technical Specification 3/4.8.1, 
``AC Sources''. The requirements are related to the emergency diesel 
generators (EDGs). The proposed changes would:
    1. Increase the EDG fuel storage system minimum volume requirements 
specified in Limiting Condition for Operation 3.8.1.1.b.2;
    2. Add a footnote applicable to Surveillance Requirement 
4.8.1.1.2.f to qualify the words during shutdown. The footnote would 
allow the option of performing selected surveillances, or portions 
thereof, during conditions or modes other than shutdown;
    3. Delete from Surveillance Requirement 4.8.1.1.2.f.14 the 
requirement to verify that the cooling tower fans start automatically 
on a Tower Actuation signal; and
    4. Delete Surveillance Requirement 4.8.1.1.2.h.2 which specifies 
performing a periodic pressure test on the ASME Code Class 3 diesel 
fuel oil piping.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration. The NRC staff has reviewed the licensee's analysis 
against the standards of 10 CFR 50.92(c). The NRC staff's review is 
presented below.
    A. The changes do not involve a significant increase in the 
probability or consequences of an accident previously evaluated (10 CFR 
50.92(c)(1)).
1. Limiting Condition for Operation  3.8.1.1.b.2
    The proposed change increases the minimum EDG fuel oil storage 
requirement to account for various factors that may affect the fuel 
consumption rate. The revised storage requirement reflects actual EDG 
test data and accounts for external variables including fuel oil 
specific gravity, heating value of the fuel, and ambient conditions. 
The proposed increase in the minimum volume storage requirement is 
conservative and ensures that there will be at least a 7 day supply of 
fuel oil stored for each EDG to meet the maximum Engineered Safety 
Feature load requirements following a loss of power and a design basis 
accident as described in Updated Final Safety Analysis Report (UFSAR) 
Section 9.5.4.1, Diesel Generator Fuel Oil Storage and Transfer 
System--Design Basis. Therefore, the proposed change does not involve a 
significant increase in the probability or consequences of an accident 
previously evaluated.
2. Surveillance Requirement  4.8.1.1.2.f
    The proposed change qualifies the requirement to perform EDG 
surveillance requirements ``during shutdown''. Because the terms Hot 
Shutdown and Cold Shutdown are defined in the TSs as operating modes or 
conditions, the requirement to perform certain surveillances during 
shutdown may be misinterpreted, as noted in NRC Generic Letter 91-04. 
The proposed footnote would permit certain maintenance and testing 
activities to be performed during conditions or modes other than 
shutdown. The proposed footnote to Surveillance Requirement 4.8.1.1.2.f 
would not alter the intent or the method by which the surveillances are 
conducted, and the acceptance criteria for the surveillances would be 
unchanged. The footnote would not degrade the ability of the EDGs to 
perform their intended function, and it would not affect the response 
of the EDGs to a loss of power as described in the UFSAR. Since plant 
response to an accident would not change and since failure of an EDG 
could not initiate any of the accidents evaluated in the UFSAR, the 
proposed footnote would not alter the probability or consequences of an 
accident previously analyzed.
3. Surveillance Requirement  4.8.1.1.2.f.14
    The cooling tower functions as the ultimate heat sink following a 
seismic event which results in blockage of the circulating water 
tunnels and therefore a loss of service water. Amendment 18 eliminated 
the requirement for automatic start of the cooling tower fans; 
therefore, the automatic-start function for the cooling tower fans has 
been defeated by placing the control switch in ``Pull-to-Lock''. The 
proposed change to delete the automatic fan start reference from 
Surveillance Requirement 4.8.1.1.2.f.14 is administrative only to 
correct an oversight since the requirement should have been deleted 
with the issuance of Amendment 18. The proposed deletion does not 
affect the manner by which the facility is operated or involve any

[[Page 66712]]

changes to equipment or features which affect the operational 
characteristics of the facility. Since there is no change to the 
facility or operating procedures, there is no affect upon the 
probability or consequences of any accident previously analyzed.
4. Surveillance Requirement  4.8.1.1.2.h.2
    The ASME Code, Section XI, including applicable ASME Code Cases as 
authorized by the NRC, provides alternate test methods to use in lieu 
of a 110% hydrostatic pressure test that is not practical to perform on 
the EDG fuel oil system as currently designed. With the proposed 
deletion of Surveillance Requirement 4.8.1.1.2.h.2, the provisions of 
Surveillance Requirement 4.0.5 and the ASME Code along with NRC-
authorized Code Cases would be utilized as an equivalent testing 
requirement to ensure the continued integrity of the diesel fuel oil 
system. Therefore, since the reliability of the EDG fuel oil system 
will not be reduced, the probability or consequences of any accident 
previously evaluated is not increased.
    B. The changes do not create the possibility of a new or different 
kind of accident from any accident previously evaluated (10 CFR 
50.92(c)(2)).
1. Limiting Condition for Operation 3.8.1.1.b.2
    The proposed minimum fuel storage requirement has been developed 
using actual EDG performance data and accounting for possible 
variations in fuel oil specific gravity, heating value of the fuel, and 
ambient conditions. The proposed change will provide additional 
assurance that there will be at least a 7 day supply of fuel oil to 
meet the maximum Engineered Safety Feature load requirements following 
a loss of power and a design basis accident. The amount of fuel oil 
stored has no effect upon the initiation of any accident sequence, 
therefore, the proposed change does not create the possibility of a new 
or different kind of accident from any previously analyzed.
2. Surveillance Requirement 4.8.1.1.2.f
    The proposed change to allow the option (as supported by a 10 CFR 
50.59 safety evaluation) of performing selected surveillance tests, or 
portions thereof, during conditions or modes other than during shutdown 
does not affect the operation or response of any plant equipment, 
including the EDGs, or introduce any new failure mechanism. Therefore, 
the proposed change does not create the possibility of a new or 
different kind of accident from any previously analyzed.
3. Surveillance Requirement 4.8.1.1.2.f.14
    Amendment 18 to the Seabrook Station Operating License approved the 
change in the cooling tower operating mode from automatic actuation to 
manual actuation. The proposed change to Surveillance Requirement 
4.8.1.1.2.f.14 does not create the possibility of a new or different 
kind of accident from any accident previously evaluated (10 CFR 
50.92(c)(2)) because it does not affect the manner by which the 
facility has been operated since Amendment 18 was issued, involve any 
changes to equipment or features which affect the operational 
characteristics of the facility, or introduce a new failure mode. The 
proposed change merely corrects an oversight in that the requirement 
should have been deleted when Amendment 18 was issued.
4. Surveillance Requirement 4.8.1.1.2.h.2
    The change does not create the possibility of a new or different 
kind of accident from any accident previously evaluated (10 CFR 
50.92(c)(2)) because it does not affect the manner by which the 
facility is operated as assumed in the design analysis or Safety 
Evaluation, involve any changes to equipment or features which affect 
the operational characteristics of the facility, or introduce a new 
failure mode. The proposed change merely provides a practical alternate 
test method using methods acceptable per Section XI of the ASME Code, 
applicable ASME Code Cases as authorized by the NRC, and Regulatory 
Guide (RG) 1.137, ``Fuel-Oil Systems at Nuclear Power Plants,'' 
Revision 1, October 1979. Therefore, the proposed change does not 
create the possibility of a new or different kind of accident from any 
previously analyzed.
    C. The changes do not involve a significant reduction in a margin 
of safety (10 CFR 50.92(c)(3)).
1. Limiting Condition for Operation 3.8.1.1.b.2
    The proposed change does not reduce the ability of the EDGs to 
provide sufficient power for at least 7 days to meet the maximum 
Engineered Safety Feature load requirements following a loss of power 
and a design basis accident as described in UFSAR Section 9.5.4.1.
2. Surveillance Requirement 4.8.1.1.2.f
    The proposed change does not reduce the ability of the EDGs to 
provide sufficient power to meet the maximum Engineered Safety Feature 
load requirements following a loss of power and a design basis accident 
as described in the UFSAR. Performing certain surveillances during 
conditions or modes other than shutdown (as supported by a 10 CFR 50.59 
safety evaluation) does not involve a significant reduction in a margin 
of safety (10 CFR 50.92(c)(3)) because it does not affect the manner by 
which the facility is operated as assumed in the design analysis or 
Safety Evaluation, involve any changes to equipment or features which 
affect the operational characteristics of the facility. The proposed 
change will continue to ensure the reliability of the EDGs to perform 
their intended function.
3. Surveillance Requirement  4.8.1.1.2.f.14
    The change does not create the possibility of a new or different 
kind of accident from any accident previously evaluated (10 CFR 
50.92(c)(2)) because it does not affect the manner by which the 
facility has operated since Amendment 18 was issued, involve any 
changes to equipment or features which affect the operational 
characteristics of the facility, or introduce a new failure mode. The 
proposed change merely corrects an oversight in that the requirement 
should have been deleted when Amendment 18 was issued.
4. Surveillance Requirement  4.8.1.1.2.h.2
    The change does not involve a significant reduction in a margin of 
safety (10 CFR 50.92(c)(3)) because it does not affect the manner by 
which the facility is operated or involve any changes to equipment or 
features which affect the operational characteristics of the facility. 
The proposed change will continue to ensure the reliability of the EDG 
fuel oil system. The proposed change merely provides a practical 
alternate test method using methods acceptable per Section XI of the 
ASME Code, applicable ASME Code Cases as authorized by the NRC, and 
Regulatory Guide (RG) 1.137, ``Fuel-Oil Systems at Nuclear Power 
Plants,'' Revision 1, October 1979.
    Based on this review, it appears that the three standards of 10 CFR 
50.92(c) are satisfied. Therefore, the NRC staff proposes to determine 
that the amendment request involves no significant hazards 
consideration.
    Local Public Document Room location: Exeter Public Library, 
Founders Park, Exeter, NH 03833.
    Attorney for licensee: Lillian M. Cuoco, Esquire, Northeast 
Utilities

[[Page 66713]]

Service Company, Post Office Box 270, Hartford CT 06141-0270.
    NRC Project Director: S. Singh Bajwa, Acting.

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

    Date of amendment request: October 17, 1996.
    Description of amendment request: The proposed amendment would 
delete certain instrumentation requirements stated in Technical 
Specification (TS) 3/4.3, Instrumentation. The deleted requirements 
would be relocated to the Seabrook Station Technical Requirements 
Manual (SSTR). The associated Bases for the deleted TS requirements 
will be deleted also, but they will not be incorporated into the SSTR. 
The following Limiting Conditions for Operation (LCO) and associated 
Surveillance Requirements (SRs) would be relocated to the SSTR:

------------------------------------------------------------------------
         Technical  specification                       Title           
------------------------------------------------------------------------
LCO--3.3.3.2..............................  Incore Detector System.     
LCO--3.3.3.3 and associated SRs & Tables..  Seismic Instrumentation.    
LCO--3.3.3.4 and associated SRs & Tables..  Meteorological              
                                             Instrumentation            
LCO--3.3.4 and associated SRs.............  Turbine Overspeed           
                                             Protection.                
------------------------------------------------------------------------

    The proposed amendment would also delete (without relocating to the 
SSTR) the reference to the location of the meteorological tower from TS 
5.5.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration. The NRC staff has reviewed the licensee's analysis 
against the standards of 10 CFR 50.92(c). The NRC staff's review is 
presented below.
    A. The changes do not involve a significant increase in the 
probability or consequences of an accident previously evaluated (10 CFR 
50.92(c)(1)) because the proposed changes do not involve any physical 
changes to the plant, do not alter the way any structure, system or 
component functions, do not modify the manner in which the plant is 
operated, do not impact the physical protective boundaries of the 
plant, and do not decrease the effectiveness of administrative controls 
for assuring safe operation of the facility. The instrumentation-
related systems are not considered a design feature or an operating 
restriction that is an initial condition of a design basis accident or 
transient analysis, nor do they function in any way to mitigate the 
consequences of a design basis accident or transient.
    B. The changes do not create the possibility of a new or different 
kind of accident from any accident previously evaluated (10 CFR 
50.92(c)(2)) because the proposed changes do not involve any physical 
changes to the plant, do not alter the way any structure, system or 
component functions, do not modify the manner in which the plant is 
operated, do not impact the physical protective boundaries of the 
plant, and do not decrease the effectiveness of administrative controls 
for assuring safe operation of the facility.
    C. The changes do not involve a significant reduction in a margin 
of safety (10 CFR 50.92(c)(3)) because the proposed changes do not 
involve any physical changes to the plant, do not alter the way any 
structure, system or component functions, do not modify the manner in 
which the plant is operated, do not impact the physical protective 
boundaries of the plant, and do not decrease the effectiveness of 
administrative controls for assuring safe operation of the facility. 
Further, the proposed changes do not affect the ability of systems, 
structures or components important to safety to perform their intended 
function.
    Based on this review, it appears that the three standards of 10 CFR 
50.92(c) are satisfied. Therefore, the NRC staff proposes to determine 
that the amendment request involves no significant hazards 
consideration.
    Local Public Document Room location: Exeter Public Library, 
Founders Park, Exeter, NH 03833.
    Attorney for licensee: Lillian M. Cuoco, Esquire, Northeast 
Utilities Service Company, Post Office Box 270, Hartford CT 06141-0270.
    NRC Project Director: S. Singh Bajwa, Acting.

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

    Date of amendment request: March 20, 1996 and as supplemented on 
July 25, 1996.
    Description of amendment request: The amendments would modify the 
Susquehanna Steam Electric Station (SSES), Units 1 and 2, Technical 
Specifications to change the ``open'' logic for the high pressure core 
injection (HPCI) suction valves HV-155/255-F042 in order to eliminate 
the HPCI pump auto-transfer on high suppression pool level.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. The proposed change does not involve a significant increase 
in the probability or consequences of an accident previously 
evaluated.
    Based on the following discussion for the containment, reactor 
building, HPCI and RCIC [reactor core isolation cooling] systems, 
and the safety-related valves in piping connected to the suppression 
pool, the proposed action does not increase the probability or 
consequences of an accident previously evaluated. Primary 
Containment and Reactor Building Safety-Related Systems, Structures, 
and Components Affected by LOCA/SRV [Loss-of-coolant-accident/safety 
relief valve] Hydrodynamic Loads
    As discussed in the Safety Assessment for this change, 
elimination of the HPCI auto suction transfer on high suppression 
pool level will allow higher suppression pool water levels in 
accidents and transients which involve HPCI operation. The impact of 
the higher suppression pool levels were examined for the following 
design-basis accidents and transients:
    Loss of Coolant Accidents inside containment (FSAR [Final Safety 
Analysis Report] *6.2.1.1.3.3),
    Inadvertent Safety/Relief valve opening (FSAR *15.1.4),
    Primary system break outside containment (FSAR *3.6A),
    Inadvertent HPCI initiation (FSAR *15.5.1),
    Loss of feedwater flow (FSAR *15.2.7),
    Loss of Offsite AC Power (FSAR *15.2.6),
    Loss of Main Condenser vacuum (FSAR *15.2.5),
    Inadvertent MSIV closure (FSAR *15.2.4),
    Turbine trip (with and without bypass) (FSAR *15.2.3),
    Generator Load Rejection (with and without bypass), (FSAR 
*15.2.2), and
    Pressure regulator failure-closed/open (FSAR *15.2.1 & 15.1.3).
    These accidents and transients were selected for evaluation 
because they involve an initiation of the HPCI system either 
inadvertently or as a result of a decrease in vessel inventory and/
or coolant level. Two special events, ATWS and SBO, are also 
considered along with the design basis events listed above.
    It was concluded that design-basis SRV and LOCA loads envelop 
the loads expected with the proposed change. Therefore, the proposed 
change does not increase the failure probability of any primary 
containment or reactor building structure, system or component which 
is affected by LOCA/SRV hydrodynamic loads. The major findings which 
lead to this conclusion about SRV and LOCA loads are summarized 
below:
    DBA [design basis accident] dynamic pressure loads are based on 
a maximum initial suppression pool level of 24 feet. The proposed 
modification to the HPCI suction

[[Page 66714]]

transfer logic does not affect the initial pool level or the initial 
suppression chamber air space volume. During normal plant operation, 
suppression pool level (and hence suppression chamber air space 
volume) is controlled by Technical Specification requirements.
    For LOCAs other than the DBA, the containment is designed for 
ADS [automatic depressurization system] blowdown loads in 
combination with the LOCA loads. For an intermediate break, the 
proposed HPCI modification does allow suppression pool level to 
exceed 24 feet by a small amount. ADS loads are, however, 
independent of suppression pool level when the downcomer vents are 
cleared. Therefore, the proposed modification has no influence on 
ADS hydrodynamic loads for an intermediate break.
    For small breaks, HPCI injection prevents ADS actuation. 
Nevertheless, SRV actuations occur during the RPV [reactor pressure 
vessel] cooldown. Downcomer vents are opened in the beginning part 
of the accident, but close later on as the break enthalpy decreases. 
When the downcomer vents are cleared, the level inside the SRV 
tailpipe is not influenced by pool level, and therefore, the SRV 
hydrodynamic loads are unaffected by the proposed modification. 
During the phase of the accident in which the downcomer vents are 
sealed with water, there are no wetwell LOCA hydrodynamic loads, but 
the SRV loads are dependent on SP [suppression pool] water level. In 
this case, SRV loads are acceptable because SP water level is always 
below the Load Limit curve.
    ADS actuation would be required in the event of a HPCI failure 
during a small-break accident. If HPCI fails during the phase of the 
accident in which the downcomer vents are cleared, then ADS loads 
would be acceptable because water level (and air volume) within the 
SRV tailpipes is independent of pool level. Even if HPCI failure 
occurs in the latter part of the accident where the downcomer vents 
are sealed, ADS loads are acceptable because water level is always 
well below the Load Limit curve.
    Under non-LOCA conditions, the containment is designed for 
simultaneous actuation of all 16 SRVs. The Load Limit Line defines 
the acceptable operating region, in terms of reactor pressure and 
suppression pool level, for actuation of all 16 SRVs. Following a 
plant transient involving HPCI operation, the suppression pool level 
is always below the Load Limit curve, and only a small number of 
SRVs actuate to remove decay heat from the reactor.

HPCI System

    The proposed change does not increase the probability of an 
equipment malfunction in the HPCI system. In fact, the change 
eliminates the potential failure of the HPCI suction auto-transfer 
on high suppression pool level since that logic is removed. 
Potential spurious auto-transfer associated with high suppression 
pool logic is also eliminated. HPCI suction auto-transfer on low CST 
[condensate storage tank] level and its potential to fail are 
unchanged by this change. Also, the change does not affect the 
manual suction transfer from the CST to the suppression pool.
    As discussed in the safety assessment for this change, the 
proposed change has no adverse effects on HPCI valves, pump, or 
turbine. Therefore, elimination of the HPCI suction auto transfer 
logic (on high suppression pool level) does not increase the 
probability of a HPCI malfunction. The consequence of a HPCI failure 
in a design-basis accident is evaluated in NEDC-32071P Rev.1, 
``Susquehanna Steam Electric Station Units 1 and 2 SAFER/GESTR-LOCA 
Loss-of-Coolant Accident Analysis.'' With regard to the fuel, the 
consequence of a HPCI failure is unaffected by the proposed change.
    If HPCI fails in a design-basis small break accident, ADS 
actuation would be required. ADS loads continue to be enveloped by 
design loads with the proposed change. Therefore, the proposed 
change does not increase the consequences of a HPCI failure.
    HPCI Relay Panel 1C620(2C620) & 250 V DC Control Center 
1D264(2D264)
    On a component level, the failure probability and consequences 
of failure associated with the AX [auxiliary] relay in 250 VDC 
Control Center 1D264 (2D264) are eliminated because the relay is 
disconnected and removed by this modification. Since the control 
functions of K19 in panel 1C620 (2C620) have been eliminated, the 
failure of the relay has no effect on HPCI suction valve F042 
operation.
    The 250 VDC Control Center 1D264 (2D264) and HPCI Relay Panel 
1C620 (2C620) both receive power from battery systems during Station 
Blackout. Removal of the relay from 250 VDC Control Center 1D264 
(2D264) and the replacement of the relay in HPCI Relay Panel 1C620 
(2C620) decreases the load on the battery systems by a small amount. 
The change in battery load and line voltage drop is negligible and 
is documented in applicable calculations. Dynamic qualification of 
the subject equipment is not adversely affected by this modification 
as documented in applicable calculations.

RCIC Turbine

    As discussed in the safety assessment for this change, RCIC is 
used to provide coolant makeup following a reactor vessel isolation 
and for an Appendix R shutdown scenario. The Appendix R event also 
assumes the reactor vessel is isolated. These events are discussed 
in Section 15.2.4 of the FSAR and in the FPRR [fire protection 
review report]. The proposed change has no adverse effects on RCIC 
turbine operation following a MSIV [main steam isolation valve] 
closure (see discussion in the safety assessment for this change 
[letter dated March 20, 1996, as supplemented July 25, 1996]). 
Therefore, there is no increase in the RCIC failure probability for 
the MSIV-closure event or the Appendix R shutdown scenario. The 
consequence of RCIC failure is unchanged by the proposed 
modification; if RCIC fails, HPCI is available as a backup 
system.\1\ [All footnotes are listed at the end of the no signficant 
hazards basis section.]
    Although RCIC is not designed for mitigation of a small break 
accident, the effect of the proposed change on RCIC turbine 
operation for such an accident was evaluated in the safety 
assessment for this change. The assessment concludes that the 
proposed change has no adverse effects on RCIC operation, and 
therefore, there is no increase in RCIC failure probability during a 
small break accident. Failure of RCIC in a small break accident 
would require ADS initiation only for a particular break flow which 
is slightly greater than HPCI injection capability. But ADS 
initiation has already been considered when evaluating the 
consequences of HPCI failure during a small break accident.

Safety-Related Valves on Piping Connected to Suppression Chamber

    MOVs [motor operated valves]--The proposed change could 
potentially lead to a maximum suppression pool level of 26 feet in a 
design-basis accident. This is 2 feet above the maximum design level 
of 24 feet. As discussed in the safety assessment for this change, 
this is equivalent to a pressure increase of 0.86 psi at the bottom 
of the suppression pool. This small pressure increase has negligible 
effect on valve operation, and therefore, there is no increase in 
the probability of a failure or malfunction of valves in piping 
connected to the suppression pool.
    Vacuum Breakers--Allowing suppression pool level to potentially 
increase to 26 feet in a design-basis accident does not affect the 
failure probability of downcomer-vent vacuum breakers because the 
level is well below the vacuum breaker elevation of 42 feet.
    SRVs/Tailpipes--As discussed in the safety assessment for this 
change, the increased suppression pool level associated with the 
proposed change does not have any adverse effect on SRV operation or 
on the structural integrity of the SRV tailpipe.
    2. The proposed change does not create the possibility of a new 
or different kind of accident from any accident previously 
evaluated.
    Based on the following discussion for the containment, reactor 
building, HPCI and RCIC systems, and the safety-related valves in 
piping connected to the suppression pool, the proposed action does 
not create the possibility of a new or different kind of accident 
from any accident previously evaluated.
    The following discussion concerning the impact of the change on 
the primary containment, the reactor building, the HPCI system, and 
safety-related valves, provides the basis for this conclusion.

Primary Containment and Reactor Building Safety-Related Systems, 
Structures, and Components Affected by LOCA/SRV Hydrodynamic Loads

    The HPCI suction transfer logic is not necessary to maintain 
LOCA loads within design limits because these dynamic pressure loads 
are characterized in terms of the SP level at the initiation of the 
accident. That is, LOCA blowdown tests were conducted without the 
removal of water from the suppression chamber section of the test 
tank.2 The increase in pool level realized during these tests 
was proto-typical of the pool level increase expected at 
Susquehanna. Removal of the HPCI suction transfer logic on high pool 
level does not affect suppression pool level at the initiation of a 
DBA.3

[[Page 66715]]

    In addition, the HPCI suction transfer logic is not necessary to 
maintain SRV/ADS blowdown loads within design limits. SRV dynamic 
pressure loads consist of two components: air clearing loads and 
steam condensation loads. The steam condensation loads are bounded 
by the more severe air clearing loads which are caused by gas bubble 
oscillations following the expulsion of noncondensible gas from the 
SRV tailpipe. Air clearing loads are a function of reactor pressure 
and water level inside the SRV tailpipe.
    Depending on the break size and location, the downcomer vents 
may be cleared for the entire time that HPCI is operating, or they 
may reseal in the latter part of the accident. When the downcomer 
vents are cleared, the level inside the SRV tailpipe is depressed to 
the elevation coinciding with the bottom of the downcomer pipes, and 
it is therefore decoupled from the rising suppression pool level. In 
this situation SRV air-clearing loads are unaffected by the proposed 
change.
    When the downcomer vents are sealed with water, the Load Limit 
line can be used to determine if SRV/ADS loads are enveloped by 
design loads. For the most limiting event, which is the small break 
LOCA, the overall safety margin increases as pool level rises during 
the event. This is because the decrease in reactor pressure more 
than offsets the adverse effects associated with the rise in pool 
level.
    Since LOCA and SRV dynamic loads remain bounded by design loads, 
dynamic loading of primary containment and reactor building 
structures, systems, and components are unaffected by the proposed 
change. Therefore, with respect to dynamic loads, the proposed 
change does not create the possibility for an accident or 
malfunction of a different type than any evaluated in the SAR 
[Safety Analysis Report].

HPCI System

    There are no new HPCI turbine failure modes introduced by the 
higher suppression pool levels which can occur with the proposed 
change. Turbine exhaust pressure remains well below the design limit 
of 65 psia. In addition, the higher pool level does not create the 
possibility of water hammer damage to the turbine discharge piping. 
If the operator fails to control RPV level less than +54'' (single 
operator error) in the long-term part of the small-break accident 
when suppression pool level is greater than 25.6 feet, leakage 
through check valve F049 is such that it will be contained well 
within the volume of the turbine-discharge-line drain pot. Note that 
suppression pool level is limited to 26 feet by operator action. 
Furthermore, suppression pool level can reach 26 feet only for a 
particular range of small breaks, and for this range of small 
breaks, suppression pool level would exceed 25.6 feet for only 
approximately 10 minutes of the accident duration. This corresponds 
to about 10% of the time that HPCI is operating. Thus it is very 
unlikely that HPCI would trip with pool level greater than 25.6 
feet.
    If check valve F049 is failed during the small-break accident 
(single equipment failure), the turbine exhaust line would become 
flooded if the HPCI system tripped during the 10 minute interval 
when suppression pool level greater than 26 feet; however, it is not 
necessary to postulate an operator error (failure to control RPV 
level less than +54'') along with the check valve failure. A small 
break accident with failure of check valve F049 and failure of the 
operator to control RPV level as required by the EOPs [emergency 
operating procedures], in a narrow time interval during the long-
term part of the accident, is beyond the plant design basis.
    A new type of malfunction does not occur even in the beyond-
design-basis condition where failure of check valve F049 is 
considered along with failure of the operator to control RPV level 
less than 54'' in the narrow time interval when pool level is 
greater than 25.6. With these failures, the turbine exhaust piping 
will become flooded, and the system may fail on restart. The General 
Electric Company has performed an analysis to determine the 
consequences of a HPCI start with flooding of the turbine and 
adjacent exhaust line.\4\ The analysis, which addresses a potential 
design deficiency in the HPCI barometric condenser, shows that the 
containment penetration head fitting and interface piping will not 
fail as a result of the water hammer associated with the HPCI start. 
Since failure of the HPCI system is already considered in the plant 
design-basis accident analysis; this is not a different type of 
malfunction than that already considered.

HPCI Relay Panel 1C620(2C620) & 250 V DC Control Center 1D264(2D264)

    No new failure modes are introduced by the hardware changes in 
the 250 VDC Control Center 1D264 (2D264) and HPCI Relay Panel 1C620 
(2C620). Some failure modes are eliminated by the proposed change. 
Specifically, the potential failure of the HPCI suction auto-
transfer on high suppression pool level is eliminated since that 
logic is removed. Potential spurious auto-transfer associated with 
high suppression pool logic is also eliminated. HPCI suction auto-
transfer on low CST level and its potential to fail are unchanged by 
this change.
    On a component level, potential failure modes for the AX relay 
in 250 VDC Control Center 1D264 (2D264) are eliminated by this 
modification because the relay is disconnected and removed by this 
change. The potential failure modes for the relay K19 in panel 1C620 
(2C620) are unchanged. Since the control functions of K19 have been 
eliminated, the failure of the relay has no effect on HPCI suction 
valve F042 operation.
    Removal of the relay from 250 VDC Control Center 1D264 (2D264) 
and the replacement of the relay in the HPCI Relay Panel 1C620 
(2C620) changes the load on the battery systems by a small amount. 
The change in battery load and change in line voltage drop are 
negligible and they do not adversely affect the performance of the 
panels or battery systems. In addition, seismic qualification of the 
panels is not adversely affected by this change.

RCIC Turbine

    As discussed in the safety assessment for this change, the 
proposed change has no adverse effects on RCIC turbine operation. 
Therefore, the proposed change cannot result in a new RCIC failure 
mode.

Safety-Related Valves on Piping Connected to Suppression Chamber

    MOVs--The increased suppression pool water level which can occur 
as a result of the proposed change does not create a failure 
mechanism for safety-related valves on piping connected to the 
suppression pool. The pressure differential for any valve on piping 
connected to the suppression pool will increase by at most 0.86 psi. 
This change in differential pressure has negligible effect on valve 
operation.
    Vacuum Breakers--The proposed change cannot lead to malfunction 
of the downcomer-vent vacuum breakers as the maximum level expected 
in a design-basis event is 26 feet, and the vacuum breakers are 
located at 42 feet above the suppression pool floor.
    SRVs/Tailpipes--There is no interaction between increased 
suppression pool level and SRV operation since the flow through the 
SRVs is choked and therefore decoupled from downstream conditions. 
Also, the increased suppression pool level cannot lead to failure of 
the SRV tailpipe because the potential level increase is well below 
the SRV Tailpipe Level Limit.\5\ If suppression pool water level is 
below this limit, there is no concern of tailpipe failure due to 
overpressurization. The minimum value of the SRV Tailpipe Level 
Limit is 35 feet.\6\ This is 9 feet above the maximum level expected 
in a design-basis accident. For beyond-design-basis events, SRV 
tailpipe integrity is protected by the EOP requirement to 
depressurize the reactor on the SRV Tailpipe Level Limit.
    3. The proposed change does not involve a significant reduction 
in a margin of safety.
    Based on the following discussion for the containment, reactor 
building, HPCI and RCIC system, and the safety-related valves in 
piping connected to the suppression pool, the proposed action does 
not involve a significant reduction in a margin of safety.

HPCI System

    The HPCI Technical Specifications ensure that the system is 
capable of providing adequate core cooling to limit clad 
temperatures in the event of a small break LOCA which does not 
result in rapid depressurization of the RPV (Technical Specification 
Section 3/4.5.1 & 3/4.5.2). The proposed change has no adverse 
affects on the injection capability of the HPCI system. Therefore, 
the safety function of the system is not degraded, and there is no 
reduction in the margin of safety as defined in the basis for the 
HPCI Technical Specifications.
    Primary Containment and Reactor Building Safety-Related Systems, 
Structures, and Components Affected by LOCA/SRV Hydrodynamic Loads
    Removal of the HPCI auto suction transfer on high suppression 
pool level does not affect the Technical Specification requirement 
to maintain suppression pool water level between 22 and 24 feet 
(Technical Specification 3.6.2.1). Therefore, the maximum 
containment pressure during the design-basis accident is unaffected 
by the proposed change, and there can be no reduction in the margin 
of safety as defined in the basis for Technical Specification

[[Page 66716]]

3.6.2.1. Furthermore, a detailed examination of the reactor and 
containment response under accident and transient conditions 
involving HPCI operation found no situations where the auto suction 
transfer was necessary to maintain LOCA and SRV loads within the 
design basis envelope. Therefore, from the standpoint of LOCA/SRV 
hydrodynamic loads, the proposed change does not reduce the margin 
of safety for any primary containment or reactor building structure, 
system, or component.

RCIC Turbine

    The basis for Technical Specification 3.7.3 states that the RCIC 
system is provided to assure adequate core cooling in the event of a 
reactor isolation with loss of feedwater flow. The proposed change 
does not prohibit RCIC from performing this function, nor does it 
degrade in any way the core cooling capability of RCIC. Therefore, 
there is no reduction in the margin of safety as defined in the 
basis for Technical Specification 3.7.3.

Safety-Related Valves on Piping Connected to Suppression Pool

    MOVs--The increase in suppression pool water level which can 
occur as a result of the proposed change does not reduce the margin 
of safety for safety-related valves on piping connected to the 
suppression pool. The pressure differential for any valve on piping 
connected to the suppression pool will increase by at most 0.86 psi. 
This change in differential pressure has negligible effect on valve 
operation.
    Vacuum Breakers--The proposed change cannot reduce the margin of 
safety as discussed in the basis for Technical Specification 3.6.4 
because the maximum level expected in a design-basis event is 26 
feet which is well below the downcomer-vent vacuum breaker elevation 
of 42 feet.
    SRVs/Tailpipes--There is no interaction between increased 
suppression pool level and SRV operation since the flow through the 
SRVs is choked and therefore decoupled from downstream conditions. 
Consequently, there is no reduction in the margin of safety as 
defined in the bases for Technical Specifications 3.4.2 (safety 
valve function) and 3.5.1.d (ADS function). Also, the increased 
suppression pool level does not lead to a reduction in the margin of 
safety for the SRV tailpipes because the tailpipes can operate 
safely with pool levels up to 35 feet. This is nine feet above the 
maximum suppression pool level that can occur in a design-basis 
accident with the proposed change. For beyond-design-basis events, 
SRV tailpipe integrity is protected by the EOP requirement to 
depressurize the reactor on the SRV Tailpipe Level Limit.\7\

HPCI Relay Panel 1C620(2C620) & 250 V DC Control Center 1D264(2D264)

    As discussed previously, removal of the relay from 250 VDC 
Control Center 1D264 (2D264) and the replacement of the relay in the 
HPCI Relay Panel 1C620 (2C620) changes the load on the battery 
systems by a small amount. The change in battery load and change in 
line voltage drop are negligible and therefore they do not reduce 
the margin of safety for the panels or battery systems. In addition, 
seismic qualification of the panels is not adversely affected by 
this change so there is no reduction in the margin of safety for 
seismic events.
    1. DBD041, Rev. 0, p. 1. [design basis document for RCIC system]
    2. SSES DAR [design assessment report for suppression pool 
hydrodynamic loads], Section 9.4.1
    3. Suppression pool level must be maintained less than 24 feet 
in accordance with Technical Specification 3.6.2.1.a.
    4. GKR-03-001, ``NRC and Utility Notification of Closeout of GE 
PRC92-05, Potential Design Deficiency on HPCI,'' January 6, 1993 [GE 
letter to PP&L regarding closure of HPCI design issue].
    5. This limit is defined in EO-100/200-103 [emergency operating 
procedure]
    6. Bechtel Calculations PUP-15598-S2 & PUP-15598-S6, and PLE-
15315 (March 2, 1992)
    7. The limit is defined in EO-100/200-103 [emergency operating 
procedure]

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

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

    Date of amendment request: October 7, 1996.
    Description of amendment request: The amendments would modify the 
Susquehanna Steam Electric Station, Units 1 and 2, Technical 
Specifications by revising the trip setpoints and allowable values for 
the secondary containment isolation ``Refuel Floor High Exhaust Duct 
Radiation--High'' monitor, the ``Railroad Access Shaft Exhaust Duct 
Radiation--High'' monitor, and the ``Refuel Floor Wall Exhaust Duct 
Radiation--High'' monitor in Table 3.3.2-2. The change would enhance 
the operational efficiency of plant operations by eliminating 
compensatory measures which prevent spurious secondary containment 
isolations, and initiation of the standby gas treatment system (SGTS) 
and recirculation system during refueling activities. This change would 
also allow for the use of the hydrogen water chemistry system during 
operation.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. This proposal does not involve a significant increase in the 
probability or consequences of an accident previously evaluated.
    The proposed change to the trip setpoints and allowable values 
to the ``Refuel Floor High Exhaust Duct Radiation--High'' monitor, 
the ``Railroad Access Shaft Exhaust Duct Radiation--High'' monitor, 
and the ``Refuel Floor Wall Exhaust Duct Radiation--High'' monitor 
does not involve a significant increase in the probability or 
consequences of an accident previously evaluated. The design basis 
for the monitors is to monitor radiation in the unfiltered air from 
the Zone III exhaust system to provide signals which isolate the 
Zone III of the secondary containment on a high radiation condition, 
and to initiate SGTS and the Recirculation system to limit offsite 
doses to maintain regulatory requirements.
    The original setpoints for these monitors were based upon normal 
radiological operating conditions and were set at a value to 
preclude spurious design actuations by these monitors during normal 
plant operations. However, the monitors are designed to detect 
radiation associated with certain postulated accident conditions. As 
required by the Technical specifications the monitors are operable 
when conditions exist that may result in fuel damage events, and 
therefore, will perform their design basis function. Consequently, 
an increase to the trip setpoints and allowable values is warranted 
since the existing setpoints, which are conservatively based on 
normal radiological operating conditions, are not related to the 
design basis of the monitors. Therefore, based upon the design basis 
of the monitors, an increase to the trip setpoints and allowable 
values will not result in a decrease of the safety function of the 
monitors but will make the trip setpoints and allowable values 
consistent with the design basis.
    Based on the design basis of these monitors, revised analytical 
limits were derived reflecting the accident function of the 
monitors. The analytical limit calculations utilized FSAR realistic 
source terms, instead of the worst case source terms utilized for 
10CFR [Part] 100 compliance. Use of the realistic source terms 
results in conservative analytical limits.
    The ``Refuel Floor High Exhaust Duct Radiation--High'' monitor, 
and the ``Refuel Floor Wall Exhaust Duct Radiation--High'' [monitor] 
are required to be OPERABLE during CORE ALTERATIONS (except for 
single control rod movements unless performing TS 3.10.3), 
operations with the potential for draining the reactor vessel, and 
handling of irradiated fuel in the secondary containment. The 
``Railroad Access Shaft Exhaust Duct Radiation--High'' monitor is 
required to be operable during handling of irradiated fuel. These 
Technical Specification

[[Page 66717]]

applicable operational conditions for the monitors are not affected 
since this proposed revision only revises the trip setpoints and 
allowable values to be consistent with the design bases of the 
monitors.
    For the reasons stated above the revisions to the trip setpoints 
and allowable values to the ``Refuel Floor High Exhaust Duct 
Radiation--High'' monitor, the ``Railroad Access Shaft Exhaust Duct 
Radiation--High'' monitor, and the ``Refuel Floor Wall Exhaust Duct 
Radiation--High'' monitor in Technical Specification.
    Table 3.3.2-2 can be implemented without a significant increase 
in the probability or consequence of an accident previously 
evaluated.
    II. This proposal does not create the possibility of a new or 
different kind of accident previously evaluated.
    The proposed change to the trip setpoints and allowable values 
for the ``Refuel Floor High Exhaust Duct Radiation--High'' monitor, 
the ``Railroad Access Shaft Exhaust Duct Radiation--High'' monitor, 
and the ``Refuel Floor Wall Exhaust Duct Radiation--High'' monitor 
does not create the possibility of a new or different kind of 
accident from any accident previously evaluated.
    The monitors are designed to limit the release of airborne 
radioactivity in the secondary containment Zone III exhaust system 
by isolating Zone III, initiating [the] SGTS and initiating the 
Recirculation System on high radiation resulting from fuel handling 
accidents. Therefore, the design basis for these monitors is to 
monitor radiation in the unfiltered air from the Zone III exhaust 
system, and provide signals to limit offsite doses to maintain 
regulatory requirements. Zone III includes the Refueling Floor and 
can include the Railroad Access Shaft during certain alignments. 
These radiation monitors are not provided for occupational 
protection associated with operational radiation doses. The proposed 
revision does not affect the design basis of the monitors nor the 
kind of accident associated with the basis; therefore, no potential 
to create a new or different accident exists.
    For the reasons stated above the revisions to the trip setpoints 
and allowable values to the ``Refuel Floor High Exhaust Duct 
Radiation--High'' monitor, the ``Railroad Access Shaft Exhaust Duct 
Radiation--High'' monitor, and the ``Refuel Floor Wall Exhaust Duct 
Radiation--High'' monitor in Technical Specification Table 3.3.2-2 
can be implemented without creating the possibility of a new or 
different kind of accident from any accident previously evaluated.
    III. This proposal does not involve a significant reduction on a 
margin of safety.
    The proposed change to the trip setpoints and allowable values 
for the ``Refuel Floor High Exhaust Duct Radiation--High'' monitor, 
the ``Railroad Access Shaft Exhaust Duct Radiation--High'' monitor, 
and the ``Refuel Floor Wall Exhaust Duct Radiation--High'' monitor 
does not involve a significant reduction in a margin of safety.
    The monitors are designed to limit the release of airborne 
radioactivity in the secondary containment Zone III exhaust system 
by isolating Zone III, initiating [the] SGTS and initiating the 
Recirculation System on high radiation resulting from fuel handling 
accidents. Therefore, the design basis for these monitors is to 
monitor radiation in the unfiltered air from the Zone III exhaust 
system, and provide signals to limit offsite doses to maintain 
regulatory requirements. Zone III includes the Refueling Floor and 
can include the Railroad Access Shaft during certain alignments. 
These radiation monitors are not provided for occupational 
protection associated with operational radiation doses. However, the 
original setpoints for these monitors were conservatively based upon 
normal radiological operating conditions and were set at a value to 
preclude spurious design actuation by these monitors during normal 
plant operations. The calculations performed to support the trip 
setpoint and allowable value revisions concluded that the change 
will maintain offsite doses within the 10CFR100 limits. The ``Refuel 
Floor High Exhaust Duct Radiation--High'' monitor, and the ``Refuel 
Floor Wall Exhaust Duct Radiation--High'' are required to be 
OPERABLE during CORE ALTERATIONS (except for single control rod 
movements unless performing TS 3.10.3), operations with the 
potential for draining the reactor vessel, and handling of 
irradiated fuel in the secondary containment. The ``Railroad Access 
Shaft Exhaust Duct Radiation--High'' monitor is required to be 
operable during handling of irradiated fuel. These Technical 
Specification applicable operational conditions for the monitors are 
not affected since the proposed revision only revises the trip 
setpoints and allowable values to be consistent with the design 
bases of the monitors.
    The proposed revisions to the trip setpoints and allowable 
values, in addition to being based on the appropriate accident 
conditions, were also developed utilizing standard setpoint change 
methodologies that consider instrument and calibration accuracies 
and instrument drift tolerances. This provides added conservatism to 
assure that the revised trip setpoints and allowable values are not 
exceeded.
    For the reasons stated above the revisions to the trip setpoints 
and allowable values to the ``Refuel Floor High Exhaust Duct 
Radiation--High'' monitor, the ``Railroad Access Shaft Exhaust Duct 
Radiation--High'' monitor, and the ``Refuel Floor Wall Exhaust Duct 
Radiation--High'' monitor in Technical Specification Table 3.3.2-2 
can be implemented without involving a significant reduction in a 
margin of safety.

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

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

    Date of amendment request: November 25, 1996.
    Description of amendment request: The proposed Technical 
Specifications (TS) changes would revise the wording in TS Section 
4.8.1.1.2.e.2 and the associated TS Bases Section 3/4.8 to remove the 
specific reference to the Residual Heat Removal pump motor and its 
corresponding kW rating value, and replace it with wording consistent 
with that specified in the Improved TS (i.e., NUREG-1433, Revision 1, 
``Standard Technical Specifications General Electric Plants,'' dated 
April 1995).
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. The proposed Technical Specifications (TS) changes do not 
involve a significant increase in the probability or consequences of 
an accident previously evaluated.
    The proposed TS changes do not make any physical alterations or 
modifications to the plant systems or equipment. The proposed 
changes do not adversely impact the operation of any plant 
equipment. The EDGs will continue to function as designed to ensure 
that the necessary electrical power is provided to essential plant 
equipment to mitigate the consequences of an accident, e.g., Loss-
of-Offsite-Power (LOOP) and Loss-of-Coolant Accident LOCA) 
coincident with a LOOP (LOCA/LOOP). The proposed TS changes do not 
impact the performance testing requirements associated with the 
EDGs. The accident mitigating capabilities of the diesel generators 
and emergency loads will remain the same.
    The proposed TS changes are consistent with the guidance 
stipulated in NUREG-1433, Revision [1], ``Standard Technical 
Specification General Electric Plants,'' regarding single load 
rejection testing of the EDGs. Specifically, the proposed changes 
involve revising the wording in TS Surveillance Requirement (SR) 
4.8.1.1.2.e.2 to remove the specific reference to the Residual Heat 
Removal (RHR) pump motor and associated kW loading value (992 kW), 
and replace it with wording indicating that the EDGs must be capable 
of rejecting the single largest post-accident load, which is 
consistent with NUREG-1433, Revision 1, guidance. The proposed 
changes will also provide additional flexibility for future plant 
maintenance activities.
    Each EDG will continue to be tested by rejecting a load of 
greater than or equal to that of its single largest post-accident 
load while maintaining voltage and frequency

[[Page 66718]]

within the current specified parameters. The RHR pump motors are 
currently used in performing the EDG single load rejection testing. 
The RHR pump motors will continued [sic] [continue] to be used in 
performing the surveillance testing since they are the single 
largest post-accident electrical load. The consequences of a 
malfunction of equipment are not affected. Failure of a EDG or its 
safety-related loads is bounded by the loss of a Class 1E electrical 
power division which has been previously evaluated as discussed in 
LGS Updated Final Safety Analysis Report (UFSAR) Sections 8.1.5.2.e 
and 8.3.1.1.3.
    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 proposed TS changes do not make any physical alterations or 
modifications to the plant systems or equipment. The proposed 
changes do not adversely impact the operation of any plant 
equipment. The EDGs will continue to function as designed to provide 
essential electrical power to mitigate the consequences of an 
accident. The proposed TS changes are consistent with the guidance 
stipulated in NUREG-1433, Revision 1, regarding single load 
rejection testing of the EDGs. The proposed changes do not introduce 
any new accidents or transients. The proposed TS changes will 
provide additional flexibility for future maintenance activities. 
The proposed changes do not alter any EDG testing requirements or 
frequencies. The RHR pump motors are currently used in performing 
the EDG single load rejection testing. The RHR pump motors will 
continue to be used in performing the surveillance testing since 
they are the single largest post-accident electrical load. The 
operation of the EDGs and their corresponding safety-related 
electrical loads remain unchanged as a result of the proposed TS 
changes.
    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 proposed TS changes do not involve any physical changes to 
plant systems or equipment. The proposed TS changes are consistent 
with the guidance stipulated in NUREG-1433, Revision 1, ``Standard 
Technical Specification General Electric Plants,'' regarding single 
load rejection testing of the EDGs. The proposed TS changes will 
provide additional flexibility for future plant maintenance 
activities. The EDGs will continue to function as designed to 
provide essential electrical power to mitigate the consequences of 
an accident. The operation of the EDGs and their corresponding 
safety-related electrical loads remain unchanged as a result of the 
proposed TS changes.
    Therefore, the proposed TS changes do not involve a significant 
reduction in a margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: Pottstown Public Library, 500 
High Street, Pottstown, PA 19464.
    Attorney for licensee: J. W. Durham, Sr., Esquire, Sr. V.P. and 
General Counsel, Philadelphia Electric Company, 2301 Market Street, 
Philadelphia, PA 19101.
    NRC Project Director: John F. Stolz.

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

    Date of amendments request: November 15, 1996.
    Description of amendments request: The amendments would eliminate 
the containment systems Technical Specification 3.6.2.2. ``Spray 
Additive System.'' The specification would be replaced with a new 
emergency core cooling system Technical Specification 3.5.6 ``ECCS 
Recirculation Fluid pH Control System.''
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. The proposed change does not involve a significant increase 
in the probability or consequences of an accident previously 
evaluated. The proposed change involves replacement of concentrated 
NaOH injected via the containment spray system with trisodium 
phosphate (TSP) stored in the containment and dissolved in the sump 
recirculation solution to maintain acceptable post accident spray/
recirculation solution chemistry. Deletion of the concentrated NaOH 
will eliminate a personnel hazard. The pH control system functions 
in response to an accident and does not involve or have any effect 
on any initiating event for any accident previously evaluated. 
Operation under the proposed amendments will continue to ensure that 
iodine potentially released post-LOCA [loss-of-coolant accident] is 
retained in the sump solution, and resultant offsite and control 
room thyroid doses are within the limits of 10 CFR [Part] 100 and 10 
CFR [Part] 50, Appendix A, General Design Criterion [GDC] 19, 
respectively.
    2. The proposed change does not create the possibility of a new 
or different kind of accident from any accident previously 
evaluated. The deleted equipment is isolated from the remaining 
equipment by cut-and-capped piping, determinated and/or spared 
cables; and interfaces are analyzed to ensure the remaining required 
equipment meets applicable original design requirements. The new 
equipment (TSP and baskets) is a passive pH control system and is 
supported and analyzed to ensure there are no adverse interfaces 
(e.g., pipe break, jet impingement, seismic) with existing 
equipment, system, or structures.
    3. The proposed change does not involve a significant reduction 
in a margin of safety. The slight change in recirculation solution 
pH maintains adequate protection against chloride and caustic 
induced stress corrosion cracking on mechanical systems and 
components, and maintains the capability of the solution to retain 
iodine. It does not result in a change to the hydrogen generation 
analysis for containment. The increased mass inside containment will 
have no significant impact on post-accident flood levels, 
recirculation solution boron concentration, or peak clad 
temperatures. No other operating parameters for systems, structures, 
or components assumed to operate in the safety analysis are changed. 
The offsite and control room doses meet the limits of 10 CFR [Part] 
100 and GDC 19, respectively. Because the trisodium phosphate is 
nonvolatile and the baskets are protected with solid covers and are 
located slightly above the floor in the containment where access is 
strictly controlled, a surveillance interval of once per refueling 
outage provides assurance that the TSP will be available.

    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.

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

[[Page 66719]]

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.

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

    Date of amendment request: June 3, 1996, as supplemented October 
23, 1996.
    Description of amendment request: The proposed amendment would 
clarify a restriction on shutdown margin monitor operability while 
changing modes so that it only limits reactivity changes caused by 
boron dilution and rod withdrawal.
    Date of publication of individual notice in Federal Register: June 
20, 1996 (61 FR 31559).
    Expiration date of individual notice: July 22, 1996.
    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.

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.

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

    Date of application for amendments: October 4, 1996 and 
supplemented on November 6, 1996.
    Brief description of amendments: The amendments add a Mode of 
Applicability to Technical Specification 3.2.3.D, Inoperable Rod 
Position Indicator Channels.
    Date of issuance: November 25, 1996.
    Effective date: Immediately, to be implemented within 30 days.
    Amendment Nos.: 176 and 163.
    Facility Operating License Nos. DPR-39 and DPR-48: The amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: October 17, 1996 (61 FR 
54240).
    The November 6, 1996, submittal provided additional clarifying 
information that did not affect the Commission's initial proposed 
finding of no significant hazards consideration determination.
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated November 25, 1996.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: Waukegan Public Library, 128 
N. County Street, Waukegan, Illinois 60085.

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

    Date of application for amendments: June 21, 1996.
    Brief description of amendments: The amendments revise Technical 
Specification (TS) Section 3/4.9.6, ``Manipulator Crane,'' to make the 
wording consistent with the TS Bases description and consistent with 
the design of the load handling equipment.
    Date of issuance: November 25, 1996.
    Effective date: As of the date of issuance to be implemented within 
30 days.
    Amendment Nos.: 156 and 148.
    Facility Operating License Nos. NPF-35 and NPF-52: Amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: October 23, 1996 (61 FR 
55031) The Commission's related evaluation of the amendments is 
contained in a Safety Evaluation dated November 25, 1996.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: York County Library, 138 East 
Black Street, Rock Hill, South Carolina 29730.

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

    Date of application of amendments: September 17, 1996 (TSC 96-01) 
as supplemented October 23, 1996.
    Brief description of amendments: The amendments lower the maximum 
allowable reactor building pressure, lower the actuation setpoint for 
actuation of the reactor building spray system, and modify the 
associated TS Bases requirements.
    Date of Issuance: November 25, 1996.
    Effective date: As of the date of issuance to be implemented within 
30 days.
    Amendment Nos.: 219, 219, 216.
    Facility Operating License Nos. DPR-38, DPR-47, and DPR-55: The 
amendments revised the Technical Specifications.
    Date of initial notice in Federal Register: October 23, 1996 (61 FR 
55031). The October 23, 1996, letter provided clarifying information 
that did not change the scope of the September 17, 1996, application 
and the initial proposed no signficant hazards consideration 
determination.
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated November 25, 1996.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: Oconee County Library, 501 
West South Broad Street, Walhalla, South Carolina 29691.

[[Page 66720]]

Entergy Operations, Inc., System Energy Resources, Inc., South 
Mississippi Electric Power Association, and Entergy Mississippi, Inc., 
Docket No. 50-416, Grand Gulf Nuclear Station, Unit 1, Claiborne 
County, Mississippi

    Date of application for amendment: July 31, 1996, as supplemented 
by letters of September 5, October 22, and November 15, 20, and 21, 
1996, which supersede the application submitted in the letter of May 9, 
1996.
    Brief description of amendment: The amendment (1) increased the 
safety limit minimum critical power ratio (MCPR) for two loop operation 
and single loop operation to 1.12 and 1.14, respectively, and (2) added 
two General Electric topical reports to the list of documents 
describing the analytical methods used to determine the core operating 
limits. The changes are to Section 2.1.1, Reactor Core Safety Limits, 
and Section 5.6.5, Core Operating Limits Report (COLR), respectively, 
of the Technical Specifications. This amendment would go into effect in 
Operating Cycle 9, at the end of the current Refueling Outage 8, and 
the plant will have a mixed core of Siemens Power Corporation (SPS) 
9 x 9-5 and General Electric (GE) GE11 reload fuel. The licensee also 
changed the Bases of the Technical Specifications associated with the 
above amendment.
    Date of issuance: November 21, 1996.
    Effective date: November 21, 1996.
    Amendment No: 131.
    Facility Operating License No. NPF-29: Amendment revises the 
Technical Specifications.
    Date of initial notice in Federal Register: September 25, 1996.The 
October 22, and November 15, 20, and 21, 1996, submittals provide 
clarifying information that did not change the initial determination. 
The Commission's related evaluation of the amendment is contained in a 
Safety Evaluation dated November 21, 1996.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: Judge George W. Armstrong 
Library, 220 S. Commerce Street, Natchez, MS 39120.

GPU Nuclear Corporation, et al., Docket No. 50-219, Oyster Creek 
Nuclear Generating Station, Ocean County, New Jersey

    Date of application for amendment: April 15, 1996 (TSCR No. 244).
    Brief description of amendment: The amendment revises Specification 
5.3.1.B to allow the shield plug and the associated lifting hardware to 
be moved over irradiated fuel assemblies that are in a dry shielded 
canister within the transfer cask in the cask drop protection system.
    Date of Issuance: November 7, 1996.
    Effective date: November 7, 1996, to be implemented within 30 days 
of issuance.
    Amendment No.: 187.
    Facility Operating License No. DPR-16: Amendment revises the 
Technical Specifications.
    Date of initial notice in Federal Register: May 8, 1996 (61 FR 
20849). The Commission's related evaluation of this amendment and final 
determination of no significant hazards consideration addressing 
comments received on the proposed no significant hazards consideration 
determination are contained in a Safety Evaluation dated November 7, 
1996.
    No significant hazards consideration comments received: Yes.
    Local Public Document Room location: Ocean County Library, 
Reference Department, 101 Washington Street, Toms River, NJ 08753.

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: February 22, 1996, and as 
supplemented by letters dated July 24, October 4, November 19 and 
November 25, 1996.
    Brief description of amendment: The amendment changes Clinton Power 
Station Technical Specification (TS) 3.3.8.1, ``Loss of Power 
Instrumentation,'' and TS 3.8.1, ``AC Sources-Operating,'' by revising 
the setpoint for the degraded voltage protection instrumentation and 
modifying or deleting other Loss of Power Instrumentation TS 
requirements. In addition, changes were also made to the minimum 
required diesel generator voltage specified for certain diesel 
generator surveillances.
    Date of issuance: December 4, 1996.
    Effective date: December 4, 1996.
    Amendment No.: 110.
    Facility Operating License No. NPF-62: The amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: April 24, 1996 (61 FR 
18168).The letters of July 24, October 4, November 19 and November 25, 
1996, provided clarifying information and did not represent significant 
changes from the original Federal Register notice.
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated December 4, 1996.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: The Vespasian Warner Public 
Library, 120 West Johnson Street, Clinton, Illinois 61727.

Southern California Edison Company, et al., Docket Nos. 50-361 and 50-
362, San Onofre Nuclear Generating Station, Unit Nos. 2 and 3, San 
Diego County, California

    Date of application for amendments: October 11, 1996.
    Brief description of amendments: These amendments revise Technical 
Specification (TS) 3.9.6, ``Refueling Water Level,'' for San Onofre 
Nuclear Generating Station (SONGS), Units 2 and 3. The proposed change 
is required to restore certain provisions of the SONGS Units 2 and 3 
operating practice that were not incorporated during the conversion to 
the improved TS (Amendment Nos. 127 and 116, dated February 9, 1996).
    Date of issuance: December 3, 1996.
    Effective date: December 3, 1996, to be implemented within 30 days 
from the date of issuance.
    Amendment Nos.: Unit 2--134; Unit 3--123.
    Facility Operating License Nos. NPF-10 and NPF-15: The amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: October 31, 1996 (61 FR 
56251) The Commission's related evaluation of the amendments is 
contained in a Safety Evaluation dated December 3, 1996.
    No significant hazards consideration comments received: No.
    Temporary Local Public Document Room location: Science Library, 
University of California, P.O. Box 19557, Irvine, California 92713.

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

    Date of amendment request: March 24, 1995, as supplemented by 
letter dated July 26, 1996.
    Brief description of amendment: The amendment revised Technical 
Specification (TS) Surveillance Requirement 4.5.1.1.a.1 to base 
accumulator operability on actual parameters (i.e., borated water 
volume and nitrogen cover-pressure in the tanks) vs. the absence of 
alarms.
    Date of issuance: November 22, 1996.
    Effective date: November 22, 1996, to be implemented within 30 days 
of issuance.
    Amendment No.: 103.
    Facility Operating License No. NPF-42: The amendment revised the 
Technical Specifications.

[[Page 66721]]

    Date of initial notice in Federal Register: April 12, 1995 (60 FR 
18632) The July 26, 1996, letter provided additional clarifying 
information and did not change the initial no significant hazards 
consideration determination. The Commission's related evaluation of the 
amendment is contained in a Safety Evaluation dated November 22, 1996.
    No significant hazards consideration comments received: No.
    Local Public Document Room locations: Emporia State University, 
William Allen White Library, 1200 Commercial Street, Emporia, Kansas 
66801 and Washburn University School of Law Library, Topeka, Kansas 
66621.

    Dated at Rockville, Maryland, this 11th day of December 1996.

    For the Nuclear Regulatory Commission.
Steven A. Varga,
Director, Division of Reactor Projects--I/II, Office of Nuclear Reactor 
Regulation.
[FR Doc. 96-31944 Filed 12-17-96; 8:45 am]
BILLING CODE 7590-01-P