[Federal Register Volume 60, Number 187 (Wednesday, September 27, 1995)]
[Notices]
[Pages 49929-49963]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: X95-10927]



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NUCLEAR REGULATORY COMMISSION
Biweekly Notice


Applications and Amendments to Facility Operating 
LicensesInvolving No Significant Hazards Considerations

I. Background

    Pursuant to Public Law 97-415, the U.S. Nuclear Regulatory 
Commission (the Commission or NRC staff) is publishing this regular 
biweekly notice. Public Law 97-415 revised section 189 of the Atomic 
Energy Act of 1954, as amended (the Act), to require the Commission to 
publish notice of any amendments issued, or proposed to be issued, 
under a new provision of section 189 of the Act. This provision grants 
the Commission the authority to issue and make immediately effective 
any amendment to an operating license upon a determination by the 
Commission that such amendment involves no significant hazards 
consideration, notwithstanding the pendency before the Commission of a 
request for a hearing from any person.
    This biweekly notice includes all notices of amendments issued, or 
proposed to be issued from August 30, 1995, through September 15, 1995. 
The last biweekly notice was published on Wednesday, September 13, 1995 
(60 FR 47613).

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 


[[Page 49930]]
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 Rules Review and 
Directives Branch, Division of Freedom of Information and Publications 
Services, Office of Administration, U.S. Nuclear Regulatory Commission, 
Washington, DC 20555, and should cite the publication date and page 
number of this Federal Register notice. Written comments may also be 
delivered to Room 6D22, Two White Flint North, 11545 Rockville Pike, 
Rockville, Maryland from 7:30 a.m. to 4:15 p.m. Federal workdays. 
Copies of written comments received may be examined at the NRC Public 
Document Room, the Gelman Building, 2120 L Street, NW., Washington, DC. 
The filing of requests for a hearing and petitions for leave to 
intervene is discussed below.
    By October 27, 1995, the licensee may file a request for a hearing 
with respect to issuance of the amendment to the subject facility 
operating license and any person whose interest may be affected by this 
proceeding and who wishes to participate as a party in the proceeding 
must file a written request for a hearing and a petition for leave to 
intervene. Requests for a hearing and a petition for leave to intervene 
shall be filed in accordance with the Commission's ``Rules of Practice 
for Domestic Licensing Proceedings'' in 10 CFR Part 2. Interested 
persons should consult a current copy of 10 CFR 2.714 which is 
available at the Commission's Public Document Room, the Gelman 
Building, 2120 L Street, NW., Washington, DC and at the local public 
document room for the particular facility involved. If a request for a 
hearing or petition for leave to intervene is filed by the above date, 
the Commission or an Atomic Safety and Licensing Board, designated by 
the Commission or by the Chairman of the Atomic Safety and Licensing 
Board Panel, will rule on the request and/or petition; and the 
Secretary or the designated Atomic Safety and Licensing Board will 
issue a notice of a hearing or an appropriate order.
    As required by 10 CFR 2.714, a petition for leave to intervene 
shall set forth with particularity the interest of the petitioner in 
the proceeding, and how that interest may be affected by the results of 
the proceeding. The petition should specifically explain the reasons 
why intervention should be permitted with particular reference to the 
following factors: (1) the nature of the petitioner's right under the 
Act to be made a party to the proceeding; (2) the nature and extent of 
the petitioner's property, financial, or other interest in the 
proceeding; and (3) the possible effect of any order which may be 
entered in the proceeding on the petitioner's interest. The petition 
should also identify the specific aspect(s) of the subject matter of 
the proceeding as to which petitioner wishes to intervene. Any person 
who has filed a petition for leave to intervene or who has been 
admitted as a party may amend the petition without requesting leave of 
the Board up to 15 days prior to the first prehearing conference 
scheduled in the proceeding, but such an amended petition must satisfy 
the specificity requirements described above.
    Not later than 15 days prior to the first prehearing conference 
scheduled in the proceeding, a petitioner shall file a supplement to 
the petition to intervene which must include a list of the contentions 
which are sought to be litigated in the matter. Each contention must 
consist of a specific statement of the issue of law or fact to be 
raised or controverted. In addition, the petitioner shall provide a 
brief explanation of the bases of the contention and a concise 
statement of the alleged facts or expert opinion which support the 
contention and on which the petitioner intends to rely in proving the 
contention at the hearing. The petitioner must also provide references 
to those specific sources and documents of which the petitioner is 
aware and on which the petitioner intends to rely to establish those 
facts or expert opinion. Petitioner must provide sufficient information 
to show that a genuine dispute exists with the applicant on a material 
issue of law or fact. Contentions shall be limited to matters within 
the scope of the amendment under consideration. The contention must be 
one which, if proven, would entitle the petitioner to relief. A 
petitioner who fails to file such a supplement which satisfies these 
requirements with respect to at least one contention will not be 
permitted to participate as a party.
    Those permitted to intervene become parties to the proceeding, 
subject to any limitations in the order granting leave to intervene, 
and have the opportunity to participate fully in the conduct of the 
hearing, including the opportunity to present evidence and cross-
examine witnesses.
    If a hearing is requested, the Commission will make a final 
determination on the issue of no significant hazards consideration. The 
final determination will serve to decide when the hearing is held.
    If the final determination is that the amendment request involves 
no significant hazards consideration, the Commission may issue the 
amendment and make it immediately effective, notwithstanding the 
request for a hearing. Any hearing held would take place after issuance 
of the amendment.
    If the final determination is that the amendment request involves a 
significant hazards consideration, any hearing held would take place 
before the issuance of any amendment.
    A request for a hearing or a petition for leave to intervene must 
be filed with the Secretary of the Commission, U.S. Nuclear Regulatory 
Commission, Washington, DC 20555, Attention: Docketing and Services 
Branch, or may be delivered to the Commission's Public Document Room, 
the Gelman Building, 2120 L Street, NW., Washington DC, by the above 
date. Where petitions are filed during the last 10 days of the notice 
period, it is requested that the petitioner promptly so inform the 
Commission by a toll-free telephone call to Western Union at 1-(800) 
248-5100 (in Missouri 1-(800) 342-6700). The Western Union operator 
should be given Datagram Identification Number N1023 and the following 
message addressed to (Project Director): petitioner's name and 
telephone number, date petition was mailed, plant name, and publication 
date and page number of this Federal Register notice. A copy of the 
petition 

[[Page 49931]]
should also be sent to the Office of the General Counsel, U.S. Nuclear 
Regulatory Commission, Washington, DC 20555, and to the attorney for 
the licensee.
    Nontimely filings of petitions for leave to intervene, amended 
petitions, supplemental petitions and/or requests for a hearing will 
not be entertained absent a determination by the Commission, the 
presiding officer or the Atomic Safety and Licensing Board that the 
petition and/or request should be granted based upon a balancing of 
factors specified in 10 CFR 2.714(a)(1)(i)-(v) and 2.714(d).
    For further details with respect to this action, see the 
application for amendment which is available for public inspection at 
the Commission's Public Document Room, the Gelman Building, 2120 L 
Street, NW., Washington, DC, and at the local public document room for 
the particular facility involved.

Carolina Power & Light Company, et al., Docket No. 50-324, 
Brunswick Steam Electric Plant, Unit 2, Brunswick County, North 
Carolina

    Date of amendment request: August 4, 1995
    Description of amendment request: The proposed amendment will allow 
the loading and use of GE13 fuel assemblies in the Brunswick Steam 
Electric Plant (BSEP), Unit 2, during Cycle 12 operation. The use of 
GE13 fuel assemblies requires that the safety limit value for minimum 
critical power ratio be revised. This safety limit is established to 
maintain fuel cladding integrity. Use of GE13 fuel also requires an 
increase in the concentration of sodium pentaborate solution required 
by the Technical Specifications (TS) for the standby liquid control 
system. This change provides the additional shutdown reactivity 
necessary to permit use of this fuel.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:
    1. The proposed amendment does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    Proposed Change 1:
    The proposed amendment will allow the loading and use of GE13 
fuel assemblies in the Brunswick Unit 2 reactor core. The use of 
GE13 fuel assemblies requires that the safety limit minimum critical 
power ratio value also be revised. The safety limit minimum critical 
power ratio is established to maintain fuel cladding integrity. The 
GE13 fuel assembly design has been analyzed using methods that have 
been previously approved by the Nuclear Regulatory Commission and 
documented in General Electric Nuclear Energy's reload licensing 
methodology Topical Report (NEDE-24011-P-A-10, ``General Electric 
Standard Application for Reactor Fuel (GESTAR II)'' dated February 
1991).
    The proposed revision of the safety limit minimum critical power 
ratio does not alter any plant safety-related equipment, safety 
function, or plant operations that could change the probability of 
an accident. The change does not affect the design, materials, or 
construction standards applicable to the fuel bundles in a manner 
that could change the probability of an accident.
    A methodology that has been previously reviewed and accepted by 
the Nuclear Regulatory Commission was used to derive the both 
existing and updated safety limit minimum critical power ratio 
value. The same methodology criteria have been applied to derive the 
existing safety limit minimum critical power ratio of 1.07 as that 
used to derive the updated safety limit minimum critical power ratio 
value of 1.09. The updated safety limit minimum critical power ratio 
assures that fuel cladding protection equivalent to that provided 
with the existing safety limit minimum critical power ratio value is 
maintained. This ensures that the consequences of previously 
evaluated accidents are not significantly increased.
    Proposed Change 2:
    The standby liquid control system provides a means of reactivity 
control that is independent of the normal reactivity control system. 
The standby liquid control system must be capable of assuring that 
the reactor core can be placed in a subcritical condition at any 
time during reactor core life. Technical Specification Figure 3.1.5-
1 specifies the acceptable range of concentrations and volumes for 
sodium pentaborate solution used as a neutron absorber (i.e., for 
reactivity control). The portion of the sodium pentaborate 
concentration range shown in Technical Specification Figure 3.1.5-1 
applicable to the lower range of tank volumes is being revised to 
increase the required concentration of sodium pentaborate solution. 
This change is needed to account for the additional shutdown 
reactivity needed based on the planned use of GE13 fuel assemblies 
as reload fuel for the Unit 2 reactor core. Since the standby liquid 
control system is independent from the normal means of controlling 
reactor core reactivity and not used to control core reactivity 
during normal plant operations, the proposed revision to the sodium 
pentaborate concentration curve for the standby liquid control 
system does not alter any plant safety-related equipment, safety 
function, or plant operations that could change the probability of 
an accident.
    The current volume-concentration range of sodium pentaborate 
used in the standby liquid control system will achieve a sufficient 
concentration of boron in the reactor vessel to ensure reactor 
shutdown. Based on the increased reactivity of the new GE13 reload 
fuel assemblies, the required sodium pentaborate volume-
concentration range is being revised to ensure sufficient neutron 
absorbing solution is available to achieve reactor shutdown; 
therefore, the consequences of an accident previously evaluated are 
not significantly increased.
    2. The proposed amendment would not create the possibility of a 
new or different kind of accident from any accident previously 
evaluated.
    Proposed Change 1:
    The GE13 fuel assembly has been designed and complies with the 
acceptance criteria contained in General Electric Nuclear Energy's 
standard application for reactor fuel (GESTAR-II), which provides 
the latest acceptance criteria for new General Electric fuel 
designs. The GE13 fuel assembly complies with GESTAR-II acceptance 
criteria that have been previously reviewed and accepted by the 
Nuclear Regulatory Commission. The similarity of the GE13 fuel 
design to the previously accepted GE11 fuel design, in conjunction 
with the increased critical power capability of the GE13 fuel 
design, ensure that no new mode or condition of plant operation is 
being authorized by the loading and use of the GE13 fuel type. The 
proposed revision of the safety limit minimum critical power ratio 
from 1.07 to 1.09 does not modify any plant controls or equipment 
that will change the plant's responses to any accident or transient 
as given in any current analysis. Therefore, the proposed change to 
allow the loading and use of the GE13 fuel type and the revision of 
the safety limit minimum critical power ratio value from 1.07 to 
1.09 will not create the possibility for a new or different kind of 
accident from any accident previously evaluated.
    Proposed Change 2:
    As discussed above, the standby liquid control system provides a 
means of reactivity control that is independent of the normal 
reactivity control system and is capable of assuring that the 
reactor core can be placed in a subcritical condition at any time 
during reactor core life. The proposed revision to the sodium 
pentaborate concentration range does not modify the standby liquid 
control system or its controls, does not modify other plant systems 
and equipment, and does not permit a new or different mode of plant 
operation. As such, the proposed revision to the minimum pentaborate 
concentration value does not create the possibility of a new or 
different kind of accident from any accident previously evaluated.
    3. The proposed license amendment does not involve a significant 
reduction in a margin of safety.
    Proposed Change 1:
    As previously discussed, the GE13 fuel assembly design has been 
analyzed using methods that have been previously approved by the 
Nuclear Regulatory Commission and documented in General Electric 
Nuclear Energy's reload licensing methodology Topical Report (NEDE-
24011-P-A-10, ``General Electric Standard Application for Reactor 
Fuel (GESTAR II)'' dated February 1991). The safety limit minimum 
critical power ratio value is selected to maintain the fuel cladding 
integrity safety limit (i.e., that 99.9 percent of all fuel rods in 
the core be expected to avoid boiling transition). 

[[Page 49932]]
Appropriate operating limit minimum critical power ratio values are 
established, based on the safety limit minimum critical power ratio 
value, to ensure that the fuel cladding fuel integrity safety limit 
is maintained. The operating limit minimum critical power ratio 
values are incorporated in the Core Operating limits Report as 
required by Technical Specification 6.9.3.1. The new GE13 safety 
limit minimum critical power ratio value of 1.09 is based on the 
same fuel cladding integrity safety limit criteria at that for the 
GE11 safety limit minimum critical power ratio value of 1.07 (i.e., 
that 99.9 percent of all fuel rods in the core be expected to avoid 
boiling transition); therefore, the proposed change does not result 
in a significant reduction in the margin of safety.
    Proposed Change 2:
    As previously stated, the purpose of the standby liquid control 
is to inject a neutron absorbing solution into the reactor in the 
event that a sufficient number of control rods cannot be manually 
inserted to maintain subcriticality. Sufficient solution is to be 
injected such that the reactor will be brought from maximum rated 
power conditions to subcritical over the entire reactor temperature 
range from maximum operating to cold shutdown conditions. General 
Electric reactor fuel methodology establishes a fuel type dependent 
standby liquid control system shutdown margin to account for 
calculational uncertainties. General Electric calculations show that 
an in-vessel concentration of 660 ppm will provide an estimated 
standby liquid control system minimum shutdown margin of 4.1% delta 
k. To achieve an in-vessel concentration of 660 ppm, the acceptable 
range of standby liquid control system tank concentrations is being 
revised for the lower range of tank volumes. Thus, proposed revision 
of the standby liquid control system sodium pentaborate volume-
concentration range ensures that there will not be a significant 
reduction in the amount of available shutdown margin and, therefore, 
not 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: University of North Carolina 
at Wilmington, William Madison Randall Library, 601 S. College Road, 
Wilmington, North Carolina 28403-3297.
    Attorney for licensee: R. E. Jones, General Counsel, Carolina Power 
& Light Company, Post Office Box 1551, Raleigh, North Carolina 27602
    NRC Project Director: David B. Matthews

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 amendment request: April 10, 1995

    Description of amendment request: The requested amendment would 
modify Technical Specification 4.6.4.3 to allow a reduction in the 
number of hydrogen mitigation system igniters that must be maintained 
Operable. This would allow removal of the hydrogen igniters in the 
incore instrument tunnel.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:
    Criterion 1
    The requested amendments will not involve a significant increase 
in the probability or consequences of an accident previously 
evaluated. No impact upon accident probabilities will be created, 
since the EHM System is not an accident initiating system. In 
addition, it has been demonstrated that based on the results of 
computer analysis, and the review of results of an external study 
performed for a similar type containment, that hydrogen 
concentrations in the cavity during degraded core accidents will 
remain within acceptable limits. No impact on the plant response to 
any accident will be created (either design basis or beyond-design 
basis).
    Criterion 2
    The requested amendments will not create the possibility of a 
new or different kind of accident from any accident previously 
evaluated. As stated previously, the EHM System is not an accident 
initiating system. No new accident causal mechanisms will be created 
as a result of deleting the affected igniters. Plant operation will 
not be affected by the proposed amendments and no new failure modes 
will be created.
    Criterion 3
    The requested amendments will not involve a significant 
reduction in a margin of safety. No adverse impact upon any plant 
safety margins will be created. As shown previously, applicable 
computer analysis has successfully demonstrated that the affected 
igniters could be removed with no adverse consequences. No fission 
product barriers are being degraded. No change to the manner in 
which the units are operated is being made.
    Based upon the preceding analyses, Duke Power Company concludes 
that the requested amendments do not involve a significant hazards 
consideration.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 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: York County Library, 138 East 
Black Street, Rock Hill, South Carolina 29730
    Attorney for licensee: Mr. Albert Carr, Duke Power Company, 422 
South Church Street, Charlotte, North Carolina 28242
    NRC Project Director: Herbert N. Berkow

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 amendment request: September 1, 1995
    Description of amendment request: Generic Letter 88-16 provided 
guidance on removing cycle-specific parameters which are calculated 
using NRC approved methodologies from Technical Specifications (TS). 
The parameters are replaced in TS with a reference to a named report 
which contains the parameters, and a requirement that the parameters 
remain within the limits specified in the report. The proposed changes 
incorporate NRC approved methodologies, approved revisions to 
previously approved methodologies, or republished versions of 
previously approved methodologies into Section 6.9 of the Catawba TS. 
For Catawba, the limits to which these methodologies are applied are 
explicitly listed in the TS. Since the proposed changes only 
incorporate NRC approved methodologies into the TS the licensee 
proposed that the changes are administrative in nature and can be 
assumed to have no impact, or potential impact, on the health and 
safety of the public.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:
    The proposed changes will not create a significant hazards 
consideration, as defined by 10 CRF 50.92, because:
    1) The proposed changes will not involve a significant increase 
in the probability or consequences of an accident previously 
evaluated.
    The proposed changes are administrative in nature, and do not 
affect any system, procedure, or manipulation of any equipment which 
could affect the probability or consequences of any accident.
    2) The proposed changes will not create the possibility of any 
new or different kind of accident from any accident previously 
evaluated.
    The proposed changes are administrative in nature, and cannot 
introduce any new failure mode or transient which could create any 
accident.

[[Page 49933]]

    3) The proposed changes will not involve a significant reduction 
in a margin of safety.
    The proposed changes are administrative in nature, and will not 
affect any operating parameters or limits which could result in a 
reduction in a margin of safety.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: York County Library, 138 East 
Black Street, Rock Hill, South Carolina 29730
    Attorney for licensee: Mr. Albert Carr, Duke Power Company, 422 
South Church Street, Charlotte, North Carolina 28242
    NRC Project Director: Herbert N. Berkow

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 amendment request: September 13, 1995
    Description of amendment request: The proposed amendments modify 
the notation for the overpower delta-temperature (OPDT) reactor trip 
heatup setpoint penalty coefficient to be consistent with NUREG-0452, 
Revision 4, ``Standard Technical Specifications for Westinghouse 
Pressurized Water Reactors'' (STS). This change is necessary in order 
to allow implementation of the modification to reduce the reactor 
coolant system hot leg temperature as planned during the Unit 2 end-of-
cycle 7 refueling outage.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:
    As required by 10CFR50.91, this analysis is provided concerning 
whether the requested amendments involve significant hazards 
considerations, as defined by 10CFR50.92. Standards for 
determination that an amendment request involves no significant 
hazards considerations are if operation of the facility in 
accordance with the requested 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.
    Criterion 1
    The proposed amendments will not involve a significant increase 
in the probability or consequences of an accident previously 
evaluated. The amendments will have no impact whatsoever upon the 
probability of any accident being initiated, since the reactor trip 
system is an accident mitigating system. The amendments will have no 
adverse impact upon any accident consequences or upon the function 
of the OPDT setpoint. The reactor trip heatup setpoint penalty will 
continue to be applied anytime T-avg is greater than T [double 
prime] and will not be applied when T-avg is less than or equal to T 
[double prime]. This is consistent with the intent of this function.
    Criterion 2
    The proposed amendments will not create the possibility of a new 
or different kind of accident from any accident previously 
evaluated. The function of the OPDT setpoint will not be altered by 
the proposed changes. As stated previously, the reactor trip system 
is an accident mitigating system, so no new failure modes can be 
created. No change to any aspect of plant operation will result from 
NRC approval of the proposed amendments.
    Criterion 3
    The proposed amendments will not involve a significant reduction 
in a margin of safety. The changes are necessary to allow full 
implementation of the T-hot reduction modification on Catawba Unit 
2. The proposed changes are consistent with the terminology of both 
NUREG-0452, Revision 4 and NUREG-1431, Revision 1. OPDT setpoint 
behavior will not be adversely impacted by the proposed changes; 
therefore, no impact upon any plant safety margins will result.
    Based upon the preceding analyses, Duke Power Company concludes 
that the requested amendments do not involve a significant hazards 
consideration.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 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: York County Library, 138 East 
Black Street, Rock Hill, South Carolina 29730
    Attorney for licensee: Mr. Albert Carr, Duke Power Company, 422 
South Church Street, Charlotte, North Carolina 28242
    NRC Project Director: Herbert N. Berkow

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

    Date of amendment request: March 29, 1995
    Description of amendment request: The amendments would revise the 
Technical Specification 3.4.9.3 requirements for the Low Temperature 
Overpressure Protection (LTOP) system and update the heatup and 
cooldown curves. The intent of the proposed amendments is to enhance 
overpressure protection during low temperature operations. These 
enhancements can be fully implemented, improving startup and shutdown 
operation of McGuire Units 1 and 2.
    Specifically, these changes are categorized into five groups 
identified as follows:
    1) Revisions to the LCO requirements, the Action Statements and the 
SR for the Reactor Coolant System Overpressure Protection System during 
low temperature conditions,
    2) A reduction in the Reactor Coolant System (RCS) vent requirement 
from 4.5 square inches to 2.75 square inches,
    3) The use of the Residual Heat Removal suction relief valve (1ND3 
and 2ND3) for overpressure protection under restricted conditions. (RCS 
greater than 107 deg.F and cooldown rate less than 20 deg.F/hr; or RCS 
greater than 167 deg.F),
    4) Revisions of the Pressure/Temperature curves to 16 EFPY, 
including the incorporation of the latest radiation surveillance 
capsule results and removal of instrumentation margins from the 
Technical Specification figures, and
    5) Changes to format and consistency.
    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 for each of the five groups listed above.
    FIRST STANDARD
    (Amendment would not) involve a significant increase in the 
probablility or consequences of an accident previously evaluated.
    1) Revised LCO [limiting conditions for operation] and SR 
[surveillance requirements] for LTOP:
    The reduced maximum setpoint will prevent the violation of the 
10 CFR 50 Appendix G pressure/temperature curves (as modified by the 
provisions of ASME Code Case N-514) during overpressure transients 
at low temperatures. Since the maximum setpoint is reduced, the peak 
pressure for LTOP [low-temperature overpressure protection] events 
will be reduced as well. Accordingly, the consequences of an LTOP 
event would not change as result of the proposed changes.
    The analysis performed to determine the setpoint is, in 
accordance with the methods used in previous evaluations, found 
acceptable by the NRC. The three possible transients evaluated are; 
1) a mass input from an operable safety injection pump; 2) a mass 
input from an operable centrifugal charging pump; and 3) a heat 
input from a 50 deg.F temperature difference between the steam 
generators and the NC system. The LTOP setpoint of the PORV [power-
operated relief valve] proposed by this technical specification 
change is not considered to be an initiator of any of these three 
transients. As such, the probability of an accident 

[[Page 49934]]
previously evaluated would not be increased as a result of the proposed 
changes.
    Two additional conditions for operability of the LTOP system are 
defined (accumulator isolation and only one NV or NI pump operable) 
and new surveillance requirements are specified as well. They 
provide additional limitations, requirements and restrictions that 
currently do not exist within the technical specifications for 
McGuire. The incorporation of these proposed changes are consistent 
with what is specified within NUREG-1341. Therefore, these changes 
do not increase the probability of consequences of an accident 
previously evaluated.
    2) Reduction in NC vent opening:
    The bases for the size of the vent to be established per the 
technical specifications is to ensure that the 10 CFR 50, Appendix G 
pressure/temperature limits are not exceeded during an LTOP event. 
The determination of the size of the opening continues to preserve 
the above design basis. The evaluation performed demonstrated that a 
2.75 square inch opening would provide adequate overpressure 
protection for the combined capacity of a centrifugal charging pump 
and a safety injection pump.
    The only time that the vent path is to be established is when 
the PORVs may not be available. Defining the size of the vent is not 
considered to be an initiator of any LTOP events that have been 
previously evaluated. As such, this change in the size of the vent 
opening does not increase the probability of an overpressure event 
during low temperature conditions. The analysis performed verifies 
that the size opening specified is sufficient to mitigate the 
consequences of an LTOP event. Accordingly, the change in the size 
of the opening for the vent will not impact the consequences of LTOP 
events.
    3) Use of RHR [residual heat removal] suction relief valves:
    By letter dated September 11, 1990, the NRC authorized the 
deletion of the RHR autoclosure interlock circuitry. A modification 
which removed the RHR system suction isolation valve autoclosure 
interlocks has been completed. As such, the RHR suction relief valve 
can be exposed to NC system pressure and would be available to 
mitigate LTOP events.
    The proposed amendments specify the necessary requirements and 
controls to ensure proper ND system alignments and conditions will 
exist to protect the pressure/temperature limits. This added 
relieving capacity will enhance the current LTOP system at McGuire 
in mitigating overpressure events at low temperatures. As such, the 
mitigation of previously evaluated LTOP events would be improved by 
the proposed technical specification changes. Further, the proposed 
changes would not esult in the initiation of an LTOP event or cause 
an overpressure transient. Accordingly, the proposed amendment would 
not involve an increase in the consequences or the probability of an 
accident previously evaluated.
    4) Revised pressure/temperature curves to 16 EFPY [effective 
full-power year]:
    The proposed pressure/temperature curves, provided by this 
amendment request, satisfy all regulatory required material 
embrittlement considerations including: ASME Section XI Appendix G, 
10 CFR 50 Appendix G, and Regulatory Guide 1.99, Revision 2. In 
addition, the margins for instrument error have been removed from 
the curves. Instrument error will be administratively handled by 
incorporating them into the LTOP system setpoint selection 
calculations and into appropriate controlling procedures for unit 
operations.
    The proposed changes to the pressure/temperature curves are not 
considered to be an initiator of LTOP events. The changes to the 
curves proposed by this amendment request will not cause an LTOP 
event. The curves define the new limits that have been defined in 
accordance with regulatory requirements by which both units are to 
be operated within. Accordingly, the proposed amendment will not 
increase the probability or the consequences of previously evaluated 
accidents.
    5) Format and consistency:
    The changes associated within this group are considered to be 
administrative in nature. They do not affect station operability or 
require any modifications to the facility. Accordingly, the proposed 
amendment request does not increase the probability or consequences 
of any previously evaluated accident.
    SECOND STANDARD
    (Amendment would not) create the possibility of a new or 
different kind of accident from any kind of accident previously 
evaluated.
    1) Revised LCO and SR for LTOP:
    The only potential impact to plant systems, structures and 
components, as a result of the proposed changes associated with this 
group, would be the setting of the PORV low pressure setpoint. No 
other changes to plant systems, structures or components would 
occur. The proposed amendments, also, would not impact the plant 
operation. Although the value for the PORV pressure setting 
specified within the technical specification would be reduced per 
the proposed amendment, the actual settings of the PORV are now 
currently set low enough to comply with the proposed lower setpoint 
value. As such, the proposed lower setpoint would not require any 
changes to the plant nor how the plant is operated.
    The additional requirements for LTOP operability will not 
require any modifications to the plant nor how the plant is 
operated. Currently, when entering LTOP conditions, the accumulators 
are isolated and only one NV or NI pump is capable of injecting into 
the reactor vessel. these actions are currently controlled and are 
specified within the operating procedures for heatup and cooldown of 
the respective units. The proposed changes will now specify these 
current operating requirements within the technical specifications 
as well.
    Accordingly, the proposed revisions will not create a new or 
different kind of accident than what has already been previously 
evaluated.
    2) Reduction in NC vent opening:
    The proposed changes to the technical specifications associated 
with this group involves the size of the vent opening. The proposed 
amendment reduces the size of the vent opening from 4.5 square 
inches to 2.75 square inches. The analysis that was performed has 
determined that the proposed size for the vent opening is adequate 
for overpressure events. Therefore, this proposed revision to the 
technical specifications will not result in a new or different kind 
of accident from any kind of accident previously evaluated.
    3) Use of RHR suction relief valves;
    The proposed amendment associated with this group will specify 
the necessary requirements and controls to ensure the appropriate 
use of the RHR suction relief valve for overpressure protection. 
This added relieving capacity will enhance the current LTOP system 
in mitigating overpressure events during low temperature conditions. 
The analysis that has been performed demonstrates the adequacy of 
the RHR suction relief valve, in conjunction with a PORV, in 
mitigating overpressure events at low temperatures, assuming a worst 
case single failure as well. As such, the use of the RHR suction 
relief valve in the manner prescribed by the proposed technical 
specification amendment will not create a new or different kind of 
accident from those accidents that have been previously evaluated.
    4) Revised pressure/temperature curves to 16 EFPY:
    The changes associated with this group, provide new heatup and 
cooldown curves for both Units 1 and 2, which will extend the 
service period from 10 EFPY to 16 EFPY and will remove the 
instrument error as well. The proposed [heatup] and cooldown curves 
were developed in accordance with all regulatory required material 
embrittlement criteria. Thus, operation of the units in accordance 
with the proposed new pressure/temperature curves will not create 
the possibility of a new or different kind of accident from those 
accident[s] that have been previously evaluated.
    5) Format and consistency:
    The changes associated within this group are considered to be 
administrative in nature. They do not affect station operability or 
require any modifications to the facility. Accordingly, the proposed 
amendment will create the possibility of a new or different kind of 
accident from that previously evaluated.
    THIRD STANDARD
    (Amendment would not) involve a significant reduction in a 
margin of safety.
    1) Revised LCO and SR for LTOP:
    This proposed change will reduce the maximum PORV setpoint such 
that, for LTOP events, the maximum pressure in the vessel would not 
exceed 110% of the pressure/temperature limits that have been 
established in accordance with ASME Appendix G. This is congruous 
with the provisions of ASME Code Case N-514. Currently, the maximum 
PORV setpoint for LTOP events ensure that the maximum pressure would 
not exceed 100% of the pressure/temperature curves. As such, the 
proposed change appears to involve a slight reduction in a margin of 
safety.
    Although the proposed change may involve a slight reduction in a 
margin of safety, the proposed change will provide an 

[[Page 49935]]
equivalent margins of safety to the reactor vessel during LTOP 
transients and will satisfy the underlying purpose of 10 CFR 50.60 
for fracture toughness requirements. By letter dated June 28, 1994, 
an exemption request and authorization to use ASME Code Case N-514 
at McGuire was submitted to the NRC for review and approval. 
Approval for the use of the code case was granted on September 30, 
1994. The proposed change to reduce the maximum PORV setpoint, 
coupled with the September 30, 1994 NRC approval for the use of Code 
Case N-514 satisfies current regulatory acceptance criteria. 
Therefore, the proposed change would not involve a significant 
reduction in a margin of safety.
    This change group, also, defines two additional conditions for 
the operability of the LTOP system (accumulator isolation and only 
one NV or NI pump operable) and proposes new surveillance 
requirements and restrictions that currently do not exist within the 
technical specifications for McGuire. The incorporation of these 
proposed changes are consistent with what is specified within NUREG-
1341. Therefore, these changes do not involve a significant 
reduction in a margin of safety.
    2) Reduction in NC vent opening:
    The proposed changes to the technical specifications associated 
with this group involves the size of the vent opening. The proposed 
amendment reduces the size of the vent opening from 4.5 square 
inches to 2.75 square inches. The basis for the size of the vent to 
be established per the technical specifications is to ensure that 
the 10 CFR 50, Appendix G pressure/temperature limits are not 
exceeded during an LTOP event. The determination of the size of the 
opening continues to preserve the above design basis. The evaluation 
performed demonstrated that a 2.75 square inch opening would provide 
adequate overpressure protection for the combined capacity of a 
centrifugal charging pump and a safety injection pump. Accordingly, 
the proposed changes would not involve a significant reduction in a 
margin of safety.
    3) Use of RHR suction relief valves:
    The proposed amendment associated with this group will specify 
the necessary requirements and controls to ensure the appropriate 
use of the RHR suction relief valves for overpressure protection. 
This added relieving capacity will enhance the current LTOP system 
in mitigating overpressure events during low temperature conditions. 
The analysis that has been performed demonstrates the adequacy of 
the RHR suction relief valve, in conjunction with a PORV, in 
mitigating overpressure events at low temperatures.
    Further, by letter dated September 11, 1990, the NRC approved 
amendments to delete a portion of the surveillance requirements 
regarding periodic verification that the RHR suction isolation 
valves automatically close on a RCS [reactor coolant system] signal 
less than or equal to 560 psig. This action, in effect, authorizes 
the removal of the RHR autoclosure interlock circuitry. As discussed 
within the NRC Safety evaluation for the amendment, the Commission 
and industry have recognized the safety benefits of removing the ACI 
[automatic closure and interlock] circuitry from the RHR system to 
minimize, and thus reduce the risk associated with loss of decay 
heat removal events.
    Therefore, the proposed amendments associated with this change 
group will not involve a significant reduction in a margin of 
safety.
    4) Revised pressure/temperature curves to 16 EFPY:
    The changes associated with this group provide new heatup and 
cooldown curves for both Units 1 and 2, which will extend the 
service period from 10 EFPY to 16 EFPY and will relocate the 
instrument error as well. The proposed pressure/temperature curves 
provided by this amendment request satisfy all regulatory required 
material embrittlement considerations including; ASME Section XI 
Appendix G, 10 CFR 50 Appendix G, and Regulatory Guide 1.99, 
Revision 2. The instrument error will be administratively handled by 
incorporating them into the LTOP system setpoint selection 
calculations and into the controlling procedures for unit 
operations.
    The relocation of the instrument error to licensee controlled 
documents is consistent with the NRC actions proposed within NUREG-
1431, new standard technical specifications for Westinghouse plants. 
As prescribed within NUREG-1431, the pressure/temperature limit 
curves are to be relocated to a licensee controlled document 
entitled ``Pressure Temperature Limit Report (PTLR)''. Changes to 
the heatup and cooldown curves would then be performed in accordance 
with 10 CFR 50.59 criteria. For the situation proposed by this 
amendment, updates and revisions of the instrument error associated 
with the pressure/temperature limit curves will be processed in a 
similar fashion. Thus, the proposed change to relocate the 
instrument error to licensee controlled documents is analogous with 
NRC acceptable practices.
    Accordingly, the proposed changes will not reduce a margin of 
safety.
    5) Format and consistency:
    The changes associated within this group are considered to be 
administrative in nature. They do not affect station operability or 
require any modifications to the facility. Accordingly, there is no 
reduction in the margin of safety of the LTOP system due to the 
incorporation of these editorial/administrative changes.
    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: Atkins Library, University of 
North Carolina, Charlotte (UNCC Station), North Carolina 28223
    Attorney for licensee: Mr. Albert Carr, Duke Power Company, 422 
South Church Street, Charlotte, North Carolina 28242
    NRC Project Director: Herbert N. Berkow

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

    Date of amendment request: June 21, 1995
    Description of amendment request: The proposed amendments will 
revise the action statements for a single inoperable Emergency Diesel 
Generator (EDG), TS 3.8.1.1.b, to extend the allowed outage time (AOT) 
from 72 hours to 7 days, and permit a 10 day AOT to be used once per 
refueling cycle. This proposal is a result of a cooperative study by 
participating Combustion Engineering Owners Group members which 
concluded that the proposed AOT extension improves plant operational 
flexibility while adequately controlling overall plant risk.
    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 for St. Lucie Unit 1 and Unit 2 will 
extend the action completion/allowed outage time (AOT) for a single 
inoperable Emergency Diesel Generator (EDG) from 72 hours to 7 days, 
with provisions for a 10 day AOT once per refueling cycle. The EDGs 
are designed as backup AC power sources for essential safety systems 
in the event of a loss of offsite power. As such, the EDGs are not 
accident initiators, and an extended AOT to restore operability of 
an inoperable diesel generator would not increase the probability of 
occurrence of accidents previously analyzed.
    The proposed technical specification revisions involve the AOT 
for a single inoperable EDG, and do not change the conditions, 
operating configuration, or minimum amount of operating equipment 
assumed in the plant safety analyses for accident mitigation. In 
addition, a Probability Safety Assessment (PSA) was performed to 
quantitatively assess the risk impact of the proposed amendment. The 
impact on the early radiological release probability for design 
basis events was also evaluated. It was concluded that the risk 
contribution from this proposed AOT is very small, and that the 
impact will be negligible.
    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 

[[Page 49936]]
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 either Facility License. The 
changes do not involve the addition or modification of equipment, 
nor do they alter the design 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 are designed to improve EDG reliability 
by providing flexibility in the scheduling and performance of 
preventive and corrective maintenance activities. The surveillance 
intervals or the operability requirements are not changed by the 
proposal; only the AOT for a single inoperable EDG will be extended. 
The proposed changes do not alter the basis for any technical 
specification that is related to the establishment of, or the 
maintenance of, a nuclear safety margin. Moreover, an integrated 
assessment of the risk impact of extending the AOT for a single 
inoperable EDG has determined that the risk contribution is very 
small and can be offset by improvements in EDG reliability. 
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: J. R. Newman, Morgan, Lewis & Bockius, 1800 
M Street, NW., Washington, DC 20036
    NRC Project Director: David B. Matthews

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

    Date of amendment request: June 21, 1995
    Description of amendment request: The proposed amendments will 
revise TS 3.5.2 to allow up to 7 days to restore an inoperable Low 
Pressure Safety Injection train to operable status. This proposal is a 
result of a cooperative study by participating Combustion Engineering 
Owners Group members which concluded that an extension of the allowed 
outage time (AOT) from 72 hours to 7 days can improve plant operational 
flexibility and is risk beneficial.
    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 for St. Lucie Unit 1 and Unit 2 will 
extend the action completion/allowed outage time (AOT) for a single 
inoperable Low Pressure Safety Injection (LPSI) train from 72 hours 
to 7 days. A LPSI train is designed as a part of each Emergency Core 
Cooling System (ECCS) subsystem to supplement Safety Injection Tank 
(SIT) inventory during the early stages of mitigating a Design Basis 
Accident. As such, components of the LPSI system are not accident 
initiators, and an extended AOT to restore operability of an 
inoperable LPSI train would not increase the probability of 
occurrence of accidents previously analyzed.
    The safety analyses for both St. Lucie Units demonstrate that 
ECCS performance acceptance criteria are satisfied with only one of 
the two redundant ECCS subsystems operating during the postulated 
Design Basis Accident. The proposed technical specification 
revisions involve the AOT for a single inoperable LPSI train, and do 
not change the conditions assumed for the minimum amount of 
operating equipment needed for accident mitigation. Therefore, the 
consequences of an accident previously evaluated will not be 
significantly increased.
    In addition to the preceding evaluation, a Probabilistic Safety 
Analysis (PSA) was performed to quantitatively assess the risk 
impact of the proposed amendments. It was concluded from the results 
of that assessment that the risk contribution of the AOT extension 
is very small, and that the net impact of the proposed amendment can 
be risk beneficial.
    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 either Facility License. The 
changes do not involve the addition or modification of equipment nor 
do they alter the design 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 margin of safety associated with the ECCS system is 
established by acceptance criteria for system performance defined in 
10 CFR 50.46. The proposed amendments will not change this 
acceptance criteria nor the operability requirements for equipment 
that is used to achieve such performance as demonstrated in the 
plant safety analyses. Moreover, an integrated assessment of the 
risk impact of extending the AOT for a single inoperable LPSI train 
has concluded that the risk contribution is very small, LPSI system 
reliability can potentially be improved, and the net impact of the 
proposed change can be risk beneficial. 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: J. R. Newman, Morgan, Lewis & Bockius, 1800 
M Street, NW., Washington, DC 20036
    NRC Project Director: David B. Matthews

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

    Date of amendment request: June 21, 1995
    Description of amendment request: The proposed amendments will 
revise the action statements and certain surveillances of TS 3/4.5.1, 
Safety Injection Tanks (SIT). This proposal is based on the results of 
a cooperative study performed by participating Combustion Engineering 
Owners Group members which investigated the impact of a risk-based 
allowed outage time (AOT) extension, and also included recommendations 
for line-item TS improvements from NUREG-1366 and Generic Letter 93-05.
    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 license amendments proposed for St. Lucie Units 1 and 2 
incorporate certain line-

[[Page 49937]]
item Technical Specifications (TS) improvements for the Safety 
Injection Tanks (SIT), and include an extension of the required 
action completion/allowed outage time (AOT) from one hour to 72 
hours to restore an inoperable SIT (that is still able to perform 
its safety function) to operable status. In addition, an AOT of 24 
hours, based on risk assessment techniques, is proposed for an SIT 
that may be unable to perform its design function.
    The SITs are passive components of the Emergency Core Cooling 
System (ECCS). As such, they are not accident initiators for any 
transient evaluated in the plant safety analyses, and an extension 
of the AOTs for restoring an inoperable SIT to operable status would 
not increase the probability of occurrence of accidents previously 
analyzed.
    The SITs, in combination with other ECCS components, are used to 
mitigate the consequences of a loss of coolant accident. The TS 
revisions will provide a longer AOT for a single inoperable SIT, but 
do not involve a change to the ECCS configuration or method of 
operation. The proposed amendments will not change the conditions 
assumed for the minimum amount of operating equipment needed for 
accident mitigation. Therefore, the consequences of an accident 
previously evaluated will not be significantly increased.
    In addition to the preceding evaluation, a Probability Safety 
Assessment (PSA) was performed to quantitatively assess the risk 
impact of the 24 hour AOT proposal. The impact on the early 
radiological release probability for design basis events was also 
evaluated. It was concluded that the risk contribution from this AOT 
is very small, and that the impact is negligible.
    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 either Facility License. The 
changes do not involve the addition or modification of equipment, 
nor do they alter the design 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 margin of safety associated with the ECCS system is 
established by acceptance criteria for system performance defined in 
10 CFR 50.46. The proposed amendments will not change this criteria 
nor the operability requirements for equipment that is used to 
achieve such performance as demonstrated by the plant safety 
analyses. Moreover, an integrated assessment of the risk impact of 
allowing 24 hours to restore an inoperable SIT to operable status 
has concluded that this impact is very small, and can be offset by 
averting an unnecessary transition to the shutdown modes. 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: J. R. Newman, Morgan, Lewis & Bockius, 1800 
M Street, NW., Washington, DC 20036
    NRC Project Director: David B. Matthews

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

    Date of amendment request: August 16, 1995
    Description of amendment request: The revisions will modify 
Technical Specification 3.6.6.1, Shield Building Ventilation System 
(SBVS), to more effectively address the design functions performed by 
the SBVS for both the Shield Building (secondary containment) and the 
Fuel Handling Building.
    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 license amendment for St. Lucie Unit 2 will clarify 
the Applicability and the Actions required by Technical 
Specification (TS) 3.6.6.1, and explicitly account for the dual 
purpose of the Shield Building Ventilation System (SBVS) to perform 
design functions for both the Shield Building (secondary 
containment) and the Fuel Handling Building. The proposed amendment 
is administrative in nature.
    The SBVS only operates when actuated by automatic control 
signals generated by systems detecting postulated accident 
conditions. The SBVS is not an accident initiator, the proposed TS 
changes do not involve any assumptions relative to accident 
initiators used in the plant safety analyses, and the amendment, 
therefore, will not impact the probability of occurrence for 
accidents previously analyzed. Relative to accident consequences, at 
least one train of the SBVS must operate to fulfill the design 
function of evacuating filtered air from the Shield Building during 
the postulated Loss of Coolant Accident; and likewise assumed in the 
analysis for the Fuel Handling Building during a fuel handling 
accident. The proposed changes simply remove elements of ambiguity 
from TS 3.6.6.1; do not reduce the existing operability requirements 
for the system; and provide further assurance that proper 
compensatory measures will be taken in the event one or both SBVS 
trains become inoperable.
    Therefore, operation of the facility in accordance with the 
proposed amendment will 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 amendment is administrative in nature and will not 
change the physical plant or the modes of plant operation defined in 
the facility license. The changes do not involve the addition or 
modification of equipment, nor do they alter the design or methods 
of operation of plant systems. Plant configurations that are 
prohibited by TS will not be created by this amendment. Therefore, 
operation of the facility in accordance with the proposed amendment 
would not create the possibility of a new or different kind of 
accident from any accident previously evaluated.
    (3) Operation of the facility in accordance with the proposed 
amendment would not involve a significant reduction in a margin of 
safety.
    The proposed amendment will not change the SBVS operability 
requirements nor otherwise alter the basis for any technical 
specification that is related to the establishment of, or the 
maintenance of, a nuclear safety margin. The proposed changes are 
administrative in nature, and are designed to provide assurance that 
the SBVS capability to perform design functions assumed available in 
the safety analyses will remain available during the various plant 
operating modes. Therefore, operation of the facility in accordance 
with the 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: J. R. Newman, Morgan, Lewis & Bockius, 1800 
M Street, NW., Washington, DC 20036
    NRC Project Director: David B. Matthews

[[Page 49938]]


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

    Date of amendment request: August 16, 1995
    Description of amendment request: The proposed amendments revise 
St. Lucie Units 1 and 2 Technical Specifications to relocate selected 
Technical Specification Monitoring Instrumentation utilizing the Final 
Policy Statement on Technical Specification Improvement for Nuclear 
Power Reactors, 58 FR 39132, July 22, 1993. The proposed amendments 
also include relocation of Technical Specifications related to the 
Emergency and Security Plan review process utilizing the guidance 
contained in NRC Generic Letter 93-07, ``Modification of the Technical 
Specification Administrative Requirements for Emergency and Security 
Plans.''
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:
    (1) Operation of the facility in accordance with the proposed 
amendments would not involve a significant increase in the 
probability or consequences of an accident previously evaluated.
    The proposed changes to the Selected Technical Specification 
Requirements Related to Instrumentation are administrative in nature 
in that the specifications for operation and surveillance of the 
selected Technical Specification instrumentation will be relocated 
from Appendix A of the facility operating license to the Updated 
Final Safety Analysis Report (UFSAR) for each unit. Once relocated, 
future changes will be controlled by 10 CFR 50.59 and the UFSARs 
updated pursuant to 10 CFR 50.71(e). Relocation of these 
requirements to the UFSAR is consistent with the NRC ``Final Policy 
Statement on Technical Specifications Improvements for Nuclear Power 
Reactors'' published in the Federal Register (58 FR 39132) dated 
July 22, 1993.
    The selected Technical Specification instruments are not 
accident initiators nor a part of the success path(s) which function 
to mitigate accidents evaluated in the plant safety analyses. The 
proposed Technical Specification change does not involve any change 
to the configuration or method of operation of any plant equipment 
that is used to mitigate the consequences of an accident, nor do the 
changes alter any assumptions or conditions in any of the plant 
accident analyses. Therefore, 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 Technical Specifications changes associated with Emergency 
Plan and Security Plan requirements are proposed in accordance with 
Generic Letter 93 07. The changes being proposed are administrative 
in nature and do not affect assumptions contained in plant safety 
analyses, the physical design and/or operation of the plant, nor do 
they affect Technical Specifications that preserve safety analysis 
assumptions. Therefore, operation of the facility in accordance with 
the proposed amendments would not affect the probability or 
consequences of an accident previously analyzed.
    (2) Use of the modified specification would not create the 
possibility of a new or different kind of accident from any 
previously evaluated.
    The proposed amendment to relocate the existing Technical 
Specification requirements for selected Technical Specification 
instrumentation to the UFSAR will not change the physical plant or 
the modes of plant operation defined in the Facility License. The 
change does not involve the addition or modification of equipment 
nor does it alter the design or operation of plant systems. 
Therefore, operation of the facility in accordance with the proposed 
amendment would not create the possibility of a new or different 
kind of accident from any accident previously evaluated.
    The proposed amendments, in accordance with Generic Letter 93-
07, change the Technical Specifications to remove the audit of the 
emergency and security plans and implementing procedures from the 
list of responsibilities of the Facility Review Group. The changes 
being proposed are administrative in nature and will not change the 
physical plant or the modes of operation defined in the Facility 
License. The change does not involve the addition or modification of 
equipment nor does it alter the design or operation of plant 
systems. Therefore, operation of the facility in accordance with the 
proposed amendments would not create the possibility of a new or 
different kind of accident from any accident previously evaluated.
    (3) Use of the modified specification would not involve a 
significant reduction in a margin of safety.
    The proposed changes are administrative in nature in that 
operating and surveillance requirements for the selected Technical 
Specification instrumentation will be relocated from Appendix A of 
the facility license to the appropriate Updated Final Safety 
Analysis Report for each unit. These selected instruments are not 
used to actuate safety-related equipment, provide interlocks, or 
otherwise perform plant control functions. Conditions evaluated in 
plant accident and transient analyses do not involve these selected 
instruments. The proposed changes 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 
the facility in accordance with the proposed amendment would not 
involve a significant reduction in a margin of safety.
    The proposed amendments, in accordance with Generic Letter 93-
07, change the Technical Specifications to remove the audit of the 
emergency and security plans and implementing procedures from the 
list of responsibilities of the Facility Review Group. The changes 
being proposed are administrative in nature and do not alter the 
bases for assurance that safety-related activities are performed 
correctly or the basis for any Technical Specification that is 
related to the establishment of or maintenance of a safety margin. 
Therefore, operation of the facility in accordance with the proposed 
amendments 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: J. R. Newman, Morgan, Lewis & Bockius, 1800 
M Street, N.W., Washington, DC 20036
    NRC Project Director: David B. Matthews

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

    Date of amendment request: July 21, 1995
    Description of amendment request: The proposed amendment would make 
administrative changes to various sections of the Duane Arnold Energy 
Center (DAEC) Technical Specifications (TS). These changes replace a 
conditional surveillance if one emergency service water (ESW) pump or 
loop is determined to be inoperable (TS 4.8.E.2); credit successful 
emergency diesel generator (EDG) tests performed in the previous 24 
hours (TS 4.8.E.2); clarify the requirements governing spent and new 
fuel storage in Section 5.5 of the DAEC TS; and eliminate the 
Operations Committee reviews of procedures in support of the DAEC 
Emergency Plan and Security Plan, as specified in Sections 6.5 and 6.8 
of the TS. DAEC TS Section 4.8.E.2 states the surveillance requirement 
applicable when one ESW pump or loop is determined to be inoperable. 
This amendment request deletes the surveillance requirement to 
physically test the opposite train's EDG and replaces it with a 
requirement to verify OPERABILITY of the opposite train low pressure 
core and containment cooling systems and EDG.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the 

[[Page 49939]]
licensee has provided its analysis of the issue of no significant 
hazards consideration, which is presented below:
    1. The proposed revision does not involve a significant increase 
in the probability or consequences of an accident previously 
evaluated. The changes are administrative in nature and are 
consistent with previously-published NRC guidance. The proposed 
revision does not change any accident analysis, plant safety 
analysis or calculations; degrade existing plant programs; or modify 
any functions of safety related systems or accident mitigation 
functions for which the DAEC has previously been credited. The 
proposed revision to the Surveillance Requirements will continue to 
assure OPERABILITY as required, but eliminate unnecessary operation 
of an EDG.
    2. The proposed revision does not create the possibility of a 
new or different kind of accident from any accident previously 
evaluated. The proposed revision does not alter any plant 
parameters, revise any safety limit setpoint, or provide any new 
release pathways. In addition, the proposed revision does not modify 
the operation or function of any safety-related equipment, nor 
introduce any new modes of operation, failure modes, or physical 
changes to the plant.
    3. The proposed revision does not involve a significant 
reduction in a margin of safety. The proposed revision does not 
alter any plant parameters, revise any safety limit setpoint, or 
provide any new release pathways. In addition, the proposed revision 
does not modify the operation or function of any safety-related 
equipment, nor introduce any new modes of operation, failure modes, 
or physical changes to the plant.
    The NRC staff has reviewed the licensee's analysis and, based on 
thisreview, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: Cedar Rapids Public Library, 
500 First Street, S.E., Cedar Rapids, Iowa 52401
    Attorney for licensee: Jack Newman, Kathleen H. Shea, Morgan, Lewis 
& Bockius, 1800 M Street, NW., Washington, DC 20036-5869
    NRC Project Director: Gail H. Marcus

Nebraska Public Power District, Docket No. 50-298, Cooper Nuclear 
Station, Nemaha County, Nebraska

    Date of amendment request: May 5, 1995, as revised by letter dated 
July 14, 1995
    Description of amendment request: The proposed changes would amend 
the Cooper Nuclear Station (CNS) Technical Specifications (TS) sections 
3/4.5.F.1, 3.5.F.2, 3.9.B.1, 3.9.B.2, 4.9.A.2, and the associated 
bases. These changes would revise the TS to: 1) verify that the 
redundant diesel generator is operable upon the loss of one diesel 
generator, and implement provisions to verify that the operable diesel 
generator does not have a common cause failure; 2) incorporate 
provisions to allow a modified start for the diesel generators; and 3) 
remove the requirement that the reactor power level be reduced to 25% 
of rated power upon loss of both diesel generator units or both 
incoming power sources (start-up and emergency transformers). In 
addition, the period of time allowed for continued reactor operation 
with both diesels inoperable would be reduced from 24 to two hours.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:
    10 CFR 50.91(a)(1), requires that licensee requests for 
operating license amendments be accompanied by an evaluation of 
significant hazards posed by the issuance of the amendment. NPPD has 
reviewed the proposed changes in accordance with 10CFR50.92 and 
concludes that the changes do not involve a significant hazards 
consideration (SHC). The basis for this conclusion is that the three 
criteria of 10CFR50.92(c) are not compromised. The proposed changes 
do not involve a SHC because the changes would not:
    1. Involve a significant increase in the probability or 
consequences of an accident previously analyzed.
    Proposed Revision 1:
    This proposed revision serves to ensure that an emergency diesel 
generator is always available to perform on demand and that lowering 
the number of demands to demonstrate operability reduces the 
probability of equipment failure. The required action no longer 
requires the redundant emergency diesel generator to be demonstrated 
operable immediately. Therefore, this requirement has been deleted 
from TS 4.5.F.1.
    The proposed change includes provisions to determine if the 
redundant diesel generator has been made inoperable by a common 
cause failure or perform a demonstration test. The redundant 
emergency diesel generator will remain in service during the entire 
period of inoperability of the out of service emergency diesel 
generator. If a common cause failure cannot be ruled out, the 
redundant diesel generator will be tested in accordance with the 
surveillance requirements of TS section 4.9.A.2.a.1 to assure 
operability.
    Since this proposed revision does not affect the design or 
negatively affect the performance of the diesel generators, the 
change will not result in an increase in the consequences or 
probability of an accident previously analyzed. This proposed 
revision will increase diesel generator reliability and 
availability, thereby increasing overall plant safety.
    Proposed Revision 2:
    This proposed revision only affects emergency diesel generator 
periodic testing. The diesel generators are not accident initiators 
and the method of testing the diesel generators cannot initiate an 
accident and therefore will not increase the probability of an 
accident. This change to the diesel generator testing method does 
not impact any Updated Safety Analysis Report (USAR) safety 
analysis. The proposed surveillances will still provide assurance 
that the diesel generators are available to mitigate the 
consequences of accidents previously evaluated. Thus the 
consequences of an accident previously evaluated are not increased.
    The revised periodic testing will still demonstrate that the 
emergency diesel generators are ready to perform their safety 
function. An overall improvement in diesel engine reliability and 
availability can be gained by performing diesel generator starts for 
surveillance testing using engine prelubes, warmups and other 
manufacturer recommended practices to reduce engine stress and wear. 
Since this proposed revision does not affect the design or 
negatively affect the performance of the diesel generators, the 
change will not result in an increase in the consequences or 
probability of an accident previously analyzed. This proposed 
revision will increase diesel generator reliability, thereby 
increasing overall plant safety.
    Proposed Revision 3:
    This proposed revision does not affect the operation of the 
emergency diesel generators or the incoming power sources (start-up 
and emergency transformers). Both the diesel generators and the 
incoming power sources function to mitigate the consequences of 
postulated accidents. As such, removing the requirement to reduce 
power level upon the loss of both redundant components in either of 
these systems does not create an increase in the probability of an 
accident. By eliminating this requirement, the potential for plant 
transients during power reduction to 25% are also eliminated. 
Eliminating this requirement will not increase the consequences of a 
postulated accident because the redundant components will remain 
available. Additionally, the loss of both offsite power sources 
condition becomes more restrictive by requiring a plant shutdown 
instead of notification within 24 hours.
    The proposed changes do not alter the conditions or assumptions 
in any of the Updated Safety Analysis Report (USAR) accident 
analyses. Since the USAR accident analyses remains bounding, the 
radiological consequences previously evaluated are not adversely 
affected by the proposed changes. Therefore, no significant increase 
in the probability or consequences of an accident previously 
analyzed would occur.
    The proposed rearrangement of information, and rewording of some 
the TS requirements are included to enhance usability and alleviate 
any possible confusion. These changes are strictly editorial have no 
impact, and do not alter technical content or meaning of the 
specifications. These editorial changes do not involve a significant 
increase in the probability or consequences of an accident 
previously analyzed.

[[Page 49940]]

    2. Create the possibility of a new or different kind of accident 
from any previously analyzed.
    Proposed Revision 1:
    Accidents involving loss of off-site power and single failure 
have been previously evaluated, and this proposed change does not 
impact any of those assumptions. This proposed revision does not 
introduce any new mode of plant operation or new accident 
precursors, involve any physical alterations to plant 
configurations, or make changes to system setpoints which could 
initiate a new or different kind of accident. Operation of the 
facility in accordance with the proposed revised changes does not 
create the possibility of a new or different kind of accident from 
any previously evaluated.

    Proposed Revision 2:
    This proposed revision only affects emergency diesel generator 
periodic testing. The diesel generators are not accident initiators 
and the method of testing the diesel generators cannot initiate an 
accident. This revision does not relieve the operation of the diesel 
generator from existing requirements and the diesel generators 
remain bounded by the assumptions in the USAR accident analysis. The 
method of testing provides assurance that the diesel generators are 
available when needed. The proposed revision does not involve any 
changes in setpoints, plant equipment, plant operation, protective 
functions, or the design basis of the plant. Therefore, a change in 
the method of starting the diesel generators during periodic testing 
would not create a different kind of accident than previously 
evaluated.

    Proposed Revision 3:
    This proposed revision does not add or change any equipment or 
logic, nor do the changes associated with this revision alter any 
system operability requirements. The proposed changes for this 
revision do not introduce any new failure modes for any plant system 
or component important to safety nor has any new limiting failure 
been identified as a result of the proposed revision. Since there 
are no changes to the function, or operation of any system, 
equipment, or component, the possibility of a new or different kind 
of accident is not created.
    The proposed rearrangement of information, and rewording of some 
[of] the TS requirements are included to enhance usability and 
alleviate any possible confusion. These changes are strictly 
editorial have no impact, and do not alter technical content or 
meaning of the specifications. These editorial changes do not create 
the possibility of a new or different kind of accident from any 
previously analyzed.
    3. Involve a significant reduction in the margin of safety.
    Proposed Revision 1:
    This proposed revision does not result in an overall reduction 
in the margin of safety. The reduction in margin going from 
``immediately'' testing an operable diesel generator to 24 hours to 
determine no common cause, is offset by the increase in margin 
resulting from increased diesel generator reliability and 
availability associated with implementing the vendor recommendations 
for testing and not exposing the diesel generator to potential grid 
disturbances when a diesel generator is found to be inoperable. No 
physical modification to the plant or change in the procedurally 
prescribed operator actions result from the proposed changes 
associated with this revision. Operation of the facility in 
accordance with the proposed revision does not involve a significant 
reduction in a margin of safety.

    Proposed Revision 2:
    This proposed revision is made to increase the reliability and 
availability of the emergency diesel generators thus enhancing the 
safety of the plant. Changing the way periodic testing of the diesel 
generators is conducted does not involve a reduction in safety. The 
test still demonstrates the ability of the diesel generator to start 
within the time required, and reach rated voltage and frequency as 
required in the accident analysis. The test also demonstrates the 
ability of the diesel generator to start reliably, carry the 
required load, and ensures the capabilities of the cooling system 
and other support systems are operable. Therefore, assurance that 
the diesel generators operate within the limits determined to be 
acceptable continues to be provided. Implementing manufacturer's 
recommendations to minimize stress and wear of the diesel engine 
does not involve a significant reduction in the margin of safety, 
but rather enhances safety.
    Proposed Revision 3:
    This proposed revision deletes the requirement to reduce reactor 
power level to 25% of rated power upon the loss of either both 
diesel generators or both incoming power sources. The elimination of 
this requirement will allow the plant to maintain the existing power 
level rather than subject the plant to an unnecessary transient. 
Maintaining the plant at the existing power level provides a more 
stable operating environment. The equipment and components of the 
diesel generators or the incoming power sources are not impacted in 
any way as a result of the proposed revisions. The margin of safety 
for the diesel generators and the incoming power sources are not 
significantly reduced since these systems are not altered in any 
way, and will continue to be surveillance tested as required. 
Assurance of operability is provided by the normal, scheduled 
surveillances which have been established at a sufficient interval 
to provide reasonable assurance of operability. Therefore, the 
proposed changes do not involve a significant reduction in the 
margin of safety.
    The proposed rearrangement of information, and rewording of some 
[of] the TS requirements are included to enhance usability and 
alleviate any possible confusion. These changes are strictly 
editorial have no impact, and do not alter technical content or 
meaning of the specifications. These editorial 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. The licensee's July 14, 1995, letter revised the proposed 
changes in their letter of May 5, 1995, to further limit the period of 
time that continued reactor operation would be allowed with both 
emergency diesel generators inoperable from 24 to two hours. This 
revision to the proposed changes is more restrictive and does not 
impact the licensee's analysis of the criteria of 10 CFR 50.92(c). 
Therefore, the NRC staff proposes to determine that the amendment 
request involves no significant hazards consideration.
    Local Public Document Room location: Auburn Public Library, 118 
15th Street, Auburn, Nebraska 68305
    Attorney for licensee: Mr. John R. McPhail, Nebraska Public Power 
District, Post Office Box 499, Columbus, Nebraska 68602-0499
    NRC Project Director: William D. Beckner

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

    Date of amendment request: August 31, 1995
    Description of amendment request: The proposed amendment modifies 
the definition of HOT SHUTDOWN and COLD SHUTDOWN to specify that the 
definitions are not applicable during the performance of an inservice 
hydrostatic and leak test (IHLT). Technical Specification Section 3.6.B 
and 4.6.B would be modified by adding Section 3.6.B.1.b and 4.6.B.1.b 
to identify the requirements that must be satisfied to consider the 
reactor in COLD SHUTDOWN during the performance of an IHLT. In 
addition, the proposed amendment will change temperature specific 
requirements on several pages to mode or condition specific 
requirements; make several editorial changes; and change the associated 
Bases.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has rovided 
its analysis of the issue of no significant hazards consideration, 
which is presented below:
    NNECO has reviewed the proposed changes in accordance with 
10CFR50.92 and concluded that the changes do not involve a 
significant hazards consideration (SHC). The bases for this 
conclusion is that the three criteria of 10CFR50.92(c) are not 
compromised. The proposed changes do not involve an SHC because the 
changes would not:

[[Page 49941]]

    1. Involve a significant increase in the probability or 
consequences of an accident previously analyzed.
    The proposed changes will allow the reactor to be considered in 
COLD SHUTDOWN during an IHLT with the average reactor coolant 
temperature greater than 212 deg.F but less than 280 deg.F. The 
change to allow the reactor to be in COLD SHUTDOWN during the 
performance of IHLT will not increase the probability or 
consequences of an accident. The probability of a leak in the 
reactor pressure boundary during this testing is not increased by 
considering the reactor to be in COLD SHUTDOWN. The IHLT is 
performed near water solid, all control rods inserted, and with an 
appropriate availability of engineering safety features. The stored 
energy in the reactor core will be very low and the potential for 
failed fuel and a subsequent increase in coolant activity are 
minimal. In addition, secondary containment will be operable and 
capable of handling airborne radioactivity from leaks that could 
occur during the performance of an IHLT. Requiring secondary 
containment to be operable will further ensure that potential 
airborne radiation from leaks will be filtered by one or both trains 
of SBGT [standby gas treatment], thereby limiting releases to the 
environment. Therefore, the changes will not significantly increase 
the consequences of an accident.
    In the unlikely event of a large pressure boundary leak, the 
reactor vessel would rapidly depressurize, allowing one or both of 
the operable core spray systems to operate. Small system leaks would 
be detected by leakage inspections before significant inventory loss 
occurred, since leakage inspections are an integral part of the IHLT 
program.
    Based upon the above, the proposed changes do not involve a 
significant increase in the probability or consequences of an 
accident previously evaluated.
    2. Create the possibility of a new or different kind of accident 
from any previously analyzed.
    The IHLT conditions remain unchanged. The potential for a system 
leak remains unchanged since the reactor coolant system is designed 
for temperatures exceeding 500 deg.F with similar pressures. The 
change in operable engineered safety features available to mitigate 
a postulated accident does not reduce the ability to
    safely mitigate a postulated accident. Adequate ECCS [emergency 
core cooling system] equipment will be available to mitigate a LOCA 
[loss of coolant accident] with an assumed single failure. 
Therefore, this will not create the possibility of a new or 
different kind of accident from any previously evaluated.
    3. Will not involve a significant reduction in a margin of 
safety.
    The proposed changes will not have any significant impact on any 
design basis accident or safety limit. The various engineered safety 
features which are required by the proposed change will ensure 
appropriate mitigation of postulated events. Since the test is 
performed at a near water solid condition and at low decay heat 
values, no fuel damage is expected in case of an accident such as a 
LOCA. Nevertheless, secondary containment and the SBGT system will 
be maintained operable to process air-borne radioactivity from a 
steam leak that could occur during the performance of the IHLT. 
Therefore, the proposed change does not constitute a significant 
reduction in the margin of safety.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: Learning Resources Center, 
Three Rivers Community-Technical College, 574 New London Turnpike, 
Norwich, CT 06360.
    Attorney for licensee: Lillian M. Cuoco, Esq., Senior Nuclear 
Counsel, Northeast Utilities Service Company, P.O. Box 270, Hartford, 
CT 06141-0270.
    NRC Project Director: Phillip F. McKee

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

    Date of amendment request: August 31, 1995
    Description of amendment request: The proposed change to the 
Millstone 2 Technical Specifications would remove the phrase ``other 
than Millstone Unit No. 2'' from Section 6.3.1 on page 6-2.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:
    NNECO has reviewed the proposed change in accordance with 
10CFR50.92 and concluded that the change does not involve a 
significant hazards consideration (SHC). The basis for this 
conclusion is that the three criteria of 10CFR50.92(c) are not 
compromised. The proposed change does not involve an SHC because the 
change would not:
    1. Involve a significant increase in the probability or 
consequences of an accident previously analyzed.
    The proposed change does not affect any system or equipment of 
Millstone Unit No. 2. The proposed change does not affect the 
qualification of any of the licensed individuals involved in the 
day-to-day operation of Millstone Unit No. 2. The proposed change 
corrects a statement which could be interpreted such that an 
individual who once held a Millstone Unit No. 2 SRO [Senior Reactor 
Operator] license would not be eligible to be Operations Manager. 
Since this change does not affect any equipment or operating 
procedures, does not affect the level of expertise and
    training required for on-shift personnel, and does not reduce 
the level of expertise required of operations management, this 
change does not involve a significant increase in the probability or 
consequences of an accident previously analyzed.
    2. Create the possibility of a new or different kind of accident 
from any previously analyzed.
    This change does not affect any equipment or operating 
procedures, does not affect the level of expertise and training 
required for on-shift personnel, and does not reduce the level of 
expertise required of operations management. Therefore, this change 
does not create the possibility of a new or different kind of 
accident.
    3. Involve a significant reduction in the margin of safety.
    This change eliminates a phrase which could be interpreted to 
prevent an individual who had possessed a Millstone Unit No. 2 SRO 
license from becoming the Operations Manager. The training and 
experience necessary to possess a Millstone Unit No. 2 SRO license 
is equivalent to that of other PWRs. Therefore, this proposed change 
does not involve a significant reduction in the margin of safety.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: Learning Resources Center, 
Three Rivers Community-Technical College, 574 New London Turnpike, 
Norwich, CT 06360.
    Attorney for licensee: Lillian M. Cuoco, Esq., Senior Nuclear 
Counsel, Northeast Utilities Service Company, P.O. Box 270, Hartford, 
CT 06141-0270.
    NRC Project Director: Phillip F. McKee

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

    Date of amendment request: July 28, 1995
    Description of amendment request: The proposed amendment would 
eliminate the Technical Specifications requirements to perform 10 CFR 
50, Appendix J, Type C hydrostatic testing on certain valves that are 
within closed systems and are assured a water seal following a Design 
Basis Accident.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:
    1. The proposed TS changes do not involve a significant increase 
in the probability or 

[[Page 49942]]
consequences of an accident previously evaluated.
    The primary containment (drywell and suppression pool) and the 
affected closed systems are accident mitigators not accident 
initiators. The proposed change to the scope of Appendix J, Type C 
testing does not affect the probability of the DBA [Design Basis 
Accident]. The valves will continue to be maintained in an operable 
state, and in their current design configuration. There is no 
correlation between the scope of Appendix J, Type C testing and 
accident probability. There are no physical or operational changes 
to the containment structure, system or components being made as a 
result of the proposed changes. Therefore, the consequences of a 
malfunction of equipment important to safety is not increased from 
those previously evaluated.
    The consequences of loss-of-coolant accidents (LOCAs) under the 
proposed change were considered where a single active failure of a 
containment isolation valve (CIV) or a passive failure of the closed 
system were reviewed, within the limits of the existing licensing 
basis. Under the existing licensing basis, a pipe rupture of the 
seismically qualified ECCS piping does not have to be assumed 
concurrent with the LOCA, except if it is a consequence of the LOCA. 
Consideration of consequential failures can be eliminated, since a 
LOCA inside containment is separated from the affected piping by the 
containment structure. Consideration of consequential failures of 
the ECCS piping from LOCAs outside containment are outside the 
Appendix J design considerations. A single active failure of the 
CIV, under the LOCA condition, can be accommodated since the closed 
and water sealed system piping remains as the leakage barrier. The 
ECCS passive failure criterion does require consideration of system 
leaks, but not pipe breaks, beyond the initiating LOCA. The 
capability to make-up water inventory to the suppression pool is 
adequate to ensure that postulated seat leakage and pipe leakage 
does not result in a condition that jeopardizes pool level. Make-up 
capability exists for the suppression pool via the Condensate 
Storage Tank and Ultimate Heat Sink Spray Pond. Operator actions to 
make-up the suppression pool are delineated in existing Operating 
Procedures.
    The subject valves are single isolation valves associated with 
lines that penetrate the primary containment, but are not connected 
directly to the primary containment atmosphere or the reactor 
coolant pressure boundary. This configuration is described in the 
LGS UFSAR, Section 6.2.4.3.1.3.1, which states ``the systems which 
the lines from the suppression pool connect to outside containment 
are closed systems meeting the appropriate requirements of closed 
systems.'' The integrity of these closed systems are also monitored 
and controlled in accordance with TS Section 6.8.4.a. Any leakage 
that may escape the confines of the closed system will be contained 
within the Reactor Building, treated by standby gas and radwaste 
systems, and, therefore, are within the existing LGS licensing 
bases.
    Finally, the affected penetrations will continue to be subjected 
to the periodic 10 CFR 50, Appendix J, Type A test (Integrated 
Containment Leakage Rate Test).
    The suppression pool level is designed and operated so that 
water level is maintained in accordance with current TS, and the 
associated bases. The supply of water in the suppression pool is 
assured for 30 days during all DBA, post-accident modes of 
operation. The lowest water level which the suppression pool will 
reach was analyzed, and it was determined that the affected lines 
will remain below this minimum level, thereby assuring a water seal. 
The valves will continue to be tested and maintained to ensure their 
operability, and the closed systems' integrity will continue to be 
monitored and controlled in accordance with TS 6.8.4.a and the 
performance of the periodic 10 CFR 50, Appendix J, Type A test.
    Therefore, the proposed changes will not increase the 
probability or consequences of an accident previously evaluated.
    2. The proposed TS changes do not create the possibility of a 
new or different kind of accident from any accident previously 
evaluated.
    The proposed changes do not change the plant response to 
accident scenarios, and do not introduce new or different scenarios. 
The primary containment (drywell and suppression pool) and the 
affected closed systems are accident mitigators not accident 
initiators. The proposed change to the scope of Appendix J, Type C 
hydrostatic testing maintains the existing barriers to primary 
containment bypass leakage by the assurance that a water seal will 
be maintained for 30 days during all DBA, post-accident modes of 
operation. The valves will continue to be tested and maintained to 
ensure their operability, and the closed systems' integrity will 
continue to be monitored and controlled in accordance with TS 
6.8.4.a. Therefore, the proposed changes cannot cause an accident, 
and the plant response to the design basis events is unchanged, 
whereby the change does 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 water seal provided by the assurance of a minimum 
suppression pool level will prevent post-accident containment bypass 
leakage. Appendix J does not require air leak testing of the valves 
since the 30 day post-accident supply of water is maintained. In 
addition, the closed systems' integrity is monitored and controlled 
in accordance with TS 6.8.4.a. Any leakage that may escape the 
confines of the closed system will be contained within the Reactor 
Building, and is within the existing LGS licensing bases. Therefore, 
the proposed TS changes do not involve a significant reduction in a 
margin of safety.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: Pottstown Public Library, 500 
High Street, Pottstown, Pennsylvania 19464.
    Attorney for licensee: J. W. Durham, Sr., Esquire, Sr. V. P. and 
General Counsel, Philadelphia Electric Company, 2301 Market Street, 
Philadelphia, Pennsylvania 19101
    NRC Project Director: John F. Stolz

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

    Date of amendment request: July 28, 1995
    Description of amendment request: The proposed amendments, which 
are consistent with the Improved Standard Technical Specifications 
(NUREG-1433), delete the operability and surveillance requirements 
involving secondary containment differential pressure instrumentation.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:
    1. The proposed Technical Specifications (TS) changes do not 
involve a significant increase in the probability or consequences of 
an accident previously evaluated.
    Deleting the operability and surveillance requirements for the 
secondary containment differential pressure instrumentation does not 
involve any changes to the design, function, or operation of any 
plant components or safety-related systems. There are no changes to 
the separation, redundancy, qualification, quality assurance or fire 
protection requirements for the associated components and systems, 
nor are there any new failure modes created. This activity only 
removes operability and surveillance requirements from the Technical 
Specifications for selected plant components associated with the 
secondary containment differential pressure trip functions. No 
credit for operation of these trip functions is taken in any design 
basis accidents valuated in the SAR [Safety Analysis Report]. 
Thesecomponents will be maintained in accordance with the plant 
preventive maintenance program. The failure of any of these 
components does not result in the occurrence of an accident. 
Consequently, there is no increase in the probability of occurrence 
of an accident previously evaluated in the SAR.
    The Outside Atmosphere to Reactor Enclosure Delta Pressure-Low 
and Outside Atmosphere To Refueling Area Delta Pressure-Low trip 
functions are not symptomatic of a design basis accident. No credit 
for operation of the trip functions is taken in any design basis 
accidents evaluated in the SAR. Neither failure of the differential 
pressure components nor failure to generate the associated trip 
functions affects the consequences of an accident previously 
evaluated in the SAR. The appropriate 

[[Page 49943]]
accident prevention and mitigation actions are generated from other 
plant parameters symptomatic of an accident. Sufficient plant 
parameters symptomatic of a design basis accident are monitored to 
initiate the appropriate actions as evaluated in the SAR. 
Furthermore, all safety-related systems will still be able to 
perform all of their design basis safety-related functions. 
Consequently, there is no increase in the consequences of an 
accident previously evaluated in the SAR.
    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 failure of the differential pressure automatic isolation 
instrumentation components does not result in the occurrence of an 
accident. The failure to generate the associated trip functions does 
not result in the occurrence of an accident. This activity does not 
involve any changes to the design, function, or operation of any 
plant components or safety-related systems. There are no changes to 
the separation, redundancy, qualification, quality assurance or fire 
protection requirements for the associated components and systems. 
These components will be maintained in accordance with the plant 
preventative maintenance program. Consequently, there is no 
possibility of an accident of a different type than previously 
evaluated in the SAR.
    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 ability of secondary containment to minimize any ground 
level release of radioactive material which may result from any 
accident is not affected. Surveillance and operability requirements 
for secondary containment SGTS [Standby Gas Treatment System] and 
RERS [Reactor Enclosure Recirculation System] are not changed by 
this activity. Draw down time, leakage factors, secondary 
containment system ratings, and secondary containment system 
response to a LOCA [Loss of Coolant Accident] or refueling accident 
are not affected by this activity. SGTS and RERS initiation will 
continue to occur when plant parameters symptomatic of a LOCA or 
refueling accident exceed predetermined values. There are no changes 
to the inputs for the post-LOCA offsite dose analysis.
    Therefore, the proposed TS changes do not involve a significant 
reduction in a margin of safety.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: Pottstown Public Library, 500 
High Street, Pottstown, Pennsylvania 19464.
    Attorney for licensee: J. W. Durham, Sr., Esquire, Sr. V. P. and 
General Counsel, Philadelphia Electric Company, 2301 Market Street, 
Philadelphia, Pennsylvania 19101
    NRC Project Director: John F. Stolz

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

    Date of amendment request: July 28, 1995
    Description of amendment request: The proposed amendment would 
modify Technical Specifications (TS) Surveillance Requirements 4.9.1.1, 
4.9.1.2, 4.9.3, 4.9.5, and 4.9.8 to delete specific requirements to 
perform surveillances just prior to beginning or resuming core 
alterations or control rod withdrawal associated with refueling 
activities. This proposed TS change would delete the phrase ``incore 
instrumentation'' from the footnote in TS Section 3/4.9.5, 
``Communications.''
    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 involve any physical changes to 
plant systems or equipment. The proposed TS changes only delete 
those Surveillance Requirements (SRs) pertaining to the performance 
of tests just prior to beginning or resuming core alterations or 
control rod withdrawal, and revises a footnote description to be 
consistent with the current TS definition of ``Core Alteration.'' 
The proposed TS changes do not revise any of the other applicable 
periodic SRs, or modify any procedural controls currently in place 
governing fuel handling operations. The periodic surveillance test 
frequencies provide adequate assurance that the equipment will 
remain in an operable condition. The normal periodic surveillance 
intervals bound those surveillance intervals for the tests that are 
being altered by this proposed TS change. In the event that one of 
the periodic surveillances has not been performed within the 
specified time interval, entry into the specified condition (i.e., 
performance of core alterations, control rod withdrawal, or handling 
of fuel or control rods) is not permitted as required by TS 4.0.4 
until the surveillance has been satisfactorily completed.
    The consequences of an accident are not increased by the 
proposed TS changes, since the changes only involve revising the 
frequency of conducting surveillance tests. The method of operation 
or performance of plant structures, systems, or plant components are 
not affected by the proposed TS changes. The proposed TS changes 
will not impact the operation of any fuel handling equipment, and 
therefore, the potential for a Fuel Handling Accident as described 
in Section 15.7.4 of the LGS [Limerick Generating Station] Updated 
Final Safety Analysis Report (UFSAR) is not increased.
    In addition, any unexpected reduction of water level in the 
reactor cavity or fuel pool at the start of fuel handling or control 
rod handling will be immediately apparent to operators by direct 
observation. Plant procedures utilized by the refueling personnel 
require the suspension of core component transfers in the event of 
loss of water inventory.
    Therefore, the proposed TS change does 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 only involve changes to the frequency in 
which the specified surveillance tests are performed. The proposed 
TS changes do not revise any of the other applicable periodic SRs, 
or modify any procedural controls currently in place governing fuel 
handling operations. The periodic surveillance test frequencies 
provide adequate assurance that the equipment will remain in 
operable condition. The periodic surveillance intervals bound those 
surveillance intervals for the tests that are being altered by this 
proposed TS change. The refueling interlock system combined with 
strict procedural controls provide multiple barriers to preclude an 
inadvertent criticality.
    The proposed TS changes do not involve any physical changes to 
plant systems or equipment. The proposed TS changes do not alter the 
configuration of the plant or the way that the plant is operated. 
The associated plant equipment will continue to function as 
designed. This equipment is not designed to perform any other 
function than it is presently capable of, and therefore, will not 
affect the operation of any other plant equipment.
    Therefore, the proposed TS change does not create the 
possibility of a new or different kind of accident from any 
previously evaluated.
    3. The proposed TS changes do not involve a significant 
reduction in a margin of safety.
    The proposed TS changes do not involve any physical changes to 
plant systems or equipment. The reactor will continue to be 
maintained subcritical during refueling operations and reactor water 
level will be maintained at the required level (i.e., above the 
vessel flange). The proposed TS changes do not affect the operation 
of other plant systems and equipment essential in maintaining 
reactor water temperature during refueling operations, or the 
capability in responding to a postulated Fuel Handling Accident.
    The proposed changes do not adversely affect reliability of the 
refueling interlocks or refuel platform communications equipment. 

[[Page 49944]]
Since the proposed changes only impact the frequency in which certain 
surveillance tests are performed, and do not change the plant 
configuration or setpoints, there is substantial assurance that the 
reactor will be maintained subcritical during refueling.
    Therefore, the proposed TS changes do not involve a reduction in 
a margin of safety.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: Pottstown Public Library, 500 
High Street, Pottstown, Pennsylvania 19464.
    Attorney for licensee: J. W. Durham, Sr., Esquire, Sr. V. P. and 
General Counsel, Philadelphia Electric Company, 2301 Market Street, 
Philadelphia, Pennsylvania 19101
    NRC Project Director: John F. Stolz

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

    Date of amendment request: July 28, 1995
    Description of amendment request: The proposed amendment would 
revise Technical Specifications Table 4.3.1.1-1, ``Reactor Protection 
System Instrumentation Surveillance Requirements'', to reflect changes 
to the surveillance test frequency requirements for various Reactor 
Protection System [RPS] instrumentation.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:
    1. The proposed Technical Specifications (TS) changes do not 
involve a significant increase in the probability or consequences of 
an accident previously evaluated.
    In all of the applicable SAR [Safety Analysis Report] evaluated 
events, the IRM [Intermediate Range Monitor] and APRM [Average Range 
Power Monitor] instrumentation is credited for performing a 
mitigating function (i.e., initiating a scram), to terminate the 
transient prior to a safety limit being exceeded. The proposed TS 
changes do not alter the RPS configuration, or RPS instrumentation 
setpoints, nor do they change the manner in which the IRM and APRM 
instrumentation carry out the scram functions. Therefore the 
consequences of any potential malfunction of equipment important to 
safety will remain unchanged.
    In each case where a startup surveillance test requirement is 
proposed to be deleted, (i.e., IRM and APRM), the normal 
surveillance test frequency specified for the required Operational 
Condition remains unchanged (except for the APRM Upscale Setdown 
functional test). The startup surveillance requirement is 
conservatively bounded by the normal surveillance test interval 
which is greater than or equal to any interval associated with the 
startup surveillance requirement and ensures that the IRM and APRM 
instrumentation reliability is unchanged. This is in accordance with 
the Improved Standard Technical Specifications, NUREG-1433, issued 
September 28, 1992.
    The reliability of the APRM Upscale Setdown scram function will 
not be decreased due to changing the functional test frequency from 
Weekly (W), to Quarterly (Q), in Operational Conditions 2, 3, and 5 
(Startup, Hot Shutdown and Refueling, respectively). Plant 
operational data taken from each of the APRM calibration/functional 
tests performed since August 1992 until present at LGS Units 1 and 
2, shows that setpoint reliability will be maintained if the 
functional test frequency is increased to quarterly, as proposed. 
Presently, each time an APRM calibration/functional test is 
performed, both the Upscale Setdown and the Flow Reference scram 
circuits are tested. The results of the quarterly tests confirm that 
the APRM Upscale Setdown function already has over 2.5 years of 
performance without failure in Operational Condition 1, thus being 
extremely reliable.
    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 affect only the required surveillance 
test intervals, not the RPS configuration or RPS instrumentation 
setpoints. The proposed TS changes do not introduce a new failure 
mode for the IRM or APRM instrumentation. Plant operating experience 
data confirms that at LGS Units 1 and 2, the IRM and APRM 
instrumentation will continue to perform their safety function as 
currently designed, with the same degree of reliability.
    The proposed TS changes do not alter the configuration of the 
plant, nor the way the plant is operated.
    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 following TS Bases were reviewed for potential reduction in 
the margin of safety:
    B 2.2.1 Reactor Protection System Instrumentation Setpoints
     B 3/4.1.4 Control Rod Program Controls
     B 3/4.2 Power Distribution Limits
    B 3/4.3.1 Reactor Protection System Instrumentation
    B 3/4.3.6 Control Rod Block Instrumentation
    The surveillance test frequency changes proposed for the RPS 
instrumentation section of TS do not adversely affect the IRM or 
APRM instrumentation, which will continue to perform the RPS 
functions required to maintain the present margin of safety. Changes 
to the IRM instrumentation startup surveillance intervals are 
already bounded by the existing surveillance requirements, and are 
in accordance with the Improved Standard Technical Specifications, 
NUREG-1433, issued September 28, 1992. The same statement applies to 
the APRM instrumentation, with respect to deletion of the startup 
surveillance requirement. The change of the APRM Upscale Setdown 
Channel functional test surveillance interval from Weekly to 
Quarterly was evaluated to ensure that the APRM instrumentation 
would perform that function, with the same degree of reliability as 
presently experienced. A review of the plant operating experience 
data at LGS Units 1 and 2 shows that APRM instrumentation is 
extremely reliable for a quarterly surveillance test interval. The 
proposed TS changes do not modify plant configuration, RPS 
instrumentation setpoints, or RPS operation. The margin of safety 
remains unchanged.
    Therefore, the proposed TS changes do not involve a reduction in 
a margin of safety.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: Pottstown Public Library, 500 
High Street, Pottstown, Pennsylvania 19464.
    Attorney for licensee: J. W. Durham, Sr., Esquire, Sr. V. P. and 
General Counsel, Philadelphia Electric Company, 2301 Market Street, 
Philadelphia, Pennsylvania 19101
    NRC Project Director: John F. Stolz

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

    Date of amendment request: August 1, 1995
    Description of amendment request: The proposed amendment would 
modify Technical Specifications Section 3/4.9.1, ``Reactor Mode 
Switch,'' in order to provide alternate actions to allow the 
continuation of core alterations in the event certain Reactor Manual 
Control System (RMCS) and refueling interlocks are inoperable, while 
preserving the intended function of the inoperable interlocks.
    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 

[[Page 49945]]
increase in the probability or consequences of an accident previously 
evaluated.
    The refueling and one-rod-out interlocks impose barriers to 
preclude an inadvertent criticality during refueling operations. 
Section 7.7.2.15.1 of the LGS Updated Final Safety Analysis Report 
(UFSAR) clearly delineates the functions of the interlocks and the 
criteria used in assessing correct refueling and one-rod-out 
interlock operation in the following statement.
    In all cases, correct operation of the refueling interlock 
prevents either the operation of loaded refueling equipment over the 
core when any control rod is withdrawn, or the withdrawal of any 
control rod when fuel-loaded refueling equipment is operating over 
the core. In addition, when the reactor mode switch is in REFUEL 
position, only one rod can be withdrawn, and selection of a second 
control rod initiates a rod block.
    The proposed TS changes provide operational flexibility while 
strictly conforming to, and preserving, the intended function of the 
refueling and one-rod-out interlocks. The proposed TS changes that 
could affect interlock capabilities are identified below, along with 
the appropriate justification to substantiate that the proposed TS 
changes will not result in an increase in the probability or 
consequences of an accident previously evaluated.
     a.TS Section 3.9.1, ACTION Statement b. The proposed change to 
this existing TS ACTION will add a verification that all control 
rods are fully inserted, and then disabled from being withdrawn as a 
suitable alternative to placing the reactor mode switch in the 
SHUTDOWN position when the one-rod-out interlock is not operable. In 
addition, the proposed change to this TS section includes a caveat 
of non-applicability for those control rods already removed in 
accordance with requirements stipulated in TS Sections 3.9.10.1 and 
3.9.10.2. As indicated in LGS UFSAR which is described in the 
statement above, it is expected that the refuel and one-rod-out 
interlocks will permit the withdrawal of only one (1) control rod at 
a time with the reactor mode switch in the REFUEL position, and no 
control rods can be moved when fuel-loaded refueling equipment is 
operating over the core. By verifying all control rods are inserted, 
then disabling withdraw capabilities of all rods, as requested, the 
most limiting requirements for control rod motion will be met. The 
potential for having more than one (1) control rod out at a time, or 
having any control rod not fully inserted while fuel-loaded 
refueling equipment is operating over the core, does not exist when 
applying the alternative. Therefore, the intended functions of the 
refuel and one-rod-out interlocks are operationally preserved. Since 
TS Sections 3.9.10.1 and 3.9.10.2 have specific requirements for 
removing surrounding fuel prior to control rod blade removal, the 
control rods already removed are no longer required to carry out a 
safety function in the defueled cell, and as a result would not 
apply for this specific proposed TS change. From a control rod 
withdrawal perspective, there is no functional difference between 
the proposed TS change and the existing, and still remaining, TS 
ACTION of locking the reactor mode switch in SHUTDOWN position.
    b. TS Section 3.9.1, ACTION Statement c. This existing TS ACTION 
requires that core alterations be suspended in the event that a 
refueling interlock is not operable. The proposed TS change to this 
TS ACTION leaves this requirement in place, but makes this ACTION 
specifically applicable to the refueling platform, and adds three 
(3) new additional ACTION alternatives. The wording for changes to 
this TS section are such that implementation of any one of the three 
(3) new alternatives can be substituted for suspending core 
alterations. The proposed wording for these three (3) new 
alternatives and justification is provided below.
    1) Verify all control rods are fully inserted and disable 
withdraw capabilities of all control rods***.
    Since this alternative ensures all control rods are, and will 
remain fully inserted, all required conditions of the associated 
refueling and one-rod-out interlocks are met. The refueling 
interlock is satisfied since a fuel-loaded refueling platform 
operating over the core would be assured that all control rods are 
fully inserted and prevented from being withdrawn. The one-rod-out 
interlock is satisfied since control rod withdrawal is disabled for 
all control rods, which is an even more conservative requirement 
than the one-rod-out interlock itself. While operating in this 
configuration, there will be no associated travel or hoist 
restrictions for the refueling platform over the core, which is 
normal for the current refuel interlock design. The potential for 
having any control rod not fully inserted while a fuel-loaded 
refueling platform is operating over the core, does not exist when 
applying this proposed alternative. Therefore, the intended function 
of the refueling platform refuel interlocks are operationally 
preserved with the implementation of this proposed alternative, and 
there will be no increase in the probability of occurrence of an 
accident. This proposed alternative also maintains an exclusion (via 
a reference to the proposed *** footnote) for control rods removed 
in accordance [with] TS Sections 3.9.10.1 and 3.9.10.2. This 
exclusion does not apply to inadvertent criticality concerns, as 
previously discussed in Item 1.a above.
    2) Verify Refuel Platform is not over core (limit switches not 
reached) and disable refuel platform travel over core.
    As previously stated above, LGS UFSAR Section 7.7.2.15.1 
stipulates that the refueling platform position interlocks initiate 
a control rod block whenever a fuel-loaded refueling platform is 
over the core, and stop a fuel-loaded refueling platform from moving 
over the core if a control rod is already withdrawn. This specific 
proposed TS change satisfies both these requirements by precluding 
the possibility of the platform from being over the core. If a 
control rod is being withdrawn, the platform will not be over the 
core, and the withdrawal will be in accordance with the current 
design. If a control rod is already withdrawn, disabling platform 
travel over the core, before reaching the over-core limit switches, 
is performing the same function as the existing refueling platform 
reverse and forward motion blocks. Therefore, the potential for 
having any control rod not fully inserted while a fuel-loaded 
refueling platform is operating over the core, does not exist when 
applying this proposed alternative. The intended refueling interlock 
functions are operationally preserved with the implementation of 
this proposed alternative.
    3) Verify that no Refuel Platform hoist is loaded and disable 
all Refuel Platform hoists from picking up (grappling) a load.
    As previously stated above, UFSAR Section 7.7.2.15.1 stipulates 
that blocking control rod withdrawal with a refueling platform over 
the core, and restricting refueling platform travel from going over 
the core with a control rod already withdrawn, are based on the 
refueling platform hoist being fuel-loaded. An unloaded platform 
without grappling capabilities poses no threat to erroneous fuel 
bundle or control rod removal, and eliminates the potential for 
having any control rod not fully inserted while a fuel-loaded 
refueling platform is operating over the core. Therefore, 
implementing this proposed alternative operationally preserves the 
intended interlock functions.
    c. TS Section 3.9.1, ACTION Statement d. The proposed TS change 
adds this new TS ACTION section to specify the refueling interlock 
requirements for the service platform, since the applicability of 
ACTION Statement c above is being revised to specifically address 
refueling interlocks associated with the refueling platform. The 
proposed TS changes for new this TS section retain the existing 
requirement to suspend core alterations if the service platform 
associated refueling interlock is not operable, unless the service 
platform is not installed over vessel. The specific proposed TS 
changes add two (2) new additional ACTION alternatives. The proposed 
wording for these two (2) new ACTION statements are such that 
implementation of any one of the two (2) new alternatives can be 
substituted for suspending core alterations. Not enforcing 
operability requirements on the service platform refueling 
interlocks when the service platform is not over the vessel does not 
pose an inadvertent criticality concern since there is no associated 
hoist to manipulate fuel bundles or control rods. These two (2) new 
alternatives are not applicable unless the service platform is 
installed over the vessel, and are described below.
    1) Verify all control rods are fully inserted and disable 
withdraw capabilities of all control rods***.
    This alternative ensures that all control rods are, and will 
remain, fully inserted which meets the required conditions for 
proper refueling and one-rod-out interlock operation. The refueling 
interlock is satisfied since a fuel-loaded service platform hoist 
operating over-core is assured that all control rods are fully 
inserted and prevented from being withdrawn. The one-rod-out 
interlock is satisfied since all control rods are disabled, an even 
more conservative requirement than the one-rod-out interlock itself. 
While operating in this configuration, there will be no associated 
hoist restrictions for the service 

[[Page 49946]]
platform, which is normal for the current refuel interlock design. The 
potential for having any control rod not fully inserted while a 
service platform hoist is fuel-loaded over the core, does not exist 
when utilizing this proposed alternative. Therefore, the intended 
function of the service platform refuel interlocks are operationally 
preserved with the implementation of this proposed alternative. This 
proposed alternative also maintains an exclusion (via a reference to 
the proposed *** footnote) for control rods removed in accordance 
with the requirements of TS Sections 3.9.10.1 and 3.9.10.2. This 
exclusion is not applicable to inadvertent criticality concerns as 
discussed in Item 1.a above.
    2) Verify Service Platform hoist is not loaded and disable 
Service Platform hoist from picking up (grappling) a load.
    As previously described above, UFSAR Section 7.7.2.15.1 
stipulates that blocking control rod withdrawal with the service 
platform over the core is based on the service platform hoist being 
fuel-loaded. An unloaded hoist without grappling capabilities poses 
no threat to erroneous fuel bundle or control rod removal and 
eliminates the potential for having any control rod not fully 
inserted while a fuel-loaded service platform is operating over the 
core. Therefore, implementing this proposed alternative 
operationally preserves the intended refueling interlock functions.
    As discussed in the LGS UFSAR, the use of the refueling and one-
rod-out interlocks are evaluated from a prevention, not a 
mitigation, perspective. A Rod Withdrawal Error (RWE) transient 
event during refueling is concerned with an inadvertent criticality, 
and assumes the reactor vessel head is off, and the plant is 
shutdown (i.e, Operating State A). As described in the LGS UFSAR 
under Nuclear Safety Operational Analysis (NSOA) Event 16, it is 
assumed that the Reactor Protection System (RPS) terminates the 
event should the reactor actually reach Operating State B (i.e., 
head off and not shut down), which is conditional on the reactor 
mode switch being in the STARTUP position. The proposed TS changes 
only pertain to the refueling and one-rod-out interlocks. Since 
these interlocks act only in a preventive mode, the consequences of 
an inadvertent criticality accident during refueling remain 
unchanged.
    Since the proposed TS changes are limited to the one-rod-out and 
refueling interlocks, they do not affect the reliability of the 
associated equipment. The proposed TS changes specify alternative 
actions that can be taken in the event that an interlock is 
inoperable. These alternative actions serve to ensure the failed 
interlock function is preserved, and do not affect the probability 
of malfunction of the interlocks.
    The one-rod-out and refueling interlocks, as evaluated in the 
LGS UFSAR, are designed to preclude an inadvertent criticality 
during refueling operations by placing strict controls on fuel 
bundle and control rod manipulations, using the following methods.
    a. Preventing operation of a fuel-loaded refueling platform or 
service platform hoist while over the core if a control rod is 
already withdrawn.
    b. Preventing a fuel-loaded refueling platform from traveling 
over the core if a control rod is already withdrawn.
    c. Preventing any control rod from being withdrawn if a fuel-
loaded refueling platform or service platform is already operating 
over the core.
    d. Preventing the withdrawal of more than one control rod at a 
time with the reactor mode switch in the REFUEL position.
    The LGS UFSAR indicates that a single component failure does not 
cause an interlock failure and that a single interlock failure does 
not cause an accident. The proposed TS changes provide alternative 
actions that can be taken in the event of an associated component or 
interlock malfunction. Implementing the proposed TS changes will 
continue to ensure that the intended interlock functions are 
maintained and operationally preserved, as described in the LGS 
UFSAR.
    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 only pertain to the refueling and one-
rod-out interlocks. The refueling and one-rod-out interlocks impose 
barriers to preclude an inadvertent criticality during refueling 
operations. The proposed TS changes provide operational flexibility, 
while strictly conforming to, and preserving, the intended function 
of the refueling and one-rod-out interlocks. There is no other 
potential failure mode for these interlocks than has already been 
evaluated and described in the LGS UFSAR. Implementation of these 
proposed changes will maintain and operationally preserve the 
intended interlock functions. Therefore, the malfunction of any 
associated component or interlock will not adversely impact the 
plant and any other equipment important to safety, directly or 
indirectly.
    Therefore, the proposed TS changes do not create the possibility 
of a new or different kind of accident from any previously 
evaluated.
    3. The proposed TS changes do not involve a significant 
reduction in a margin of safety.
    The proposed TS changes only affect the TS associated with the 
one-rod-out and refueling interlocks. The associated TS Bases 
Section 3/4.9, ``Refueling Operations,'' states that the one-rod-out 
and refueling interlocks maintain conditions during refueling 
activities that reinforce refueling procedures and reduce the 
potential for the probability of occurrence of each of the following 
conditions:
    a. Inadvertent criticality,
    b. Damage to reactor internals or fuel assemblies, and
    c. Exposure of personnel to excessive radioactivity.
    The proposed TS changes do not adversely affect the one-rod-out 
or refueling interlocks. The associated interlocks will continue to 
perform the refueling functions required to maintain the present 
margin of safety. The proposed TS changes only contain alternative 
actions that can be taken in the event an interlock is inoperable. 
These proposed alternative actions ensure that the intent of the 
interlocks is preserved, and that there is no reduction in the 
ability of the interlocks to maintain adequate refueling conditions.
    The proposed TS changes will preserve the intended interlock 
functions, and maintain the existing level of protection against 
refueling errors that could lead to an inadvertent criticality, 
damage to reactor internals or fuel assemblies, or excessive 
personnel radiation exposure. The one-rod-out and refueling 
interlocks will continue to function with their present degree of 
reliability. The proposed TS changes will continue to maintain 
strict controls on fuel bundle and control rod manipulations to 
avoid inadvertent criticality. The proposed TS changes provide the 
same level of assurance regar[d]ing the manipulation of control rods 
during refueling operations as that currently described in the LGS 
UFSAR, and as discussed below.
    a. Preventing operation of a fuel-loaded refueling platform or 
service platform hoist while over the core if a control rod is 
already withdrawn.
    b. Preventing a fuel-loaded refueling platform from traveling 
over the core if a control rod is already withdrawn.
    c. Preventing any control rod from being withdrawn if fuel-
loaded refueling platform or service platform is already operating 
over the core.
    d. Preventing the withdrawal of more than one control rod at a 
time with the reactor mode switch in the REFUEL position.
    Therefore, the proposed TS changes do not involve a reduction in 
a margin of safety.
    The NRC staff has reviewed the licensee's analysis and, based on 
this eview, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: Pottstown Public Library, 500 
High Street, Pottstown, Pennsylvania 19464.
    Attorney for licensee: J. W. Durham, Sr., Esquire, Sr. V. P. and 
General Counsel, Philadelphia Electric Company, 2301 Market Street, 
Philadelphia, Pennsylvania 19101
    NRC Project Director: John F. Stolz

South Carolina Electric & Gas Company (SCE&G), South Carolina 
Public Service Authority, Docket No. 50-395, Virgil C. Summer 
Nuclear Station, Unit No. 1, Fairfield County, South Carolina

    Date of amendment request: February 21, 1995, as revised on August 
31, 1995
    Description of amendment request: The proposed amendment would 
revise the Technical Specifications (TS) to reflect changes to 10 CFR 
Part 20 (including Appendix B, Table 2 concentrations) and provide 
additional administrative corrections.

[[Page 49947]]

    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 probability or consequences of an accident previously 
evaluated does not involve a significant increase.
    The proposed TS changes showing the relocation of the old 10 CFR 
20.106 requirements to the new 10 CFR 20.1302, the old 10 CFR 
20.203(c)(2) requirements to the new 10 CFR 20.1601(a), and the old 
10 CFR 20.407 requirements to the new 10 CFR 20.2206(b) will not 
involve a significant increase in the probability or consequences of 
an accident previously evaluated because there will be no change in 
the types and amounts of effluents that will be released, nor will 
there be an increase in individual or cumulative occupational 
radiation exposures.
    The proposed revision to the liquid and gaseous release rate 
limits will not involve a significant increase in the probability or 
consequences of an accident previously evaluated because there will 
be no change in the types and amounts of effluents that will be 
released, nor will there be an increase in individual or cumulative 
occupational radiation exposures. This is only a change to the 
method of (algorithm) determining release rate limits and will not 
change net limits or change the more restrictive 10 CFR 50 Appendix 
I dose limits.
    The proposed revision to the radioactive material quantity in 
the settling pond and its associated TS Bases will not involve a 
significant increase in the probability or consequences of an 
accident previously evaluated because there will be no change in the 
types of effluents that will be released, nor will there be an 
increase in individual or cumulative occupational radiation 
exposures. This is only a change to the quantity of radioactive 
material in the settling pond and will conservatively lower net 
limits.
    The proposed revision to the TS bases for the liquid holdup tank 
activity limit will not involve a significant increase in the 
probability or consequences of an accident previously evaluated 
because there will be no change in the types and amounts of 
effluents that will be released, nor will there be an increase in 
individual or cumulative occupational radiation exposures. The curie 
limit is not affected, therefore, the change does not represent a 
decrease in the level of control previously evaluated.
    The proposed revision to the distance at which dose rates are 
measured from the radiation source or surface will not involve a 
significant increase in the probability or consequences of an 
accident previously evaluated because there will be no increase in 
the individual or cumulative occupational radiation exposures. The 
change in distance is conservative in its effect on worker 
protection and is in conformance with 10 CFR 20.1601 requirements.
    2. The possibility of a new or different kind of accident from 
any previously evaluated is not created.
    The proposed TS changes showing the relocation of the old 10 CFR 
20.106 requirements to the new 10 CFR 20.1302, relocation of the old 
10 CFR 20.203(c)(2) requirements to the new 10 CFR 20.1601(a), and 
relocation of the old 10 CFR 20.407 requirements to the new 10 CFR 
20.2206(b) will not create the possibility of a new or different 
kind of accident from any previously evaluated because the revisions 
are administrative and will not change the types and amounts of 
effluents that will be released.
    The proposed revision to the liquid and gaseous release rate 
limits will not create the possibility of a new or different kind of 
accident from any previously evaluated because the revision is 
administrative and will not change the types and amounts of 
effluents that will be released.
    The proposed revision to the quantity of radioactive material in 
the settling pond and its associated TS Bases will not create the 
possibility of a new or different kind of accident from any 
previously evaluated because the revision will not change the types 
of effluents that will be released. This is only a change to the 
quantity of radioactive material in the settling pond and will 
conservatively lower net limits.
    The proposed revision to the TS bases for the liquid holdup tank 
activity limit will not create the possibility of a new or different 
kind of accident from any previously evaluated because the revision 
is administrative and will not change the types and amounts of 
effluents that will be released.
    Implementation of the more conservative distance at which dose 
rates are measured will not create the possibility of a new or 
different kind of accident from any previously evaluated.
    3. A significant reduction in a margin of safety is not 
involved.
    The proposed revisions due to the location of requirements will 
not reduce a margin of safety because they are administrative in 
nature. No equipment or procedural changes are postulated. There is 
no impact on any margin of safety.
    The proposed revision to liquid and gaseous release rate limits 
will not reduce a margin of safety because it is administrative in 
nature. These revisions preserve the existing level of effluent 
control. No changes to the more restrictive 10 CFR 50 Appendix I 
dose limits are made. There are no equipment or operational 
procedure changes, therefore, no accidents of any kind will be 
created by this change.
    The proposed revision to the quantity of radioactive material in 
the settling pond and its associated TS Bases will not reduce a 
margin of safety because it is conservative in nature and preserves 
the existing level of effluent control. There are no equipment or 
operational procedure changes required, therefore, no accidents of 
any kind will be created by this change.
    The proposed revision to the TS bases for the liquid holdup tank 
activity limit will not reduce a margin of safety because it is 
administrative in nature and preserve[s] the existing level of 
effluent control. No equipment or procedural changes are postulated. 
There is no impact on any margin of safety.
    The change in distance for a High Radiation Area classification 
from 18 in.(45 cm) to (30 cm)12 in. from the radiation source or 
surface will not reduce the margin of safety because this change 
will reduce the worker's stay time in the area and therefore 
minimize exposure.
    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: Fairfield County Library, 300 
Washington Street, Winnsboro, SC 29180
    Attorney for licensee: Randolph R. Mahan, South Carolina Electric & 
Gas Company, Post Office Box 764, Columbia, South Carolina 29218
    NRC Project Director: Frederick J. Hebdon

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 amendment requests: July 19, 1995
    Description of amendment requests: The licensee proposes to revise 
technical specifications (TSs) to (1) support modifications to the 
containment area radiation monitors, to either upgrade or replace 
existing equipment with state-of-the-art equipment, (2) relocate the 
setpoint and allowable values for the control room airborne radiation 
monitors to be consistent with the containment airborne radiation 
monitors TS, and (3) make minor editorial changes to the TS pertaining 
to the control room airborne radiation monitors and the containment 
airborne radiation monitors. The proposed changes affect TS Tables 3.3-
3, 3.3-4, 3.3-5, 3.3-6, 4.3-2, and 4.3-3.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:
    1. The proposed change does not involve a significant increase 
in the probability or consequences of an accident previously 
evaluated.
    Control Room Airborne Radiation Monitors
    The proposed change would permit relocation of the setpoint and 
allowable values for the monitors from the Technical Specifications 
(TSs) to the administrative control procedures. This change is 
consistent with the existing Containment Airborne Radiation Monitor 
TSs. This change will not prevent the radiation monitors from 

[[Page 49948]]
performing their intended function following a design basis accident. 
Therefore, operation of the facility in accordance with this change 
will not involve a significant increase in the probability or 
consequences of an accident previously evaluated.
    Containment Area Radiation Monitors
    The proposed change deletes the existing Containment Area 
Radiation Monitors RE-7856-1 and RE-7857-2 and their Engineered 
Safety Feature Actuation System (ESFAS) function to initiate 
containment purge isolation on high radiation in containment. The 
deletion of this ESFAS function does not create a precursor to any 
analyzed accident since these monitors are for accident mitigation 
only.
    Currently, no release of radioactivity is assumed during a Fuel 
Handling Accident in containment since the Containment Area 
Radiation Monitors detect and isolate containment purge prior to 
release. The proposed deletion will cause some release prior to 
detection and isolation of purge by the remaining noble gas 
containment monitors. The consequences of a Fuel Handling Accident 
inside containment were previously re-evaluated, assuming no 
containment purge isolation, to resolve inconsistencies in the 
original analysis assumptions and methodology. The results of the 
calculation indicated off-site doses well within the limits of 10 
CFR 100 and Control Room doses that met the limits of 10 CFR 50 
Appendix A General Design Criterion 19. Containment purge isolation 
on high gaseous activity during a Fuel Handling Accident will still 
be available with this proposed change but is not required for the 
dose consequences to remain within the dose criteria. Therefore, the 
proposed change will not significantly increase the consequences of 
a Fuel Handling Accident inside containment.
    The Loss of Coolant Accident (LOCA) function of the Containment 
Purge Isolation System (CPIS) signal will be essentially unaffected 
by this proposed change. Currently, containment purge isolation 
(containment minipurge) on high radiation signals is a diverse 
signal with Safety Injection Actuation System (SIAS) and Containment 
Isolation Actuation System (CIAS). In a LOCA event, containment 
purge isolation is expected to occur on either SIAS or CIAS prior to 
a CPIS signal on high radiation in containment. While this proposed 
change reduces the diversity of radiation monitoring inputs, the 
diversity of parameters measured (pressure and radiation) is still 
preserved. Therefore, the proposed change will not increase the 
consequences of a LOCA.
    2. The proposed change does not create the possibility of a new 
or different kind of accident from any accident previously 
evaluated.
    Control Room Airborne Radiation Monitors
    Relocating the monitor setpoint and allowable values from the 
TSs to the administrative procedures would not alter the design and 
operational interface between the Control Room Isolation System 
instrumentation and existing plant equipment. As such, the monitors 
would continue to operate and perform their intended safety function 
to isolate the control room following a design basis accident as 
before. Therefore, operation of the facility in accordance with this 
proposed change will not create the possibility of a new or 
different kind of accident from any accident previously evaluated.
    Containment Area Radiation Monitors
    The deletion of the Containment Area Radiation Monitors will not 
alter the operation of CPIS. The remaining interface between CPIS 
and existing plant equipment will continue to perform their intended 
safety function to isolate containment purge by closing the 
containment purge valves. This function will continue to be 
performed by Containment Airborne Radiation Monitors 2(3) RT-7804-1 
and 2(3) RT-7807-2. Therefore, operation of the facility in 
accordance with this proposed change will not create the possibility 
of a new or different kind of accident from any accident previously 
evaluated.
    3. The proposed change does not involve a significant reduction 
in a margin of safety.
    Control Room Airborne Radiation Monitors
    Relocating the monitor setpoint and allowable values to the 
administrative procedures would not alter the existing margin of 
safety. The relocation would only relinquish control of the setpoint 
and allowable values from the TSs to quality-affecting (changes will 
require a 10 CFR 50.59 evaluation) procedures. Therefore, operation 
of the facility will not involve a significant reduction in a margin 
of safety.
    Containment Area Radiation Monitors
    The proposed change does not affect the margin of safety in 
Modes 1 through 4 since the diversity of the parameters measured is 
maintained for minipurge isolation. Either SIAS, CIAS, CPIS, or 
manual operation will close the containment mini purge valves. The 
main purge is sealed closed during Modes 1 through 4 with the purge 
valves closed and deactivated.
    The diversity of the parameters measured is not maintained for 
the containment main purge isolation. The main purge is only 
applicable during Modes 5 and 6 and main purge isolation is 
initiated only by either CPIS or manual operation. This proposed 
change along with the previously submitted PCN-299 reduces the 
diversity of radiation sensing in containment for CPIS generation 
from four types (gaseous, iodine, particulate, and gamma) to one 
type (gaseous activity). Since the consequences of a Fuel Handling 
Accident inside containment without purge isolation have been 
calculated to be well within 10 CFR 100 dose limits, the loss of 
diversity for this accident does not result in a significant 
reduction in a margin of safety. Therefore, this proposed change 
will not involve a significant reduction in a margin of safety.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment requests involve no significant hazards consideration.
    Local Public Document Room location: Main Library, University of 
California, P. O. Box 19557, Irvine, California 92713
    Attorney for licensee: T. E. Oubre, Esquire, Southern California 
Edison Company, P. O. Box 800, Rosemead, California 91770
    NRC Project Director: William H. Bateman

Tennessee Valley Authority, Docket No. 50-328, Sequoyah Nuclear 
Plant, Unit 2, Hamilton County, Tennessee

    Date of amendment request: May 19, 1995; revised September 11, 1995 
(TS 95-13)
    Description of amendment request: The proposed change would revise 
License Condition 2.C.(17) to extend the required surveillance interval 
to May 18, 1996, for Surveillance Requirement 4.3.2.1.3. The proposed 
change would extend the Engineered Safety Features Response Time 
instrument tests required at 36-month intervals shown in Table 3.3-3 
associated with safety injection, feedwater isolation, containment 
isolation Phase A, auxiliary feedwater pump, essential raw cooling 
water system, emergency gas treatment system, containment spray, 
containment isolation Phase B, turbine trip, 6.9-kilovolt shutdown 
board-degraded voltage or loss of voltage, and automatic switchover to 
containment sump actuations. The proposed extension will limit the 
interval past the allowable extension provided by TS 4.0.2 to 5 months.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:
    TVA has evaluated the proposed technical specification (TS) 
change and has determined that it does not represent a significant 
hazards consideration based on criteria established in 10 CFR 
50.92(c). Operation of Sequoyah Nuclear Plant (SQN) in accordance 
with the proposed amendment will not:
    1. Involve a significant increase in the probability or 
consequences of an accident previously evaluated.
    The proposed change is temporary and allows a one-time extension 
of Surveillance Requirement 4.3.2.1.3 for Cycle 7 to allow 
surveillance testing to coincide with the seventh refueling outage. 
The proposed surveillance interval extension will not cause a 
significant reduction in system reliability nor affect the ability 
of the systems to perform their design function. Current monitoring 
of plant conditions and continuation of the surveillance testing 
required during normal plant operation will continue to be performed 
to ensure conformance with TS operability requirements. Therefore, 
this change does not involve a significant increase in the 

[[Page 49949]]
probability or consequences of an accident previously evaluated.
    2. Create the possibility of a new or different kind of accident 
from any previously analyzed.
    Extending the surveillance interval for the performance of 
specific testing will not create the possibility of a new or 
different kind of accidents. No changes are required to any system 
configurations, plant equipment, or analyses. Therefore, this change 
will not create the possibility of a new or different kind of 
accident from any previously evaluated.
    3. Involve a significant reduction in a margin of safety.
    Surveillance interval extensions will not impact any plant 
safety analyses since the assumptions used will remain unchanged. 
The safety limits assumed in the accident analyses and the design 
function of the equipment required to mitigate the consequences of 
any postulated accidents will not be changed since only the 
surveillance test interval is being extended. Historical performance 
generally indicates a high degree of reliability, and surveillance 
testing performed during normal plant operation will continue to be 
performed to verify proper performance. Therefore, the plant will be 
maintained within the analyzed limits, and the proposed extension 
will not significantly reduce the margin of saety.
    The NRC has reviewed the licensee's analysis and, based on 
thisreview, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: Chattanooga-Hamilton County 
Library, 1101 Broad Street, Chattanooga, Tennessee 37402
    Attorney for licensee: General Counsel, Tennessee Valley Authority, 
400 West Summit Hill Drive, ET 11H, Knoxville, Tennessee 37902
    NRC Project Director: Frederick J. Hebdon

Virginia Electric and Power Company, Docket Nos. 50-338 and 50-339, 
North Anna Power Station, Units No. 1 and No. 2, Louisa County, 
Virginia

    Date of amendment request: September 1, 1995
    Description of amendment request: The proposed changes to the 
Technical Specifications (TS) for the North Anna Power Station, Units 
1&2 (NA-1&2) would allow a single outage of up to 14 days for each 
emergency diesel generator (EDG) once every 18 months. The purpose of 
the outage is the performance of a preventive maintenance inspection, 
appropriate for diesels used for this class of standby service, which 
requires disassembly of the EDG. Currently this maintenance inspection 
is performed during refueling outages. The proposed changes would 
permit this maintenance inspection to be performed during Modes 1 to 4 
in addition to the current allowance during Modes 5 or 6.
    A probabilistic safety analysis (PSA) has been performed which 
demonstrates that a fourteen (14) day maintenance inspection outage, 
once every eighteen (18) months for each EDG, results in no significant 
change in core damage frequency assuming adequate compensatory measures 
are in place. The compensatory measures include requirements that the 
other EDGs, off-site power supply, and the alternate A.C. diesel (AAC 
DG) be operable during the preventive maintenance inspection outage.
    The effect of the proposed change has been calculated to be an 
increase in core damage frequency of approximately 1E-6 per year, which 
is not considered to be a significant change (i.e., an acceptable 
change in risk, or a non-risk significant change) from the baseline 
core damage frequency of 4.1E-5.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:
    Specifically, operation of North Anna Power Station in 
accordance with the [proposed] Technical Specifications changes will 
not:
    a. involve a significant increase in the probability or 
consequences of an accident previously evaluated. The probabilistic 
safety analysis (PSA) demonstrates that the increase in core damage 
frequency due to performing the EDG maintenance inspection over a 
fourteen day period once every 18 months is not significant as long 
as the AAC DG is operable to act as a source of emergency power to 
replace the EDG. The period of time during which the EDG is 
unavailable is short enough to limit the impact of using the 
manually operated AAC DG as a replacement for the automatically 
operated EDG.
    b. create the possibility of a new or different kind of accident 
from any accident previously evaluated. The proposed Technical 
Specifications changes only modify the operability of an EDG for a 
limited and defined period of time. The UFSAR [Updated Final Safety 
Analysis Report] accidents are analyzed assuming that the EDG is the 
worst single failure. This assumption is more severe than the 
proposed Technical Specifications changes which replaces the EDG 
with the AAC DG. Similarly, the PSA performed to evaluate the 
proposed Technical Specifications changes considered all of the 
initiating events defined for the PSA performed for the Individual 
Plant Examination. No new initiators were defined as a result of a 
review of the PSA model. Therefore, it is concluded that no new or 
different kind of accident from any previously evaluated has been 
created.
    c. The proposed Technical Specifications changes do not result 
in a reduction in margin of safety as defined in the basis for any 
Technical Specifications. The PSA was performed to evaluate the 
concept of a one time outage. The results of the analyses show no 
significant change in the core damage frequency. As described above 
the proposed Technical Specifications changes only modify the 
operability of an EDG for a limited and defined period of time. 
Thus, operation with slightly increased EDG unavailability due to 
maintenance, and the AAC DG operable is acceptable.
    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: The Alderman Library, Special 
Collections Department, University of Virginia, Charlottesville, 
Virginia 22903-2498.
    Attorney for licensee: Michael W. Maupin, Esq., Hunton and 
Williams, Riverfront Plaza, East Tower, 951 E. Byrd Street, Richmond, 
Virginia 23219.
    NRC Project Director: David B. Matthews

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 August 16, 1995.
    Description of amendment request: This request proposes to revise 
Technical Specification 1.7, ``Containment Integrity,'' Technical 
Specification 3/4.6.1, ``Containment Integrity,'' Technical 
Specification 3/4.6.3, ``Containment Isolation Valves,'' and their 
associated Bases. These proposed changes will remove Technical 
Specification Table 3.6-1 ``Containment Isolation Valves,'' to Wolf 
Creek Generating Station (WCGS) procedures. This proposed change is in 
accordance with the guidance provided in Generic Letter 91-08, 
``Removal of Component Lists from Technical Specifications,'' dated May 
6, 1991. In addition, this request proposes to add a footnote to 
Technical Specification 3.6.3 extending the allowed outage time for the 
component cooling water (CCW) system reactor coolant pump seal water 
supply and return valves. This determination supersedes the staff's 
proposed no significant hazards consideration determination evaluation 
for the requested changes that was published on April 26, 1995 (60 FR 
20532).

[[Page 49950]]

    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:
    1. The proposed change does not involve a significant increase 
in the probability or consequences of an accident previously 
evaluated.
    The proposed changes simplify the technical specifications, meet 
the regulatory requirements for control of containment isolation, 
and are consistent with the guidelines of GL 91-08. The procedural 
details of Technical Specification Table 3.6-1 have not been 
changed, but only relocated to a different controlling document. The 
proposed changes are administrative in nature, should result in 
improved administrative practices, and do not affect plant 
operations. The addition of the footnote to allow up to 12 hours for 
valve testing the CCW MOVs [motor-operated valves] does not affect 
the severity of any accident previously evaluated. This footnote 
does not impact plant safety since the second isolation device in 
the affected penetrations would still be available to provide 
isolation between the RCS and the outside atmosphere.
    The probability of occurrence of a previously evaluated accident 
is not increased because this change does not introduce any new 
potential accident initiating conditions. The consequences of an 
accident previously evaluated is not increased because the ability 
of containment to restrict the release of any fission product 
radioactivity to the environment will not be degraded by this 
change.
    2. The proposed change does not create the possibility of a new 
or different kind of accident from any accident previously 
evaluated.
    The proposed changes are administrative in nature, do not result 
in physical alterations or changes to the operation of the plant, 
and cause no change in the method by which any safety-related system 
performs its function. The addition of the footnote to allow up to 
12 hours for valve testing the CCW MOVs does not affect the severity 
of any accident previously evaluated. The additional time provides 
assurance that the inoperable valve is in proper working order prior 
to returning it to OPERABLE status. Therefore, this proposed change 
will not create the possibility of a new or different kind of 
accident from any previously evaluated.
    3. The proposed change does not involve a significant reduction 
in a margin of safety.
    The administrative change to relocate Technical Specification 
Table 3.6-1 to appropriate plant procedures does not alter the basic 
regulatory requirements for containment isolation and will not 
adversely affect containment isolation capability for credible 
accident scenarios. Adequate control of the content of the table is 
assured by existing plant procedures. The additional footnote to 
extend the allowed outage time to 12 hours for the CCW MOVs does not 
affect containment isolation capability since the second isolation 
device in the affected penetrations would still be available to 
provide isolation between the RCS and the outside atmosphere, and to 
ensure that a release of radioactive material to the environment 
following an accident will not exceed the assumptions used in the 
LOCA Analyses.
    The proposed relocation of the Technical Specification Table 
3.6-1 does not alter the requirements for containment isolation 
valve operability currently in the technical specifications. The LCO 
and Surveillance Requirements would be retained in the revised 
technical specifications. Therefore, the proposed change will not 
affect the meaning, application, and function of the current 
technical specification requirements for the valves in Table 3.6-1.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room locations: Emporia State University, 
William Allen White Library, 1200 Commercial Street, Emporia, Kansas 
66801 and Washburn University School of Law Library, Topeka, Kansas 
66621
    Attorney for licensee: Jay Silberg, Esq., Shaw, Pittman, Potts and 
Trowbridge, 2300 N Street, N.W., Washington, D.C. 20037
    NRC Project Director: William H. Bateman

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

    Date of amendment request: August 22, 1995
    Description of amendment request: The proposed license amendment 
request would relocate Technical Specification Tables 3.3-2, ``Reactor 
Trip System Instrumentation Response Times,'' and 3.3-5, ``Engineered 
Safety Features Response Times,'' and applicable Bases discussions, to 
Updated Safety Analysis Report (USAR) Chapter 16. The NRC has already 
implemented this line-item technical specification improvement in the 
new Standard Technical Specifications (NUREG-1431 for Westinghouse 
plants). This amendment request follows the guidance provided by the 
NRC in Generic Letter 93-08, ``Relocation of Technical Specification 
Tables of Instrument Response Time Limits,'' for relocating instrument 
response time tables.
    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.
    This license amendment request does not change any Reactor Trip 
System (RTS) or Engineered Safety Features Actuation System (ESFAS) 
instrument response times or surveillance intervals currently 
prescribed in Technical Specification Tables 3.3-2 and 3.3-5. The 
RTS and ESFAS will continue to function in a manner consistent with 
the assumptions in the Updated Safety Analysis Report Chapter 15 
accident analyses and the plant design basis. Therefore, overall 
protection system performance will remain within the bounds of the 
accident analyses documented in USAR Chapter 15. As such, there will 
be no degradation in system performance, nor will there be an 
increase in the number of challenges to equipment assumed to 
function during an accident situation.
    The proposed technical specification revision does not involve 
any hardware changes or changes to any instrumentation setpoints, 
system operating parameters, or system accident mitigation 
capabilities, nor do the changes affect the probability of any event 
initiators. Thus, the proposed change will not result in an increase 
in the consequences of or the probability of occurrence of any 
accident or safety-related equipment malfunction.
    2. The proposed change does not create the possibility of a new 
or different kind of accident from any accident previously 
evaluated.
    As discussed above, there are no hardware changes associated 
with this proposed amendment request, nor are there any changes in 
the method by which any safety-related plant system performs its 
safety function. The normal manner of plant operation is not 
affected by this proposed change.
    No new accident scenarios, transient precursors, failure 
mechanisms, or limiting single failures are introduced as a result 
of the proposed changes. There will be no adverse effect or 
challenges imposed on any safety-related system as a result of these 
changes. Therefore, the possibility of a new or different kind of 
accident is not created by the proposed changes.
    3. The proposed change does not involve a significant reduction 
in a margin of safety.
    No response times will be changed by this amendment request. The 
proposed request only changes the document where the response times 
will be listed. This proposed amendment request will not affect the 
manner in which safety limits or limiting safety system settings are 
determined, nor will there by [be] any effect on plant systems 
necessary to assure the accomplishment of protection functions. The 
proposed change will not impact any margin of safety defined in the 
basis for any Technical Specification.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 


[[Page 49951]]
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room locations: Emporia State University, 
William Allen White Library, 1200 Commercial Street, Emporia, Kansas 
66801 and Washburn University School of Law Library, Topeka, Kansas 
66621
    Attorney for licensee: Jay Silberg, Esq., Shaw, Pittman, Potts and 
Trowbridge, 2300 N Street, N.W., Washington, D.C. 20037
    NRC Project Director: William H. Bateman

Previously Published Notices Of Nonsideration Of Issuance Of 
Amendments To Facility Operating Licenses, Proposed No Significant 
Hazards Consideration Determination, and opportunity for a hearing

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

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

    Date of application for amendment: August 4, 1995 (AEP:NRC:1129E)
    Description of amendment request: The proposed amendment would 
modify Technical Specification 4.4.5.4 and 4.4.5.5, on steam 
generators, to allow for repair of hybrid expansion joint sleeves under 
redefined repair boundary limits.
    Date of publication of individual notice in the Federal Register: 
August 14, 1995 (60 FR 41904)
    Expiration date of individual notice: For comments: August 29, 
1995; hearing requests: September 13, 1995
    Local Public Document Room location: Maud Preston Palenske Memorial 
Library, 500 Market Street, St. Joseph, Michigan 49085

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

    Date of application for amendments:  August 21, 1995Description of 
amendments request: Amend technical specification 3.7.5.c to allow an 
increase in the average essential raw cooling water supply header 
temperature from 84.5 deg.F to 87 deg.F until September 30, 1995.
    Date of publication of individual notice in the Federal Register: 
August 28, 1995 (60 FR 44517)
    Expiration date of individual notice: September 12, 1995
    Local Public Document Room location:  Chattanooga-Hamilton County 
Library, 1101 Broad Street, Chattanooga, Tennessee 37402

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.

Arizona Public Service Company, et al., Docket Nos. STN 50-528, STN 
50-529, and STN 50-530, Palo Verde Nuclear Generating Station, 
Units 1, 2, and 3, Maricopa County, Arizona

    Date of application for amendments:  June 17, 1994
    Brief description of amendments: These amendments revise the 
surveillance requirement and Bases section of TS 4.7.1.6 to increase 
the minimum nitrogen accumulator pressure for the atmospheric dump 
valves (ADVs).
    Date of issuance: September 6, 1995
    Effective date: September 6, 1995
    Amendment Nos.: Unit 1 - Amendment No. 99; Unit 2 - Amendment No. 
87; Unit 3 - Amendment No. 70
    Facility Operating License Nos. NPF-41, NPF-51, and NPF-74: The 
amendments revised the Technical Specifications.
    Date of initial notice in Federal Register: August 17, 1994 (59 FR 
42333) The Commission's related evaluation of the amendments is 
contained in a Safety Evaluation dated September 6, 1995.No significant 
hazards consideration comments received: No
    Local Public Document Room location: Phoenix Public Library, 12 
East McDowell Road, Phoenix, Arizona 85004

Arizona Public Service Company, et al., Docket Nos. STN 50-528, STN 
50-529, and STN 50-530, Palo Verde Nuclear Generating Station, 
Units 1, 2, and 3, Maricopa County, Arizona

    Date of application for amendments: March 31, 1995
    Brief description of amendments: The amendments clarify the 
shutdown margin definition, change the shutdown margin applicability 
and surveillance requirements to comply with the safety analysis 
assumptions for subcritical inadvertent control element assembly 
withdrawal (UFSAR Section 15.4, and expand the applicability for core 
protection calculator (CPC) operability. In addition, the amendments 
add a reference to the Core Operating Limits Report for the MODE 6 
refueling boron concentration limit. The amendments also change the 
power calibration requirements for the linear power level, the CPC 
delta T power, and CPC nuclear power signals to allow more conservative 
settings than previously required.
    Date of issuance: September 1, 1995

[[Page 49952]]

    Effective date: September 1, 1995
    Amendment Nos.: Unit 1 - Amendment No. 98; Unit 2 - Amendment No. 
86; Unit - Amendment No. 69
    Facility Operating License Nos. NPF-41, NPF-51, and NPF-74: The 
amendments revised the Technical Specifications.
    Date of initial notice in Federal Register: June 6, 1995 (60 FR 
29871) The Commission's related evaluation of the amendments is 
contained in a Safety Evaluation dated September 1, 1995.No significant 
hazards consideration comments received: No
    Local Public Document Room location: Phoenix Public Library, 1221 
N. Central, Phoenix, Arizona 85004

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

    Date of application for amendment: June 3, 1995, as supplemented on 
August 7, 1995. The supplemental submittal did not expand the scope of 
the original Federal Register notice or change the no significant 
hazards determination.
    Brief description of amendment: The amendment clarifies the 
definition of operability of the charging pumps by adding a footnote to 
TS Section 3.2.2.a that states that the connectibility of the emergency 
power sources is not required for charging pump operability. The bases 
statement for TS 3.2.2 is also changed for clarification.
    Date of issuance: September 5, 1995
    Effective date: September 5, 1995
    Amendment No.: 166
    Facility Operating License No. DPR-23. Amendment revises the 
Technical Specifications.
    Date of initial notice in Federal Register: July 5, 1995 (60 FR 
35063) The Commission's related evaluation of the amendment is 
contained in a Safety Evaluation dated September 5, 1995.No significant 
hazards consideration comments received: No
    Local Public Document Room location: Hartsville Memorial Library, 
147 West College Avenue, Hartsville, South Carolina 29550

Commonwealth Edison Company, Docket Nos. STN 50-454 and STN 50-455, 
Byron Station, Unit Nos. 1 and 2, Ogle County, IllinoisDocket Nos. 
STN 50-456 and STN 50-457, Braidwood Station, Unit Nos. 1 and 2, 
Will County, Illinois

    Date of application for amendments: February 21, 1995
    Brief description of amendments: The amendments revise the 
technical specifications to permit replacement of the reactor coolant 
resistance temperature detector (RTD) bypass manifold system with fast 
response RTDs mounted in thermowells welded directly into the reactor 
coolant system piping.
    Date of issuance: September 5, 1995
    Effective date: September 5, 1995
    Amendment Nos.: 74 and 66
    Facility Operating License Nos. NPF-37, NPF-66, NPF-72 and NPF-77: 
The amendments revised the Technical Specifications.
    Date of initial notice in Federal Register: July 5, 1995 (60 FR 
35063) The Commission's related evaluation of the amendments is 
contained in a Safety Evaluation dated September 5, 1995.No significant 
hazards consideration comments received: No
    Local Public Document Room location: For Byron, the Byron Public 
Library District, 109 N. Franklin, P.O. Box 434, Byron, Illinois 61010; 
for Braidwood, the Wilmington Public Library, 201 S. Kankakee Street, 
Wilmington, Illinois 60481

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

    Date of application for amendments: February 23, 1995
    Brief description of amendments: The amendment revises the Quad 
Cities Nuclear Power Station, Units 1 and 2, operating licenses to 
reflect the transfer of the Iowa-Illinois Gas and Electric Company's 25 
percent undivided ownership to MidAmerican Energy Company.
    Date of issuance: September 11, 1995
    Effective date: As of the consummation of the merger between Iowa-
Illinois Gas and Electric Company, Midwest Power Systems, Inc., 
MidAmerican Energy Company, and Midwest Resources, Inc.
    Amendment Nos.: 159 and 155
    Facility Operating License Nos. DPR-29 and DPR-30: The amendments 
revised the operating licenses.
    Date of initial notice in Federal Register: July 5, 1995 (60 FR 
35054) The Commission's related evaluation of the amendments is 
contained in an Environmental Assessment and Finding of No Significant 
Impact dated March 21, 1995, and in a Safety Evaluation dated September 
11, 1995.No significant hazards consideration comments received: No
    Local Public Document Room location: Dixon Public Library, 221 
Hennepin Avenue, Dixon, Illinois 61021

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: May 31, 1995
    Brief description of amendments: The amendments authorize an 
alternative repair criteria for defects found in the tube expansion 
region within the tubesheet.
    Date of issuance: September 11, 1995
    Effective date: September 11, 1995
    Amendment Nos.: 168 and 155
    Facility Operating License Nos. DPR-39 and DPR-48: The amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: July 5, 1995 (60 FR 
35067) The Commission's related evaluation of the amendments is 
contained in a Safety Evaluation dated August 11, 1995.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:  September 19, 1994, as 
supplemented April 26 and June 19, 1995
    Brief description of amendments: These changes to the Technical 
Specifications (TS) increase the enrichment limits for fuel stored in 
the fuel pools and establish restricted loading patterns and associated 
burnup criteria for qualifying fuel in the spent fuel pools. In 
addition, several administrative changes have been included in order to 
provide clarity to the TS and bring them more in line with the Standard 
Technical Specifications format. These changes are as follows: (1) The 
TS index is changed to add TS 3/4.9.12 and 3/4.9.13, Tables 3.9-1 and 
3.9-2 and Figure 3.9-1; (2) TS 3/4.9.12, Spent Fuel Pool (SFP) Boron 
Concentration is added to establish a boron concentration limit and to 
establish a Limiting Condition for Operation (LCO) for all modes of 
operation and to allow the numerical value of the limit to be specified 
in the Core Operating Limits Report (COLR); (3) TS 3/4.9.13, Tables 
3.9-1 and 3.9-2 and Figure 3.9-1 are being added to establish 
restricted loading patterns for spent fuel storage and associated 
burnup criteria; (4) Corresponding BASES for TS 3/4.9.12 and 3/4.9.13 
are added to explain the basis for each LCO, Action Statement and 
Surveillance 

[[Page 49953]]
Requirement covered by the subject TS; (5) TS 5.6, Fuel Storage, is 
changed to reflect limits for criticality analysis for fuel storage; 
and (6) TS 6.9, Reporting Requirements, is changed to reflect the 
inclusion of the SFP boron concentration limit values in the COLR as 
established by TS 3/4.9.12.
    Date of issuance: August 31, 1995
    Effective date: As of the date of issuance to be implemented within 
30 days
    Amendment Nos.: 134 and 128
    Facility Operating License Nos. NPF-35 and NPF-52: Amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: May 23, 1995 (60 FR 
27338) The June 19, 1995, letter provided clarifying information that 
did not change the scope of the September 19, 1994, application and the 
initial proposed no significant hazards consideration determination. 
The Commission's related evaluation of the amendments is contained in a 
Safety Evaluation dated August 31, 1995, and Environmental Assessment 
dated August 15, 1995.No significant hazards consideration comments 
received: No
    Local Public Document Room location: York County Library, 138 East 
Black Street, Rock Hill, South Carolina 29730

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

    Date of application for amendments: August 31, 1994, as 
supplemented May 18, 1995.
    Brief description of amendments: These amendments delete Beaver 
Valley Power Station, Unit 2, License Conditions 2.C.(3), 2.C.(5), 
2.C.(7), 2.C.(8), 2.C.(9) and 2.C.(10) to reflect completion of 
activities required by these license conditions and make the following 
revisions to the Beaver Valley Power Station, Units 1 and 2, TSs:
    1. Eliminate references to specific frequencies for each of the TS 
required audits (TS 6.2.2.8).
    2. Eliminate references to reviews and audits of the Emergency plan 
and Security Plant (TSs 6.5.2.8 and 6.8.1).
    3. Include Offsite Dose Calculation Manual and Process Control 
Program and associated implementing procedures into the list of 
required audits (TS 6.5.2.8).
    4. Editorial changes which were necessitated by a reorganization 
(TS 6.2.1, 6.2.3.1, 6.2.3.4, 6.5.2.2, 6.5.2.8, 6.5.2.9, and 6.5.2.10).
    5. Eliminate reference to Appendix A of 10 CFR Part 55 (TS 6.4.1).
    6. Separate the Inservice Inspection (ISI) and Inservice Testing 
(IST) Programs surveillance requirements and remove the requirement 
that relief requests be granted before they are implemented for both 
IST and ISI (TS 4.0.5).
    The May 18, 1995, letter requested withdrawal of the proposed 
changes to TS 6.5.2.8 dealing with audits of the Beaver Valley Power 
Station, Units 1 and 2, fire protection program and withdrawal of a 
proposed 25-percent grace period for all audit frequencies (Item 6 in 
August 31, 1994 application).
    Date of issuance: August 31, 1995
    Effective date: Units 1 and 2, as of the date of issuance and shall 
be implemented within 60 days.
    Amendment Nos.: 191 and 74
    Facility Operating License Nos. DPR-66 and NPF-73: Amendments 
revised the Units 1 and 2 Technical Specifications, and the Unit 2 
License.
    Date of initial notice in Federal Register: (59 FR 65812) December 
21, 1994. The May 18, 1995, letter did not change the original no 
significant hazards consideration determination or expand the scope of 
the December 21, 1994, Federal Register notice. The Commission's 
related evaluation of the amendments is contained in a Safety 
Evaluation dated August 31, 1995.No significant hazards consideration 
comments received: No
    Local Public Document Room location: B. F. Jones Memorial Library, 
663 Franklin Avenue, Aliquippa, Pennsylvania 15001

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 application for amendments: October 11, 1994, as 
supplemented June 23, 1995, and August 24, 1995
    Brief description of amendments: These amendments revise Beaver 
Valley Power Station Technical Specifications (TSs) 1.18, ``Quadrant 
Power Tilt Ratio,'' 3/4.2.4, ``Quadrant Power Tilt Ratio,'' the table 
Notation of TS Table 3.3-1, ``Reactor Trip System Instrumentation,'' 
and associated Bases to incorporate the guidance provided in the NRC's 
Improved Standard Technical Specifications (NUREG-1431, Revision 1) to 
these TSs. The amendments clarify the requirements of the subject TSs 
with regard to the use of excore power range neutron flux detectors to 
monitor quadrant power tilt ratio when an excore power range neutron 
flux instrument is inoperable. The changes also make several minor 
editorial changes in the subject TSs.
    Date of issuance: September 15, 1995
    Effective date: As of date of issuance, to be implemented within 60 
days.
    Amendment Nos.: 192 and 75
    Facility Operating License Nos. DPR-66 and NPF-73: Amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: August 2, 1995 (60 FR 
39436) The August 24, 1995, letter provided typed final TS pages, with 
minor editorial changes, for issuance of these amendments. The August 
24, 1995, letter did not change the initial proposed no significant 
hazards consideration determination or expand the scope of the August 
2, 1995, Federal Register notice. The Commission's related evaluation 
of the amendments is contained in a Safety Evaluation dated September 
15, 1995. No significant hazards consideration comments received: No
    Local Public Document Room location: B. F. Jones Memorial Library, 
663 Franklin Avenue, Aliquippa, Pennsylvania 15001

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

    Date of amendment request: June 22, 1994
    Brief description of amendment: The amendment changes the Appendix 
A Technical Specifications by removing the seismic and meteorological 
monitoring instrumentation requirements. These requirements are to be 
relocated in the Updated Final Safety Analysis Report.
    Date of issuance: September 5, 1995
    Effective date: September 5, 1995
    Amendment No.: 112
    Facility Operating License No. NPF-38. Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: August 3, 1994 (59 FR 
39585) The Commission's related evaluation of the amendment is 
contained in a Safety Evaluation dated September 5, 1995.No significant 
hazards consideration comments received: No
    Local Public Document Room location: University of New Orleans 
Library, Louisiana Collection, Lakefront, New Orleans, LA 70122

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

    Date of amendment request: June 22, 1994, and December 9, 1994
    Brief description of amendment: The amendment changes the Appendix 
A TSs by revising the plant protection system trip setpoints and 
allowable 

[[Page 49954]]
values such that they will be consistent with the current setpoint/
uncertainty methodology being implemented at Waterford 3.
    Date of issuance: September 5, 1995
    Effective date: September 5, 1995
    Amendment No.: 113
    Facility Operating License No. NPF-38. Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: August 3, 1994 (59 FR 
39586) and February 1, 1995 (60 FR 6300)The Commission's related 
evaluation of the amendment is contained in a Safety Evaluation dated 
September 5, 1995.No significant hazards consideration comments 
received: No
    Local Public Document Room location: University of New Orleans 
Library, Louisiana Collection, Lakefront, New Orleans, LA 70122

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

    Date of amendment request: December 9, 1994, as supplemented by 
letter dated July 25, 1995
    Brief description of amendment: The requested changes revised the 
allowable opening tolerances on the pressurizer safety valves (PSVs) 
and the main steam line safety valves (MSSVs) from plus or minus 1 
percent to plus or minus 3 percent. However, following testing, the as-
left lift setting of the PSVs and MSSVs will be within plus or minus 1 
percent of the pressure specified in the Technical Specifications.
    Date of issuance: September 11, 1995
    Effective date: September 11, 1995
    Amendment No.: 111
    Facility Operating License No. NPF-38. Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: February 1, 1995 (60 FR 
6300) The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated September 11, 1995.No significant hazards 
consideration comments received: No
    Local Public Document Room location: University of New Orleans 
Library, Louisiana Collection, Lakefront, New Orleans, Louisiana 70122

Florida Power and Light Company, Docket No. 50-335, St. Lucie 
Plant, Unit No. 1, St. Lucie County, Florida

    Date of application for amendment: May 23, 1994
    Brief description of amendment: The amendment revises Technical 
Specification 3.5.2 for Emergency Core Cooling Systems (ECCS) by 
removing the option that allows High Pressure Safety Injection (HPSI) 
Pump 1C to be used as an alternative to the preferred pump for 
subsystem operability. HPSI pump 1C is an installed spare which is not 
required to be maintained in an operable status, and this change 
upgrades the ECCS operability requirements consistent with actual plant 
operating needs.
    Date of issuance: September 11, 1995
    Effective date: September 11, 1995
    Amendment No.: 139
    Facility Operating License No. DPR-67: Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: July 6, 1994 (59 FR 
34663) The Commission's related evaluation of the amendment is 
contained in a Safety Evaluation dated September 11, 1995.No 
significant hazards consideration comments received: No
    Local Public Document Room location: Indian River Junior College 
Library, 3209 Virginia Avenue, Fort Pierce, Florida 34954-9003

Florida Power and Light Company, et al., Docket No. 50-389, St. 
Lucie Plant, Unit No. 2, St. Lucie County, Florida

    Date of application for amendment: February 27, 1995
    Brief description of amendment: This amendment will change Table 
3.3-3 and 3.3-4 to accommodate an improved coincidence logic and relay 
replacement for the 4.16 kV Loss of Voltage Relays. Actions required 
for certain trip units with the number of operable channels one less 
than the total number of channels will also be changed. In addition, 
the format used to state the time delay for the 4.16 kV Degraded 
Voltage trip unit will be revised.
    Date of issuance: September 1, 1995
    Effective date: September 1, 1995
    Amendment No.: 79
    Facility Operating License No. NPF-16: Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: March 29, 1995 (60 FR 
16187) The Commission's related evaluation of the amendment is 
contained in a Safety Evaluation dated September 1, 1995.No significant 
hazards consideration comments received: No
    Local Public Document Room location: Indian River Junior College 
Library, 3209 Virginia Avenue, Fort Pierce, Florida 34954-9003

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

    Date of application for amendments: January 13, 1995, as 
supplemented by letters dated April 5 and June 20, 1995.
    Brief description of amendments: The amendments modify
    Facility Operating License Nos. DRP-57 and NPF-5 and the 
corresponding TS for Hatch Units 1 and 2, respectively, to authorize an 
increase in the maximum power level from 2436 megawatts thermal (MWt) 
to 2558 MWt. The amendments also approve changes to the Technical 
Specification to implement uprated power operation.
    Date of issuance: August 31, 1995
    Effective date: As of the date of issuance to be implemented prior 
to startup in Cycle 17 for Unit 1; and prior to startup in Cycle 13 for 
Unit 2
    Amendment Nos.: 197 and 138
    Facility Operating License Nos. DPR-57 and NPF-5. Amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: July 5, 1995 (60 FR 
35072) The Commission's related evaluation of the amendments is 
contained in a Safety Evaluation dated August 31, 1995 and an 
Environmental Assessment dated July 21, 1995.No significant hazards 
consideration comments received: No
    Local Public Document Room location: Appling County Public Library, 
301 City Hall Drive, Baxley, Georgia 31513

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

    Date of application for amendments: June 6, 1995, as supplemented 
August 9, 1995.
    Brief description of amendments: The amendments revise Technical 
Specification Surveillance Requirements (SR) 3.6.4.1.3 and 3.6.4.1.4 
for the secondary containment drawdown. The revision reduces the SR 
acceptance criteria to greater than or equal to 0.20 inch water gauge 
(wg) negative pressure from greater than or equal to O.25 inch wg 
negative pressure. The appropriate TS Bases pages are also changed to 
reflect the TS revision.
    Date of issuance: September 11, 1995
    Effective date: As of the date of issuance to be implemented within 
60 days
    Amendment Nos.: 198 and 139

[[Page 49955]]

    Facility Operating License Nos. DPR-57 and NPF-5. Amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: June 21, 1995 (60 FR 
32364) The August 9, 1995, letter provided clarifying information that 
did not change the scope of the June 6, 1995, application and the 
initial proposed no significant hazards consideration determination. 
The Commission's related evaluation of the amendments is contained in a 
Safety Evaluation dated September 11, 1995.No significant hazards 
consideration comments received: No
    Local Public Document Room location: Appling County Public Library, 
301 City Hall Drive, Baxley, Georgia 31513

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

    Date of application for amendment: June 26, 1995
    Brief description of amendment: The amendment revises the snubber 
visual inspection intervals to match the schedule developed by the NRC 
staff for use with a 24-month refueling interval. This schedule was 
documented in Generic Letter 90-09. The amendment also revises the 
bases for the snubber visual inspection interval to be consistent with 
the bases described in Generic Letter 90-09.
    Date of issuance: September 6, 1995
    Effective date: September 6, 1995
    Amendment No.: 182
    Facility Operating License No. DPR-16. Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: August 2, 1995 (60 FR 
39440). The Commission's related evaluation of this amendment is 
contained in a Safety Evaluation dated September 6, 1995. No 
significant hazards consideration comments received: No
    Local Public Document Room location: Ocean County Library, 
Reference Department, 101 Washington Street, Toms River, NJ 08753

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

    Date of amendment request: May 25, 1995
    Brief description of amendment: The amendment revises the Physical 
Security Plan vital island requirements.
    Date of issuance: September 12, 1995
    Effective date: September 12, 1995
    Amendment No.: 83
    Facility Operating License No. NPF-47. The amendment revised the 
operating license.
    Date of initial notice in Federal Register: July 19, 1995 (60 FR 
37091) The Commission's related evaluation of the amendment is 
contained in a Safety Evaluation dated September 12, 1995.No 
significant hazards consideration comments received. No
    Local Public Document Room location: Government Documents 
Department, Louisiana State University, Baton Rouge, LA 70803

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

    Date of amendment request: May 25, 1995, as supplemented by letter 
dated August 3, 1995.
    Brief description of amendments: The amendments revised the 
technical specifications (TSs) on containment leakage, making the 
action statement consistent with the need to perform Type C testing at 
power, and replacing the surveillance requirements with a single 
requirement to apply the requirements of Appendix J as modified by 
approved exemptions. The amendments also revised the TSs on containment 
integrity, containment leakage, and containment air locks, to eliminate 
the numerical value of calculated peak containment internal pressure 
related to the design basis accident.
    Date of issuance: September 7, 1995
    Effective date: September 7, 1995
    Amendment Nos.: Unit 1 - Amendment No. 80; Unit 2 - Amendment No. 
69
    Facility Operating License Nos. NPF-76 and NPF-80. The amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: July 19, 1995 (60 FR 
37092) The August 3, 1995, supplement provided clarifying information 
and did not change the original no significant hazards consideration 
determination. The Commission's related evaluation of the amendments is 
contained in a Safety Evaluation dated September 7, 1995.No significant 
hazards consideration comments received: No
    Local Public Document Room location: Wharton County Junior College, 
J. M. Hodges Learning Center, 911 Boling Highway, Wharton, Texas 77488

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

    Date of amendment request: May 31, 1995, as supplemented by letter 
dated August 2, 1995
    Brief description of amendments: The amendments modified (by 
relocation to the Technical Requirements Manual) TS 3/4.1.2.1, Boration 
Systems/Flow Paths - Shutdown, TS 3/4.1.2.2, Boration Systems/Flow 
Paths - Operating, TS 3/4.1.2.3, Charging Pumps - Shutdown, TS 3/
4.1.2.4, Charging Pumps - Operating, TS 3/4.1.2.5, Borated Water 
Sources - Shutdown, TS 3/4.1.2.6, Borated Water Sources - Operating, TS 
3/4.4.2.1, Safety Valves - Shutdown, and the associated Bases.
    Date of issuance: September 5, 1995
    Effective date: September 5, 1995
    Amendment Nos.: Unit 1 - Amendment No. 79; Unit 2 - Amendment No. 
68
    Facility Operating License Nos. NPF-76 and NPF-80. The amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: August 2, 1995 (60 FR 
39441) The additional information contained in the supplemental letter 
dated August 2, 1995, was clarifying in nature and thus, within the 
scope of the initial notice and did not affect the staff's proposed no 
significant hazards consideration determination.The Commission's 
related evaluation of the amendments is contained in a Safety 
Evaluation dated September 5, 1995.No significant hazards consideration 
comments received: No
    Local Public Document Room location: Wharton County Junior College, 
J. M. Hodges Learning Center, 911 Boling Highway, Wharton, TX 77488

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

    Date of application for amendment: February 3, 1995, as 
supplemented April 25, 1995.
    Brief description of amendment: The amendment modifies the 
technical specifications to extend the interim steam generator tube 
plugging criteria used in Cycle 14 to the next operating cycle (Cycle 
15).
    Date of issuance: September 13, 1995
    Effective date: September 13, 1995
    Amendment No.: 200

[[Page 49956]]

    Facility Operating License No. DPR-58. Amendment revises the 
Technical Specifications.
    Date of initial notice in Federal Register: July 19, 1995 (60 FR 
37093) The Commission's related evaluation of the amendment is 
contained in a Safety Evaluation dated September 13, 1995.No 
significant hazards consideration comments received: No
    Local Public Document Room location: Maud Preston Palenske Memorial 
Library, 500 Market Street, St. Joseph, Michigan 49085

Indiana Michigan Power Company, Docket No. 50-316, Donald C. Cook, 
Nuclear Plant, Unit No. 2, Berrien County, Michigan

    Date of application for amendment: February 15, 1994, as 
supplemented June 29, 1995
    Brief description of amendment: The amendment deletes Technical 
Specification section 3/4.3.4, associated bases, and associated index 
listings for the Unit 2 turbine overspeed protection system.
    Date of issuance: September 1, 1995
    Effective date: September 1, 1995
    Amendment No.: 185
    Facility Operating License No. DPR-74. Amendment revises the 
Technical Specifications.
    Date of initial notice in Federal Register: March 30, 1994 (59 FR 
14890) The licensee's submittal of June 29, 1995, did not change the 
basis for the proposed no significant hazards consideration 
determination.The Commission's related evaluation of the amendment is 
contained in a Safety Evaluation dated September 1, 1995.No significant 
hazards consideration comments received: No
    Local Public Document Room location: Maud Preston Palenske Memorial 
Library, 500 Market Street, St. Joseph, Michigan 49085

Nebraska Public Power District, Docket No. 50-298, Cooper Nuclear 
Station, Nemaha County, Nebraska

    Date of amendment request: June 15, 1995
    Brief description of amendment: The amendment changes the Technical 
Specifications to revise the definition for logic system functional 
test and revises the surveillance interval for emergency core cooling 
system logic system functional testing from 6 months to 18 months.
    Date of issuance: September 7, 1995
    Effective date: September 7, 1995
    Amendment No.: 171
    Facility Operating License No. DPR-46. Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: July 19, 1995 (60 FR 
37096) The Commission's related evaluation of the amendment is 
contained in a Safety Evaluation dated September 7, 1995.No significant 
hazards consideration comments received: No
    Local Public Document Room location:  Auburn Public Library, 118 
15th Street, Auburn, NE 68305

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

    Date of application for amendment: January 6, 1995
    Brief description of amendment: The amendment incorporates Limiting 
Condition for Operation 3.3.3.1 from Standard Technical Specifications 
into Technical Specification (TS) 3/4.3.7.5, Accident Monitoring 
Instrumentation and make associated changes in TS 3/4.4.2, Safety 
Relief Valves.
    Date of issuance: September 11, 1995
    Effective date: As of the date of issuance to be implemented within 
30 days.
    Amendment No.: 69
    Facility Operating License No. NPF-69: Amendment revises the 
Technical Specifications.
    Date of initial notice in Federal Register: February 15, 1995 (60 
FR 8748)The Commission's related evaluation of the amendment is 
contained in a Safety Evaluation dated September 11, 1995.No 
significant hazards consideration comments received: No
    Local Public Document Room location:  Reference and Documents 
Department, Penfield Library, State University of New York, Oswego, New 
York 13126

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

    Date of application for amendment: January 6, 1995, as supplemented 
April 18, 1995
    Brief description of amendment: The amendment revises Technical 
Specifications (TSs) Sections 3.8.1.1 and 3.8.1.2; TS Surveillance 
Requirements Section 4.8.1.1.2; TS Bases Section 3/4.8.1.3; and TS 
Administrative Controls Section 6.8.4. The changes include: updating 
the minimum day tank and storage tank oil inventory, specific actions 
required if oil level fall below minimum required, revising and 
relocating the fuel oil sampling and testing criteria to the associated 
Bases, and specific action to be taken if the fuel oil properties do 
not meet the specified limits. In addition, a requirement was added for 
a diesel fuel oil testing program. These changes are consistent with 
guidance provided in NUREG-1434.
    Date of issuance: September 15, 1995
    Effective date: As of the date of issuance to be implemented within 
30 days.
    Amendment No.: 70
    Facility Operating License No. NPF-69: Amendment revises the 
Technical Specifications.
    Date of initial notice in Federal Register: February 15, 1995 (60 
FR 8747) The Commission's related evaluation of the amendment is 
contained in a Safety Evaluation dated September 15, 1995.No 
significant hazards consideration comments received: No
    Local Public Document Room location:  Reference and Documents 
Department, Penfield Library, State University of New York, Oswego, New 
York 13126

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

    Date of amendment request: April 16, 1995.
    Description of amendment request: The amendment revises the 
Appendix A Technical Specifications (TS) relating to containment 
building penetrations. Specifically, the amendment modifies Limiting 
Conditions for Operation 3.9.4 to permit both doors of one personnel 
airlock to be open during core alterations or irradiated fuel movement 
if certain conditions are met and to add equivalent and alternate 
penetration closure methodologies. Surveillance Requirement 4.9.4 is 
changed to reflect that the penetrations are to be verified to be in 
the condition required. Bases Section 3/4 9.4 also is revised to 
reflect the changes described above.
    Date of issuance: August 31, 1995
    Effective date: As of the date of issuance, to be implemented 
within 60 days.
    Amendment No.: 40
    Facility Operating License No. NPF-86. Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: June 21, 1995 (60 FR 
32369) The Commission's related evaluation of the amendment is 
contained in a Safety Evaluation dated August 31, 1995.No significant 
hazards consideration comments received: No
    Local Public Document Room location: Exeter Public Library, 
Founders Park, Exeter, NH 03833

[[Page 49957]]


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

    Date of amendment request: May 30, 1995.
    Description of amendment request: The amendment revises the 
Appendix A Technical Specifications (TS) relating to Moderator 
Temperature Coefficient. The amendment changes the upper limit for the 
moderator temperature coefficient (MTC) for certain operating 
conditions. Additionally, a reference for the analytical method used to 
determine the cycle-specific MTC upper limit is added to TS 6.8.1.6.b.
    Date of issuance: September 14, 1995
    Effective date: As of the date of issuance to be implemented within 
60 days.
    Amendment No.: 41
    Facility Operating License No. NPF-86. Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: July 5, 1995 (60 FR 
35082). The Commission's related evaluation of the amendment is 
contained in a Safety Evaluation dated September 14, 1995. No 
significant hazards consideration comments received: No
    Local Public Document Room location: Exeter Public Library, 
Founders Park, Exeter, NH 03833

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

    Date of amendment request: June 16, 1995
    Description of amendment request: The amendment revises the 
Appendix A Technical Specifications (TS) relating to core reactivity 
control available from borated water sources. The amendment changes the 
minimum boron concentration specified for the refueling water storage 
tank (RWST) in Limiting Condition for Operation (LCO) in TS 3.1.2.5 and 
replaces the minimum specified concentration for boron with an 
acceptable range of boron concentration for the RWST and the 
accumulators in the LCOs for TS 3.1.2.6, 3.5.1.1, and 3.5.4.
    Date of issuance: September 14, 1995
    Effective date: As of the date of issuance, to be implemented prior 
to entering MODE 4 following the fourth refueling outage.
    Amendment No.: 42
    Facility Operating License No. NPF-86. Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: August 2, 1995 (60 FR 
39442). The Commission's related evaluation of the amendment is 
contained in a Safety Evaluation dated September 14, 1995. No 
significant hazards consideration comments received: No
    Local Public Document Room location: Exeter Public Library, 
Founders Park, Exeter, NH 03833

Northeast Nuclear Energy Company, Docket No. 50-245, 
MillstoneNuclear Power Station, Unit 1, New London County, 
Connecticut

    Date of application for amendment: July 11, 1995
    Brief description of amendment: The amendment modifies Technical 
Specification 3.5.F.7 to also allow the use of pull-to-lock switches to 
defeat the automatic initiation of the emergency core cooling system 
while in the refuel condition. The amendment also makes editorial 
corrections and makes changes to the associated Bases section.
    Date of issuance: September 13, 1995
    Effective date: As of the date of issuance to be implemented within 
30 days.
    Amendment No.: 88
    Facility Operating License No. DPR-21. Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: August 2, 1995 (60 FR 
39442). The Commission's related evaluation of the amendment is 
contained in a Safety Evaluation dated September 13, 1995.No 
significant hazards consideration comments received: No
    Local Public Document Room location: Learning Resources Center, 
Three Rivers Community-Technical College, 574 New London Turnpike, 
Norwich, CT 06360

Northeast Nuclear Energy Company, Docket No. 50-245, 
MillstoneNuclear Power Station, Unit 1, New London County, 
Connecticut

    Date of application for amendment: July 18, 1995
    Brief description of amendment: The amendment adds operability and 
surveillance requirements for reactor pressure vessel overfill 
protection instrumentation. The amendment also adds the associated 
Bases.
    Date of issuance: September 13, 1995
    Effective date: As of the date of issuance to be implemented within 
60 days.
    Amendment No.: 87
    Facility Operating License No. DPR-21. Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: August 2, 1995 (60 FR 
39443) The Commission's related evaluation of the amendment is 
contained in a Safety Evaluation dated September 13, 1995.No 
significant hazards consideration comments received: No
    Local Public Document Room location: Learning Resources Center, 
Three Rivers Community-Technical College, 574 New London Turnpike, 
Norwich, CT 06360

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

    Date of amendment request: April 7, 1995
    Brief description of amendment: The amendment revises the technical 
specifications (TS) to relocate the axial power distribution limits to 
the Core Operating Limits Report (COLR).
    Date of issuance: September 1, 1995
    Effective date: September 1, 1995
    Amendment No.: 170
    Facility Operating License No. DPR-40. Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: May 23, 1995 (60 FR 
27339) The Commission's related evaluation of the amendment is 
contained in a Safety Evaluation dated September 1, 1995.No significant 
hazards consideration comments received: No
    Local Public Document Room location: W. Dale Clark Library, 215 
South 15th Street, Omaha, Nebraska 68102

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

    Date of amendment request: May 8, 1995, as supplemented by letter 
dated July 11, 1995.
    Brief description of amendment: The amendment changes Sections 2.3, 
3.1, 3.2, 3.3, and 3.6 of the Technical Specifications in accordance 
with the guidance of Generic Letter (GL) 93-05, ``Line Item Technical 
Specifications Improvements to Reduce Surveillance Requirements for 
Testing During Power Operation,'' dated September 27, 1993. The changes 
are consistent with Station operating experience and NUREG-1366, 
``Improvements to Technical Specifications Surveillance Requirements,'' 
dated December 1992. In addition, a change was made to TS Section 3.1 
in accordance with the Commission's Final Policy Statement on Technical 
Specifications Improvements for Nuclear Power Reactors. Also, changes 
were made to the TS sections identified above for clarity and to 
correct administrative errors.
    Date of issuance: September 7, 1995
    Effective date: September 7, 1995
    Amendment No.: 171

[[Page 49958]]

    Facility Operating License No. DPR-40. Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: June 6, 1995 (60 FR 
29883) The Commission's related evaluation of the amendment is 
contained in a Safety Evaluation dated September 7, 1995.No significant 
hazards consideration comments received: No
    Local Public Document Room location: W. Dale Clark Library, 215 
South 15th Street, Omaha, Nebraska 68102

Pennsylvania Power and Light Company, Docket No. 50-387, 
Susquehanna Steam Electric Station, Unit 1, Luzerne County, 
Pennsylvania

    Date of application for amendment:  April 11, 1995
    Brief description of amendment: This amendment extends on a one-
time basis the allowed outage time from 3 to 7 days for one offsite 
circuit being out of service.
    Date of issuance: August 31, 1995
    Effective date: As of the date of issuance and is to be implemented 
within 30 days.
    Amendment No.: 153
    Facility Operating License No. NPF-14: This amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: June 6, 1995 (60 FR 
29886). The Commission's related evaluation of the amendment is 
contained in a Safety Evaluation dated August 31, 1995.No significant 
hazards consideration comments received: No
    Local Public Document Room location: Osterhout Free Library, 
Reference Department, 71 South Franklin Street, Wilkes-Barre, 
Pennsylvania 18701

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

    Date of application for amendments: February 2, 1995
    Brief description of amendments: These amendments change the 
Technical Specifications for the two Susquehanna units to increase the 
licensed discharge fuel assembly for SPC 9X9-2 fuel from 40 to 45 GWD/
MTU. This change is consistent with the Commissions approval of Topical 
Report PL-NF-94-005-P, ``Technical Basis for SPC 9X9-2 Extended Fuel 
Exposure at Susquehanna SES,'' documented in a letter to PP&L dated 
December 15, 1994.
    Date of issuance: September 12, 1995
    Effective date: September 12, 1995
    Amendment Nos.: 154 and 124
    Facility Operating License Nos. NPF-14 and NPF-22. The amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: March 29, 1995 (60 FR 
16194) The Commission's related evaluation of the amendments is 
contained in a Safety Evaluation dated September 12, 1995.No 
significant hazards consideration comments received: No
    Local Public Document Room location: Osterhout Free Library, 
Reference Department, 71 South Franklin Street, Wilkes-Barre, 
Pennsylvania 18701

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

    Date of application for amendment: March 31, 1995
    Brief description of amendment: This amendment changes Technical 
Specification Section 6.9.3.2 to allow four GE demonstration assemblies 
to be loaded into Susquehanna Unit 2, Cycle 8 core.
    Date of issuance: September 13, 1995
    Effective date: As of date of issuance and shall be implemented 
within 30 days.
    Amendment No.: 125
    Facility Operating License No. NPF-22. This amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: April 26, 1995 (60 FR 
20523) The Commission's related evaluation of the amendment is 
contained in a Safety Evaluation dated September 13, 1995.No 
significant hazards consideration comments received: No
    Local Public Document Room location:  Osterhout Free Library, 
Reference Department, 71 South Franklin Street, Wilkes-Barre, 
Pennsylvania 18701

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

    Date of application for amendment: May 5, 1995, and supplemented by 
letter dated August 18, 1995
    Brief description of amendment: This amendment deletes from SSES 
Technical Specification Table 3.6.3-1, ``Primary Containment Isolation 
Valves,'' three relief valves in the residual heat removal system. 
These specific valves which were originally intended to support the 
steam condensing mode, were previously eliminated from the plant 
design. The valves are being replaced during the September Unit 2 
refueling outage and will be replaced by blind flanges.
    Date of issuance: September 11, 1995
    Effective date: September 11, 1995
    Amendment No.: 123
    Facility Operating License No. NPF-22. This amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: July 5, 1995 (60 FR 
35083 and July 17, 1995 (60 FR 36449)The supplemental letter provided 
clarifying information that did not change the initial proposed no 
significant hazards consideration determination or expand the scope of 
the original Federal Register notice. The Commission's related 
evaluation of the amendment is contained in a Safety Evaluation dated 
September 11, 1995.No significant hazards consideration comments 
received: No
    Local Public Document Room location: Osterhout Free Library, 
Reference Department, 71 South Franklin Street, Wilkes-Barre, 
Pennsylvania 18701

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

    Date of application for amendment:  February 23, 1995, as 
supplemented July 28, 1995
    Brief description of amendment: The amendment revised the minimum 
emergency diesel generator (EDG) fuel oil requirements, as indicated in 
Technical Specification (TS) Section 3.7 (Auxiliary Electrical 
Systems), from 7056 to 6671 gallons. The actual minimum fuel oil level 
had always been 6671 gallons; however, the previous TS limit of 7056 
gallons was based on a level indicator that had an accuracy of +/- 385 
gallons. This revision clarified the TS such that any level indicator 
can now be used as long as an actual minimum level of 6671 gallons is 
assured.
    Date of issuance: August 30, 1995
    Effective date: As of the date of issuance to be implemented within 
30 days.
    Amendment No.: 161
    Facility Operating License No. DPR-64: Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: March 29, 1995 (60 FR 
16196) The Commission's related evaluation of the amendment is 
contained in a Safety Evaluation dated August 30, 1995.No significant 
hazards consideration comments received: No
    Local Public Document Room location: White Plains Public Library, 
100 Martine Avenue, White Plains, New York 10610

[[Page 49959]]


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

    Date of application for amendment: March 2, 1995
    Brief description of amendment: The amendment revised the titles of 
several management positions as described in Technical Specifications 
Section 6.0 (Administrative Controls). Specifically, the title of 
Executive Vice President and Chief Nuclear Officer and the title of 
Shift Supervisor were changed to Chief Nuclear Officer and Shift 
Manager, respectively. In addition, the position titles of Senior 
Reactor Operator and Reactor Operator were deleted and replaced with 
qualification requirements.
    Date of issuance: August 31, 1995
    Effective date: As of the date of issuance to be implemented within 
30 days.
    Amendment No.: 162
    Facility Operating License No. DPR-64: Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: March 29, 1995 (60 FR 
16197) The Commission's related evaluation of the amendment is 
contained in a Safety Evaluation dated August 31, 1995.No significant 
hazards consideration comments received: No
    Local Public Document Room location: White Plains Public Library, 
100 Martine Avenue, White Plains, New York 10610

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

    Date of application for amendment: October 3, 1994
    Brief description of amendment: The amendment proposed changes to 
FitzPatrick TSs which will extend the instrumentation functional test 
interval and allowable out-of-service times, remove the average power 
range monitor downscale scram function and the instrument response time 
values, and incorporate several editorial, clarification, and 
correction changes.
    Date of issuance: September 11, 1995
    Effective date: As of the date of issuance to be implemented within 
30 days.
    Amendment No.: 227
    Facility Operating License No. DPR-59: Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: November 9, 1994 (59 FR 
55887) The Commission's related evaluation of the amendment is 
contained in a Safety Evaluation dated September 11, 1995.No 
significant hazards consideration comments received: No
    Local Public Document Room location: Reference and Documents 
Department, Penfield Library, State University of New York, Oswego, New 
York 13126

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

    Date of application for amendment: September 29, 1994
    Brief description of amendment: This amendment revises Table 4.3.6-
1, ``Control Rod Block Instrumentation Surveillance,'' of the Hope 
Creek TS. The channel calibration frequencies for the Source Range 
Monitor (SRM) and the Intermediate Range Monitor (IRM), in TS Table 
4.3.6-1, are changed for the up-scale and the down-scale trip functions 
on each instrument from ``SA'' (once-per-184 days) to ``R'' (once-per-
refueling cycle).
    Date of issuance: September 12, 1995
    Effective date: As of the date of issuance and shall be implemented 
within 60 days.
    Amendment No.: 78
    Facility Operating License No. NPF-57: This amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: June 18, 1995 (60 FR 
3676). The Commission's related evaluation of the amendment is 
contained in a Safety Evaluation dated September 12, 1995. No 
significant hazards consideration comments received: No
    Local Public Document Room location: Pennsville Public Library, 190 
S. Broadway, Pennsville, New Jersey 08070

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

    Date of application for amendments: May 2, 1995
    Brief description of amendments: The amendments eliminate the 
monthly manual initiation of auxiliary feedwater from Technical 
Specification Tables 3.3-3, 3.3.-4 and 4.3-2.
    Date of issuance: September 6, 1995
    Effective date: Units 1 and 2, as of the date of issuance, to be 
implemented within 60 days.
    Amendment Nos. 175 and 156
    Facility Operating License Nos. DPR-70 and DPR-75. The amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: June 6, 1995 (60 FR 
29887)The Commission's related evaluation of the amendments is 
contained in a Safety Evaluation dated September 6, 1995.No significant 
hazards consideration comments received: No
    Local Public Document Room location: Salem Free Public Library, 112 
West Broadway, Salem, New Jersey 08079

South Carolina Electric & Gas Company, South Carolina Public 
ServiceAuthority, Docket No. 50-395, Virgil C. Summer Nuclear 
Station, Unit No. 1, Fairfield County, South Carolina

    Date of application for amendment: June 19, 1995
    Brief description of amendment: The amendment restructures the 
primary containment and primary containment leakage technical 
specifications to reduce the repetition of those requirements contained 
in NRC regulations such as Appendix J to 10 CFR Part 50.
    Date of issuance: September 5, 1995
    Effective date: September 5, 1995
    Amendment No.: 126
    Facility Operating License No. NPF-12: Amendment revises the 
Technical Specifications.
    Date of initial notice in Federal Register: July 19, 1995 (60 FR 
37099)The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated September 5, 1995.No significant hazards 
consideration comments received: No
    Local Public Document Room location: Fairfield County Library, 300 
Washington Street, Winnsboro, SC 29180

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: May 20, 1994
    Brief description of amendments: The amendments revise Technical 
Specification 3/4.7.3, ``Component Cooling Water System,'' and the 
corresponding Bases to support the addition of the component cooling 
water surge tank backup nitrogen supply (BNS) system.
    Date of issuance: September 13, 1995
    Effective date: September 13, 1995
    Amendment Nos.: Unit 2 - Amendment No. 125; Unit 3 - Amendment No. 
114
    Facility Operating License Nos. NPF-10 and NPF-15: The amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: August 31, 1994 (59 FR 

[[Page 49960]]
    45034)The Commission's related evaluation of the amendments is 
contained in a Safety Evaluation dated September 13, 1995.No 
significant hazards consideration comments received: No
    Local Public Document Room location: Main Library, University of 
California, P.O. Box 19557, Irvine, California 92713

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

    Date of application for amendments:  March 31, 1995, supplemented 
July 14, 1995 (TS 349)
    Brief description of amendment: These amendments revise the Browns 
Ferry Nuclear Plant (BFN) Units 1, 2, and 3 reactor vessel pressure-
temperature curves and bolt-up temperatures.
    Date of issuance: September 13, 1995
    Effective date: September 13, 1995
    Amendment Nos.: 224, 239, 198
    Facility Operating License Nos. DPR-33, DPR-52 and DPR-68: 
Amendments revised the Technical Specifications.
    Date of initial notice in Federal Register: June 6, 1995 (60 FR 
29888)The July 14, 1995 letter provided clarifying information that did 
not change the initial proposed no significant hazards consideration 
determination. The Commission's related evaluation of the amendment is 
contained in a Safety Evaluation dated September 13, 1995.No 
significant hazards consideration comments received: No
    Local Public Document Room location: Athens Public library, South 
Street, Athens, Alabama 35611

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

    Date of application for amendments:  May 11, 1995, supplemented 
June 30, 1995 (TS 359)
    Brief description of amendment: The amendments provide for the 
addition of a reactor trip on low scram pilot air header pressure for 
BFN Unit 3, and revise a note regarding instrumentation requirements 
for all three BFN reactors.
    Date of issuance: August 29, 1995
    Effective date: August 29, 1995
    Amendment Nos.: 223, 228 and 197
    Facility Operating License Nos. DPR-33, DPR-52 and DPR-68: 
Amendments revised the Technical Specifications.
    Date of initial notice in Federal Register: June 6, 1995 (60 FR 
29889)The June 30, 1995 letter provided clarifying information that did 
not change the initial proposed no significant hazards consideration 
determination. The Commission's related evaluation of the amendment is 
contained in a Safety Evaluation dated August 29, 1995.No significant 
hazards consideration comments received: No
    Local Public Document Room location: Athens Public library, South 
Street, Athens, Alabama 35611

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

    Date of application for amendments: June 29, 1995 (TS 95-14)
    Brief description of amendments: The amendments revise Technical 
Specification 3.9.4, Containment Building Penetrations, to allow both 
sets of containment personnel airlock doors to be open during core 
alterations and fuel movement provided one door is capable of closure 
and one train of auxiliary building gas treatment remains operable.
    Date of issuance: September 6, 1995
    Effective date: September 6, 1995
    Amendment Nos.: 209 and 199
    Facility Operating License Nos. DPR-77 and DPR-79: Amendments 
revise the technical specifications.
    Date of initial notice in Federal Register: July 19, 1995 (60 FR 
37100)The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated September 6, 1995. No significant hazards 
consideration comments received: No
    Local Public Document Room location: Chattanooga-Hamilton County 
Library,1101 Broad Street, Chattanooga, Tennessee 37402

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

    Date of application for amendments: August 7, 1995 (TS 95-11)
    Brief description of amendments: The amendments revise the time 
constant used in the overtemperature delta temperature and the 
overpower delta temperature trip equations of Technical Specification 
Table 2.2-1.
    Date of issuance: September 15, 1995
    Effective date: September 15, 1995
    Amendment Nos.: 211 and 201
    Facility Operating License Nos. DPR-77 and DPR-79: Amendments 
revise the technical specifications.
    Date of initial notice in Federal Register: April 26, 1995 (60 FR 
20527); superseded August 15, 1995 (60 FR 42187) The Commission's 
related evaluation of the amendment is contained in a Safety Evaluation 
dated September 15, 1995.No significant hazards consideration comments 
received: No
    Local Public Document Room location: Chattanooga-Hamilton County 
Library,1101 Broad Street, Chattanooga, Tennessee 37402

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

    Date of application for amendments: August 21, 1995 (TS 95-21)
    Brief description of amendments: The amendments change Technical 
Specification 3.7.5.c to allow an increase in the average essential raw 
cooling water supply header temperature from 84.5 deg.F to 87 deg.F 
untilSeptember 30, 1995.
    Date of issuance: September 13, 1995
    Effective date: September 13, 1995
    Amendment Nos.: 210 and 200
    Facility Operating License Nos. DPR-77 and DPR-79: Amendments 
revise the technical specifications.Public comments requested as to 
proposed no significant hazards consideration: Yes (August 28, 1995, 60 
FR 44517). That notice provided an opportunity to submit comments on 
the Commission's proposed no significant hazards determination. No 
comments have been received. The notice also provided an opportunity to 
request a hearing, by September 12, 1995, but indicated that if the 
Commission makes a final no significant hazards consideration 
determination before the expiration of the notice period, any such 
hearing would take place after issuance of the amendments.The 
Commission's related evaluation of the amendment, finding of exigent 
circumstances, and final determination of no significant hazards 
consideration are contained in a Safety Evaluation dated September 13, 
1995.
    Local Public Document Room location: Chattanooga-Hamilton County 
Library,1101 Broad Street, Chattanooga, Tennessee 37402

The Cleveland Electric Illuminating Company, Centerior Service 
Company, Duquesne Light Company, Ohio Edison Company, Pennsylvania 
Power Company, Toledo Edison Company, Docket No. 50-440, Perry 
Nuclear Power Plant, Unit No. 1, Lake County, Ohio

    Date of application for amendment: April 28, 1995
    Brief description of amendment: The amendment extends for one 
additional operating cycle the exception to Limiting Condition for 
Operation 3.0.4 as it applies to the main steam isolation 

[[Page 49961]]
valve leakage control system Technical Specification.
    Date of issuance: September 8, 1995
    Effective date: September 8, 1995
    Amendment No.: 71
    Facility Operating License No. NPF-58: This amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: May 23, 1995 (60 FR 
27344)The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated September 8, 1995. No significant hazards 
consideration comments received: No
    Local Public Document Room location: Perry Public Library, 3753 
Main Street, Perry, Ohio 44081

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

    Date of application for amendment:  April 10, 1995
    Brief description of amendment: This amendment changes auxiliary 
feedwater system, motor driven feedwater pump, and condensate system 
Technical Specifications to increase clarity and changes format to more 
closely follow improved standard technical specifications and increases 
content of Bases discussions.
    Date of issuance: September 5, 1995
    Effective date: September 5, 1995
    Amendment No.: 200
    Facility Operating License No. NPF-3. Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: August 2, 1995 (60 FR 
39453) The Commission's related evaluation of the amendment is 
contained in a Safety Evaluation dated September 5, 1995.No significant 
hazards consideration comments received: No
    Local Public Document Room location: University of Toledo Library, 
Documents Department, 2801 Bancroft Avenue, Toledo, Ohio 43606

Virginia Electric and Power Company, Docket Nos. 50-280 and 50-281, 
Surry Power Station, Unit Nos. 1 and 2, Surry County, Virginia.

    Date of application for amendments: November 10, 1994
    Brief description of amendments: Clarify the surveillance 
requirementsfor the Reactor Protection and Engineered Safeguards 
Systems instrumentation and actuation logic.
    Date of issuance: September 14, 1995
    Effective date: September 14, 1995
    Amendment Nos.: 205 and 205
    Facility Operating License Nos. DPR-32 and DPR-37: Amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: April 12, 1995 (60 FR 
18630) The Commission's related evaluation of the amendment is 
contained in a Safety Evaluation dated September 14, 1995.No 
significant hazards consideration comments received: No
    Local Public Document Room location: Swem Library, College of 
William and Mary, Williamsburg, Virginia 23185

Virginia Electric and Power Company, Docket Nos. 50-280 and 50-281, 
Surry Power Station, Unit Nos. 1 and 2, Surry County, Virginia.

    Date of application for amendments: July 14, 1995
    Brief description of amendments: These amendments would provide a 
2-hour allowed outage time for one residual heat removal loop to 
accommodate plant safety and emergency power systems surveillance 
testing, permit depressurizing safety injection accumulators in lieu of 
accumulator isolation, and make administrative changes.
    Date of issuance: September 1, 1995
    Effective date: September 1, 1995
    Amendment Nos.: 204 and 204
    Facility Operating License Nos. DPR-32 and DPR-37: Amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: August 2, 1995 (60 FR 
39455) The Commission's related evaluation of the amendment is 
contained in a Safety Evaluation dated September 1, 1995.No significant 
hazards consideration comments received: No
    Local Public Document Room location: Swem Library, College of 
William and Mary, Williamsburg, Virginia 23185

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

    Date of application for amendment: December 16, 1994.
    Brief description of amendment: The amendment revises the Kewaunee 
Nuclear Power Plant (KNPP) Technical Specifications Sections 3.4 and 
4.1 by removing the limiting conditions for operation (LCO) and the 
surveillance requirements for the turbine overspeed protection system 
(TOPS). The TOPS requirements will be relocated to the Updated Safety 
Analysis Report (USAR).
    Date of issuance: August 31, 1995
    Effective date: August 31, 1995
    Amendment No.: 121
    Facility Operating License No. DPR-43. Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: January 18, 1995 (60 FR 
3676). The Commission's related evaluation of the amendment is 
contained in a Safety Evaluation dated August 31, 1995.No significant 
hazards consideration comments received: No
    Local Public Document Room location: University of Wisconsin 
Library Learning Center, 2420 Nicolet Drive, Green Bay, Wisconsin 54301

Notice Of Issuance Of Amendments To Facility Operating Licenses And 
Final Determination Of No Significant Hazards Consideration And 
Opportunity For A Hearing (Exigent Public Announcement Or Emergency 
Circumstances)

    During the period since publication of the last biweekly notice, 
the Commission has issued the following amendments. The Commission has 
determined for each of these amendments that the application for the 
amendment complies with the standards and requirements of the Atomic 
Energy Act of 1954, as amended (the Act), and the Commission's rules 
and regulations. The Commission has made appropriate findings as 
required by the Act and the Commission's rules and regulations in 10 
CFR Chapter I, which are set forth in the license amendment.
    Because of exigent or emergency circumstances associated with the 
date the amendment was needed, there was not time for the Commission to 
publish, for public comment before issuance, its usual 30-day Notice of 
Consideration of Issuance of Amendment, Proposed No Significant Hazards 
Consideration Determination, and Opportunity for a Hearing.
    For exigent circumstances, the Commission has either issued a 
Federal Register notice providing opportunity for public comment or has 
used local media to provide notice to the public in the area 
surrounding a licensee's facility of the licensee's application and of 
the Commission's proposed determination of no significant hazards 
consideration. The Commission has provided a reasonable opportunity for 
the public to comment, using its best efforts to make available to the 
public means of communication for the public to respond quickly, and in 
the case of telephone comments, the comments have been recorded or 
transcribed as appropriate and the licensee has been informed of the 
public comments.
    In circumstances where failure to act in a timely way would have 
resulted, for 

[[Page 49962]]
example, in derating or shutdown of a nuclear power plant or in 
prevention of either resumption of operation or of increase in power 
output up to the plant's licensed power level, the Commission may not 
have had an opportunity to provide for public comment on its no 
significant hazards consideration determination. In such case, the 
license amendment has been issued without opportunity for comment. If 
there has been some time for public comment but less than 30 days, the 
Commission may provide an opportunity for public comment. If comments 
have been requested, it is so stated. In either event, the State has 
been consulted by telephone whenever possible.
    Under its regulations, the Commission may issue and make an 
amendment immediately effective, notwithstanding the pendency before it 
of a request for a hearing from any person, in advance of the holding 
and completion of any required hearing, where it has determined that no 
significant hazards consideration is involved.
    The Commission has applied the standards of 10 CFR 50.92 and has 
made a final determination that the amendment involves no significant 
hazards consideration. The basis for this determination is contained in 
the documents related to this action. Accordingly, the amendments have 
been issued and made effective as indicated.
    Unless otherwise indicated, the Commission has determined that 
these amendments satisfy the criteria for categorical exclusion in 
accordance with 10 CFR 51.22. Therefore, pursuant to 10 CFR 51.22(b), 
no environmental impact statement or environmental assessment need be 
prepared for these amendments. If the Commission has prepared an 
environmental assessment under the special circumstances provision in 
10 CFR 51.12(b) and has made a determination based on that assessment, 
it is so indicated.
    For further details with respect to the action see (1) the 
application for amendment, (2) the amendment to Facility Operating 
License, and (3) the Commission's related letter, Safety Evaluation 
and/or Environmental Assessment, as indicated. All of these items are 
available for public inspection at the Commission's Public Document 
Room, the Gelman Building, 2120 L Street, NW., Washington, DC, and at 
the local public document room for the particular facility involved.
    The Commission is also offering an opportunity for a hearing with 
respect to the issuance of the amendment. By October 27, 1995, the 
licensee may file a request for a hearing with respect to issuance of 
the amendment to the subject facility operating license and any person 
whose interest may be affected by this proceeding and who wishes to 
participate as a party in the proceeding must file a written request 
for a hearing and a petition for leave to intervene. Requests for a 
hearing and a petition for leave to intervene shall be filed in 
accordance with the Commission's ``Rules of Practice for Domestic 
Licensing Proceedings'' in 10 CFR Part 2. Interested persons should 
consult a current copy of 10 CFR 2.714 which is available at the 
Commission's Public Document Room, the Gelman Building, 2120 L Street, 
NW., Washington, DC and at the local public document room for the 
particular facility involved. If a request for a hearing or petition 
for leave to intervene is filed by the above date, the Commission or an 
Atomic Safety and Licensing Board, designated by the Commission or by 
the Chairman of the Atomic Safety and Licensing Board Panel, will rule 
on the request and/or petition; and the Secretary or the designated 
Atomic Safety and Licensing Board will issue a notice of a hearing or 
an appropriate order.
    As required by 10 CFR 2.714, a petition for leave to intervene 
shall set forth with particularity the interest of the petitioner in 
the proceeding, and how that interest may be affected by the results of 
the proceeding. The petition should specifically explain the reasons 
why intervention should be permitted with particular reference to the 
following factors: (1) the nature of the petitioner's right under the 
Act to be made a party to the proceeding; (2) the nature and extent of 
the petitioner's property, financial, or other interest in the 
proceeding; and (3) the possible effect of any order which may be 
entered in the proceeding on the petitioner's interest. The petition 
should also identify the specific aspect(s) of the subject matter of 
the proceeding as to which petitioner wishes to intervene. Any person 
who has filed a petition for leave to intervene or who has been 
admitted as a party may amend the petition without requesting leave of 
the Board up to 15 days prior to the first prehearing conference 
scheduled in the proceeding, but such an amended petition must satisfy 
the specificity requirements described above.
    Not later than 15 days prior to the first prehearing conference 
scheduled in the proceeding, a petitioner shall file a supplement to 
the petition to intervene which must include a list of the contentions 
which are sought to be litigated in the matter. Each contention must 
consist of a specific statement of the issue of law or fact to be 
raised or controverted. In addition, the petitioner shall provide a 
brief explanation of the bases of the contention and a concise 
statement of the alleged facts or expert opinion which support the 
contention and on which the petitioner intends to rely in proving the 
contention at the hearing. The petitioner must also provide references 
to those specific sources and documents of which the petitioner is 
aware and on which the petitioner intends to rely to establish those 
facts or expert opinion. Petitioner must provide sufficient information 
to show that a genuine dispute exists with the applicant on a material 
issue of law or fact. Contentions shall be limited to matters within 
the scope of the amendment under consideration. The contention must be 
one which, if proven, would entitle the petitioner to relief. A 
petitioner who fails to file such a supplement which satisfies these 
requirements with respect to at least one contention will not be 
permitted to participate as a party.
    Those permitted to intervene become parties to the proceeding, 
subject to any limitations in the order granting leave to intervene, 
and have the opportunity to participate fully in the conduct of the 
hearing, including the opportunity to present evidence and cross-
examine witnesses. Since the Commission has made a final determination 
that the amendment involves no significant hazards consideration, if a 
hearing is requested, it will not stay the effectiveness of the 
amendment. Any hearing held would take place while the amendment is in 
effect.
    A request for a hearing or a petition for leave to intervene must 
be filed with the Secretary of the Commission, U.S. Nuclear Regulatory 
Commission, Washington, DC 20555, Attention: Docketing and Services 
Branch, or may be delivered to the Commission's Public Document Room, 
the Gelman Building, 2120 L Street, NW., Washington, DC, by the above 
date. Where petitions are filed during the last 10 days of the notice 
period, it is requested that the petitioner promptly so inform the 
Commission by a toll-free telephone call to Western Union at 1-(800) 
248-5100 (in Missouri 1-(800) 342-6700). The Western Union operator 
should be given Datagram Identification Number N1023 and the following 
message addressed to (Project Director): petitioner's name and 
telephone number, date petition was mailed, plant name, and publication 
date and page number of this Federal Register notice. A copy of the 
petition should also be sent to the Office of the General Counsel, U.S. 
Nuclear 

[[Page 49963]]
Regulatory Commission, Washington, DC 20555, and to the attorney for 
the licensee.
    Nontimely filings of petitions for leave to intervene, amended 
petitions, supplemental petitions and/or requests for a hearing will 
not be entertained absent a determination by the Commission, the 
presiding officer or the Atomic Safety and Licensing Board that the 
petition and/or request should be granted based upon a balancing of the 
factors specified in 10 CFR 2.714(a)(1)(i)-(v) and 2.714(d).

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

    Date of application for amendment: August 28, 1995
    Brief description of amendment: The amendment revises Primary 
Containment Purge System Technical Specification Section 3.6.1.7, 
Limiting Condition for Operation. The revision extends the amount of 
time the 12-inch and 14-inch purge system supply and exhaust lines may 
be used for venting or purging from 90 to 135 hours per 365 days. In 
addition, expired footnotes were deleted as an editorial change and the 
associated Bases section was revised.
    Date of issuance: August 31, 1995
    Effective date: As of the date of issuance to be implemented upon 
receipt.
    Amendment No.: 68
    Facility Operating License No. NPF-69: Amendment revises the 
Technical Specifications.Public comments requested as to proposed no 
significant hazards consideration: NoThe Commission's related 
evaluation of the amendment, emergency circumstances and consultation 
with the State, and final determination of no significant hazards 
consideration are contained in a Safety Evaluation dated August 31, 
1995.
    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: Ledyard B. Marsh
    For the Nuclear Regulatory Commission
John N. Hannon,
Acting Director, Division of Reactor Projects - III/IV Office of 
Nuclear Reactor Regulation
[Doc. 95-23806 Filed 9-26-95; 8:45 am]
BILLING CODE 7590-01-F