[Federal Register Volume 60, Number 177 (Wednesday, September 13, 1995)]
[Notices]
[Pages 47613-47630]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: 95-22616]



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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 18, 1995, through August 30, 1995. 
The last biweekly notice was published on Wednesday, August 30, 1995 
(60 FR 45172). 

[[Page 47614]]


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

    The Commission has made a proposed determination that the following 
amendment requests involve no significant hazards consideration. Under 
the Commission's regulations in 10 CFR 50.92, this means that operation 
of the facility in accordance with the proposed amendment would not (1) 
involve a significant increase in the probability or consequences of an 
accident previously evaluated; or (2) create the possibility of a new 
or different kind of accident from any accident previously evaluated; 
or (3) involve a significant reduction in a margin of safety. The basis 
for this proposed determination for each amendment request is shown 
below.
    The Commission is seeking public comments on this proposed 
determination. Any comments received within 30 days after the date of 
publication of this notice will be considered in making any final 
determination.
    Normally, the Commission will not issue the amendment until the 
expiration of the 30-day notice period. However, should circumstances 
change during the notice period such that failure to act in a timely 
way would result, for example, in derating or shutdown of the facility, 
the Commission may issue the license amendment before the expiration of 
the 30-day notice period, provided that its final determination is that 
the amendment involves no significant hazards consideration. The final 
determination will consider all public and State comments received 
before action is taken. Should the Commission take this action, it will 
publish in the Federal Register a notice of issuance and provide for 
opportunity for a hearing after issuance. The Commission expects that 
the need to take this action will occur very infrequently.
    Written comments may be submitted by mail to the 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 13, 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 

[[Page 47615]]
Document Room, the Gelman Building, 2120 L Street, NW., Washington DC, 
by the above date. Where petitions are filed during the last 10 days of 
the notice period, it is requested that the petitioner promptly so 
inform the Commission by a toll-free telephone call to Western Union at 
1-(800) 248-5100 (in Missouri 1-(800) 342-6700). The Western Union 
operator should be given Datagram Identification Number N1023 and the 
following message addressed to (Project Director): petitioner's name 
and telephone number, date petition was mailed, plant name, and 
publication date and page number of this Federal Register notice. A 
copy of the petition should also be sent to the Office of the General 
Counsel, U.S. Nuclear Regulatory Commission, Washington, DC 20555, 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, Docket No. 50-261, H. B. Robinson 
Steam Electric Plant, Unit No. 2, Darlington County, South Carolina

    Date of amendment request: July 17, 1995
    Description of amendment request: The requested change to Technical 
Specification (TS) section 3.8 would specify that the spent fuel 
building refueling filter fan and at least one containment purge fan 
shall be shown to operate within plus or minus 10 percent of the design 
flow.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:The proposed change to TS is to 
revise Section 3.8.2.c. This TS section currently states ``All filter 
system fans shall be shown to operate within [plus or minus] 10% of the 
design flow.'' The proposed requirements are as follows:
    c.1 The Spent Fuel Building refueling filter fan shall be shown 
to operate within [plus or minus] 10% of the design flow.
    c.2 At least one Containment purge filter fan shall be shown to 
operate within [plus or minus] 10% of the design flow and must be 
operable during core alterations or movement of irradiated fuel 
assemblies, or at least one automatic containment isolation valve in 
each line penetrating the containment which provides a direct path 
from the containment atmosphere to the outside atmosphere shall be 
securely closed.
    This proposed change does not involve a significant hazards 
consideration for the following reasons.
    1. The proposed change does not involve a significant increase 
in the probability or consequences of an accident previously 
evaluated. The proposed change clarifies the operating requirements 
for the Containment purge and Spent Fuel Building refueling filter 
systems. This proposed change to the TS specifically delineates the 
fan filter systems required for refueling operations and does not 
change the physical operation of the filter systems. The affected 
systems are not involved in the initiation of any accident. The 
system response to previously analyzed accidents, including system 
flows and filter efficiencies will not be altered by the proposed 
change. These changes are enhancements to clarify existing TS 
requirements that will not increase the probability or consequences 
of a previously analyzed accident.
    2. The proposed change does not create the possibility of a new 
or different kind of accident from any accident previously 
evaluated. The proposed change merely clarifies the specific filter 
systems that are necessary to mitigate a fuel handling accident 
during core alterations or the movement of irradiated fuel 
assemblies and is consistent with the accident analysis in Section 
15.7.4 of the Updated Final Safety Analysis Report (UFSAR). This 
proposed change does not involve the addition or modification of 
plant equipment, nor does it alter the design or operation of plant 
systems. Therefore, operation of the facility in accordance with the 
proposed TS change will not create the possibility of a new or 
different kind of accident from any accident previously evaluated.
    3. The proposed change clarifies which filter systems that must 
be capable of mitigating a design basis fuel handling accident 
during core alterations or the movement of irradiated fuel 
assemblies and is consistent with the accident analysis in Section 
15.7.4 of the UFSAR. The proposed change will not result in an 
increase in the Control Room or offsite radiation doses. The 
performance of the filtration systems, including adsorption 
efficiencies, will not change. Therefore, the proposed change does 
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: Hartsville Memorial Library, 
147 West College Avenue, Hartsville, SC 29550
    Attorney for licensee: R. E. Jones, General Counsel, Carolina Power 
& Light Company, Post Office Box 1551, Raleigh, NC 27602
    NRC Project Director: David B. Matthews

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

    Date of amendment request: June 30, 1995
    Description of amendment request: The proposed amendments would 
modify the emergency diesel generator testing requirements in the 
Technical Specifications.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:
    1. The proposed changes do not involve a significant increase in 
the probability or consequences of occurrence of any accident 
previously evaluated.
    The proposed changes to the Technical Specifications will change 
the scope of EDG [Emergency Diesel Generator] testing that is 
performed on a refueling cycle frequency. The proposed changes will 
eliminate the requirement to perform sequenced loading of the EDG as 
part of the hot restart test, and will allow the hot restart test to 
be initiated from any EDG start signal. The revised requirements 
will eliminate testing that is redundant, provides no additional 
meaningful information, significantly constrains scheduling of 
refueling outage maintenance and testing, and impacts the 
availability of systems and components important to safety. The 
proposed testing requirements satisfy the underlying purpose of the 
EDG hot restart test. The testing in accordance with the proposed 
requirements will verify the ability of each EDG to complete the 
start up sequence from an equilibrium temperature immediately 
following operation at full load for a period of time long enough to 
stabilize operating temperature.
    A two hour period for operation at full load has been chosen to 
ensure that full load operating temperature has stabilized prior to 
shutdown preceding the hot restart test. Momentary transients 
outside the full load operating band of 3600 to 4000 kW will not 
invalidate the two hour run since momentary transient will not 
significantly affect operating temperature. Brief operation 
subsequent to a momentary transient will normalize operating 
temperature. Since the proposed changes impact only surveillance 
requirements used to periodically verify the operability of a 
required safety system, and since the proposed changes provide an 

[[Page 47616]]
equivalent level of testing and eliminate redundant testing, the 
proposed changes will not impact the operability or availability of 
a required system.
    Operation in accordance with the revised requirements will not 
increase the likelihood that a transient initiating event will occur 
since transients are initiated by equipment malfunction and/or 
catastrophic system failure. The revised requirements affect testing 
that is performed on a Refueling Cycle frequency. Testing in 
accordance with the proposed requirements will not increase the 
probability of failure of the EDGs since the testing will provide an 
equivalent level of testing to verify the operability of the EDGs. 
In addition, failure of an EDG to start or failure of an EDG while 
operating is not assumed to be an initiating event of an accident 
considered in the Updated Final Safety Analysis Report (UFSAR). 
Based on the above, operation in accordance with the proposed 
requirements will not significantly increase the probability of 
occurrence of any accident previously evaluated.
    The proposed requirements will meet the underlying purposed of 
the existing testing requirements. The proposed testing will ensure 
the ability of the EDG to start from a hot condition in the unlikely 
event of an accident. The proposed changes will eliminate testing 
requirements that are redundant and unnecessarily challenge the 
reliability of the EDGs by requiring unnecessary wear and cycling of 
the diesel engine and auxiliary systems. Since the proposed changes 
will not adversely affect the operability or availability of the 
EDGs, the ability of the EDGs to operate and power equipment 
important to safety will not be impacted and the ability to mitigate 
the consequences of accidents previously evaluated will not be 
affected. Based on the preceding discussion, the consequences of 
accidents previously evaluated will not significantly increase.
    2. The proposed changes do not create the possibility of a new 
or different kind of accident from any accident previously 
evaluated.
    The proposed changes to the Technical Specifications do not 
involve the addition of any new or different types of safety related 
equipment, nor do they involve the operation of equipment required 
for safe operation of the facility in a manner different from those 
addressed in the UFSAR. No safety related equipment or function will 
be altered as a result of the proposed changes. Also, the procedures 
that govern normal operation and recovery from an accident are not 
affected by the proposed changes.
    The proposed changes will eliminate testing requirements that 
are redundant and provide no additional meaningful information. 
Testing in accordance with the revised requirements will provide an 
equivalent level of confidence in the reliability of the EDG systems 
to complete the start up sequence from a hot condition. The proposed 
testing requirements satisfy the purpose Regulatory Guide 1.108 in 
that the testing requirements will ensure EDG operability and 
reliability. In addition, the proposed changes are consistent with 
the changes recommended by the NRC in Generic Letter 93-05. Since no 
new failure modes or mechanisms are introduced by the proposed 
changes, the possibility of a new or different kind of accident is 
not created.
    3. The proposed changes do not involve a significant reduction 
in a margin of safety.
    Plant safety margins are established through LCOs, limiting 
safety system settings, and safety limits specified in the Technical 
Specifications. There will be no changes to either the physical 
design of the plant or to any of these settings or limits as a 
result of the proposed changes. The proposed changes will eliminate 
testing requirements that are redundant and provide no additional 
information. Testing in accordance with the revised requirements 
will verify the ability of the EDGs to complete the start up 
sequence from a hot condition as is intended by the recommended 
testing in Regulatory Guide 1.108. In addition, the proposed changes 
are consistent with the changes recommended by the NRC in Generic 
Letter 93-05. Since the proposed changes will not impact the 
availability or operability of the EDGs to perform their intended 
function and since no LCOs, safety limits, or safety system settings 
are affected by the proposed changes, there is no significant 
reduction in a margin of safety.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
requested amendments involve no significant hazards consideration.
    Local Public Document Room location: Waukegan Public Library, 128 
N. County Street, Waukegan, IL 60085
    Attorney for licensee: Michael I. Miller, Esquire; Sidley and 
Austin, One First National Plaza, Chicago, IL 60603
    NRC Project Director: Robert A. Capra

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

    Date of amendment request: July 26, 1995.
    Description of amendment request: The licensee proposes to change 
Turkey Point Units 3 and 4 Technical Specifications to allow rod 
misalignment of +/- 18 steps at or below 90% of rated thermal power. In 
addition, a change is proposed to the Limiting Condition for Operation 
range of rod travel from 228 to ``All Rods Out.'' The introduction of 
``All Rods Out'' is consistent with Amendment 167/161 which approved 
the removal of Technical Specification 3.1.3.6, ``Rod Insertion Limit'' 
from the Technical Specifications and placement into the Core Operating 
Limits Report (COLR).
    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 any accident previously evaluated.
    The proposed limits on rod misalignment do not increase the 
probability of an accident. The Technical Specifications' allowed 
increase in peaking factor limits as power is reduced accommodates 
an increase in rod misalignment of [plus or minus] 18 steps below 
90% of RTP [rated thermal power]. The initial conditions remain 
unchanged from that assumed in the Updated Final Safety Analysis 
Report (UFSAR). Therefore, this proposed change poses no significant 
increase in the probability or consequences of any 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.
    No new accident scenarios, failure mechanisms or limiting single 
failure are introduced as a result of implementing the proposed rod 
misalignment criteria. The institution of the proposed rod 
misalignment criteria will have no adverse effect, nor does it 
challenge, the performance of any other safety related system. 
Therefore, the proposed amendment does not in any way 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 the margin of 
safety.
    Operation of the facility in accordance with the proposed 
amendment would not involve a significant reduction in the margin of 
safety. The margin of safety, as defined in the BASES for the 
Technical Specifications, is not significantly affected by the 
changes to the rod misalignment limit. The Technical Specifications' 
allowed increase in peaking factor limits as power is reduced 
accommodates an increase in rod misalignment of [plus or minus] 18 
steps below 90% of RTP. The initial conditions remain unchanged from 
that assumed in the UFSAR. Since the peaking factor limits are not 
modified, 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 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: Florida International 
University, University Park, Miami, FL 33199 

[[Page 47617]]

    Attorney for licensee: J. R. Newman, Esquire, Morgan, Lewis & 
Bockius, 1800 M Street, NW., Washington, DC 20036
    NRC Project Director: David B. Matthews

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

    Date of amendment request: July 26, 1995.
    Description of amendment request: The licensee proposes to change 
Turkey Point Units 3 and 4 Technical Specifications to delete the 
requirement to adjust the Nuclear Instrumentation System (NIS) downward 
when operating at less than 70% of rated thermal power (RTP).
    At reduced power levels (i.e., less than 70% of RTP), calorimetric 
power measurement uncertainties are most influenced by the feedwater 
flow measurements, which have the potential for large flow 
uncertainties under low flow conditions. These calorimetric 
uncertainties create the potential for a non-conservative gain 
adjustment of the NIS when the NIS is adjusted downward to match 
calorimetric power at reduced power levels, and may result in a non-
conservative NIS power level indication when operating at higher power 
levels. Inappropriate gain adjustments could cause the NIS Power Range 
High Neutron Flux trip to occur at power levels beyond that assumed in 
the plant safety analyses. The proposed changes would correct this 
situation.
    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 change does not involve any physical changes to the 
NIS. Implementation of the proposed change does not affect the 
probability of failure of the NIS and does not alter the method in 
which protection is afforded by the NIS for the reactor and primary 
system. Therefore, the proposed change does not result in an 
increase in the severity or consequences of any accident previously 
evaluated.
    The proposed change in Technical Specifications to remove the 
requirement which could result in non-conservative gain adjustments 
of the NIS at reduced power levels (below 70% of RTP), will have no 
significant effect on the probability or consequences of licensing 
basis events; and the probability or consequences of an accident 
previously evaluated for Turkey Point has not been significantly 
increased. Therefore, 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.
    (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 change does not result in a change in the method in 
which the NIS provides plant protection. No change is being made 
which alters the function of the NIS. Therefore, the proposed change 
does not create the possibility of a new or different kind of 
accident nor involve a reduction in a margin of safety as defined in 
the Safety Analysis Report.
    The change in Technical Specifications associated with the 
removal of the requirement which could result in non-conservative 
gain adjustments of the NIS at reduced power levels (below 70% of 
RTP) will 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 the margin of 
safety.
    This change in Technical Specifications only affects the removal 
of the requirement which has the potential for non-conservative gain 
adjustments of the NIS at reduced power levels (below 70% of RTP); 
these changes do not alter the manner in which protection is 
afforded for the reactor and primary system. In addition, the 
fundamental process for implementation of the calorimetric power/NIS 
comparison remains the same.
    The changes in Technical Specifications associated with the 
removal of the requirement, which could lead to non-conservative 
gain adjustments of the NIS at reduced power levels (below 70% of 
RTP), will not involve a significant reduction in the margin of 
safety.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: Florida International 
University, University Park, Miami, FL 33199
    Attorney for licensee: J. R. Newman, Esquire, Morgan, Lewis & 
Bockius, 1800 M Street, NW., Washington, DC 20036
    NRC Project Director: David B. Matthews

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

    Date of amendment request: July 26, 1995.
    Description of amendment request: The licensee proposes to change 
Turkey Point Units 3 and 4 Technical Specifications (TS) to incorporate 
certain changes which are consistent with guidance provided by NUREG-
1366 and NRC Generic Letter (GL) 93-05, ``Line-Item Technical 
Specification Improvements to Reduce Surveillance Requirements for 
Testing During Power Operation.'' The following proposed changes are 
requested:
    (1) TS SR 4.1.3.1.2: Change the frequency interval for control rod 
movement test from monthly to quarterly.
    (2) TS SR 4.6.5.1: Change the hydrogen monitor calibration from 
quarterly to each refueling interval, and the analog channel 
operational test from monthly to quarterly.
    (3) TS SR Table 4.3-3: Change the analog channel functional test 
from monthly to quarterly for radiation monitors. Correct spelling of 
'Radioactivity' in Item 1.a.
    (4) TS SR 4.4.6.2.2: Increase the time allowed in COLD SHUTDOWN 
before leak testing the Reactor Coolant System (RCS) isolation valves 
is required, from 72 hours to 7 days.
    (5) TS SR 4.10.1.2: Changes the requirement for a rod drop test 
prior to reducing SHUTDOWN MARGIN from ``within 24 hours'' to ``within 
7 days''.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:
    (1) Operation of the facility in accordance with the proposed 
amendments would not involve a significant increase in the 
probability or consequences of an accident previously evaluated.
    The proposed amendments do not involve a significant increase in 
the probability or consequences of an accident previously evaluated 
because the proposed amendments conform to the guidance given in 
Enclosure 1 of the NRC Generic Letter 93-05. The overall functional 
capabilities of the rod control system, RCS pressure isolation 
valves, the hydrogen monitoring system, and the radiation monitoring 
systems will not be modified by the proposed change. These 
amendments will not involve a significant increase in the 
probability or consequences of an accident previously evaluated for 
the following reasons:
    (1) Increasing the interval of control rod movement testing will 
reduce the possibility of testing-related reactor trips and dropped 
rods, and result in fewer challenges to safety systems and plant 
transients.
    (2) Increasing the interval of hydrogen monitor calibration and 
operational tests will result in a reduction in equipment 
degradation and reduce a burdensome task on personnel resources.

[[Page 47618]]

    (3) Increasing the interval of radiation monitor functional 
tests will result in less equipment degradation as well as reducing 
the potential for testing-related isolations of the control room, 
fuel handling building, auxiliary buildings, and various process 
lines.
    (4) Increasing the time allowed in COLD SHUTDOWN prior to leak 
testing RCS isolation valves will permit plant personnel to focus on 
short notice outage recovery and minimize personnel radiation 
exposure. Since the methods and the acceptance criteria used for the 
leak test are not altered, increasing the time from 72 hours to 7 
days will not significantly alter the associated risk.
    (5) Increasing the time required to perform rod tests prior to 
reducing the SHUTDOWN MARGIN will result in only one rod drop test 
vice two following a refueling outage, which will in turn reduce 
plant transients and personnel resource requirements.
    (2) Operation of the facility in accordance with the proposed 
amendments would not create the possibility of a new or different 
kind of accident from any accident previously evaluated.
    The use of the proposed changes to the TS can not create the 
possibility of a new or different kind of accident from any accident 
previously evaluated since the proposed amendments will not change 
the physical plant or the modes of plant operation defined in the 
facility operating license. No new failure mode is introduced due to 
the surveillance interval changes and clarifications, since the 
proposed changes do not involve the addition or modification of 
equipment nor do they alter the design or operation of affected 
plant systems.
    (3) Operation of the facility in accordance with the proposed 
amendments would not involve a significant reduction in a margin of 
safety.
    The operating limits and functional capabilities of the affected 
systems are unchanged by the proposed amendments. The proposed 
changes to the TS which establish new or clarify old surveillance 
intervals consistent with the NRC Generic Letter 93-05 line-item 
improvement guidance do not significantly reduce any of the margins 
of safety even though the number of surveillances is decreased. 
These requested amendments are justified by the following reasoning 
from NUREG-1366:
    (1) The surveillances could lead to plant transients which would 
challenge safety systems unnecessarily as in the cases of control 
rod movement tests and post-refueling rod drop tests.
    (2) The surveillances result in the unnecessary wear to 
equipment as in the cases of the hydrogen and radiation monitor 
surveillances.
    (3) The surveillance result in radiation exposure to plant 
personnel which is not justified by the safety significance of the 
surveillances as in the case of the time requirement for leak-
testing RCS isolation valves when in COLD SHUTDOWN.
    (4) The surveillances place an unnecessary burden on plant 
personnel because the time required is not justified by the safety 
significance of the surveillance, i.e. hydrogen monitor and post-
refueling rod drop tests.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: Florida International 
University, University Park, Miami, FL 33199
    Attorney for licensee: J. R. Newman, Esquire, Morgan, Lewis & 
Bockius, 1800 M Street, NW., Washington, DC 20036
    NRC Project Director: David B. Matthews

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

    Date of amendment request: July 26, 1995.
    Description of amendment request: The licensee proposes to change 
Turkey Point Units 3 and 4 Technical Specification Administrative 
Controls Section 6.9.1.7 to reflect the use of the Westinghouse NOTRUMP 
model in the Small Break Loss-of-Coolant Accident (SBLOCA) analysis 
used in determining the K(z) curve contained in the Core Operating 
Limits Report (COLR). The following references would be added to 
Section 6.9.1.7 (COLR) of the Administrative Controls section of Turkey 
Point Units 3 and 4 TS: WCAP-10054-P-A, (proprietary) and 
WCAP-10081-NP-A, (non-proprietary), ``Westinghouse Small Break ECCS 
Evaluation Model Using the NOTRUMP Code'', October, 1985.'' WCAP-10054-
P-A Addendum 2, (proprietary), ``Addendum to the Westinghouse Small 
Break ECCS Evaluation Model Using the NOTRUMP Code: Safety Injection 
into the Broken Loop and COSI Condensation Model'', August, 1994.''
    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 modification to the current Section 6.9.1.7 of the 
Administrative Controls section of the Turkey Point Technical 
Specifications to include the references to WCAP-10054-P-A, ``Small 
Break ECCS Evaluation Model Using the NOTRUMP Code'', and WCAP-
10054-P-A Addendum 2 for the COSI model, does not involve an 
increase in the probability or consequences of an accident 
previously evaluated. This modification to the Technical 
Specification does not change the probability of occurrence 
previously evaluated.
    This change does not affect the integrity of the fission product 
barriers utilized for mitigation of radiological dose consequences 
as a result of an accident. The addition of the new methodology used 
for Turkey Point uprating analysis does not change, degrade, or 
prevent the response of safety related mitigation systems to 
accident scenarios, as described in the Updated Final Safety 
Analysis Report (UFSAR) Chapter 14. Therefore, the licensee 
concluded that the probability or consequences of an accident 
previously evaluated are not increased.
    (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 modification to the current Section 6.9.1.7 of the 
Administrative Controls section of the Turkey Point Technical 
Specifications to include the references to WCAP-10054-P-A, ``Small 
Break ECCS Evaluation Model Using the NOTRUMP Code'', and WCAP-
10054-P-A Addendum 2 for the COSI model, will not create the 
possibility of a new or different kind of accident from any accident 
previously evaluated. No new operating configuration is being 
imposed by the addition of the references to the Technical 
Specification. Therefore, no new failure modes or limiting single 
failures have been identified. The licensee concludes that no new or 
different kind of accidents from those previously evaluated have 
been created as a result of this revision.
    (3) Operation of the facility in accordance with the proposed 
amendment would not involve a significant reduction in the margin of 
safety.
    The modification to the current Section 6.9.1.7 of the 
Administrative Controls section of the Turkey Point Technical 
Specifications to include the references for the Small Break ECCS 
Evaluation Model Using the NOTRUMP Code will not involve a reduction 
in the margin of safety. The SBLOCA analysis results show that the 
limits of 10 CFR 50.46 are maintained as follows. The new calculated 
value of worst-case PCT will be 1688 deg.F, which is less than the 
limit of 2200 deg.F. There is significant margin in the current 
SBLOCA analysis such that the total cladding oxidation limit of 17 
percent will not be challenged. Further, the calculated total amount 
of hydrogen generated has been determined to remain less than 1 
percent. The SBLOCA hydraulic forces are not affected by the K(z) 
curve and the licensee concludes that the core will remain amenable 
to cooling. Additionally, post-LOCA long term core cooling and hot 
leg switchover evaluations are not impacted by the K(z) curve. 
Therefore, there is no significant reduction in the margin of 
safety.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request 

[[Page 47619]]
involves no significant hazards consideration.
    Local Public Document Room location: Florida International 
University, University Park, Miami, FL 33199
    Attorney for licensee: J. R. Newman, Esquire, Morgan, Lewis & 
Bockius, 1800 M Street, NW., Washington, DC 20036
    NRC Project Director: David B. Matthews

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

    Date of amendment request: July 26, 1995.
    Description of amendment request: The licensee proposes to change 
Turkey Point Units 3 and 4 Technical Specifications to achieve 
consistency throughout these documents by (a) removing outdated 
material, (b) incorporating administrative clarifications and 
corrections, and (c) correcting typographical errors.
    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 do not involve a significant increase in 
the probability or consequences of an accident previously evaluated 
because the proposed amendments are purely administrative in nature. 
These amendments will not involve a significant increase in the 
probability or consequences of an accident previously evaluated 
because they 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, the proposed changes do not affect the 
probability or consequences of accidents previously analyzed.
    (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 use of the modified specifications can not create the 
possibility of a new or different kind of accident from any 
previously evaluated since the proposed amendments will not change 
the physical plant or the modes of plant operation defined in the 
facility operating license. No new failure mode is introduced due to 
the administrative changes and clarifications, since the proposed 
changes do not involve the addition or modification of equipment nor 
do they alter the design or operation of affected plant systems, 
structures, or components.
    (3) Operation of the facility in accordance with the proposed 
amendment would not involve a significant reduction in the margin of 
safety.
    The operating limits and functional capabilities of the affected 
systems, structures, and components are unchanged by the proposed 
amendments. The modified specifications which correct administrative 
errors and clarify existing Technical Specification requirements do 
not significantly reduce any of the margins of safety.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: Florida International 
University, University Park, Miami, FL 33199
    Attorney for licensee: J. R. Newman, Esquire, Morgan, Lewis & 
Bockius, 1800 M Street, NW., Washington, DC 20036
    NRC Project Director: David B. Matthews

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: August 17, 1995
    Description of amendment request: The proposed amendment would 
allow the containment to be opened after about 11 days following 
shutdown during refueling and would redefine the operability 
requirements for selected engineered safety feature systems such that 
these systems are only required to be operable during the calculated 
decay period. The proposed changes will not remove requirements for 
systems to mitigate potential vessel draindown events, will not remove 
requirements for systems required for decay heat removal, and will 
continue to require high water level over the vessel during fuel 
movement. Programs are in place to close the containment, if needed, to 
address shutdown risk concerns.
    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 limits on recently irradiated fuel is used to 
establish operational conditions where specific activities represent 
situations where significant radioactive releases can be postulated. 
These operational conditions are consistent with the design basis 
analysis. Because the equipment affected by the revised operational 
conditions is not considered an initiator to any previously analyzed 
accident, inoperability of the equipment cannot increase the 
probability of any previously evaluated accident.
    The proposed applicability in conjunction with existing 
administrative controls on light loads, bounds the conditions of the 
current design basis fuel handling accident analysis. The analysis 
also concludes the limiting offsite radiological consequences are 
well within the acceptance criteria of NUREG 0800, Section 15.7.4 
and GDC 19. The analysis is also conducted in a conservative manner 
containing margins in the calculation of mechanical analysis, iodine 
inventory and iodine decontamination factor. Each of these 
conservatisms will further decrease the consequences. Therefore, the 
proposed changes do not significantly increase the probability or 
consequences of any previously evaluated accident.
    The proposed limits are used to establish operational conditions 
where specific activities represent situations where significant 
radioactive releases can be postulated. In addition, the changes to 
operation are consistent with previous limits -- only allowing 
increased flexibility after the radiological consequences are 
assured to remain within accepted limits. Therefore, these 
operational conditions are consistent with the design basis 
analysis. The proposed changes do not introduce any new modes of 
plant operation and do not involve physical modifications to the 
plant. Therefore, the proposed changes do not create the possibility 
of a new or different kind of accident from any previous analyzed.
    The revised limits are used to establish operational conditions 
where specific activities represent situations where significant 
radioactive release can be postulated. These operational conditions 
are consistent with the design basis analysis and are established 
such that the radiological consequences are at or below the current 
RBS licensing limit. Safety margins and analytical conservatisms 
have been evaluated and are well understood. Conservative methods of 
analysis are maintained through the use of accepted methodology and 
benchmarking the proposed methods to previous analysis. Margins are 
retained to ensure that the analysis adequately bounds all 
postulated event scenarios. The proposed change only eliminates some 
excess conservatism from the analysis.
    EOI has implemented NUMARC 91-06 guidelines for shutdown 
operations at RBS. Shutdown Operations Protection Plan and Primary-
Secondary Containment Integrity procedures presently include 
guidance for closure of the containment hatch and other significant 
opening in containment, in addition to the requirements contained in 
the license and design basis. This additional protection will 
enhance the ability to limit offsite effects.
    Acceptance limits for the fuel handling accident are provided in 
10CFR100 with additional guidance provided in NUREG 0800, Section 
15.7.4 Excess margin is the difference between the postulated doses 
and the corresponding licensing limit. In the 

[[Page 47620]]
initial review of River Bend Station for operation (NUREG-0989, Section 
15.7.4), the NRC accepted the design and analysis based on meeting 
the guideline dose limits of 10CFR100 and SRP 15.7.4. The proposed 
applicability continues to ensure that the whole-body and thyroid 
doses at the exclusion area and low population zone boundaries, as 
well as control room doses, are below the corresponding licensing 
limit. These margins are unchanged; therefore, the proposed changes 
do not involve a significant reduction in a margin of safety.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: Government Documents 
Department, Louisiana State University, Baton Rouge, LA 70803
    Attorney for licensee: Mark Wetterhahn, Esq., Winston & Strawn, 
1400 L Street, N.W., Washington, DC 20005
    NRC Project Director: William D. Beckner

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

    Date of amendment requests: June 20, 1995 (AEP:NRC:0692CX)
    Description of amendment requests: The proposed amendments would 
remove the requirements for fire protection systems from the licenses 
and the Technical Specifications (T/S) in accordance with the 
provisions and guidance of Generic Letters (GL) 86-10, ``Implementation 
of Fire Protection Requirements,'' 88-12, ``Removal of Fire Protection 
Requirements from Technical Specifications,'' and 93-07, Modification 
of the Technical Specification Administrative Control 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:
    We have evaluated the proposed T/S changes and have determined 
that the changes should involve no significant hazards consideration 
based on the criteria established in 10 CFR 50.92(c). Operation of 
CNP [Cook Nuclear Plant] in accordance with the proposed amendment 
will not satisfy any of the following criteria.
    (a) Involve a significant increase in the probability or 
consequences of an accident previously evaluated.
    The proposed changes are administrative in nature, in that it 
moves the T/Ss portion of the Fire Protection Program from the T/Ss 
to the UFSAR [Updated Final Safety Analysis Report] and the 
implementing procedures. This is accomplished by referencing in the 
UFSAR and the documents which address the Fire Protection Program in 
greater detail. Thus, the proposed changes will not revise the 
requirements for fire protection equipment operability, testing, or 
inspection, but only moves the T/Ss portion of the Fire Protection 
Program to implementing procedures.
    As fire protection requirements are only being relocated 
following the guidance of GLs 86-10, 88-12, and 93-07, the proposed 
changes do not involve a significant increase in the probability or 
consequences of an accident previously evaluated.
    (b) Create the possibility of a new or different kind of 
accident from any previously analyzed.
    The proposed changes do not involve any physical alteration of 
plant configurations, changes to setpoints, or operating parameters. 
[These] are administrative changes that retain the existing fire 
protection requirements and relocate these requirements from the T/S 
to the UFSAR; therefore, these changes do not create the possibility 
of a new or different kind of accident.
    (c) Involve a significant reduction in a margin of safety.
    The proposed changes follow guidance contained in GLs 86-10, 88-
12, and 93-07 for incorporating the Fire Protection Program into the 
UFSAR. A license condition will be implemented that will require 
that no changes can be made to the Fire Protection Program that will 
adversely affect the ability to achieve or maintain safe shutdown in 
the event of a fire without prior NRC approval. Therefore, the 
proposed changes do not involve a significant reduction in a margin 
of safety.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment requests involve no significant hazards consideration.
    Local Public Document Room location: Maud Preston Palenske Memorial 
Library, 500 Market Street, St. Joseph, Michigan 49085
    Attorney for licensee: Gerald Charnoff, Esq., Shaw, Pittman, Potts 
and Trowbridge, 2300 N Street, NW, Washington, DC 20037
    NRC Project Director: John N. Hannon

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 23, 1995
    Description of amendment request: The proposed amendment would 
revise Technical Specifications Section 3.8.1.1 and the Bases for 
Section 3/4.8. The proposed amendment would extend the Allowed Outage 
Time (AOT) for an Emergency Diesel Generator (EDG) from 72 hours to 7 
days.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration (SHC), which is presented below:
    ... NNECO concludes that these changes do not involve a 
significant hazards consideration since the proposed change 
satisfies the criteria of 10 CFR 50.92(c). That is, the proposed 
changes do not:
    1. Involve a significant increase in the probability or 
consequences of an accident previously analyzed.
    The EDGs supply backup power to the essential safety systems in 
the event of a Loss of Normal (offsite) Power. EDGs are not accident 
initiators. Therefore, this change does not involve an increase in 
the probability of any accident previously evaluated.
    Although the EDGs provide backup power to components that help 
mitigate the consequences of accidents previously evaluated, the 
extension in the AOT does not affect any of the assumptions used in 
the deterministic evaluations of these accidents. Thus, this change 
will not increase the consequences of any accident previously 
analyzed.
    The increase in the EDG AOT introduces the potential to increase 
the risk to the public since a longer time window provides an 
opportunity to perform additional preventive maintenance to the EDG 
while the plant is on-line. However, the extended AOT, by itself, 
does not necessarily increase risk. The increase in the risk depends 
on the total time during which an EDG was out of service and the 
other equipment that is concurrently out of service with the EDG. 
The total risk increase due to the EDG being out of service will not 
be significant since that risk increase is monitored and kept at 
acceptable levels in accordance with the risk monitor program.
    Based on the above, the proposal to extend the AOT for the EDGs 
(Technical Specification 3.8.1) 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.
    The proposed change to extend the AOT for the EDGs (Technical 
Specification 3.8.1) does not alter the physical design, 
configuration, or method of operation of the plant. Therefore, the 
proposal does 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.
    The proposed change to extend the AOT for the EDGs (Technical 
Specification 3.8.1) do not affect the Limiting Conditions for 
Operations or their bases. As a result, the deterministic analyses 
performed to establish the margin of safety are unaffected. Thus, 
the change does not involve a significant reduction in the margin of 
safety. 

[[Page 47621]]

    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 23, 1995
    Description of amendment request: The proposed amendment would 
extend the Allowed Outage Time (AOT) for an inoperable Safety Injection 
Tank (SIT) from 1 hour to 24 hours, unless the SIT is inoperable due to 
either boron concentration not within its limits or an inoperable level 
or pressure instrument. For these two special cases, the proposed 
change extends the AOT for an inoperable SIT to 72 hours. In addition, 
the proposed amendment clarifies the completion times and conditions 
for action statements and the criteria for surveillance requirements.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration (SHC), which is presented below:
    ... NNECO concludes that these changes do not involve a 
significant hazards consideration since the proposed change 
satisfies the criteria in 10 CFR 50.92(c). That is, the proposed 
changes do not:
    1. Involve a significant increase in the probability or 
consequences of an accident previously analyzed.
    The Safety Injection Tanks (SITs) are passive components in the 
Emergency Core Cooling System that mitigate the consequences of a 
Loss of Coolant Accident (LOCA). As such, the SITs are not accident 
initiators. Therefore, this change does not involve an increase in 
the probability of any accident previously evaluated.
    The increase in the AOT has the potential to increase the risk 
if it becomes necessary to stay on-line longer than one (1) hour 
with an inoperable SIT. However, the estimated risk impact is 
negligible.
    The SITs inject borated water into the reactor vessel (via the 
cold legs) during the blowdown phase of a large break LOCA. The 
introduction of the inventory of borated water from all four (4) 
SITs is needed to ensure adequate reflooding of the core (i.e., 
minimize core damage) until the Engineered Safety Feature (ESF) 
pumps can provide adequate core cooling. The SITs also provide 
makeup water for the Reactor Coolant System (RCS) for smaller break 
LOCAs. The extension of the AOT does not affect any of the 
assumptions used in the deterministic evaluations of these 
accidents. Thus, this change will not increase the consequences of 
any accident previously evaluated.
    The increased AOT extension to 72 hours, based solely on 
instrumentation (level and pressure) malfunction, also does not 
involve a significant increase in the consequences of an accident 
previously evaluated as endorsed by the NRC in NUREG-1366, 
``Improvements to Technical Specifications Surveillance 
Requirements.''
    The modification to the completion times and the modification of 
the Surveillance Requirements for volumetric changes in the SIT as a 
result of addition from the Refueling Water Storage Tank (RWST) also 
do not involve a significant increase in the consequences of any 
accident previously evaluated by the NRC in NUREG-1432, ``Standard 
Technical Specifications for Combustion Engineering Plants.''
    Based on the above, the proposed changes to extend the AOT for 
an inoperable SIT, clarify action statements, and modify the 
criteria for surveillance requirements, do 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.
    The proposed changes to extend the AOT for an inoperable SIT, 
clarify action statements, and modify the criteria for surveillance 
requirements, do not alter the physical design, configuration, or 
method of operation of the plant. Therefore, the proposal does 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.
    The proposed changes to extend the AOT for an inoperable SIT, 
clarify action statements, and modify the criteria for surveillance 
requirements, do not affect the Limiting Conditions for Operations 
(LCOs) of the SITs or the bases of the LCOs. As a result, the 
deterministic analyses performed to establish the margin of safety 
are unaffected. Thus, the 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

Northern States Power Company, Docket Nos. 50-282 and 50-306, 
Prairie Island Nuclear Generating Plant, Unit Nos. 1 and 2, Goodhue 
County, Minnesota

    Date of amendment requests: May 4, 1995.
    Description of amendment requests: The proposed amendments would 
revise the pressurizer and main steam safety valve lift setting 
tolerances from plus or minus 1% to plus or minus 3%, revise the Safety 
Limit curves and revise the Technical Specification Section 2 to 
conform to Standard Technical Specifications.
    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 will not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated
    The proposed changes increase the ``as-found'' setpoint 
tolerances for the Pressurizer Safety Valves and Main Steam Safety 
Valves from [plus or minus] 1% to [plus or minus] 3%. The proposed 
changes do not involve any hardware modifications to plant 
structures, systems, or components. Analyses have determined that 
the proposed changes do not significantly affect the structural 
integrity of either the Reactor Coolant System or the Main Steam 
system.
    The proposed setpoint tolerance of [plus or minus] 3% was 
included in the assumptions for the performance of the reload safety 
evaluations for the current fuel cycles, PI1-17 and PI2-16, and 
subsequent Prairie Island fuel cycle analyses. These analyses 
concluded that the minimum acceptable DNBR [departure from nucleate 
boiling ratio] is maintained, over-pressure protection is 
maintained, LOCA [loss-of-coolant accident] acceptance criteria are 
met and offsite dose limits are not exceeded. These changes are 
consistent with the guidance provided by Section III and XI of the 
ASME [American Society of Mechanical Engineers] Code and Standard 
Technical Specifications.
    The proposed change to Technical Specification Figure TS.2.1-1 
does not affect any existing accident analyses. This revision 
ensures that the design bases and safety limits are accurately and 
appropriately reflected in the Technical Specifications and will 
ensure that plant operations are properly evaluated for DNBR 
encroachment.
    Therefore, the probability or consequences of an accident 
previously evaluated are not affected by any of the proposed 
amendments.
    2. The proposed amendment will not create the possibility of a 
new of different 

[[Page 47622]]
kind of accident from any accident previously analyzed The lift 
setpoint the Pressurizer Safety Valves and Main Steam Safety Valves 
will be restored to [plus or minus] 1% following testing, thus the 
``as-left'' setpoint tolerance for the Pressurizer Safety Valves and 
Main Steam Safety Valves are unchanged. Evaluations of plant normal 
operation, transient and accident conditions have been performed 
assuming these safety valve lift settings are [plus or minus] 3% of 
the nominal setpoint and demonstrated that new or different kinds of 
accidents are not created by the proposed changes.
    The proposed changes to Technical Specification Figure TS.2.1-1 
do not affect the design, function or operation of any Prairie 
Island structures, systems or components. The curves show the loci 
of points of reactor core differential temperature (an indication of 
thermal power), Reactor Coolant System pressure, and average 
temperature for which the minimum DNBR is not less than the safety 
analysis limit, that fuel centerline temperature remains below 
melting, that the average enthalpy in the hot leg is less than or 
equal to enthalpy of saturated liquid, or that the exit quality is 
within the limits defined by the applicable DNBR correlation. There 
are no new failure modes introduced by the proposed changes to the 
Figure. The changes conservatively adjust Figure TS.2.1-1 to current 
plant conditions and ensure that the design is accurately reflected 
and that the plant is operated in accordance with its design bases.
    Therefore, the possibility of a new or different kind of 
accident from any accident previously evaluated would not be created 
[by] these amendments.
    3. The proposed amendment will not involve a significant 
reduction in the margin of safety
    The proposed changes to the safety valve lift setting tolerances 
are consistent with the guidance provided by Section III and XI of 
the ASME Code and Standard Technical Specifications. Analyses have 
demonstrated these valves will continue to perform their function of 
protecting their respective system from over-pressurization under 
all postulated transients and accidents. The changed setting 
tolerances do not cause a reduction in any other safety margin such 
as DNBR. SAFETY LIMIT curves are provided to define minimum 
allowable safety margin for plant steady state operation, normal 
operational transients and anticipated operational occurrences. The 
SAFETY LIMITs represent a design requirement for establishment of 
many of the RPS [reactor protection system] trip setpoints which 
prevent reactor conditions from approaching the SAFETY LIMITs. The 
proposed revision of the SAFETY LIMIT curves provide the minimum 
safety margins with somewhat more conservatism than previously 
included. No RPS trips setpoints are changed.
    Therefore, a significant reduction in the margin of safety would 
not be involved with these amendments.
    Based on the evaluation described above, and pursuant to 10 CFR 
Part 50, Section 50.91, Northern States Power Company has determined 
that operation [of] the Prairie Island Nuclear Generating Plant in 
accordance with the proposed license amendment request does not 
involve any significant hazards considerations as defined by Nuclear 
Regulatory Commission regulations in 10 CFR Part 50, Section 50.92.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment requests involve no significant hazards consideration.
    Local Public Document Room location: Minneapolis Public Library, 
Technology and Science Department, 300 Nicollet Mall, Minneapolis, MN 
55401
    Attorney for licensee: Jay Silberg, Esq., Shaw, Pittman, Potts, and 
Trowbridge, 2300 N Street, NW, Washington, DC 20037
    NRC Project Director: John N. Hannon

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

    Date of amendment request: July 28, 1995
    Description of amendment request: The proposed amendment would 
revise the 250 volt DC [direct current] profiles in Technical 
Specifications Surveillance 4.8.2.1 (d) (2c) to reflect the new load 
profile calculations.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:
    1. Involve a significant increase in the probability or 
consequences of an accident previously evaluated.
    [Final Safety Analysis Report] FSAR Section 8.3 states that the 
station batteries have sufficient capacity without the charger to 
independently supply the required loads for four hours. The 
Technical Specifications require that the batteries be surveilled to 
dummy loads which are greater than the design loads. The load 
profiles for the 250 VDC batteries were recalculated using discrete 
increments of time when the loads would be in use for each of five 
design basis events. The Technical Specification load profiles are a 
composite of the worst case loads for the events plus margin. The 
required ampere-hours for each battery using the new load profiles 
is less than the ampere-hours required using the existing load 
profiles. Therefore, since the load profiles envelop the actual 
loads on the batteries, the change to the 250 VDC battery load 
profiles does not involve a significant increase in the probability 
or consequence of an accident previously evaluated.
    2. Create the possibility of a new or different kind of accident 
from any accident previously evaluated.
    As stated above, the 250 VDC batteries have sufficient capacity 
to power the actual battery loads thus enabling them to perform 
their intended function. Any postulated accident resulting from this 
change is bounded by previous analysis. Therefore, the change to the 
250 VDC battery load profiles does 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.
    The Class 1E 250 VDC batteries are required to have sufficient 
capacity and capability to ensure sufficient power is available to 
supply the safety related equipment for (1) the safe shutdown of the 
facilities and (2) the mitigation and control of accident conditions 
within the facilities. The proposed load profiles envelope the worst 
case loads plus margin.
    The ampere-hours removed from the Class 1E 250 VDC batteries are 
less for the proposed load profiles than the existing load profiles. 
The ampere-hours available in the batteries after the batteries have 
supplies[d] the emergency loads for 4 hours are: [See table in 
subject application].
    * * * * * * *
    Engineering calculation shows that the Class 1E 250 VDC 
batteries maintain at least 210 VDC at the Class 1E 250 VDC MCCs 
while supplying the proposed loads, corrected for temperature and 
aging. Since the Class 1E 250 VDC circuits are designed to operate 
properly with a minimum of 210 VDC at the Class 1E MCCs, all the 
Class 1E emergency equipment supplied from the Class 1E batteries 
have sufficient voltage to operate for 4 hours after the loss of ac 
power.
    The Class 1E 250 VDC batteries and Class 1E 250 VDC battery 
chargers have been sized using the proposed load profiles. The 
Engineering calculation shows that the 120 cell, 12 positive plates 
per cell battery banks are sufficient to supply the proposed load 
profiles, corrected for temperature and aging. The same calculation 
also shows that the Class 1E 250 VDC battery chargers have 
sufficient capacity to re-charge the batteries from the proposed 
emergency discharged conditions to the fully charged condition in 12 
hours while continuing to supply the plant normal continuous loads.
    Base upon the above discussion, the proposed changes to the 
Technical Specification load profiles do not reduce the margin of 
safety as defined in the Technical Specifications.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: Osterhout Free Library, 
Reference Department, 71 South Franklin Street, Wilkes-Barre, 
Pennsylvania 18701
    Attorney for licensee: Jay Silberg, Esquire, Shaw, Pittman, Potts 
and 

[[Page 47623]]
Trowbridge, 2300 N Street NW., Washington, DC 20037
    NRC Project Director: John F. Stolz

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

    Date of amendment request: August 11, 1995
    Description of amendment request: The proposed amendment would 
revise Susquehanna Unit 2 Technical Specification Table 3.3.7.5-1 as 
follows:a.
    Revise Item 13, Required Number of Channels from 1 to 2;b.
    Revise Item 13, Minimum Channel Operable from 0 to 1;c.
    Delete Footnote .
    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:
    I. This proposal does not involve a significant increase in the 
probability or consequences of an accident previously evaluated.
    Reestablishing the channel operability values in Item 
13 of Technical Specification Table 3.3.7.5-1, and deleting 
footnote , has no impact on the 
probability or consequences of an accident previously evaluated. The 
proposed change in the channel operability values is a return to the 
values which were reviewed as part of the licensing basis.
    II. This proposal does not create the possibility of a new or 
different kind of accident from any accident previously evaluated.
    Reestablishing the channel operability values in Item 
13 of Technical Specification Table 3.3.7.5-1, and deleting 
footnote , does not create the 
possibility of a new or different kind of accident from any accident 
previously evaluated. The change in the channel operability values 
increases the required number of channels available for accident 
monitoring. There is no correlation between increasing the number of 
neutron flux accident monitoring channels available and the creation 
of accident scenarios.
    III. This change does not involve a significant reduction in a 
margin of safety.
    Reestablishing the channel operability values in Item 
13 of Technical Specification Table 3.3.7.5-1, and deleting 
footnote , does not involve a reduction 
in a margin of safety. The proposed change increases the number of 
required channels from current levels, and restores the values to 
those which have historically been required. At the present time, 
the number of required channels is being administratively controlled 
at the proposed levels to ensure conservatism in operability.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: Osterhout Free Library, 
Reference Department, 71 South Franklin Street, Wilkes-Barre, 
Pennsylvania 18701
    Attorney for licensee: Jay Silberg, Esquire, Shaw, Pittman, Potts 
and Trowbridge, 2300 N Street NW., Washington, DC 20037
    NRC Project Director: John F. Stolz

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

    Date of amendment request: May 12, 1995
    Description of amendment request: The proposed change would extend 
the surveillance test intervals for the emergency service water (ESW) 
system to support 24 month operating cycles.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:
    Operation of the FitzPatrick plant in accordance with the 
proposed Amendment would not involve a significant hazards 
consideration as defined in 10 CFR 50.92 since it would not:
    1. involve a significant increase in the probability or 
consequences of an accident previously evaluated.
    The proposed changes increase the interval between ESW system 
surveillance tests. These changes are consistent with the guidance 
provided in Generic Letter 91-04. These changes do not involve any 
physical changes to the plant, nor do they alter the typical way the 
ESW system functions. On-line testing will continue to assure 
equipment availability. The type of testing and the corrective 
actions required if the subject ESW surveillances fail remain the 
same. As such, the proposed changes create no new impacts on 
accidents previously evaluated.
    Therefore, 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 accident previously evaluated.
    The proposed changes increase the interval between ESW system 
surveillance tests. These changes are consistent with the guidance 
provided in Generic Letter 91-04. The proposed changes do not change 
the ability of the ESW system to provide heat removal for the ECCS 
[emergency core cooling system] components and other equipment 
essential to reactor shutdown. Past equipment performance and on-
line testing indicate the longer test intervals will not degrade ESW 
equipment. No changes are proposed to the type of testing performed, 
only to the length of the surveillance interval. The proposed 
changes do not modify the design or operation of plant equipment, 
therefore, no new or different failure modes are introduced.
    Therefore, the proposed changes do not create the possibility of 
a new or different kind of accident from any accident previously 
evaluated.
    3. involve a significant reduction in a margin of safety.
    The proposed changes increase the interval between ESW system 
surveillance tests. These changes are consistent with the guidance 
provided in Generic Letter 91-04. The proposed changes do not alter 
the configuration of the ESW system nor change the manner in which 
the ESW equipment functions. Past equipment performance and on-line 
testing indicate the longer test intervals will not degrade ESW 
equipment. Operation of the plant remains unchanged by the proposed 
changes.
    Therefore, the proposed changes do not involve a significant 
reduction in a margin of safety.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: Reference and Documents 
Department, Penfield Library, State University of New York, Oswego, NY 
13126
    Attorney for licensee: Mr. Charles M. Pratt, 1633 Broadway, New 
York, NY 10019
    NRC Project Director: Ledyard B. Marsh

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

    Date of amendment request: June 15, 1995
    Description of amendment request: The proposed change would extend 
the surveillance test intervals for the control rod system to support 
24 month operating cycles.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:
    Operation of the FitzPatrick plant in accordance with the 
proposed Amendment would not involve a significant hazards 
consideration as defined in 10 CFR 50.92, since it would not:
11. involve a significant increase in the probability or consequences 
of an accident previously evaluated.

[[Page 47624]]

    The proposed changes increase the interval between control rod 
system surveillance tests. These changes are consistent with the 
guidance provided in Generic Letter 91-04. These changes do not 
involve any physical changes to the plant, nor do they alter the way 
the control rod system functions. The type of testing and the 
corrective actions required if the subject control rod surveillances 
fail remain the same. As such, the proposed changes create no new 
impacts on accidents previously evaluated.
    The reactivity margin - core loading test can be safely extended 
to accommodate the 24 month operating cycle. The calculation of 
reactivity margin takes into account the longer operating cycle.
    The control rod scram time test can be safely extended to 
accommodate a 24 month operating cycle. Operating experience has 
indicated that control rod scram times do not significantly change 
over an operating cycle. Additional on-line testing provides 
adequate assurance of equipment operability.
    The SDIV [Scram Discharge Instrument Volume] vent and drain 
valve operability test can be safely extended to accommodate a 24 
month operating cycle. Evaluation of past surveillance performance 
and additional on-line testing assure valve operability. The 
operability of the mode switch and the reset switch is demonstrated 
during shutdowns.
    Therefore, the proposed changes do not involve a significant 
increase in the probability and do not change the consequences of an 
accident previously evaluated.
    2. create the possibility of a new or different kind of accident 
from any accident previously evaluated.
    The proposed changes increase the interval between control rod 
system surveillance tests. These changes are consistent with the 
guidance provided in Generic Letter 91-04. The proposed changes do 
not change the ability of the control rod system to provide rapid 
reactivity control in order that no fuel damage results from any 
abnormal operating transient. Past equipment performance and on-line 
testing indicate the longer test intervals will not degrade control 
equipment. No changes are proposed to the type of testing performed, 
only to the surveillance interval length. The proposed changes do 
not modify the design or operation of plant equipment, therefore, no 
new or different failure modes are introduced.
    Therefore, the proposed changes do not create the possibility of 
a new or different kind of accident from any accident previously 
evaluated.
    3. involve a significant reduction in a margin of safety.
    The proposed changes increase the interval between control rod 
system surveillance tests. These changes are consistent with the 
guidance provided in Generic Letter 91-04. The proposed changes do 
not alter the configuration of the control rod system nor change the 
manner in which the control rod system functions. Past equipment 
performance and on-line testing indicate the longer test intervals 
will not degrade control rod equipment. Operation of the plant 
remains unchanged by the proposed changes.
    Therefore, the proposed changes do not involve a significant 
reduction in a margin of safety.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: Reference and Documents 
Department, Penfield Library, State University of New York, Oswego, NY 
13126
    Attorney for licensee: Mr. Charles M. Pratt, 1633 Broadway, New 
York, NY 10019
    NRC Project Director: Ledyard B. Marsh

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

    Date of amendment request: July 21, 1995
    Description of amendment request: The proposed changes would 
replace the title-specific list of members on the Plant Operating 
Review Committee (PORC) with a more general statement of membership 
requirements, similar to that used to define Safety Review Committee 
membership; expand the scope of disciplines represented on the PORC to 
include Nuclear Licensing and Quality Assurance; change several 
management position titles; and, make several editorial corrections to 
the Technical Specifications.
    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:
    Replacing the title specific list of PORC members with a 
statement of membership requirements for the committee does not 
reduce the effectiveness of the committee to advise the Resident 
Manager (Site Executive Officer) on matters regarding nuclear 
safety.
    The proposed title changes for the Chief Nuclear Officer, Site 
Executive Officer, Shift Manager, and Control Room Supervisor are 
changes in title only and do not affect the responsibilities, 
authority, qualification requirements, or reporting relationships of 
these positions.
    The change proposed for Specification 6.12 is administrative in 
nature, reflecting a change previously approved elsewhere in 
Technical Specifications.
    The Radiological and Environmental Services Manager title change 
proposed for Specification 6.11(A)2 is administrative in nature, 
reflecting a change previously approved elsewhere in Technical 
Specifications.
    The remainder of proposed changes correct grammar or improve 
consistency in Technical Specification formatting and do not affect 
the meaning or intent of the specifications involved.
    Operation of the James A. FitzPatrick Nuclear Power Plant in 
accordance with the proposed amendment would not involve a 
significant hazards consideration as defined in 10 CFR 50.92. The 
changes are administrative in nature and would not:
    1. 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 those previously evaluated, or
    3. involve a significant reduction in the margin of safety.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: Reference and Documents 
Department, Penfield Library, State University of New York, Oswego, New 
York 13126.
    Attorney for licensee: Mr. Charles M. Pratt, 1633 Broadway, New 
York, NY 10019
    NRC Project Director: Ledyard B. Marsh

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

    Date of amendment request: July 27, 1995
    Description of amendment request: The proposed change to the 
Technical Specifications (TS) would incorporate updated pressure vs. 
temperature operating limit curves contained in TS Figure 3.4.6.1-1 and 
revise TS Surveillance Requirement 4.4.6.1.3 based on implementation of 
Regulatory Guide 1.99, Rev. 2 in accordance with Generic Letter 88-11. 
The changes are a result of data obtained from the first set of 
specimen capsules removed during Refueling Outage 5.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:
    1. Will not involve a significant increase in the probability or 
consequences of an accident [...] previously evaluated.
    The proposed changes assure that the existing safety limits are 
not exceeded due to changing Reactor Vessel conditions. These 
changes reflect the latest material testing 

[[Page 47625]]
results in accordance with 10CFR50, Appendix G. The proposed changes to 
the pressure and temperature limits do not increase the probability 
of nonductile failures. The proposed changes to the surveillance 
requirement and the associated changes to the Bases to include a 
commitment to the methodology of Regulatory Guide 1.99, Rev. 2 
ensures that the most limiting Reactor Vessel material is used in 
the determination of the pressure-temperature operating limits.
    Therefore, it may be concluded that the proposed changes do not 
involve a significant increase in the probability or consequences of 
an accident or malfunction of equipment important to safety 
previously evaluated.
    2. Will not create the possibility of a new or different kind of 
accident from any previously evaluated.
    No physical plant modifications or new operating configurations 
result from these changes. These changes do not adversely affect the 
design or operation of any system or component important to safety, 
rather they establish limits to assure that operations remain within 
acceptable safety boundaries.
    Therefore, the possibility of a new or different kind of 
accident from any previously evaluated will not be created.
    3. Will not involve a significant reduction in a margin of 
safety. Analysis of the capsule specimens has concluded that the 
Reactor Vessel has sufficient fracture toughness for continued safe 
operation, provided that operation remains within acceptable 
pressure-temperature limits. The revised pressure-temperature curves 
define these acceptable pressure-temperature limits during plant 
operation. The proposed changes maintain the existing margins of 
safety by modifying the operating limits based on the most limiting 
of the actual reference temperature shifts. This new limit 
considered analytical results of the capsule specimens, or a 
predicted shift considering the most limiting pressure vessel 
material. Changes to the Surveillance Requirement criteria and the 
associated Bases to include a commitment to the methodology 
contained in Regulatory Guide 1.99, Rev. 2 will ensure that the most 
limiting plate or beltline weld material will be utilized in the 
determination of the pressure-temperature limits.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: Pennsville Public Library, 190 
S. Broadway, Pennsville, NJ 08070
    Attorney for licensee: M. J. Wetterhahn, Esquire, Winston and 
Strawn, 1400 L Street, NW., Washington, DC 20005-3502
    NRC Project Director: John F. Stolz

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: August 1, 1995
    Description of amendment requests: The amendment request proposes 
to change Technical Specification (TS) 3/4.3.2, Table 3.3-3, 
``Engineered Safety Features Actuation System Instrumentation.'' TS 3/
4.3.2 includes the requirements for the minimum number of toxic gas 
isolation system (TGIS) trains operable. The TS change request is to 
extend the allowed TGIS outage times during replacement of TGIS 
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 change does not involve a significant increase 
in the probability or consequences of an accident previously 
evaluated.
    The Toxic Gas Isolation System (TGIS) is designed to monitor and 
mitigate the effects of toxic gas releases on control room 
habitability. TGIS unavailability is not a precursor to any accident 
previously evaluated in Chapter 15 of the San Onofre Updated Final 
Safety Analysis Report (UFSAR). A risk assessment of the TGIS 
instrumentation replacement activity was performed and found that 
the likelihood of a loss of control room habitability beyond that 
permitted by the Technical Specifications (TS) will not exceed 1E-6 
over the duration of this TS change. In addition, a loss of control 
room habitability does not necessarily lead to an accident or core 
damage event. However, if a loss of control room habitability was 
conservatively assumed to lead to a core damage event, this increase 
in risk would still not constitute a significant increase in the 
consequences or probability of any accident previously evaluated 
since the increase is less than 3% of the average annual core damage 
risk from internal events as reported in the San Onofre Individual 
Plant Examination. Therefore, operation of the facility in 
accordance with this proposed change does not involve a significant 
increase in the probability or consequences of any accident 
previously evaluated.
    2. The proposed change does not create the possibility of a new 
or different kind of accident from any accident previously 
evaluated.
    This change extends the allowed outage times of the TGIS system. 
The change does not affect the design or operation of any other 
plant systems. An increase in TGIS unavailability is not a precursor 
to any accident previously evaluated in Chapter 15 of the San Onofre 
UFSAR. Therefore, operation of the facility in accordance with this 
proposed change does not create the possibility of a new or 
different kind of accident from any previously evaluated.
    3. The proposed change does not involve a significant reduction 
in a margin of safety.
    During replacement of TGIS instrumentation a single channel of 
TGIS will be maintained operable except during periods when 
construction activity may result in spurious TGIS alarms. During 
these periods the control room will normally be isolated except for 
brief periods when the control room will be open to allow for air 
exchange or to allow for CREACUS equipment repair. These periods, 
when the control room is open without a TGIS channel available, will 
not exceed 54 hours during the entire period when this change is in 
effect. Operation with control room ventilation in the normal mode 
with a single channel of TGIS operable for 44 days and no TGIS 
channel available for up to 54 hours has been analyzed, and results 
in an increase in the probability of a loss of control room 
habitability which does not exceed 1E-6 over the duration of this TS 
change. Therefore, this proposed change does not involve a 
significant reduction in a margin of safety.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 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, CA 92713
    Attorney for licensee: T. E. Oubre, Esquire, Southern California 
Edison Company, P. O. Box 800, Rosemead, CA 91770
    NRC Project Director: William H. Bateman

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.

[[Page 47626]]

    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.

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

    Date of application for amendment: October 24, 1994, as 
supplemented July 21, 1995. The July 21, 1995, letter provides 
clarification information and did not change the scope of the October 
24, 1994, letter, or the initial no significant hazards consideration 
determination.
    Brief description of amendment: The proposed amendment would revise 
the TS to allow the relocation of TS 3/4.3.7.12, Area Temperature 
Monitoring; and the associated Bases in the TS to licensee-controlled 
documents.
    Date of issuance: August 28, 1995
    Effective date:  August 28, 1995
    Amendment No.: 62
    Facility Operating License No. NPF-63. Amendment revises the 
Technical Specifications.
    Date of initial notice in Federal Register: November 23, 1994 (59 
FR 60379) The Commission's related evaluation of the amendment is 
contained in a Safety Evaluation dated August 28, 1995.No significant 
hazards consideration comments received: No
    Local Public Document Room location: Cameron Village Regional 
Library, 1930 Clark Avenue, Raleigh, NC 27605

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: May 18, 1995, as supplemented 
May 31, 1995
    Brief description of amendments: The amendments revise the 
frequency for conducting the Catawba Unit 2 Integrated Leak Rate Test 
(ILRT) from a nominal frequency of once per 40 months to less than or 
equal to 70 months. This also involves the granting of an exemption 
from the requirements of 10 CFR Part 50, Appendix J, which is addressed 
by separate correspondence.
    Date of issuance: August 18, 1995
    Effective date: As of the date of issuance to be implemented within 
30 days from the date of issuance.
    Amendment Nos.: 133 and 127
    Facility Operating License Nos. NPF-35 and NPF-52: Amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: June 21, 1995 (60 FR 
32362) The Commission's related evaluation of the amendments is 
contained in a Safety Evaluation dated August 18, 1995.No significant 
hazards consideration comments received: No
    Local Public Document Room location:  York County Library, 138 East 
Black Street, Rock Hill, SC 29730

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

    Date of application for amendment:  April 26, 1995
    Brief description of amendment: This amendment adds a requirement 
to Technical Specification (TS) 4.5.2.a to periodically verify that the 
High Head Safety Injection (HHSI) pump minimum flow valve, 2CHS*MOV373, 
is maintained open during plant operation in Modes 1, 2, and 3. Valve 
2CHS*MOV373, must be maintained open to provide a minimum flowpath for 
the HHSI pumps thereby minimizing the likelihood of HHSI pump damage 
due to pump operation with insufficient flow. The amendment allows 
flexibility for local verification of valve position or flow indication 
if the control room indication is not available. Several editorial 
changes to TS 3/4.5.2 are also being made to provide consistent format 
with other TSs.
    Date of issuance: August 25, 1995
    Effective date: August 25, 1995
    Amendment No.: 73
    Facility Operating License No. NPF-73: Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: June 6, 1995 (60 FR 
29874). The Commission's related evaluation of the amendment is 
contained in a Safety Evaluation dated August 25, 1995. No significant 
hazards consideration comments received: No.
    Local Public Document Room location: B. F. Jones Memorial Library, 
663 Franklin Avenue, Aliquippa, PA 15001

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

    Date of application for amendment: April 14, 1995, as supplemented 
by letters dated June 22 and July 18, 1995
    Brief description of amendment: The amendment eliminates response 
time testing (RTT) requirements for selected sensors and specific loop 
instrumentations for (1) the Reactor Protection System (RPS), (2) the 
Isolation System, and (3) the Emergency Core Cooling System (ECCS). In 
addition, the Note for Surveillance Requirement 3.3.6.1.7, which reads: 
``Radiation detectors may be excluded,'' is being removed since RTT is 
not required for any radiation detector that provides a primary 
containment isolation signal as indicated in Table 3.3.6.1-1 of the TS.
    Date of issuance: August 23, 1995
    Effective date: As of the date of issuance to be implemented within 
60 days from the date of issuance
    Amendment No.: 137
    Facility Operating License No. NPF-5: Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: July 5, 1995 (60 FR 
35076) The June 22 and July 18, 1995, letters provided clarifying 
information that did not change the scope of the April 14, 1995, 
application and initial proposed no significant hazards consideration 
determination.
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated August 23, 1995.No significant hazards 
consideration comments received: No
    Local Public Document Room location: Appling County Public Library, 
301 City Hall Drive, Baxley, GA 31513 

[[Page 47627]]


Georgia Power Company, Oglethorpe Power Corporation, Municipal 
Electric Authority of Georgia, City of Dalton, Georgia, Docket Nos. 
50-424 and 50-425, Vogtle Electric Generating Plant, Units 1 and 2, 
Burke County, Georgia

    Date of application for amendments:  January 3, 1995, as 
supplemented by letters dated June 14 and July 6, 1995.
    Brief description of amendments: The amendments revise the 
Technical Specifications (TS) with editorial changes to the Action 
Statements of TS 3.8.1.1 and 3.8.1.2 in order to reflect the 
availability of a third offsite ac electrical source. Technical 
Specification 4.8.1.1.1 is clarified to specify that the offsite ac 
circuits connected to the onsite Class 1E distribution system are 
required to be verified OPERABLE. A footnote is added to TS 3.8.3.1 to 
allow the connection of the third offsite ac source to the onsite 
busses.
    Date of issuance:  August 29, 1995
    Effective date:  As of the date of issuance to be implemented 
within 30 days from the date of issuance
    Amendment Nos.:  90 and 68
    Facility Operating License Nos. NPF-68 and NPF-81: Amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: February 1, 1995 (60 FR 
6301) The June 14 and July 6, 1995, letters provided clarifying 
information that did not change the scope of the January 3, 1995, 
application and initial proposed no significant hazards consideration 
determination. The Commission's related evaluation of the amendments is 
contained in a Safety Evaluation dated August 29, 1995.No significant 
hazards consideration comments received: No
    Local Public Document Room location:  Burke County Library, 412 
Fourth Street, Waynesboro, GA 30830

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

    Date of application for amendments:  May 23, 1995
    Brief description of amendments: The amendments revise the column 
format for the Reactor Protection System and Engineered Safety Feature 
Actuation System Setpoints
    Date of issuance: August 24, 1995
    Effective date: August 24, 1995
    Amendment Nos.: 176 and 170Facility Operating Licenses Nos. DPR-31 
and DPR-41: Amendments revised the Technical Specifications.
    Date of initial notice in Federal Register: June 21, 1995 (60 FR 
32364) The Commission's related evaluation of the amendments is 
contained in a Safety Evaluation dated August 24, 1995.No significant 
hazards consideration comments received: No
    Local Public Document Room location: Florida International 
University, University Park, Miami, FL 33199

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

    Date of application for amendments: December 20, 1993, as 
supplemented July 19, 1994, and February 28, 1995.
    Brief description of amendments: The amendments revise the 
surveillance requirements and load profiles for A, B, and N Train 
batteries.
    Date of issuance: August 22, 1995
    Effective date: August 22, 1995
    Amendment Nos.: 198 and 183
    Facility Operating License Nos. DPR-58 and DPR-74. Amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: February 2, 1994 (59 FR 
4939) and June 6, 1995 (60 FR 29879) The Commission's related 
evaluation of the amendments is contained in a Safety Evaluation dated 
August 22, 1995.No significant hazards consideration comments received: 
No.
    Local Public Document Room location: Maud Preston Palenske Memorial 
Library, 500 Market Street, St. Joseph, MI 49085

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

    Date of application for amendments: May 25, 1995, and supplemented 
June 30, 1995
    Brief description of amendments: The amendments allow fuel 
reconstitution when analyzed in accordance with NRC-approved 
methodologies. The amendments are line item improvements based on NRC 
Generic Letter 90-02, ``Alternative Requirements for Fuel Assemblies in 
Design Features Section of Technical Specifications,'' supplement 1.
    Date of issuance: August 22, 1995
    Effective date: August 22, 1995
    Amendment Nos.: 199 and 184
    Facility Operating License Nos. DPR-58 and DPR-74. Amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: July 5, 1995 (60 FR 
35081) The June 30, 1995, supplement provided a minor revision to the 
proposed Technical Specification pages which was within the scope of 
the original application and did not change the staff's initial 
proposed no significant hazards considerations determination. The 
Commission's related evaluation of the amendments is contained in a 
Safety Evaluation dated August 22, 1995.No significant hazards 
consideration comments received: No.
    Local Public Document Room location: Maud Preston Palenske Memorial 
Library, 500 Market Street, St. Joseph, MI 49085

Maine Yankee Atomic Power Company, Docket No. 50-309, Maine Yankee 
Atomic Power Station, Lincoln County, Maine

    Date of application for amendment: February 14, 1995
    Brief description of amendment: This amendment makes the following 
administrative changes to the Maine Yankee (MY) Technical 
Specifications (TS):
    a. Removes responsibility for audits of the emergency and security 
plans--including their implementing procedures--from the TS and assigns 
that responsibility to the emergency and security plans,
    b. Assigns review responsibility for significant, accidental, 
unplanned, or uncontrolled radioactive releases to the Nuclear Safety 
Audit and Review (NSAR) Committee,
    c. Assigns additional reporting requirements to the NSAR Committee, 
and
    d. Provides the President of MY with the authority to initiate an 
audit of any area of facility operation.
    Date of issuance: August 22, 1995
    Effective date: As of the date of issuance, to be implemented 
within 30 days.
    Amendment No.: 152
    Facility Operating License No. DPR-36: Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: March 29, 1995 (60 FR 
16191) The Commission's related evaluation of the amendment is 
contained in a Safety Evaluation dated August 22, 1995.No significant 
hazards consideration comments received: No
    Local Public Document Room location: Wiscasset Public Library, High 
Street, P.O. Box 367, Wiscasset, ME 04578

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

    Date of application for amendment:  May 18, 1995
    Brief description of amendment: The amendment revises the minimum 
temperature at which the reactor vessel head bolting studs are allowed 
to be 

[[Page 47628]]
placed under tension. In addition, the amendment revises the minimum 
reactor vessel metal temperature during core critical operation, 
revises the minimum reactor vessel metal temperature for pressure 
tests, makes editorial changes, and revises the Bases for the 
applicable section.
    Date of issuance:  August 23, 1995
    Effective date:  As of the date of issuance to be implemented 
immediately.
    Amendment No.:  85
    Facility Operating License No. DPR-21. 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 23, 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, Millstone 
Nuclear Power Station, Unit 1, New London County, Connecticut

    Date of application for amendment: June 15, 1995
    Brief description of amendment: The amendment changes the 
definition for an alteration of the reactor core to one that is 
consistent with the intent of the improved standard technical 
specifications. The amendment also makes administrative changes to 
several technical specification pages.
    Date of issuance: August 28, 1995
    Effective date: As of the date of issuance to be implemented within 
30 days.
    Amendment No.: 86
    Facility Operating License No. DPR-21. Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: July 19, 1995 (60 FR 
37097) The Commission's related evaluation of the amendment is 
contained in a Safety Evaluation dated August 28, 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, et al., Docket No. 50-423, 
Millstone Nuclear Power Station, Unit No. 3, New London County, 
Connecticut

    Date of application for amendment: April 28, 1995, as supplemented 
August 2, 1995.
    Brief description of amendment: The amendment changes Technical 
Specification (TS) Sections 3.7.5, 4.7.5, and 3/4.7.5, to permit 
Millstone Unit 3 to remain in operation with the average ultimate heat 
sink water temperature greater than 75* F (but less than or equal to 
77* F) for a period of 12 hours.
    Date of issuance: August 28, 1995
    Effective date:  As of the date of issuance.
    Amendment No.: 119
    Facility Operating License No. NPF-49. Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: June 6, 1995 (60 FR 
29881). The information in the licensee's submittal of August 2, 1995, 
did not require a change to the initial proposed no significant hazards 
consideration determination. The Commission's related evaluation of the 
amendment is contained in a Safety Evaluation dated August 28, 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

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

    Date of application for amendments: June 29, 1995
    Brief description of amendments: The amendments revise the combined 
Technical Specifications (TS) for Diablo Canyon Nuclear Power Plant, 
Unit Nos. 1 and 2 (DCPP) to add Mode 1 applicability to TS 3/4.4.2.2, 
``Safety Valves - Operating,'' and changes the low-temperature 
overpressure protection (LTOP) system enable temperature for Mode 4 
applicability from 323 degrees F to 270 degrees F in TS 3/4.3.2.1, 
``Safety Valves - Shutdown.''
    Date of issuance: August 23, 1995
    Effective date: August 23, 1995
    Amendment Nos.: Unit 1 - Amendment No. 107; Unit 2 - Amendment No. 
106
    Facility Operating License Nos. DPR-80 and DPR-82: The amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: July 19, 1995 (60 FR 
37098) The Commission's related evaluation of the amendments is 
contained in a Safety Evaluation dated August 23, 1995.No significant 
hazards consideration comments received: No.
    Local Public Document Room location: California Polytechnic State 
University, Robert E. Kennedy Library, Government Documents and Maps 
Department, San Luis Obispo, CA 93407

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 1, 1995, as 
supplemented by letter dated June 20, 1995
    Brief description of amendments: The requested changes would modify 
the applicable operational conditions for the secondary containment 
isolation radiation monitors located on the refueling floor and for the 
monitor located in the railroad access shaft.
    Date of issuance: August 24, 1995
    Effective date: Both units, as of the date of issuance and is to be 
implemented within 30 days
    Amendment Nos.: 152 and 122
    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 
16192). The Commission's related evaluation of the amendments is 
contained in a Safety Evaluation dated August 24, 1995.No significant 
hazards consideration comments received: No
    Local Public Document Room location: Osterhout Free Library, 
Reference Department, 71 South Franklin Street, Wilkes-Barre, PA 18701

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

    Date of application for amendments:  August 31, 1994, as 
supplemented by letters dated May 11, and July 3, 1995
    Brief description of amendments: This amendment revises the 
Technical Specifications to permit the relocation of the Turbine 
Overspeed Protection System to the Updated Final Safety Analysis Report 
and Controlled Plant Procedures.
    Date of issuance: August 24, 1995
    Effective date: August 24, 1995
    Amendment Nos.: 100 and 64
    Facility Operating License Nos. NPF-39 and NPF-85. The amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: November 9, 1994 (59 FR 
55884) The supplemental letters do not 

[[Page 47629]]
change the initial no significant hazards consideration determination 
nor the initial Federal Register notice. The Commission's related 
evaluation of the amendments is contained in a Safety Evaluation dated 
August 24, 1995.No significant hazards consideration comments received: 
No
    Local Public Document Room location: Pottstown Public Library, 500 
High Street, Pottstown, PA 19464

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

    Date of application for amendments:  February 22, 1995
    Brief description of amendments: The amendments revise the 
Technical Specifications Surveillance Requirements to clarify the 
Emergency Diesel Generator acceptable steady state voltage range.
    Date of issuance: August 24, 1995
    Effective date: As of the date of issuance and shall be implemented 
within 30 days of issuance.
    Amendment Nos.: 101 and 65
    Facility Operating License Nos. NPF-39 and NPF-85. The amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: April 26, 1995 (60 FR 
20525) The Commission's related evaluation of the amendments is 
contained in a Safety Evaluation dated August 24, 1995.No significant 
hazards consideration comments received: No
    Local Public Document Room location: Pottstown Public Library, 500 
High Street, Pottstown, PA 19464

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

    Date of application for amendment:  January 13, 1995
    Brief description of amendment: The amendment revised the 
Administrative Controls Section (6.0) of the Technical Specifications 
for Hope Creek Generating Station to reflect organizational changes and 
resultant management title changes.
    Date of issuance: August 22, 1995
    Effective date: August 22, 1995
    Amendment No.: 77
    Facility Operating License No. NPF-57: This amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: June 21, 1995 (60 FR 
32371) The Commission's related evaluation of the amendment is 
contained in a Safety Evaluation dated August 22, 1995.No significant 
hazards consideration comments received: No
    Local Public Document Room location: Pennsville Public Library, 190 
S. Broadway, Pennsville, NJ 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:  December 23, 1994
    Brief description of amendments: The amendments to the Technical 
Specifications revise the surveillance requirement to perform a visual 
inspection of containment areas affected by containment entry when 
containment integrity is established. They are consistent with Item 7.5 
of Generic Letter 93-05, ``Line-Item Technical Specifications 
Improvements to Reduce Surveillance Requirements for Testing During 
Power Operation.''
    Date of issuance: August 24, 1995
    Effective date: As of the date of issuance, to be implementd within 
60 days.
    Amendment Nos.: 174 and 155
    Facility Operating License Nos. DPR-70 and DPR-75. The amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: February 1, 1995 (60 FR 
6308) The Commission's related evaluation of the amendments is 
contained in a Safety Evaluation dated August 24, 1995.No significant 
hazards consideration comments received: No
    Local Public Document Room location: Salem Free Public Library, 112 
West Broadway, Salem, NJ 08079

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: September 16, 1994
    Brief description of amendments: These amendments revise Technical 
Specification (TS) 3/4.2.1, ``Linear Heat Rate.'' The linear heat rate 
(LHR) limit for steady state operation is revised from 13.9 kw/ft to 
13.0 kw/ft. The Bases for TS 3/4.2.1, ``Linear Heat Rate,'' is also 
being revised to reflect the new value.
    Date of issuance: August 23, 1995
    Effective date: August 23, 1995, to be implemented within 30 days 
of issuance.
    Amendment Nos.: Unit 2 - Amendment No. 124; Unit 3 - Amendment No. 
113
    Facility Operating License Nos. NPF-10 and NPF-15: The amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: November 9, 1994 (59 FR 
55892) The Commission's related evaluation of the amendments is 
contained in a Safety Evaluation dated August 23, 1995. No significant 
hazards consideration comments received: No.
    Local Public Document Room location: Main Library, University of 
California, P. O. Box 19557, Irvine, CA 92713

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:  May 3, 1995
    Brief description of amendments: The amendments delay 
implementation of Amendment Nos. 182 and 174 until implementation 
problems are addressed. These changes revise the setpoints and time 
delays for the auxiliary feedwater loss of power and the 6.9 kv 
shutdown board loss of voltage and degraded voltage instrumentation.
    Date of issuance: August 22, 1995
    Effective date: August 22, 1995
    Amendment Nos.: 207 and 197
    Facility Operating License Nos. DPR-77 and DPR-79: Amendments 
revise the technical specifications.
    Date of initial notice in Federal Register: May 23, 1995 (60 FR 
27343) The Commission's related evaluation of the amendment is 
contained in a Safety Evaluation dated August 22, 1995.No significant 
hazards consideration comments received: None
    Local Public Document Room location: Chattanooga-Hamilton County 
Library, 1101 Broad Street, Chattanooga, TN 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: April 6, 1995 (TS 94-18)
    Brief description of amendments: The amendments revise Surveillance 
Requirement 4.0.5 by replacing the current Inservice Inspection program 
and the Inservice Testing program requirements with the requirements 
stated in the Standard Technical Specifications (NUREG-1431). The 
amendments also delete Technical Specification 3/4.4.10, ``Structural 
Integrity ASME Code Class 1, 2 and 3 Components,'' and its related 
Bases information.
    Date of issuance: August 22, 1995
    Effective date: August 22, 1995
    Amendment Nos.: 208 and 198
    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 
20528) 

[[Page 47630]]
The Commission's related evaluation of the amendment is contained in a 
Safety Evaluation dated August 22, 1995.No significant hazards 
consideration comments received: None
    Local Public Document Room location: Chattanooga-Hamilton County 
Library, 1101 Broad Street, Chattanooga, TN 37402

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

    Date of application for amendment: April 17, 1995, as supplemented 
on June 30, 1995
    Brief description of amendment: The amendment revises Technical 
Specifications Technical Specification 2.2.1, Table 2.2-1. The changes 
address reducing repeated alarms and partial reactor trips by revising 
the Overpower Delta-T setpoint function.
    Date of issuance: August 21, 1995
    Effective date: Immediately, to be implemented within 30 days.
    Amendment No.: 102
    Facility Operating License No. NPF-30. Amendment revises the 
Technical Specifications.
    Date of initial notice in Federal Register: May 10, 1995 (60 FR 
24922). The June 30, 1995, letter provided supplemental 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 21, 1995. No 
significant hazards consideration comments received: No.
    Local Public Document Room location:  Callaway County Public 
Library, 710 Court Street, Fulton, MO 65251

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

    Date of application for amendment: September 2, 1992
    Brief description of amendment: The amendment revises the required 
signal-to-noise ratio for the source range monitors, as recommended by 
General Electric.
    Date of issuance: August 23, 1995
    Effective date: August 23, 1995
    Amendment No.: 140
    Facility Operating License No. NPF-21: The amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: July 19, 1995 (60 FR 
37101) The Commission's related evaluation of the amendment is 
contained in a Safety Evaluation dated August 23, 1995.No significant 
hazards consideration comments received: No.
    Local Public Document Room location: Richland Public Library, 955 
Northgate Street, Richland, WA 99352
    Dated at Rockville, Maryland, this 6th day of September 1995.
    For the Nuclear Regulatory Commission
Jack W. Roe,
Director, Division of Reactor Projects - III/IV Office of Nuclear 
Reactor Regulation.
[Doc. 95-22616 Filed 9-12-95; 8:45 am]
BILLING CODE 7590-01-F