[Federal Register Volume 61, Number 206 (Wednesday, October 23, 1996)]
[Notices]
[Pages 55028-55049]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: 96-27025]


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UNITED STATES NUCLEAR REGULATORY COMMISSION

Biweekly Notice


Applications and Amendments to Facility Operating Licenses 
Involving No Significant Hazards Considerations

I. Background

    Pursuant to Public Law 97-415, the U.S. Nuclear Regulatory 
Commission (the Commission or NRC staff) is publishing this regular 
biweekly notice. Public Law 97-415 revised section 189 of the Atomic 
Energy Act of 1954, as amended (the Act), to require the Commission to 
publish notice of any amendments issued, or proposed to be issued, 
under a new provision of section 189 of the Act. This provision grants 
the Commission the authority to issue and make immediately effective 
any amendment to an operating license upon a determination by the 
Commission that such amendment involves no significant hazards 
consideration, notwithstanding the pendency before the Commission of a 
request for a hearing from any person.
    This biweekly notice includes all notices of amendments issued, or 
proposed to be issued from September 30, 1996, through October 10, 
1996. The last biweekly notice was published on October 9, 1996 (61 FR 
52962).

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

    The Commission has made a proposed determination that the following 
amendment requests involve no significant hazards consideration. Under 
the Commission's regulations in 10 CFR 50.92, this means that operation 
of the facility in accordance with the proposed amendment would not (1) 
involve a significant increase in the probability or consequences of an 
accident previously evaluated; or (2) create the possibility of a new 
or different kind of accident from any accident previously evaluated; 
or (3) involve a significant reduction in a margin of safety. The basis 
for this proposed determination for each amendment request is shown 
below.
    The Commission is seeking public comments on this proposed 
determination. Any comments received within 30 days after the date of 
publication of this notice will be considered in making any final 
determination.
    Normally, the Commission will not issue the amendment until the 
expiration of the 30-day notice period. However, should circumstances 
change during the notice period such that failure to act in a timely 
way would result, for example, in derating or shutdown of the facility, 
the Commission may issue the license amendment before the expiration of 
the 30-day notice period, provided that its final determination is that 
the amendment involves no significant hazards consideration. The final 
determination will consider all public and State comments received 
before action is taken. Should the Commission take this action, it will 
publish in the Federal Register a notice of issuance and provide for 
opportunity for a hearing after issuance. The Commission expects that 
the need to take this action will occur very infrequently.
    Written comments may be submitted by mail to the Chief, Rules 
Review and Directives Branch, Division of Freedom of Information and 
Publications Services, Office of Administration, U.S. Nuclear 
Regulatory Commission, Washington, DC 20555-0001, and should cite the 
publication date and page number of this Federal Register notice. 
Written comments may also be delivered to Room 6D22, Two White Flint 
North, 11545 Rockville Pike, Rockville, Maryland from 7:30 a.m. to 4:15 
p.m. Federal workdays. Copies of written comments received may be 
examined at the NRC Public Document Room, the Gelman Building, 2120 L 
Street, NW., Washington, DC. The filing of requests for a hearing and 
petitions for leave to intervene is discussed below.
    By November 22, 1996, the licensee may file a request for a hearing 
with respect to issuance of the amendment to the subject facility 
operating license and any person whose interest may be affected by this 
proceeding and who wishes to participate as a party in the proceeding 
must file a written request for a hearing and a petition for leave to 
intervene. Requests for a hearing and a petition for leave to intervene 
shall be filed in accordance with the Commission's ``Rules of Practice 
for Domestic Licensing Proceedings'' in 10 CFR Part 2. Interested 
persons should consult a current copy of 10 CFR 2.714 which is 
available at the Commission's Public Document Room, the Gelman 
Building, 2120 L Street, NW., Washington, DC and at the local public 
document room for the particular facility involved. If a request for a 
hearing or petition for leave to intervene is filed by the above date, 
the Commission or an Atomic Safety and Licensing Board, designated by 
the Commission or by the Chairman of the Atomic Safety and Licensing 
Board Panel, will rule on the request and/or petition; and the 
Secretary or the designated Atomic Safety and Licensing Board will 
issue a notice of a hearing or an appropriate order.
    As required by 10 CFR 2.714, a petition for leave to intervene 
shall set forth with particularity the interest of the petitioner in 
the proceeding, and how that interest may be affected by the results of 
the proceeding. The petition should specifically explain the reasons 
why intervention should be permitted with particular reference to the 
following factors: (1) the nature of the petitioner's right under the 
Act to be made a party to the proceeding; (2) the nature and extent of 
the petitioner's property, financial, or other interest in the 
proceeding; and (3) the possible effect of any order which may be 
entered in the proceeding on the petitioner's interest. The petition 
should also identify the specific aspect(s) of the subject matter of 
the proceeding as to which petitioner wishes to intervene. Any person 
who has filed a petition for leave to intervene or who has been 
admitted as a party may amend the petition without requesting leave of 
the Board up to 15 days prior to the first prehearing conference 
scheduled in the proceeding, but such an amended petition must satisfy 
the specificity requirements described above.
    Not later than 15 days prior to the first prehearing conference 
scheduled in the proceeding, a petitioner shall file a supplement to 
the petition to intervene which must include a list of the contentions 
which are sought to be litigated in the matter. Each contention must 
consist of a specific statement of the issue of law or fact to be 
raised or

[[Page 55029]]

controverted. In addition, the petitioner shall provide a brief 
explanation of the bases of the contention and a concise statement of 
the alleged facts or expert opinion which support the contention and on 
which the petitioner intends to rely in proving the contention at the 
hearing. The petitioner must also provide references to those specific 
sources and documents of which the petitioner is aware and on which the 
petitioner intends to rely to establish those facts or expert opinion. 
Petitioner must provide sufficient information to show that a genuine 
dispute exists with the applicant on a material issue of law or fact. 
Contentions shall be limited to matters within the scope of the 
amendment under consideration. The contention must be one which, if 
proven, would entitle the petitioner to relief. A petitioner who fails 
to file such a supplement which satisfies these requirements with 
respect to at least one contention will not be permitted to participate 
as a party.
    Those permitted to intervene become parties to the proceeding, 
subject to any limitations in the order granting leave to intervene, 
and have the opportunity to participate fully in the conduct of the 
hearing, including the opportunity to present evidence and cross-
examine witnesses.
    If a hearing is requested, the Commission will make a final 
determination on the issue of no significant hazards consideration. The 
final determination will serve to decide when the hearing is held.
    If the final determination is that the amendment request involves 
no significant hazards consideration, the Commission may issue the 
amendment and make it immediately effective, notwithstanding the 
request for a hearing. Any hearing held would take place after issuance 
of the amendment.
    If the final determination is that the amendment request involves a 
significant hazards consideration, any hearing held would take place 
before the issuance of any amendment.
    A request for a hearing or a petition for leave to intervene must 
be filed with the Secretary of the Commission, U.S. Nuclear Regulatory 
Commission, Washington, DC 20555-0001, Attention: Docketing and 
Services Branch, or may be delivered to the Commission's Public 
Document Room, the Gelman Building, 2120 L Street, NW., Washington DC, 
by the above date. Where petitions are filed during the last 10 days of 
the notice period, it is requested that the petitioner promptly so 
inform the Commission by a toll-free telephone call to Western Union at 
1-(800) 248-5100 (in Missouri 1-(800) 342-6700). The Western Union 
operator should be given Datagram Identification Number N1023 and the 
following message addressed to (Project Director): petitioner's name 
and telephone number, date petition was mailed, plant name, and 
publication date and page number of this Federal Register notice. A 
copy of the petition should also be sent to the Office of the General 
Counsel, U.S. Nuclear Regulatory Commission, Washington, DC 20555-0001, 
and to the attorney for the licensee.
    Nontimely filings of petitions for leave to intervene, amended 
petitions, supplemental petitions and/or requests for a hearing will 
not be entertained absent a determination by the Commission, the 
presiding officer or the Atomic Safety and Licensing Board that the 
petition and/or request should be granted based upon a balancing of 
factors specified in 10 CFR 2.714(a)(1)(i)-(v) and 2.714(d).
    For further details with respect to this action, see the 
application for amendment which is available for public inspection at 
the Commission's Public Document Room, the Gelman Building, 2120 L 
Street, NW., Washington, DC, and at the local public document room for 
the particular facility involved.

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

    Date of amendment request: September 18, 1996
    Description of amendment request: Revise Technical Specification 
(TS) 4.8.1.1.2 by removing TS 4.8.1.1.2.h.2 pressure testing 
requirement since adequate testing will be completed in accordance with 
American Society of Mechanical Engineers (ASME) Boiler and Pressure 
Vessel Code, Section XI.
    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:
    This change does not involve a significant hazards consideration 
for the following reasons:
    1. The proposed amendment does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    Applying ASME Code, Section XI alternative examination/testing 
will not affect any initiators of any previously evaluated accidents 
or change the manner in which the emergency diesel generators or any 
other systems operate. The diesel fuel oil system supports the 
emergency diesel generators which serve an accident mitigating 
function. Where portions of piping are non-isolable or where 
atmospheric tanks are involved, the Section XI ASME alternatives to 
110% pressure testing continue to ensure the integrity of the fuel 
oil system without any impact on analyzed accident scenarios or 
their consequences. Therefore, the proposed amendment does not 
result in an increase in the probability or consequences of an 
accident previously evaluated.
    2. The proposed amendment does not create the possibility of a 
new or different kind of accident from any accident previously 
evaluated.
    The proposed alternative testing and surveillance will not 
involve any physical alterations or additions to plant equipment or 
alter the manner in which any safety-related system performs it 
function. Using ASME Section XI, or NRC-approved ASME Code cases, as 
guidance for pressure testing continues to provide assurance that 
the fuel oil supply system will perform its intended function. 
Therefore, the proposed changes do not create the possibility of a 
new or different kind of accident from any accident previously 
evaluated.
    3. The proposed amendment does not involve a significant 
reduction in the margin of safety.
    There are no changes being made to the safety limits or safety 
settings that would adversely impact plant safety. Further, there is 
no impact on the margin of safety as defined in the Technical 
Specifications. Utilizing ASME Section XI as guidance for 
determining those sections of piping that should be pressure-tested 
or tested at atmospheric pressure will ensure proper operation of 
the diesel generator fuel oil supply system. 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: Cameron Village Regional 
Library, 1930 Clark Avenue, Raleigh, North Carolina 27605
    Attorney for licensee: William D. Johnson, Vice President and 
Senior Counsel, Carolina Power & Light Company, Post Office Box 1551, 
Raleigh, North Carolina 27602
    NRC Project Director: F. Mark Reinhart, Acting

Detroit Edison Company, Docket No. 50-16, Enrico Fermi Atomic Power 
Plant, Unit 1, Monroe County, Michigan

    Date of amendment request: August 29, 1996 (Reference NRC-96-0111)
    Description of amendment request: The proposed amendment will: (1) 
allow certain equipment and instruments to be removed from service for 
short periods of time to allow for

[[Page 55030]]

maintenance, testing, inspection, modifications, and account for 
equipment failures; (2) reduce the frequency of environmental liquid 
effluent monitoring and eliminate one raw water sampling location; (3) 
eliminate the requirement for moisture intrusion monitoring for the 
reactor building lower level; and (4) correction of a typographical 
error.
    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 using the standards in 10 CFR 50.92(c). The licensee's 
analysis is presented below:
    (1) The operation of Enrico Fermi Atomic Power Plant, Unit 1, in 
accordance with the proposed amendment will not involve a 
significant increase in the probability or consequences of an 
accident previously evaluated.
    The proposed changes do not involve a significant increase in 
the probability or consequences of an accident. Provisions for 
removing the primary cover gas supply from service for short periods 
of time will not significantly increase the probability of an 
accident occurring as long as the probability of a significant water 
reaction with residual sodium is not significantly increased. This 
is ensured by prescribing limits on the time that carbon dioxide 
pressure can be low. The consequences of an accident would not be 
affected by provisions for removing the primary cover gas supply 
from service as this equipment does not mitigate accidents or affect 
the accident sequences. Similarly, the provisions for removing the 
moisture intrusion and cover gas pressure alarms from service for 
short period of time will not significantly increase the probability 
of an accident. The alarms provide a monitoring function to detect 
degradation in the performance of the cover gas supply and sump 
systems. Absence of these alarm functions for short periods of time 
does not increase the probability of such degradation and it does 
not significantly impact the ability for timely detection of such 
degradation. The consequences of an accident would not be affected 
by provisions for removing the moisture intrusion and cover gas 
pressure alarms from service as this equipment does not mitigate 
accidents or affect the accident sequences. Elimination of the 
moisture intrusion alarm for the reactor building lower level does 
not significantly increase the probability of an accident because 
the probability that water could accumulate in this area is 
essentially unchanged. Design features of the foundation, 
containment structure, and annulus drains are intended to prevent 
entry of water into the reactor building. These features have 
prevented any water intrusion into this area. The consequences of an 
accident would not be affected by elimination of the moisture 
intrusion alarm for the reactor building lower level because this 
equipment does not mitigate accidents or affect the accident 
sequences. The Safety Evaluation Supporting Amendment 9 to the 
referenced license did not rely on moisture intrusion monitoring and 
alarm features for any safety function or accident prevention or 
mitigation function. Environmental monitoring surveillance are 
unrelated to postulated accident sequences and cannot affect the 
probability or consequences of an accident. The correction of the 
typographical error is unrelated to accident initiation and 
sequences and cannot affect the probability or consequences of any 
accident.
    (2) The operation of Enrico Fermi Atomic Power Plant, Unit 1, in 
accordance with the proposed amendment will not create the 
possibility of a new or different kind of accident from any accident 
previously evaluated.
    The proposed changes do not create the possibility of a new or 
different accident from any previously evaluated. With the exception 
of the allowance for composite environmental samples, which are 
unrelated to any potential accident sequence, these changes propose 
no new activities or new methods for performing existing activities. 
Previous evaluations have considered the release of all of the 
radioactivity in the residual sodium due to postulated fire or other 
catastrophe and release of radioactive water stored in the liquid 
waste tanks which bound the only possible radiological accidents at 
Fermi 1. For these reasons, no new or different type of accident is 
created by these changes.
    (3) The operation of Enrico Fermi Atomic Power Plant, Unit 1, in 
accordance with the proposed amendment will not involve a 
significant reduction in a margin of safety.
    The proposed changes do not involve a significant reduction in a 
margin of safety. The changes to the primary system cover gas system 
technical specifications still ensure that any residual sodium is 
passivated by carbon dioxide. Changes to the alarms affect only 
monitoring functions and therefore do not cause a change to any 
parameter that could affect the margin of safety. Similarly, the 
environmental surveillances are unrelated to margin of safety. The 
correction of the typographical error is unrelated to margin of 
safety. For these reasons, the proposed changes do not involve a 
significant reduction in the margin of safety.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 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: Monroe County Library System, 
3700 South Custer Road, Monroe, Michigan 48161
    Attorney for licensee: John Flynn, Esquire, Detroit Edison Company, 
2000 Second Avenue, Detroit, Michigan 48226NRC Branch Chief: Michael F. 
Weber

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

    Date of amendment request: September 25, 1996 (NRC-96-0085)
    Description of amendment request: The proposed amendment would 
revise Technical Specification (TS) Surveillance Requirement 4.8.4.3 to 
remove the requirement to periodically test the thermal overload (TOL) 
devices for safety-related motor-operated valves (MOVs). The 
surveillance requirement would continue to require testing of a TOL 
device following any maintenance activity that could affect the 
performance of the device. The surveillance requirement would also be 
clarified by indicating that testing of TOL devices is required upon 
initial installation. The associated portion of the TS Bases would also 
be revised to reflect this change.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:
    1. The proposed change does not involve a significant increase 
in the probability or consequences of an accident. The deletion of 
the requirement for testing of the TOL protective devices lessens 
degradation to the components which can improve MOV reliability. 
Based on historical data through the years of testing, there is no 
significant drifting of the trip setpoints of the TOL protective 
devices. The probability of an accident would not increase since 
terminating the periodic testing or clarifying the situational 
testing requirements cannot cause equipment to operate inadvertently 
and so cannot cause an accident. The periodic testing of the TOL 
protective devices can temporarily render MOVs inoperable due to the 
removal of the components from service and can cause safety systems/
divisions to become unavailable. The deletion of the periodic 
testing requirement would increase the availability of safety 
systems insuring that they would be able to respond to accident 
conditions. The consequences of an accident will not increase since 
eliminating the periodic testing and clarifying the situational 
testing requirements will improve reliability of safety-related MOVs 
to respond to an accident and will not increase the failure rate of 
equipment. The clarification of the situational testing ensures that 
the test will be conducted after any maintenance that could affect 
the performance of the TOL protective devices. Thus, the proposed 
change increases reliability of the MOVs and increases plant safety. 
Therefore this change will not result in a significant increase in 
the probability or consequences of an accident.
    2. The proposed change does not create the possibility of a new 
or different accident from any previously evaluated. The TOL

[[Page 55031]]

protective devices are not an accident initiator, they only protect 
equipment provided to mitigate the consequences of an accident. For 
this reason, no new or different type of accident is created by this 
change.
    3. The proposed change does not involve a significant reduction 
in a margin of safety. The trip setpoints of the TOL protective 
devices depend upon both the current and the length of time the 
current is applied. The trip setpoints for TOL protective devices 
are much higher than conditions normally experienced during an MOV 
stroke and are meant to protect the motor from stall and overload 
conditions. The difference between the current of the trip setpoints 
and the normal conditions is great enough that a premature trip of 
the TOL protective device is highly unlikely, even at degraded 
voltages. The TOL protective device protects the motor from the 
stall conditions. Not conducting the periodic testing of the TOL 
protective devices would not cause the MOVs to fail, nor would the 
performance of the MOVs be adversely affected. Throughout the life 
of the plant, there has never been an instance of a safety related 
MOV failure due to degradation or failure of TOL protective devices. 
Further, based on maintenance history, the elimination of the 
periodic testing would eliminate any significant potential 
degradation of the TOL protective devices, thereby increasing their 
reliability. Finally, with the removal of the periodic testing of 
the TOL protective devices, fewer MOVs would have to be removed from 
service for testing. Since necessary components would no longer be 
inoperable due to the periodic testing, there would be an increase 
of availability time of safety systems/divisions. Deletion of the 
periodic testing could reduce the durations of online system 
outages. Clarifying the situational testing requirements would 
better define when the testing of the TOL protective devices is 
necessary which would ensure operability. The testing would be based 
on installation or any maintenance that could affect the TOL 
protective device. For these reasons, the proposed change does not 
involve a significant reduction in the margin of safety.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: Monroe County Library System, 
3700 South Custer Road, Monroe, Michigan 48161
    Attorney for licensee: John Flynn, Esq., Detroit Edison Company, 
2000 Second Avenue, Detroit, Michigan 48226
    NRC Project Director: John N. Hannon

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

    Date of amendment request: June 21, 1996
    Description of amendment request: The proposed amendments would 
administratively correct the term ``lifting load'' in Technical 
Specification 3.9.6b.2 to ``lifting force.'' This correction would 
clarify that the static loads associated with the lifting tool, drive 
rod and control rod weights are not included in the lifting force 
limit. The amendments would also more accurately define auxiliary hoist 
minimum capacities and give a more expansive description of the 
activities for which protective measures and surveillance testing are 
used.
    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:
    Question: Will the change involve a significant increase in the 
probability or consequences of an accident previously evaluated?
    No. The proposed change[s] [are] administrative in nature, and 
do[] not represent any changes to the refueling process in the 
field. It more accurately describes the components for which the 
LCO's [limiting conditions of operation] protection is intended as 
well as giving a more accurate description of the auxiliary hoist's 
minimum capacity. [They] also broaden[] the domain of activities for 
which protective measures are taken, by including drag load testing 
into monitored activities. At both MNS [McGuire Nuclear Station] and 
CNS [Catawba Nuclear Station], the auxiliary hoists and the 
manipulator cranes are rated at [greater than or equal to] 3000 
pounds and are surveillance tested to greater than 1000 pounds. This 
brackets the limit force lifting value change from 600 to 1000 
pounds in the amendment proposal.
    Question: Will the change create the possibility of a new or 
different kind of accident from any accident previously evaluated?
    No. Th[ese] proposed administrative change[s] reflect[] no 
changes in the refueling processes, or any systems, structures or 
components connected with the refueling process.
    Question: Will the change involve a significant reduction in a 
margin of safety?
    No. The proposed administrative change[s] [have] no impact on 
refueling processes, systems, structures or components, and do[] not 
result in any significant reduction in a margin of safety. The 
subject change[s] only clarif[y] the original intent of the 
specification and more accurately describe[] the involved 
components, component capacities and the domain of activities for 
which measures are taken to protect the reactor internals.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
proposed amendments involve no significant hazards consideration.
    Local Public Document Room location: York County Library, 138 East 
Black Street, Rock Hill, South Carolina 29730
    Attorney for licensee: Mr. Albert Carr, Duke Power Company, 422 
South Church Street, Charlotte, North Carolina 28242
    NRC Project Director: Herbert N. Berkow

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

    Date of amendment request: September 17, 1996 (TSC 96-01)
    Description of amendment request: The proposed changes would reduce 
the Reactor Building pressure setpoint for actuation of the Reactor 
Building Spray System in Technical Specification (TS) 3.5.3 from a 
maximum of 30 pounds per square inch gauge (psig) to 15 psig, reduce 
the maximum allowable Reactor Building internal pressure specified in 
TS 3.6.4 from 1.5 psig to 1.2 psig when the reactor is critical, revise 
the corresponding Bases of TS 3.3 to indicate that the Reactor Building 
sprays and coolers are designed to mitigate the containment temperature 
response rather than containment pressure response to a loss-of-coolant 
accident, and make other administrative changes. In addition, the lower 
Reactor Building pressure limit (a vacuum of 5 inches of mercury (Hg)) 
in Specification 3.6.4 would be changed to the corresponding value in 
terms of psig to reflect the units displayed on the control room 
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) Involve a significant increase in the probability or 
consequences of an accident previously evaluated:
    No. The analysis of the post-LOCA [loss-of-coolant accident] 
Reactor Building response to high-energy line breaks, using the new 
methodology, uses assumptions different from the requirements 
currently delineated in Technical Specifications. The new 
assumptions used for initial Reactor Building pressure and Reactor 
Building Spray system

[[Page 55032]]

actuation are 1.2 psig and 20 psig respectively. These values are 
lower, and hence more conservative, than the values currently 
specified in Technical Specifications.
    Since the new values for Reactor Building pressure and Reactor 
Building Spray actuation are more conservative and the analysis 
methodology has received approval from the NRC via [an] SER, this 
change does not involve a significant increase in the probability or 
consequences of an accident previously identified.
    (2) Create the possibility of a new or different kind of 
accident from any kind of accident previously evaluated:
    No. The methodology for Reactor Building high energy line break 
analysis is being revised. The revision of the method of analysis 
does not alter the manner by which plant systems and components 
function for accident mitigation.
    (3) Involve a significant reduction in a margin of safety.
    No. By letter dated March 15, 1995, the NRC stated that the new 
analyses described in the topical report, DPC-NE-3003-P, expand the 
scope of analyzed piping failures in containment for the Oconee 
facilities. The NRC further stated that this new analysis method has 
been used to reanalyze existing licensing basis pipe failure events 
in containment, and to examine the potential effects of previously 
unanalyzed assumptions and initial conditions which the NRC staff 
finds to be consistent with current NRC staff acceptance criteria or 
produce equally conservative results. In conclusion, the NRC 
confirmed that this methodology, with appropriate adjustments to 
reflect potential plant modifications, may be used by Duke Power to 
perform future analyses in support of licensing applications related 
to containment accident response. This proposed change to Technical 
Specifications reflects the use of this new methodology. Based on 
this new methodology, changes have been made to setpoint assumptions 
for initial Reactor Building pressure and Reactor Building Spray 
actuation. This proposed Technical Specification change reflects 
those assumption changes. This methodology has been accepted by the 
NRC. This proposed change to Technical Specifications does not 
involve a significant reduction in the margin of safety.
    The NRC has reviewed the licensee's analysis and, based on 
thisreview, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: Oconee County Library, 501 
West South Broad Street, Walhalla, South Carolina 29691
    Attorney for licensee: J. Michael McGarry, III, Winston and Strawn, 
1200 17th Street, NW., Washington, DC 20036
    NRC Project Director: Herbert N. Berkow

Duquesne Light Company, et al., Docket No. 50-334, Beaver Valley 
Power Station, Unit No. 1, Shippingport, Pennsylvania

    Date of amendment request: September 9, 1996
    Description of amendment request: The proposed amendment would 
revise the Minimum Channels Operable requirement of Item 4.c (Steam 
Line Isolation, Containment Pressure Intermediate -- High-High) of 
Technical Specification (TS) Table 3.3-3 from 3 to 2. This proposed 
change would make this Unit 1 TS consistent with the comparable Unit 2 
TS.
    The proposed amendment would also revise the minimum charging pump 
discharge pressure in TS 3.5.5 from 2311 psig to 2397 psig. This change 
is required to ensure that safety analysis assumptions for safety 
injection flow are met. Conforming changes would also be made to the 
Bases for TS 3/4.5.5 to reflect the proposed changes to TS 3.5.5.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:
    1. Does the change involve a significant increase in the 
probability or consequences of an accident previously evaluated?
    The proposed amendment does not add or modify any existing plant 
equipment. Since normal charging pump discharge pressure is greater 
than or equal to approximately 2440 psig, no additional plant 
configuration changes or modifications will be required to comply 
with this revised charging pump discharge pressure value. The 
proposed amendment does not change the design or function of the 
containment pressure intermediate-high-high channels.
    The consequences of an accident previously evaluated are not 
significantly increased. The ability of the containment pressure 
intermediate-high-high function to initiate steam line isolation 
will not be affected. Since steam line isolation will continue to 
occur at the same required trip setpoint, the amount of mass and 
energy released to containment along with the ability to maintain at 
least one unfaulted steam generator (SG) as a heat sink for the 
reactor remains unchanged. The amount of seal injection flow will 
continue to be adequately limited to ensure sufficient flow to the 
reactor core during accident conditions. The Bases changes are 
editorial in nature and do not involve a change to probability or 
consequences of an accident previously evaluated.
    Based on the above discussion, it is concluded that this change 
does not involve a significant increase in the probability or 
consequences of an accident previously evaluated.
    2. Does the change create the possibility of a new or different 
kind of accident from any accident previously evaluated?
    The proposed amendment does not change the plant configuration 
in a way which introduces a new potential hazard to the plant. Since 
design requirements continue to be met and the integrity of the 
reactor coolant system pressure boundary is not challenged, no new 
failure mode has been created. As a result, an accident which is 
different than already evaluated in the Updated Final Safety 
Analysis Report will not be created due to this change.
    Therefore, the proposed change does not create the possibility 
of a new or different kind of accident from any accident previously 
evaluated.
    3. Does the change involve a significant reduction in a margin 
of safety?
    The margin of safety is not significantly reduced by this 
proposed change. The trip setpoint for the containment pressure 
intermediate-high-high function remains unchanged. With one channel 
inoperable, the remaining two channels will continue to initiate the 
protective function on a two-out-of-two logic. The action statement 
limits this condition to 6 hours after which time the inoperable 
channel must be placed in the trip condition. This action restores 
the function to be able to meet single failure criteria on a one-
out-of-two logic basis.
    The proposed revision to the charging pump discharge pressure 
will not change the flow limit on seal injection. The specification 
will continue to ensure that seal injection flow is limited. This 
will ensure that sufficient flow to the reactor core is provided 
during accident conditions.
    The proposed changes to the Bases for seal injection flow are 
editorial in nature and do not affect the margin of safety.
    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 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: B. F. Jones Memorial Library, 
663 Franklin Avenue, Aliquippa, PA 15001
    Attorney for licensee: Jay E. Silberg, Esquire, Shaw, Pittman, 
Potts & Trowbridge, 2300 N Street, NW., Washington, DC 20037
    NRC Project Director: John F. Stolz

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

    Date of amendment request: August 29, 1996
    Description of amendment request: The proposed amendment would 
revise the technical specifications (TSs) to reflect the elimination of 
T-factor adjustments in the Average Power

[[Page 55033]]

Range Monitors (APRM) setpoints, a decrease in the calibration 
frequency of the Local Power Range Monitors (LPMR), and an improvement 
in the calculation of Reactivity Anomaly.
    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 request does not involve a significant increase in the 
probability or consequences of an accident previously evaluated.
    This change replaces the APRM setpoints T-factor limit with 
power and flow-dependent minimum critical power ratio (MCPR) and 
linear heat generation rate (LHGR) limits. These new power and flow-
dependent thermal limits eliminate the need for manual setpoint 
adjustment resulting from power peaking conditions. The new power 
and flow-dependent thermal limits are automatically applied by 
computer software during the calculation of the core thermal limits 
and, therefore, do not require manual setpoint adjustments based on 
the power peaking conditions in the reactor. Extensive transient 
analyses at a variety of power and flow conditions have been 
performed and were utilized to study the trend of transient severity 
without the setpoints T-factor limit. A large data base was 
established by analyzing limiting transients over a range of power 
and flow conditions. The data base included evaluations 
representative of a variety of plant configurations and parameters 
such that the conclusions drawn from the studies would be applicable 
to the broad range of boiling water reactors (BWRs). This data base 
was utilized to develop plant specific operating limits (MCPR and 
LHGR), which assures that margins to fuel safety limits are equal to 
or larger than those currently in existence with the APRM setpoints 
T-factor limit applied. Therefore, this change does not involve an 
increase in the probability of any event previously evaluated.
    The consequences of an accident previously evaluated have not 
been increased because, in all cases, the new power and flow-
dependent thermal limits (MCPR and LHGR) assure that margins to fuel 
safety limits are equal to or larger than those currently in 
existence with the APRM setpoints T-factor limit applied. Protection 
of other thermal limits for all previously analyzed events is 
accomplished by specific limits that are independent of the APRM 
setpoints T-factor. These are the power and flow-dependent MCPR 
Operating Limits which provide protection from fuel dryout and the 
rated maximum average planner linear heat generation rate (MAPLHGR) 
limit which provides protection of the peak clad temperature for the 
design basis accident-loss of coolant accident (DBA LOCA). 
Therefore, the proposed change does not involve a significant 
increase in the consequences of any event previously evaluated.
    No new equipment is introduced by the change in the local power 
range monitor (LPRM) calibration frequency and, therefore, the 
probability for an accident previously evaluated is unchanged. The 
consequences of an accident can be affected by the thermal limits 
prior to the accident but LPRM chamber and cycle exposure have no 
significant effect on the calculated thermal limits. The thermal 
limit calculation is not significantly effected because the LPRM 
sensitivity versus exposure function is well defined. This allows 
accurate LPRM end-of-life calculations so that detectors can be 
replaced before their behavior significantly deteriorates. In the 
event deterioration is noted late in the cycle for a few chambers, 
they can be bypassed with no significant effect on uncertainties. 
Also, the total nodal power uncertainty remains less than the 
uncertainty assumed in the General Electric BWR Thermal Analysis 
Basis (GETAB) safety limit. Therefore, the thermal limit calculation 
is not affected by the LPRM calibration frequency and the 
consequences of an accident previously evaluated are not changed.
    The change in the parameters used to measure reactivity for 
calculation of the reactivity anomaly has no affect on either the 
consequences or the probability of an accident previously evaluated 
because the allowed reactivity anomaly criteria is unchanged. The 
only change is the parameters used to measure reactivity.
    Therefore, the proposed elimination of the APRM setpoints T-
factor maintains adequate off-rated MCPR and LHGR margin for all 
operating conditions. Also, the change in the LPRM calibration 
frequency continues to maintain the accuracy of the thermal limit 
calculation. Therefore, the consequences of an accident previously 
evaluated are not affected by this change. Finally, the change in 
the parameters used to measure reactivity for calculation of the 
reactivity anomaly has no affect on either the consequences nor the 
probability of an accident previously evaluated. Since no new plant 
equipment is introduced by any of the proposed changes, the 
probability of accidents previously evaluated are not changed. 
Therefore, none of the proposed changes involve an increase in the 
probability or consequences of any event previously evaluated.
    2. The request does not create the possibility of occurrence of 
a new or different kind of accident from any accident previously 
evaluated.
    This change only replaces the APRM setpoints T-factor limit with 
power and flow-dependent MCPR and LHGR limits, changes the LPRM 
calibration frequency, and a change to the parameter(s) used to 
measure reactivity. None of the proposed changes involve any new 
modes of operation or any plant modifications. Therefore, the 
proposed changes do not create the possibility of a new or different 
type of accident from any accident previously analyzed.
    3. The request does not involve a significant reduction in a 
margin of safety.
    The replacement of the APRM setpoints T-factor limit with power 
and flow-dependent thermal limits has been confirmed to provide 
adequate MCPR and LHGR protection at all reactor operation 
conditions. Operation with higher peaking without APRM gains or flow 
bias trip setpoints adjustment does not involve a reduction in a 
margin of safety because the higher power peaking resulting from 
elimination of the APRM setpoints T-factor has been analyzed to 
assure that the margins to fuel safety limits are equal to or larger 
than those currently in existence with the APRM setpoints T-factor 
limit applied. Therefore, the replacement of the APRM setpoint T-
factor with power and flow-dependent thermal limits does not involve 
a reduction in the margin of safety.
    Protection of other thermal limits for all previously analyzed 
events is accomplished by specific limits that are independent of 
the APRM setpoint T-factor limit. These are the power and flow-
dependent
    MCPR Operating Limits which provide protection from fuel dryout 
and the rated MAPLHGR limit which provides protection of the peak 
clad temperature for the DBA LOCA.
    The margin of safety can be affected by the thermal limits prior 
to an accident but LPRM chamber exposure and cycle exposure have no 
significant effect on the calculated thermal limits. The thermal 
limit calculation is not significantly affected because the LPRM 
sensitivity versus exposure function is well defined. This allows 
accurate LPRM end of life calculations so that detectors can be 
replaced before their behavior significantly deteriorates. In the 
event deterioration is noted late in the cycle for a few chambers, 
they can be bypassed with no significant effect on uncertainties. 
Also, the total nodal power uncertainty remains less than the 
uncertainty assumed in the GETAB safety limit. Therefore neither the 
thermal limit calculation nor the margin of safety are affected by 
the LPRM calibration.
    The change in the parameters used to measure reactivity for 
calculation of the reactivity anomaly has no affect on the margin of 
safety because the allowed reactivity anomaly criteria is unchanged. 
The only change is the parameters used to measure reactivity.
    Neither the change to APRM setpoints T-factor nor the change to 
the LPRM calibration frequency significantly effects the thermal 
limits calculation, and, therefore, do not result in an increase in 
core damage frequency. The change in the parameters used to measure 
reactivity for calculation of the reactivity anomaly has no affect 
on the core damage frequency because the allowable reactivity 
anomaly criteria remains unchanged. Therefore, the proposed changes 
do not involve a 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 Documenmt Room location: Government Documents 
Department, Louisiana State University, Baton Rouge, LA 70803
    Attorney for licensee: Mark Wetterhahn, Esq., Winston & Strawn,

[[Page 55034]]

1400 L Street, N.W., Washington, D.C. 20005
    NRC Project Director: William D. Beckner

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

    Date of amendment request: August 29, 1996
    Description of amendment request: The proposed amendment would 
provide a revision to the reactor pressure vessel (RPV) surveillance 
capsule withdrawal schedule for the River Bend Station. The first 
surveillance capsule would be withdrawn at 10.4 effective full power 
years (EFPY) rather than at 6EFPY.
    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 an accident previously evaluated.
    Pressure-temperature (P-T) limits (RBS Technical Specifications 
Figure 3.4.11-1) are imposed on the reactor coolant system to ensure 
that adequate safety margins against nonductile or rapidly 
propagating failure exist during normal operation, anticipated 
operational occurrences, and system hydrostatic tests. The P-T 
limits are related to the nil-ductility reference temperature, 
RTNDT, as described in ASME Section III, Appendix G. Changes in 
the fracture toughness properties of RPV beltline materials, 
resulting from the neutron irradiation and the thermal environment, 
are monitored by a surveillance program in compliance with the 
requirements of 10CFR50, Appendix H. The effect of neutron fluence 
on the shift in the nil-ductility reference temperature of pressure 
vessel steel is predicted by methods give in Regulatory Guide 1.99, 
Rev. 2.
    River Bend's current P-T limits were established based on 
adjusted reference temperatures developed in accordance with the 
procedures prescribed in Reg. Guide 1.99, Rev. 2, Regulatory 
Position 1. Calculation of adjusted reference temperature by these 
procedures includes a margin term to ensure conservative, upper-
bound values are used for the calculation of the P-T limits. 
Revision of the first capsule withdrawal schedule will not affect 
the P-T limits because they will continue to be established in 
accordance with Regulatory Position 1 (or other NRC-approved) 
procedures. When permitted (two or more credible surveillance data 
sets available), Regulatory Position 2 (or other NRC-approved) 
methods for determining adjusted reference temperature will be 
followed.
    This change is not related to any accidents previously 
evaluated. The proposed change is a revision of the Withdrawal Time 
for the first surveillance capsule as given in Technical 
Requirements (TR) Table 3.4.11-1 from 6 EFPY to 10.4 EFPY. This 
change will not affect P-T limits as given in RBS Technical 
Specifications Figure 3.4.11-1 or USAR Figures 5.3-4a and 5.3-4b. 
This change will not affect any plant safety limits or limiting 
conditions of operation. The proposed change will not affect reactor 
pressure vessel performance as no physical changes are involved and 
RBS vessel P-T limits will remain conservative in accordance with 
Reg. Guide 1.99, Rev. 2 requirements. The proposed change will not 
cause the reactor pressure vessel or interfacing systems to be 
operated outside of their design or testing limits. Also, the 
proposed change will not alter any assumptions previously made in 
evaluating the radiological consequences of accidents. Therefore, 
the probability or consequences of accidents previously evaluated 
will not be increased by the proposed change.
    2. The proposed changes do not create the possibility of a new 
or different kind of accident from any accident previously 
evaluated.
    The proposed change is a revision of the Withdrawal Time in TR 
Table 3.4.11 for the first RPV material surveillance capsule from 6 
EFPY to 10.4 EFPY. This proposed change does not involve a 
modification of the design of plant structures, systems, or 
components. The proposed change will not impact the manner in which 
the plant is operated as plant operating and testing procedures will 
not be affected by the change. The proposed change will not degrade 
the reliability of structures, systems or components important to 
safety (ITS) as equipment protection features will not be deleted or 
modified, equipment redundancy or independence will not be reduced, 
supporting system performance will not be downgraded, the frequency 
of operation of ITS equipment will not be increased, and increased 
or more severe testing of ITS equipment will not be imposed. No new 
accident types or failure modes will be introduced as a result of 
the proposed change. Therefore, the proposed change does not create 
the possibility of a new or different kind of accident from that 
previously evaluated.
    3. The proposed changes do not involve a significant reduction 
in a margin of safety.
    As stated in the River Bend SER, ``Appendices G and H of 10CFR50 
describe the conditions that require pressure-temperature limits and 
provide the general bases for these limits. These appendices 
specifically require that pressure-temperature limits must provide 
safety margins at least as great as those recommended in the ASME 
Code, Section III, Appendix G. .... Until the results from the 
reactor vessel surveillance program become available, the staff will 
use RG 1.99, Revision 1 [now Revision 2] to predict the amount of 
neutron irradiation damage. ... The use of operating limits based on 
these criteria--as defined by applicable regulations, codes, and 
standards--will provide reasonable assurance that nonductile or 
rapidly propagating failure will not occur, and will constitute an 
acceptable basis for satisfying the applicable requirements of GDC 
31.''
    Bases for RBS Technical Specification 3/4/11 states: ``The P/T 
limits are not derived from Design Basis Accident (DBA) analyses. 
They are prescribed during normal operation to avoid encountering 
pressure, temperature, and temperature rate of change conditions 
that might cause undetected flaws to propagate and cause nonductile 
failure of the RCPB [Reactor Coolant Pressure Boundary], a condition 
that is unanalyzed. ... Since the P/T limits are not derived from 
any DBA, there are no acceptance limits related to the P/T limits. 
Rather, the P/T limits are acceptance limits themselves since they 
preclude operation in an unanalyzed condition.''
    The proposed change will not affect any safety limits, limiting 
safety system settings, or limiting conditions of operation. The 
proposed change does not represent a change in initial conditions, 
or in a system response time, or in any other parameter affecting 
the course of an accident analysis supporting the Bases of any 
Technical Specification. The proposed change does not involve 
revision of the P-T limits but rather a revision of the Withdrawal 
Time for the first surveillance capsule. The current P-T limits were 
established based on adjusted reference temperatures for vessel 
beltline materials calculated in accordance with Regulatory Position 
1 of Reg. Guide 1.99, Rev. 2. P-T limits will continue to be revised 
as necessary for changes in adjusted reference temperature due to 
changes in fluence according to Regulatory Position 1 until two or 
more credible surveillance data sets become available. When two or 
more credible surveillance data sets become available, P-T limits 
will be revised as prescribed by Regulatory Position 2 of Reg. Guide 
1.99, Rev. 2 or other NRC-approved guidance. Therefore, the proposed 
changes do not involve a significant reduction in any margins 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, D.C. 20005
    NRC Project Director: William D. Beckner

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

    Date of amendment request: September 23, 1996
    Description of amendment request: The proposed amendment would 
revise the Crystal River Unit 3 (CR 3) technical specifications (TS) to 
delete a note

[[Page 55035]]

associated with Surveillance Requirement (SR) 3.3.7.1 for the 
Engineered Safeguard Actuation System (ESAS) Automatic Actuation Logic. 
Applicable TS Bases will also be revised to reflect the proposed TS 
change.
    SR 3.3.7.1 requires periodic testing of the ESAS automatic 
actuation logic matrix to demonstrate that the required logic 
combinations are operable. When the ESAS automatic actuation logic is 
placed in an inoperable status solely for performing of this 
surveillance, the note associated with the SR 3.3.7.1 provides relief 
in that it allows not entering into applicable Conditions and Required 
Actions for up to 8 hours, provided the associated engineering 
safeguards (ES) function is maintained. The licensee has determined 
that because of the CR 3 design of the ESAS System and the way the test 
is performed, maintenance of the ``associated ES function'' is not 
possible. Thus, the note does not provide the relief intended and 
therefore, the licensee proposes to delete the note. During the 
performance of the ESAS test and bypassing the associated ES function, 
the licensee proposes to enter into applicable TS Conditions.
    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 will not significantly increase the 
probability or consequences of an accident previously evaluated 
because unavailability of equipment is recognized in the design of 
the plant and in the Technical Specifications. The probability and 
consequences of accidents previously evaluated are bounded by the 
evaluations done for the allowed outage time of the associated 
functions.
    2. The proposed change will not create the possibility of a new 
or different kind of accident from any accident previously evaluated 
because the bypassing of ES functions for testing purposes does not 
place the plant in a configuration which would allow the possibility 
of a new or different kind or accident to be created.
    3. The proposed change will not involve a significant reduction 
to the margin of safety because deleting the NOTE does not effect 
the way the test is performed. The test is required by the Technical 
Specifications and will still be performed in the same manner. Thus, 
there is no change in the unavailability of the system as a result 
of this change and the margin of safety is not reduced.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 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: Coastal Region Library, 8619 
W. Crystal Street, Crystal River, Florida 32629
    Attorney for licensee: A. H. Stephens, General Counsel, Florida 
Power Corporation, MAC - A5D, P. O. Box 14042, St. Petersburg, Florida 
33733
    NRC Project Director: Frederick J. Hebdon

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

    Date of amendment request: September 27, 1996
    Description of amendment request: The proposed amendment would 
revise the Crystal River 3 (CR3) post-accident monitoring (PAM) 
instrumentation technical specification (TS). Specifically, the 
following TS changes are proposed:
    A. Table 3.3.17-1, Function 8: The descriptor is changed from 
``Containment Pressure (Narrow Range)'' to ``Containment Pressure 
(Expected Post-Accident Range).''
    B. Table 3.3.17-1, Function 18: The required channels for Core Exit 
Temperature (Backup) is changed from ``2 sets of 5'' to ``3 per core 
quadrant.''
    C. Table 3.3.17-1: A new Function 20 is added and designated as 
``Low Pressure Injection Flow.''
    D. Table 3.3.17-1: A new Function 21 is added and designated as 
``Degrees of Subcooling.''
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration (the letters A, B, C and D correspond to the proposed TS 
changes), which is presented below:
    1. The proposed changes will not significantly increase the 
probability or consequences of an accident previously evaluated 
because:
    A/B. The changes in containment pressure and core exit 
thermocouple nomenclature do not reflect any physical changes to the 
facility.
    C/D.The addition of low pressure injection flow and degrees of 
subcooling to the Post-Accident Monitoring Instrumentation LCO is 
being done to comply with a commitment made during the technical 
specification improvement program to include in the technical 
specifications, that instrumentation which monitors variables 
classified as Type A in accordance with Regulatory Guide 1.97. These 
two variables have recently been re-classified as Type A. The 
associated instruments are used after an accident occurs to prompt 
the operators to take certain mitigative actions. Therefore, the 
probability of an accident occurring is unaffected. As part of the 
re-classification of these variables to Type A, the associated 
monitoring instrumentation will be under more strict surveillance 
and control, which provides additional assurance that the prescribed 
manual operator actions will be implemented when necessary. This, in 
turn, assures the previously evaluated accident consequences remain 
valid.
    2. The proposed changes will not create the possibility of a new 
or different kind of accident from any accident previously evaluated 
because:
    A/B. The changes in containment pressure and core exit 
thermocouple nomenclature do not reflect any physical changes to the 
facility. The changes provide clarification for the instruments 
which are required to comply with the LCO.
    C/D. The addition of low pressure injection flow and degrees of 
subcooling to the Post-Accident Monitoring instrumentation LCO is 
being done to comply with a commitment made during the technical 
specification improvement program to include in the technical 
specifications, that instrumentation which monitors variables 
classified as Type A in accordance with Regulatory Guide 1.97. These 
two variables have been re-classified as Type A. The associated 
instruments are used after an accident occurs to prompt the 
operators to take certain mitigative actions. Since the 
instrumentation is used only post-accident, these changes do not 
create the possibility of a new or different kind of accident.
    3. The proposed change will not involve a significant reduction 
to the margin of safety because:
    A/B. The changes in containment pressure and core exit 
thermocouple nomenclature have no affect on the margin of safety. 
The changes provide clarification of the technical specifications. 
This reduces the potential for confusion regarding this 
instrumentation.
    C/D. The addition of low pressure injection flow and degrees of 
subcooling to the post-accident monitoring instrumentation table 
adds controls on the OPERABILITY of post-accident monitoring 
instrumentation providing greater assurance it will be available 
should an accident occur.
    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: Coastal Region Library, 8619 
W. Crystal Street, Crystal River, Florida 32629
    Attorney for licensee: A. H. Stephens, General Counsel, Florida 
Power Corporation, MAC - A5D, P. O. Box 14042, St. Petersburg, Florida 
33733
    NRC Project Director: Frederick J. Hebdon

[[Page 55036]]

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

    Date of amendment request: September 5, 1996
    Description of amendment request: The proposed change deletes 
License Condition 2.C.5, Integrated Implementation Schedule.
    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:
    In accordance with 10CFR50.92, NNECO has reviewed the attached 
proposed change and has concluded that it does not involve a 
significant hazards consideration (SHC). The basis for this is that 
the three criteria of 10CFR50.92(c) are not compromised. The 
proposed change does not involve an SHC because the change would 
not:
    1. Involve a significant increase in the probability or 
consequences of an accident previously evaluated.
    Operation of the facility in accordance with the proposed change 
would result in a change in an administrative process for 
prioritizing and scheduling projects and engineering evaluations. 
With the limited number of NRC required projects remaining to be 
implemented, the IIS [Integrated Implementation Schedule] is no 
longer required to schedule resources for the remaining topics. 
Since this license condition only involves an administrative 
process, it does not directly affect the design or operation of the 
plant. Therefore, no accident analyses are affected by the change, 
and the change does not increase the probability or consequences of 
any previously evaluated accident.
    2. Create the possibility of a new or different kind of accident 
from any previously evaluated.
    The proposed license modification removes a requirement relating 
to the scheduling of modifications and engineering evaluations. 
Because the license condition addresses only an administrative 
scheduling mechanism, it does not affect directly the design or 
operation of the plant. Therefore, the proposed change does not 
create a different kind of accident from those previously analyzed.
    3. Involve a significant reduction in a margin of safety.
    The proposed license modification removes a requirement relating 
to the scheduling of modifications and engineering evaluations. The 
original purpose of the IIS and the ISAP [Integrated Safety 
Assessment Program] was to prioritize and schedule modifications and 
engineering evaluations in a manner that was agreed upon by both 
NNECO and the NRC. These programs were especially important to 
Millstone Unit No. 1 for priorization of topics associated with the 
SEP [Systematic Evaluation Program] and the TMI [Three Mile Island] 
Action Plan. This program is considered to be no longer necessary. 
Modifications and engineering evaluations will be scheduled and 
prioritized using other methodologies. Since this change involves an 
administrative process only, there is no direct impact on the design 
or operation of the plant, and therefore, no significant reduction 
in the margin of safety.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: Learning Resources Center, 
Three Rivers Community-Technical College, 574 New London Turnpike, 
Norwich, CT 06360, and the Waterford Library, ATTN: Vince Juliano, 49 
Rope Ferry Road, Waterford, CT 06385
    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

PECO Energy Company, Public Service Electric and Gas Company, 
Delmarva Power and Light Company, and Atlantic City Electric 
Company, Dockets Nos. 50-277 and 50-278, Peach Bottom Atomic Power 
Station, Units Nos. 2 and 3, York County, Pennsylvania

    Date of application for amendments: August 27, 1996
    Description of amendment request: The proposed amendment revises 
the required value of control rod drive (CRD) system pressure in 
technical specification (TS) 3.10.8, ``Shutdown Margin (SDM) Test-
Refueling.''
    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 any accident previously 
evaluated.
    The proposed changes are purely administrative and do not 
involve any physical changes to plant SSC [systems, structures and 
components]. The change in the minimum CRD charging water header 
pressure from 955 psig to 940 psig was previously approved in TS 
Amendments Nos. 211 and 216 for PBAPS [Peach Bottom Atomic Power 
Station], Units 2 and 3. TS Change Request 95-12 was incomplete by 
inadvertently failing to identify the need to change requirement (f) 
of LCO [Limiting Condition for Operation] 3.10.8. Therefore, the 
proposed changes will not increase the probability of occurrence or 
the consequences of an accident previously evaluated in the SAR 
[safety analysis report].
    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 are purely administrative and do not 
involve any physical changes to plant SSC. The proposed changes do 
not allow plant operation in any mode that is not already evaluated 
in the SAR. Therefore, the possibility of a different type of 
accident than previously evaluated in the SAR is not created.
    3) The proposed changes do not result in a significant reduction 
in the margin of safety.
    The proposed changes are purely administrative and have no 
impact on any safety analysis assumptions or margins of safety. A 
change to SR 3.10.8.6 was approved by the NRC by TS Amendment Nos. 
211 and 216. LCO 3.10.8 requirement (f) should have been changed at 
the same time to reflect a minimum CRD charging water pressure of 
940 psig. Changing LCO 3.10.8 requirement (f) to reflect TS 
Amendment Nos. 211 and 216 is purely administrative, and therefore, 
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: Government Publications 
Section, State Library of Pennsylvania, (REGIONAL DEPOSITORY) Education 
Building, Walnut Street and Commonwealth Avenue, Box 1601, Harrisburg, 
PA 17105
    Attorney for licensee: J. W. Durham, Sr., Esquire, Sr. V. P. and 
General Counsel, PECO Energy Company, 2301 Market Street, Philadelphia, 
PA 19101
    NRC Project Director: John F. Stolz

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

    Date of amendment request: May 20, 1996
    Description of amendment request: The proposed Technical 
Specifications (TS) changes would revise TS Sections 3/4.4.9.2, 3/
4.9.11.1, 3/4.9.11.2, and the associated TS Bases 3/4.4.9 and 3/4.9.11, 
to more clearly describe that the Residual Heat Removal (RHR) system 
Shutdown Cooling mode of operation consists of four (4) ``subsystems.'' 
These TS sections pertain to plant operations during Operational 
Conditions (OPCONs) 4, ``Cold Shutdown'' and 5, ``Refueling.'' In 
addition, the proposed TS change would make administrative changes to 
TS Section 3/4.4.9.1 to

[[Page 55037]]

ensure consistency in terminology regarding the description of Shutdown 
Cooling ``subsystems.'' The proposed TS changes are consistent with the 
guidance delineated in the Improved TS (i.e., NUREG-1433, Revision 1, 
``Standard Technical Specifications General Electric Plants, BWR/4,'' 
dated April 1995) which indicates that the RHR Shutdown Cooling mode of 
operation is comprised of two (2) loops and four (4) subsystems (i.e., 
two (2) subsystems per loop).
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:
    1. The proposed Technical Specifications (TS) changes do not 
involve a significant increase in the probability or consequences of 
an accident previously evaluated.
    The proposed TS changes do not involve any physical changes to 
plant structures systems, or components. The RHR [Residual Heat 
Removal] Shutdown Cooling mode of operation is manually controlled 
and is not required for accident mitigation. The RHR system will 
continue to function as designed in all modes of operation. The 
consequences of equipment malfunction are not changed from those in 
existing analyses, with no increase in onsite or offsite 
radiological effects. The RHR system will continue to function as 
designed to mitigate the consequences of an accident and resultant 
onsite and offsite radiological effects remain as previously 
evaluated. The proposed TS changes will revise the TS to more 
clearly describe the RHR system configuration in OPCONs 4 and 5. The 
proposed changes are consistent with the guidance stipulated in 
NUREG-1433, Revision 1.
    The four (4) ``subsystem'' Shutdown Cooling designation permits 
operability of only one (1) RHR heat exchanger for Shutdown Cooling 
service in Operational Conditions (OPCONs) 4 and 5, as long as both 
associated RHR pumps are operable and alignable for Shutdown 
Cooling. TS requirements for RHR Shutdown Cooling operation in Hot 
Shutdown, Suppression Pool Spray, and Suppression Pool Cooling 
continue to require two (2) independent loops to be operable in 
OPCONs 1, 2, and 3*, meaning both RHR heat exchangers will still be 
required to be operable throughout OPCON 3.
    The four (4) ``subsystem'' Shutdown Cooling designation has no 
effect on the required operability of the Residual Heat Removal 
Service Water (RHRSW) system. As required by TS Section 3.7.1.1, the 
RHRSW subsystem(s) associated with the required operable RHR heat 
exchanger(s) will continue to remain operable. Each operable RHRSW 
subsystem consists of two (2) operable pumps and the required 
operable flowpath to provide decay heat removal via the associated 
RHR heat exchanger.
    The RHRSW system piping is designed, fabricated, inspected, and 
tested in accordance with the requirements of ASME [American Society 
of Mechanical Engineers], Section III Class 3, and each RHRSW 
subsystem is single active failure proof in that the failure of a 
motor-operated valve, diesel generator, or pump does not prevent the 
system from performing its safety function.
    The required availability of four (4) loops of the Low Pressure 
Coolant Injection (LPCI) mode of RHR during OPCONs 1, 2, and 3 as 
required by TS Section 3.5.1 is not impacted by the four (4) 
``subsystem'' Shutdown Cooling designation. No change to any RHR 
system instrumentation logic, required Emergency Core Cooling System 
(ECCS) availability, or method of operation is involved.
    NUREG-1433, Revision 1, also re-affirms that each Shutdown 
Cooling ``subsystem'' is considered operable if it can be manually 
aligned, remotely or locally, in the shutdown cooling mode for 
removal of decay heat. Thus, a LPCI-dedicated pump can be aligned 
for LPCI automatic initiation, yet still be considered part of an 
operable shutdown cooling subsystem as long as it can be re-aligned 
for Shutdown Cooling.
    Therefore, the proposed TS changes do not involve an increase in 
the probability or consequences of an accident previously evaluated.
    2. The proposed TS changes do not create the possibility of a 
new or different kind of accident from any accident previously 
evaluated.
    The proposed TS changes do not involve any physical changes to 
plant structures, systems, or components. The RHR system will 
continue to function as designed in all modes of operation. No new 
accident type is created as a result of the proposed changes. No new 
failure mode for any equipment is created. The changes are 
consistent with the guidance provided in NUREG-1433, Revision 1, 
pertaining to RHR Shutdown Cooling operation in OPCONs 4 and 5.
    The four (4) ``subsystem'' Shutdown Cooling designation has no 
effect on the required operability of the RHRSW system. The RHRSW 
subsystem(s) associated with the required operable RHR heat 
exchanger(s) will continue to remain operable as required by TS 
Section 3.7.1.1. Each operable RHRSW subsystem consists of two (2) 
operable pumps and the required operable flowpath to provide decay 
heat removal via the associated RHR heat exchanger.
    The RHRSW system piping is designed, fabricated, inspected, and 
tested in accordance with the requirements of ASME, Section III, 
Class 3, and each RHRSW subsystem is single active failure proof in 
that the failure of a motor-operated valve, diesel generator, or 
pump does not prevent the system from performing its safety 
function.
    The required availability of four (4) loops of the LPCI mode of 
RHR during OPCONs 1, 2, and 3 as required by TS Section 3.5.1 and 
3.5.2 is not impacted by the four (4) ``subsystem'' Shutdown Cooling 
designation. No change to any RHR system instrumentation logic, 
required ECCS availability, or method of operation is involved.
    NUREG-1433, Revision 1, also re-affirms that each Shutdown 
Cooling ``subsystem'' is considered operable if it can be manually 
aligned, remotely or locally, in the Shutdown Cooling mode for 
removal of decay heat. Thus, a LPCI-dedicated pump can aligned be 
[sic] [be aligned] for automatic LPCI initiation, yet still be 
considered part of an operable shutdown cooling subsystem as long as 
it can be re-aligned for Shutdown Cooling.
    Therefore, the proposed TS changes do not create the possibility 
of a new or different kind of accident from any accident previously 
evaluated.
    3. The proposed TS changes do not involve a significant 
reduction in a margin of safety.
    Although the Bases for TS Sections 3/4.4.9.2, 3/4.9.11.1, and 3/
4.9.11.2 are being revised in support of this proposed TS change, 
the changes only involve providing clarification regarding the 
designation of the RHR Shutdown Cooling operation configuration in 
OPCONs 4 and 5. The proposed TS changes do not involve any physical 
changes to plant structures, systems, or components. The RHR system 
will continue to function as designed in all modes of operation. The 
consequences of equipment malfunction are not changed from those in 
existing analyses, with no increase in onsite or offsite 
radiological effects. The RHR system will continue to function as 
designed to mitigate the consequences of an accident and resultant 
onsite and offsite radiological effects remain as previously 
evaluated. The proposed changes are consistent with the guidance 
stipulated in NUREG-1433, Revision 1.
    The four (4) ``subsystem'' Shutdown Cooling designation has no 
effect on the required operability of the RHRSW system. As required 
by TS 3.7.1.1, the RHRSW subsystem(s) associated with the required 
operable RHR heat exchanger(s) will continue to remain operable. 
Each operable RHRSW subsystem consists of two (2) operable pumps and 
the required operable flowpath to provide decay heat removal via the 
associated RHR heat exchanger.
    The RHRSW system piping is designed, fabricated, inspected, and 
tested in accordance with the requirements of ASME, Section III, 
Class 3, and each RHRSW subsystem is single active failure proof in 
that the failure of a motor-operated valve, diesel generator, or 
pump does not prevent the system from performing its safety 
function. (In the same manner that manual action may be required for 
RHR system alignment in OPCONs 4 and 5 with one (1) RHR heat 
exchanger operable, a failure of the motor-operated RHRSW inlet or 
outlet heat exchanger isolation valves may require manual 
positioning for the required alignment.)
    The required availability of four (4) loops of the LPCI mode of 
RHR during OPCONs 1, 2, and 3* as required by TS Section 3.5.1 is 
not affected by the four (4) ``subsystem'' Shutdown Cooling 
configuration. No change to any RHR system instrumentation logic, 
required ECCS availability, or method of operation is involved.
    NUREG-1433, Revision 1, also re-affirms that each Shutdown 
Cooling ``subsystem'' is

[[Page 55038]]

considered operable if it can be manually aligned, remotely or 
locally, in the Shutdown Cooling mode for removal of decay heat. 
Thus, a LPCI-dedicated pump can be aligned for LPCI automatic 
initiation, yet still be considered part of an operable Shutdown 
Cooling ``subsystem'' as long as it can be re-aligned for Shutdown 
Cooling.
    Therefore, the proposed TS changes do not involve a significant 
reduction in a margin of safety.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: Pottstown Public Library, 500 
High Street, Pottstown, PA 19464
    Attorney for licensee: J. W. Durham, Sr., Esquire, Sr. V. P. and 
General Counsel, Philadelphia Electric Company, 2301 Market Street, 
Philadelphia, PA 19101
    NRC Project Director: John F. Stolz

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

    Date of amendment request: June 28, 1996
    Description of amendment request: The proposed Technical 
Specifications (TS) changes would incorporate performance-based 
testing, in accordance with 10 CFR Part 50, Appendix J, ``Primary 
Reactor Containment Leakage Testing For Water-Cooled Power Reactors,'' 
Option B. This option allows utilities to extend the frequencies of the 
Type A Containment (ILRT) Leak Rate Test and Type B and C Local Leak 
Rate Tests (LLRTs) based on the performance and design of the 
containment and components.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:
    1. The proposed TS changes do not involve a significant increase 
in the probability or consequences of an accident previously 
evaluated.
    Incorporation of the new 10 CFR 50, Appendix J, Option B at LGS, 
Units 1 and 2 does not increase the probability of occurrence of an 
accident previously evaluated. The containment structure including 
its isolation capability is not an accident initiator.
    These changes do not involve any changes to the containment 
structure, system or components which could increase the probability 
of occurrence of an accident previously evaluated or act as a new 
accident initiator. Implementation of the proposed changes will 
affect the manner in which these structures, systems, or components 
(SSCs) are tested; however, the new testing schedule is not an 
initiator of any analyzed event. No equipment changes are involved 
with adoption of Option B; therefore, performance-based test 
intervals for Type A, B, and C tests do not increase the probability 
of occurrence of a malfunction of equipment important to safety 
previously evaluated. No physical changes are being made to the 
plant, nor are there any changes being made in the operation of the 
plant as the result of increasing the test intervals. Additionally, 
the proposed TS changes will not alter the operation of equipment 
available for the mitigation of accidents or transients, therefore, 
this change will not result in any significant increase to onsite or 
offsite dose previously evaluated. The potential for time-based and 
activity-based failure mechanisms which could lead to excessive 
containment leakage has been determined to be minimal. Performance-
based test intervals for Type A, B, and C tests will not alter any 
safety limits which ensure the integrity of fuel barriers, and will 
not increase the primary containment leakage limits.
    Performance-based test intervals for Type A, B, and C leak tests 
do not increase the consequences of an accident previously 
evaluated. NUREG-1493 concluded that reducing the frequency of Type 
A tests from the current three per ten years to one per ten years 
was found to lead to an imperceptible increase in risk. NUREG-1493 
includes the results of a sensitivity study performed to explore the 
risk impact of several alternative leak rate test schedules. The 
estimated increase in population exposure risk ranged from 0.02% to 
0.14%. The risk impact was determined to be very small since Type B 
and C testing (local leak rate tests) detect a very large percentage 
of overall containment leakages. The percentage of leakages detected 
by Type A tests is very small. Past test results experienced at 
Limerick Units 1 and 2 concur with these determinations. NUREG-1493 
also concluded that the overall unit risk is not very sensitive to 
changes in containment leakage rates. Given the insensitivity of 
risk to containment leak rates and the small fraction of leak paths 
detected solely by the Type A tests, increasing the interval between 
Type A tests is possible with minimal impact on public risk.
    NUREG-1493 also concluded that, based on a model of component 
failure with time, the performance-based alternatives to current, 
local-leakage testing requirements are feasible without significant 
risk impact. The LGS design and past performance is bounded by the 
NUREG study. The NUREG model indicated that the number of components 
tested could be reduced by about 60% with less than a three-fold 
increase in the incremental risk due to containment leakage. Since 
under existing requirements, leakage contributes less than 0.1 
percent of overall accident risk, the overall impact is very small.
    Therefore, the proposed TS changes will not involve a 
significant increase in the probability or consequences of an 
accident previously evaluated.
    2. The proposed TS changes do not create the possibility of a 
new or different kind of accident from any accident previously 
evaluated.
    Performance-based test intervals for Type A, B, and C leak tests 
do not introduce a new or different type of accident or create the 
possibility of a different type of malfunction of equipment 
important to safety than previously evaluated. No physical changes 
are being made to the plant, nor are there any changes being made in 
the operation of the plant as the result of increasing the test 
intervals. No new failure modes of plant equipment previously 
evaluated will be introduced. Additionally, the TS changes will not 
alter the operation of equipment available for the mitigation of 
accidents or transients. The safety function of the primary 
containment will be retained since the containment will continue to 
provide an essentially leak tight barrier against the uncontrolled 
release of radioactivity to the environment for postulated accidents 
previously evaluated.
    Therefore, the proposed TS changes will not create the 
possibility of a new or different kind of accident from any accident 
previously evaluated.
    3. The proposed TS changes do not involve a significant 
reduction in a margin of safety.
    The margin of safety is not reduced as a result of adopting 10 
CFR 50, Appendix J, Option B. The effect of increasing containment 
leakage rate testing intervals was evaluated in NUREG-1493 using 
historical industry leakage rate testing results. Performance 
history at LGS is consistent with the conclusions reached in NUREG-
1493 and NEI 94-01. The results of the NUREG evaluation conclude 
that the increased safety risk corresponding to the extended test 
intervals is small (less than 0.1% of total risk). The revised TS 
will continue to maintain the allowable leakage rate for the Type A 
tests. In addition, the requirement to perform a periodic general 
visual inspection of the primary containment has been maintained at 
the original interval of three times in 10 years as part of the 
performance-based leakage rate testing program.
    The risk of a non-detectable increase of primary containment 
leakage is considered to be negligible due to the conclusion that 10 
CFR 50, Appendix J, Type B and C testing program will continue to be 
conducted between Type A tests. A review of previous LGS Type A test 
results has concluded that the only failure mechanisms are activity-
based. There is no indication of time-based failures that would not 
be identified during the performance of Type B and C tests. 
Therefore, we have concluded that the proposed adoption of the 
Option B intervals would not result in a non-detectable primary 
containment leakage rate in excess of the allowable value (i.e., 
0.5% wt/day) established by the LGS TS.
    The proposed TS will continue to maintain the allowable leakage 
rate for the combined Type B and C tests. As supported by the 
findings of NUREG-1493, the percentage of leakages detected by Type 
A tests is small (as

[[Page 55039]]

stated above) and Type B and C leakage tests are capable of 
detecting more than 97% of containment leakages and virtually all 
such leakages are identified by local leak rate tests of containment 
isolation valves. The Type B and C test intervals will be 
established through the PCLRTP for each component based on design 
and previous LGS test performance history.
    Therefore, the proposed TS changes do not involve a reduction in 
a margin of safety.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: Pottstown Public Library, 500 
High Street, Pottstown, PA 19464
    Attorney for licensee: J. W. Durham, Sr., Esquire, Sr. V. P. and 
General Counsel, Philadelphia Electric Company, 2301 Market Street, 
Philadelphia, PA 19101
    NRC Project Director: John F. Stolz

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

    Date of amendment request: September 25, 1996
    Description of amendment request: The amendments would relocate to 
the Salem Updated Final Safety Analysis Report the list of containment 
isolation valves that are currently located in Table 3.6-1 of Technical 
Specification 3.6.3. In addition, references to the table in 
specifications 1.7, 3.6.1, and 3.6.3 are being updated.
    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 consequence of an accident previously 
evaluated.
    The proposed changes simplify the TS, meet the regulatory 
requirements for control of containment isolation, and are 
consistent with the guidance provided in Generic Letter (GL) 91-08, 
``Removal of Component Lists from Technical Specifications.'' The 
procedural details of TS Table 3.6-1 have not been changed, only 
relocated to a different controlling document, the Salem Update 
[sic] [Updated] Final Safety Analysis Report (UFSAR). The proposed 
changes are administrative in nature, should result in improved 
administrative practices, and do not affect plant operations.
    The probability of occurrence of a previously evaluated accident 
is not increased because this change does not introduce any new 
potential accident initiating conditions. The consequences of an 
accident previously evaluated is not increased because the ability 
of containment to restrict the release of any fission product 
radioactivity to the environment will not be degraded by this 
change.
    2. The proposed change does not create the possibility of a new 
or different kind of accident from any accident previously 
evaluated.
    The proposed changes are administrative in nature, do not result 
in a physical alterations or changes to the operation of the plant, 
and cause no change in the method by which any safety-related system 
performs its functions. Therefore, this proposed change will not 
create the possibility of a new or different kind of accident from 
any previously evaluated.
    3. The proposed change does not involve a significant reduction 
in a margin of safety.
    The administrative change to relocate TS Table 3.6-1 to the 
UFSAR does not alter the basic regulatory requirements for 
containment isolation and will not adversely affect the containment 
isolation capability for credible accident scenarios. Adequate 
control of the content of the relocated table is assured by the 
10CFR50.59 review process.
    The proposed relocation of TS Table 3.6-1 does not alter the 
requirements for CIV operability currently in the TS. the Limiting 
Condition for Operation and the Surveillance Requirements would be 
retained in the revised TS. Therefore, the proposed changes will not 
affect the meaning, application, and function of the current TS 
requirements for the CIVs.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: Salem Free Public library, 112 
West Broadway, Salem, NJ 08079
    Attorney for licensee: Mark J. Wetterhahn, Esquire, Winston and 
Strawn, 1400 L Street, NW, Washington, DC 20005-3502
    NRC Project Director: John F. Stolz

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

    Date of amendment request: September 25, 1996
    Description of amendment request: The amendments would change 
Technical Specification 3/4.8.1, ``Electrical Power Systems,'' to 
revise the Emergency Diesel Generator (EDG) voltage and frequency 
limits as a result of updated EDG load calculations and to eliminate 
ambiguity in the testing methodology for EDG start timing.
    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.
    Since no change is being made to the offsite power supplies, or 
to any system or component that interfaces with the offsite power 
supplies, there is no change in the probability of a Loss of Offsite 
Power Accident.
    The proposed changes provide the necessary conservatism for 
voltage and frequency to ensure the EDGs are not run in an 
overloaded condition and that driven equipment is not damaged during 
steady state operation following a Loss of Offsite Power coincident 
with a Loss of Coolant Accident. Since the narrower band of voltage 
and frequency for the isochronous mode continues to ensure proper 
steady state operation of the EDG and associated driven equipment, 
there is no change in the consequences of an accident previously 
evaluated.
    Based on the above, the proposed amendment does not involve a 
significant increase in the probability or consequences of an 
accident previously evaluated.
    2. The proposed change does not create the possibility of a new 
or different kind of accident from any accident previously 
evaluated.
    The proposed amendment does not result in any design or physical 
configuration changes to the EDGs. Proposed changes made to the 
testing parameters and testing methodology will not cause a new or 
different accident since the EDGs are used for accident mitigation 
and no new failure modes are being introduced. Therefore, the 
proposed amendment will not create the possibility of a new or 
different kind of accident from any previously evaluated.
    3. The proposed change does not involve a significant reduction 
in a margin of safety.
    The proposed amendment provides further conservatism to the 
voltage and frequency band currently specified in the TSs. The 
proposed voltage and frequency changes ensure the EDG will not be 
overloaded from an over-frequency condition and driven equipment 
will not be damaged from an over-voltage condition.
    The control system is set to control the EDG voltage within the 
bands specified in the requested changes. The changes are consistent 
with current calculations and within the capability of the controls. 
Since the narrower band of voltage and frequency for the isochronous 
mode is bounded by the existing TS, there is no change in the margin 
of safety. The increased band for droop mode will ensure the EDG is 
capable of operating in accordance with normal offsite power 
parameters and does not reduce the margin of safety.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are

[[Page 55040]]

satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: Salem Free Public library, 112 
West Broadway, Salem, NJ 08079
    Attorney for licensee: Mark J. Wetterhahn, Esquire, Winston and 
Strawn, 1400 L Street, NW, Washington, DC 20005-3502
    NRC Project Director: John F. Stolz

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

    Date of amendment request: October 1, 1996
    Description of amendment request: The proposed amendments would 
change Technical Specifications (TSs) 3/4.7.1.5, ``Main Steam Line 
Isolation Valves (MSIVs),'' and 3/4.3.2, ``Engineered Safety Feature 
Actuation System Instrumentation.'' These changes are needed to 
accommodate entry into Modes 3 and 2 prior to performing MSIV closure 
time testing in Mode 2. The proposed amendments would also allow for 
the repair and testing of inoperable MSIVs in certain operating Modes, 
and would change the low steam line pressure trip setpoint value for 
safety injection to make it consistent with the previously approved 
value for steam line isolation.
    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 isolation capability of the MSIVs and the protective 
functions of the low steam line pressure channels are necessary for 
accident mitigation and do not impact the probability of an 
accident. MSIV testing in the higher modes is necessary to obtain 
conditions which enable testing of the MSIVs. These conditions are 
consistent with the current accident analyses for main steam line 
breaks and secondary system depressurization. Failure of a MSIV, 
which could be encountered during testing, is accounted for in the 
accident analyses.
    Provisions for entering Mode 2 within six hours with an 
inoperable MSIV allows operators to remove the plant from power 
generation in a more controlled manner without challenging plant 
safety systems and is consistent with other plant shutdown TS (i.e., 
TS 3.0.3). The additional six hours to Hot Shutdown, should MSIV 
closure be infeasible, does not result in a significant increase in 
the probability or consequence of an accident since this is a very 
small incremental time addition. The values for the low steam line 
pressure safety injection are higher and are bounded by the present 
accident analysis. The elimination of the obsolete stroke time of 
eight seconds is editorial in nature. As a result, the changes 
proposed do not involve a significant increase in the probability or 
consequence of an accident previously evaluated.
    2. The proposed change does not create the possibility of a new 
or different kind of accident from any accident previously 
evaluated.
    The proposed changes do not involve any modifications to 
existing plant equipment, do not alter the function of any plant 
systems, do not introduce any new operating configurations or new 
modes of plant operation, nor change the safety analyses. The 
proposed changes will, therefore, not create the possibility of a 
new or different kind of accident from any accident previously 
evaluated.
    3. The proposed change does not involve a significant reduction 
in a margin of safety.
    MSIV testing in Mode 2 is within the currently analyzed plant 
operation as discussed in the Updated Final Safety Analysis Report 
(UFSAR) Sections 10.3 and 15.4. These UFSAR sections address 
performance of the TS surveillance test at or near 1000 psig Steam 
Generator pressure to assure main steam isolation occurs within the 
accident conditions, where Steam Generator pressure may be lower 
during Mode 1 operation. The test methodology demonstrating MSIV 
operability is consistent with the accident analysis.
    Operation in Modes 2 and 3 with one or more isolation valve 
inoperable and in the closed position does not impact the margin of 
safety since the valves are already performing the safety function.
    The protective functions that occur as a result of the low steam 
line pressure initiating signal remain bounded by the values assumed 
in the safety analyses. That is, the protective functions that occur 
as a result of this initiating signal already assume a setpoint that 
is conservative for the revised value. The change to the setpoint 
eliminates conflicting information in the TS.
    Therefore, the proposed changes does not involve a significant 
reduction in a margin of safety.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: Salem Free Public library, 112 
West Broadway, Salem, NJ 08079
    Attorney for licensee: Mark J. Wetterhahn, Esquire, Winston and 
Strawn, 1400 L Street, NW, Washington, DC 20005-3502
    NRC Project Director: John F. Stolz

Public Service Electric & Gas Company, Docket No. 50-311, Salem 
Nuclear Generating Station, Unit No. 2, Salem County, New Jersey

    Date of amendment request: September 20, 1996, as supplemented 
September 30, 1996
    Description of amendment request: The proposed amendment would 
change Technical Specification 4.7.7.b.4 to indicate that the specified 
flowrate for the Auxiliary Building Exhaust Air Filtration System 
applies only to system testing.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:
    1. The proposed change does not involve a significant increase 
in the probability or consequences of an accident previously 
evaluated.
    The accident considered in this proposed change is the Loss of 
Coolant Accident (LOCA) as described in Section 15.4 of the UFSAR 
[Updated Final Safety Analysis Report]. The assumption is that: 
``The Auxiliary Building Ventilation System will discharge the vapor 
(from recirculation liquid leakage) to the atmosphere through 
charcoal filters which have an efficiency of 90 percent.'' As such 
the system acts to limit the total offsite and control room 
radiation doses following a LOCA.
    The Auxiliary Building Ventilation System [ABVS] is designed to 
maintain the Auxiliary Building at a negative pressure with respect 
to the atmosphere during normal and emergency operation. Filtration 
of radio-iodines is accomplished by administratively aligning the 
ECCS [emergency core cooling system] equipment areas exhaust flows 
to the standby charcoal adsorber bed if required. The ABVS has no 
direct impact on reactor operation or on any system connected to the 
Reactor Coolant Pressure Boundary.
    The emergency operation of the Auxiliary Building Ventilation 
System is not affected by the proposed changes. The acceptance 
criteria for system performance are not modified by the requested 
change. The change clarifies the intent of SR [surveillance 
requirement] 4.7.7.b.4 and the basis for the flowrates used for 
system acceptance testing. It has been determined that operation of 
the system at lower flow rates than those specified for surveillance 
testing is conservative with respect to the radio-iodine removal 
efficiency assumed for the charcoal adsorber. A higher removal 
efficiency results in lower total exposures at the site boundary and 
within the control room. Additionally, the system is capable of 
maintaining the required negative pressure at the reduced flowrate.
    Given the above, it is concluded that the proposed change does 
not result in an increase in the probability or consequences 
associated with previously analyzed accidents.

[[Page 55041]]

    2. The proposed change does not create the possibility of a new 
or different kind of accident from any accident previously 
evaluated.
    The proposed amendment does not result in any design or 
operational change to the ABVS, to the Nuclear Steam Supply System, 
to the ECCS System, to the Containment Building, to the fuel or to 
the electrical power supplies. Therefore, the proposed amendment 
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.
    Specification 3/4.7.7 and the associated bases were reviewed to 
determine if the proposed changes result in a reduction in the 
margin of safety. The change to SR 4.7.7.b.4 continues to assure 
that the system is operated consistent with the assumptions of the 
accident analysis. The proposed changes to Bases 3/4.7.7 clarify the 
basis for flowrates associated with ABVS surveillance test 
requirements. All changes result in ABVS operation that is just as 
conservative as that assumed in existing analyses.
    The proposed changes do not involve the addition or modification 
of plant equipment, are consistent with the design basis of the ABVS 
as described in the UFSAR, and appropriately limit operation to be 
consistent with the assumptions of the accident analysis. As such 
there is no reduction in a margin of safety.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: Salem Free Public library, 112 
West Broadway, Salem, NJ 08079
    Attorney for licensee: Mark J. Wetterhahn, Esquire, Winston and 
Strawn, 1400 L Street, NW, Washington, DC 20005-3502
    NRC Project Director: John F. Stolz

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

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

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

    Date of application for amendments: June 17, 1996
    Brief description of amendments request: The proposed amendments 
would modify the technical specifications to change (1) the reference 
method for calculating dose conversion factors (DCFs) to be used in 
dose calculations, and (2) the upper and lower limits for operating 
pressurizer pressure to account for new instrument uncertainties and to 
reduce the allowed operating band.
    Date of individual notice in Federal Register: September 11, 1996 
(61 FR 47963)
    Expiration date of individual notice: October 11, 1996
    Local Public Document Room location: Phoenix Public Library, 1221 
N. Central Avenue, Phoenix, Arizona 85004

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

    Date of application for amendments: June 28, 1996
    Brief description of amendments request: The proposed amendments 
would modify the technical specifications to increase the minimum 
required amount of anhydrous trisodium phosphate (TSP) in the 
containment baskets.
    Date of individual notice in Federal Register: September 11, 1996 
(61 FR 47962), as corrected September 26, 1996 (61 FR 50535).
    Expiration date of individual notice: October 11, 1996
    Local Public Document Room location: Phoenix Public Library, 1221 
N. Central Avenue, Phoenix, Arizona 85004

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

    Date of application for amendment: August 23, 1996
    Brief description of amendment request: The proposed amendment 
would revise Paragraph 2.B(2) of
    Facility Operating License No. DPR-40 to allow source materials in 
the form of depleted or natural uranium as reactor fuel and to revise 
Technical Specification 4.3.2 to include depleted uranium in describing 
the reactor core.
    Date of individual notice in Federal Register: August 30, 1996 (61 
FR 45995)
    Expiration date of individual notice: September 30, 1996
    Local Public Document Room location: W. Dale Clark Library, 215 
South 15th Street, Omaha, Nebraska 68102

Wisconsin Electric Power Company, Docket Nos. 50-266 and 50-301, 
Point Beach Nuclear Power Plant, Unit Nos. 1 and 2, Town of Two 
Creeks, Manitowoc County, Wisconsin

    Date of application for amendment: September 19, 1996
    Brief description of amendment request: The proposed amendments 
would change Technical Specification requirements related to the low 
temperature overpressure protection (LTOP) system. Specifically, the 
reactor coolant system (RCS) temperature below which LTOP is required 
to be enabled and one high pressure safety injection pump is required 
to be rendered inoperable would be changed from 275  deg.F to 355 
deg.F. Also, a specification would be added stating that only one 
reactor coolant pump shall be operated when the RCS temperature is less 
than or equal to 125  deg.F. Finally, editorial changes would be made 
to rename the ``Overpressure Mitigating System'' as the ``Low 
Temperature Overpressure Protection System.'' Date of individual notice 
in Federal Register: October 1, 1996 (61 FR 51308) Expiration date of 
individual notice: October 31, 1996
    Local Public Document Room location: Joseph P. Mann Library, 1516 
Sixteenth, Two Rivers, Wisconsin 54241

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

    Date of application for amendment: September 27, 1996
    Brief description of amendment request: The proposed amendment 
would change Technical Specification (TS) requirements related to the 
low temperature overpressure protection (LTOP) system. Specifically, 
the LTOP curve would be modified to define 10 CFR Part 50, Appendix G 
pressure temperature limitations for LTOP evaluation through the end of 
operating cycle (EOC) 33. In addition, the LTOP enabling temperature 
and the temperature required for starting a reactor coolant pump would 
be changed consistent with the design basis for the LTOP system. 
Finally, the TS bases would be changed consistent with he changes 
described above.
    Date of individual notice in Federal Register: October 7, 1996 (61 
FR 52472)

[[Page 55042]]

    Expiration date of individual notice: November 6, 1996
    Local Public Document Room location: University of Wisconsin, 
Cofrin Library, 2420 Nicolet Drive, Green Bay, Wisconsin 54311-7001

Notice Of Issuance Of Amendments ToFacility Operating Licenses

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

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: July 19, 1996
    Brief description of amendment: The amendment revises the 
containment spray nozzle surveillance interval in TS 3/4.6.2 from 5 to 
10 years.
    Date of issuance: October 3, 1996
    Effective date: October 3, 1996
    Amendment No.: 67
    Facility Operating License No. NPF-63. Amendment revises the 
Technical Specifications.
    Date of initial notice in Federal Register: August 28, 1996 (61 FR 
44354) The Commission's related evaluation of the amendment is 
contained in a Safety Evaluation dated October 3, 1996.No significant 
hazards consideration comments received: No
    Local Public Document Room location:  Cameron Village Regional 
Library, 1930 Clark Avenue, Raleigh, North Carolina 27605

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

    Date of amendment request: April 11, 1996, as supplemented August 
23, 1996
    Brief description of amendments: The amendments revised the 
Technical Specifications to permit implementation of 10 CFR Part 50, 
Appendix J, Option B.
    Date of issuance: October 3, 1996
    Effective date: October 3, 1996
    Amendment Nos.: 185 and 176
    Facility Operating License Nos. DPR-51 and NPF-6: Amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: May 8, 1996 (61 FR 
20846) The additional information contained in the supplemental letter 
dated August 23, 1996, was clarifying in nature and thus, within the 
scope of the initial notice and did not affect the staff's proposed no 
significant hazards consideration determination.The Commission's 
related evaluation of the amendments is contained in a Safety 
Evaluation dated October 3, 1996.No significant hazards consideration 
comments received: No.
    Public Document Room location: Tomlinson Library, Arkansas Tech 
University, Russellville, AR 72801

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

    Date of amendment request: April 29, 1996
    Brief description of amendment: The amendment relocated cycle 
specific operating parameters from the Technical Specifications to the 
Core Operating Limits Report per Generic Letter 88-16. The parameters 
being relocated by this amendment include the variable low reactor 
coolant system pressure trip and the variable low reactor coolant 
system pressure-temperature protective limits.
    Date of issuance: October 3, 1996
    Effective date: October 3, 1996
    Amendment No.: 186
    Facility Operating License No. DPR-51: Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: June 5, 1996 (61 FR 
28613) The Commission's related evaluation of the amendment is 
contained in a Safety Evaluation dated October 3, 1996.No significant 
hazards consideration comments received: No.
    Local Public Document Room location: Tomlinson Library, Arkansas 
Tech University, Russellville, AR 72801

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

    Date of amendment request: November 7, 1995, as supplemented by 
letter dated April 11, 1996.
    Brief description of amendment: The amendment modifies the Appendix 
A Technical Specifications related to Safety Injection Tank level and 
pressure setpoints.
    Date of issuance: September 27, 1996
    Effective date: September 27, 1996
    Amendment No.: 121
    Facility Operating License No. NPF-38: Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: November 27, 1995 (60 
FR 58401) The additional information contained in the supplemental 
letter dated April 11, 1996, was clarifying in nature and thus, within 
the scope of the initial notice and did not affect the staff's proposed 
no significant hazards consideration determination. The Commission's 
related evaluation of the amendment is contained in a Safety Evaluation 
dated September 27, 1996.No significant hazards consideration comments 
received: No.
    Local Public Document Room location:  University of New Orleans 
Library, Louisiana Collection, Lakefront, New Orleans, LA 70122

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

    Date of application for amendments: July 17, 1996
    Brief description of amendments: The amendments consist of changes 
to the Technical Specifications regarding containment leakage tests.
    Date of issuance: October 4, 1996
    Effective date: October 4, 1996
    Amendment Nos.: 192 and 186Facility Operating Licenses Nos. DPR-31 
and DPR-41: Amendments revised the Technical Specifications.
    Date of initial notice in Federal Register: August 28, 1996 (61 FR 
44357)

[[Page 55043]]

The Commission's related evaluation of the amendments is contained in a 
Safety Evaluation dated October 4, 1996.No significant hazards 
consideration comments received: No
    Local Public Document Room location: Florida International 
University, University Park, Miami, Florida 33199

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

    Date of application for amendments: May 21, 1996
    Brief description of amendments: The amendments revise the 
condensate storage tank level indication to ensure that the water level 
is sufficient to provide 50,000 gallons of water for core spray makeup 
to the reactor pressure vessel. On September 24, 1996, based on a 
teleconference between the licensee and the NRC project manager, it was 
mutually agreed to change the requested implementation schedule from 90 
days to 30 days.
    Date of issuance: October 2, 1996
    Effective date: As of the date of issuance to be implemented within 
30 days
    Amendment Nos.: 202 and 143
    Facility Operating License Nos. DPR-57 and NPF-5: Amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: August 28, 1996 (61 FR 
44358) The Commission's related evaluation of the amendments is 
contained in a Safety Evaluation dated October 2, 1996. No significant 
hazards consideration comments received: No
    Local Public Document Room location: Appling County Public Library, 
301 City Hall Drive, Baxley, Georgia 31513

GPU Nuclear Corporation, Docket No. 50-320, Three Mile Island 
Nuclear Station, Unit No. 2, (TMI-2), Dauphin County, Pennsylvania

    Date of application for amendment: January 16, 1995
    Brief description of amendment: This amendment revised the 
Technical Specification to incorporate an improvement from 
administrative controls section of the revised standard TS for B&W 
plants.
    Date of issuance: October 8, 1996
    Effective date: October 8, 1996
    Amendment No.: 50Possession-Only License No. DPR-73: The amendment 
revised the Technical Specifications.
    Date of initial notice in Federal Register: December 20, 1995 (60 
FR 65679). The Commission's related evaluation of the amendment is 
contained in a Safety Evaluation dated October 8, 1996No significant 
hazards consideration comments received: No.
    Local Public Document Room location: Government Publications 
Section, State Library of Pennsylvania, Walnut Street and Commonwealth 
Avenue, Box 1601, Harrisburg, Pennsylvania 17105

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

    Date of application for amendment: July 5, 1996
    Brief description of amendment: The amendment will support the 
implementation of noble metal chemical addition at the Duane Arnold 
Energy Center as a method to enhance the effectiveness of hydrogen 
water chemistry in mitigating intergranular stress corrosion cracking 
in reactor vessel internal components. Specifically, the amendment will 
permit an increase of the reactor water conductivity limit in Technical 
Specification (TS) Table 3.6.B.2-1 and several other changes in TS 
sections 4.6.B.2.c, 4.6.B.2.d, and the associated Bases.
    Date of issuance: October 3, 1996
    Effective date: October 3, 1996
    Amendment No.: 218
    Facility Operating License No. DPR-49: Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: July 31, 1996 (61 FR 
40020) The Commission's related evaluation of the amendment is 
contained in a Safety Evaluation dated October 3, 1996. No significant 
hazards consideration comments received: No.
    Local Public Document Room location: Cedar Rapids Public Library, 
500 First Street, S. E., Cedar Rapids, Iowa 52401

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

    Date of application for amendment: December 22, 1995, as 
supplemented September 20, 1996
    Brief description of amendment: The amendment revises the Duane 
Arnold Energy Center (DAEC) Technical Specifications (TS) Sections 
3.7.A and 4.7.A, ``Primary Containment,'' by deleting information also 
contained in 10 CFR Part 50, Appendix J, Option A and incorporating 
references to the Primary Containment Leakage Rate Testing Program. 
These changes allow the use of the performance based option of 
containment leak testing. The amendment also adds Operability and 
Surveillance Requirements (SRs) for the drywell air lock. Minor 
administrative changes were also made. These changes are consistent 
with comparable specifications in the Improved Standard Technical 
Specifications (ITS), NUREG-1433. In addition, the staff executed 
administrative changes and corrections to the TS Bases, as submitted in 
two letters dated February 13, 1995. Sections changed or corrected are 
Section 1.2, Bases; Section 2.2, Bases Reactor Coolant System 
Integrity; Section 3.7.H/4.7.H, Bases Containment Atmosphere Dilution; 
and Section 3.7.I/4.7.I, Bases Oxygen Concentration.
    Date of issuance: October 4, 1996
    Effective date: October 4, 1996
    Amendment No.: 219
    Facility Operating License No. DPR-49: Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: January 31, 1996 (61 FR 
3499) The September 20, 1996, submittal was clarifying in nature and 
did not affect the no significant hazards determination. The 
Commission's related evaluation of the amendment is contained in a 
Safety Evaluation dated October 4, 1996.No significant hazards 
consideration comments received: No.
    Local Public Document Room location: Cedar Rapids Public Library, 
500 First Street, S. E., Cedar Rapids, Iowa 52401

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

    Date of application for amendment: June 28, 1996 and as 
supplemented on September 17, 1996
    Brief description of amendment: The amendment will allow removal of 
the Inclined Fuel Transfer System (IFTS) primary containment blind 
flange while primary containment is required to be operable. This will 
provide flexibility to operate the IFTS for the purpose of testing and 
exercising the system during such conditions. Primary containment 
integrity will be provided by an alternate means while the blind flange 
is removed. The change will be incorporated via a provisional note into 
Technical Specification (TS) Surveillance Requirement 3.6.1.3.3, 
associated with TS 3.6.1.3, ``Primary Containment Isolation Valves 
(PCIVs).''
    Date of issuance: October 3, 1996
    Effective date: October 3, 1996
    Amendment No.: 107
    Facility Operating License No. NPF-62: The amendment revised the 
Technical Specifications.

[[Page 55044]]

    Date of initial notice in Federal Register: July 31, 1996 (61 FR 
40021) The information provided in the licensee's letter of September 
17, 1996 provided clarifying information and did not involve 
significant changes to the original Federal Register notice.The 
Commission's related evaluation of the amendment is contained in a 
Safety Evaluation dated October 3, 1996.No significant hazards 
consideration comments received: No
    Local Public Document Room location: The Vespasian Warner Public 
Library, 120 West Johnson Street, Clinton, Illinois 61727

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

    Date of application for amendment: June 21, 1996, and as 
supplemented by letter dated August 15, 1996
    Brief description of amendment: The amendment modifies Section 5.7, 
``High Radiation Areas,'' of the ``Administrative Controls'' section of 
the Clinton Power Station technical specifications (TS). The changes 
include: (1) allowing utilization of a Radiation Work Permit (RWP) ``or 
equivalent'' to control entry into a high radiation area; (2) 
clarifying the example given in the TS of individuals who are qualified 
in radiation protection procedures; (3) clarifying the requirements for 
when specified access controls and barriers for high radiation areas 
within large areas like the containment may be established; (4) 
clarifying that it is acceptable for an RWP to specify a maximum dose, 
i.e., a specified setpoint on an alarming dosimeter in lieu of a stay 
time for entry into a high radiation area (where an individual could 
receive a deep dose equivalent of 3000 mrem in one hour); (5) 
eliminating the upper dose limit for specifying the applicability of 
the requirements of Specification 5.7.1; (6) providing additional 
flexibility regarding the control of keys to locked doors for 
preventing unauthorized entry into high radiation areas; (7) providing 
alternate means of informing individuals of dose rates in immediate 
work areas; (8) reorganizing TS Sections 5.7.1, 5.7.2, and 5.7.3 into 
four sections (5.7.1, 5.7.2, 5.7.3 and 5.7.4); and (9) making minor 
edits to enhance readability.
    Date of issuance: October 3, 1996
    Effective date: October 3, 1996
    Amendment No.: 108
    Facility Operating License No. NPF-62: The amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: July 31, 1996 (61 FR 
40021) The August 21, 1996, submittal consisted of supporting technical 
information which did not change the staff's initial proposed no 
significant hazards consideration determination or expand the scope of 
the original notice. The Commission's related evaluation of the 
amendment is contained in a Safety Evaluation dated October 3, 1996.No 
significant hazards consideration comments received: No.
    Local Public Document Room location: The Vespasian Warner Public 
Library, 120 West Johnson Street, Clinton, Illinois 61727

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

    Date of application for amendment: May 2, 1996, as supplemented by 
letter dated August 30, 1996
    Brief description of amendment: The amendment removes Technical 
Specification Figure 5.1, which was used in maintaining Keff 
values, and substitutes in its place a defined requirement for maximum 
Kinfinity for any fuel placed in the Millstone Unit 1 spent fuel 
pool.
    Date of Issuance: October 4, 1996
    Effective date: As of the date of issuance, to be implemented 
within 60 days.
    Amendment No.: 97
    Facility Operating License No. DPR-21: Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: July 17, 1996 (61 FR 
37301) The August 30, 1996, letter provided additional, clarifying 
information that did not change the scope of the May 2, 1996, 
application and the initial proposed no significant hazards 
consideration determination.The Commission's related evaluation of this 
amendment is contained in a Safety Evaluation dated October 4, 1996.No 
significant hazards consideration comments received: No.
    Local Public Document Room location: Learning Resources Center, 
Three Rivers Community-Technical College, 574 New London Turnpike, 
Norwich, Connecticut and the Waterford Library, ATTN: Vince Juliano, 49 
Rope Ferry Road, Waterford, Connecticut

Northern States Power Company, Docket No. 50-282, Prairie Island 
Nuclear Generating Plant, Unit No. 1, Goodhue County, Minnesota

    Date of application for amendment: July 15, 1996, and supplemented 
August 22, 1996
    Brief description of amendment: The amendment allows the use of the 
moveable in-core detector system for measurement of the core peaking 
factors with less than 75 percent and greater than or equal to 50 
percent of the detector thimbles available. The amendment is a one-time 
only change for Prairie Island, Unit 1, to reduce the number of 
required in-core detectors necessary for continued operation for the 
remainder of Operating Cycle 18.
    Date of issuance: October 10, 1996
    Effective date: October 10, 1996, and shall remain effective for 
the remainder of Cycle 18 only
    Amendment No.: 124
    Facility Operating License No. DPR-42. Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: July 31, 1996 (61 FR 
40024) By letter dated August 22, 1996, NSP forwarded a copy of the 
results of its most recent low power physics tests to the NRC for use 
as a reference and provided additional clarifying information. This 
information was within the scope of the original application and did 
not change the staff's initial proposed no significant hazards 
considerations determination. Therefore, renoticing was not 
warranted.The Commission's related evaluation of the amendments is 
contained in a Safety Evaluation dated October 10, 1996. No significant 
hazards consideration comments received: No.
    Local Public Document Room location: Minneapolis Public Library, 
Technology and Science Department, 300 Nicollet Mall, Minneapolis, 
Minnesota 55401

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

    Date of amendment request: May 17, 1996
    Brief description of amendment: The amendment revises Technical 
Specifications (TS) 2.18, 3.14, 3.3, and 5.10 to relocate the 
operability requirements for shock suppressors (snubbers) from the TS 
to the Updated Safety Analysis Report (USAR) and incorporate snubber 
examination and testing requirements in TS 3.3.
    Date of issuance: September 27, 1996
    Effective date: September 27, 1996
    Amendment No.: 176

[[Page 55045]]

    Facility Operating License No. DPR-40: Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: August 28, 1996 (61 FR 
44360) The Commission's related evaluation of the amendment is 
contained in a Safety Evaluation dated September 27, 1996.No 
significant hazards consideration comments received: No.
    Local Public Document Room location: W. Dale Clark Library, 215 
South 15th Street, Omaha, Nebraska 68102

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

    Date of amendment request: August 23, 1996
    Brief description of amendment: The amendment modifies paragraph 
2.B.(2) of
    Facility Operating License No. DPR-40 allowing the use of source 
material, in the form of depleted or natural uranium, as reactor fuel.
    Date of issuance: October 2, 1996
    Effective date: October 2, 1996
    Amendment No.: 177
    Facility Operating License No. DPR-40: Amendment revised the 
Operating License.
    Date of initial notice in Federal Register: August 30, 1996 (61 FR 
45995) The Commission's related evaluation of the amendment is 
contained in a Safety Evaluation dated October 2, 1996.No significant 
hazards consideration comments received: No.
    Local Public Document Room location: W. Dale Clark Library, 215 
South 15th Street, Omaha, Nebraska 68102

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

    Date of application for amendment: January 25, 1996
    Brief description of amendment: The amendment would extend the 
instrumentation surveillance test intervals to support 24-month 
operating cycles. These proposed changes would eliminate the mid-cycle 
outages to perform the Technical Specification surveillance 
requirements.
    Date of issuance: October 2, 1996
    Effective date: As of the date of issuance to be implemented within 
30 days.
    Amendment No.: 233
    Facility Operating License No. DPR-59: Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: May 22, 1996 (61 FR 
25709) The Commission's related evaluation of the amendment is 
contained in a Safety Evaluation dated October 2, 1996.No significant 
hazards consideration comments received: No
    Local Public Document Room location: Reference and Documents 
Department, Penfield Library, State University of New York, Oswego, New 
York 13126

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

    Date of application for amendment: March 27, 1996, as supplemented 
April 24, 1996, and August 15, 1996
    Brief description of amendment: The proposed amendment changes 
would permit implementation of 10 CFR Part 50, Appendix J, Option B 
with an exception to the guidelines of Regulatory Guide 1.163 for Type 
C testing of primary containment isolation valves in the reverse (non-
accident) direction.
    Date of issuance: October 4, 1996
    Effective date: As of the date of issuance to be implemented within 
30 days.
    Amendment No.: 234
    Facility Operating License No. DPR-59: Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: May 8, 1996 (61 FR 
20855) The Commission's related evaluation of the amendment is 
contained in a Safety Evaluation dated October 4, 1996.No significant 
hazards consideration comments received: No
    Local Public Document Room location: Reference and Documents 
Department, Penfield Library, State University of New York, Oswego, New 
York 13126

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

    Date of application for amendment: August 9, 1996, as supplemented 
September 17, 1996
    Brief description of amendment: The amendment revises the Technical 
Specifications to revise the safety limit minimum critical power ratio 
for cycle 19 operation from its current value of 1.07 (for the fuel 
currently in the reactor for cycle 18) for two recirculation loop 
operation to 1.10, and from 1.08 to 1.12 for single recirculation loop 
operation.
    Date of issuance: October 4, 1996
    Effective date: As of the date of issuance, to be implemented 
within 30 days.
    Amendment No.: 150
    Facility Operating License No. DPR-28: Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: August 28, 1996 (61 FR 
44364) The September 17, 1996, letter provided clarifying information 
that did not change the scope of the August 9, 1996, application and 
initial proposed no significant hazards consideration determination.The 
Commission's related evaluation of the amendment is contained in a 
Safety Evaluation dated October 4, 1996.No significant hazards 
consideration comments received: No
    Local Public Document Room location: Brooks Memorial Library, 224 
Main Street, Brattleboro, VT 05301

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

    Date of application for amendment: April 16, 1996, as supplemented 
July 25, 1996
    Brief description of amendment: The amendment permits 
implementation of 10 CFR Part 50, Appendix J, Option B, ``Performance-
Based Requirements.''
    Date of issuance: October 2, 1996
    Effective date: October 2, 1996
    Amendment No.: 135
    Facility Operating License No. NPF-12: Amendment revises the 
Technical Specifications.
    Date of initial notice in Federal Register: July 3, 1996 (61 FR 
34898) The July 25, 1996, supplement provides clarifying information 
and did not change the scope of the initial notice. The Commission's 
related evaluation of the amendment is contained in a Safety Evaluation 
dated October 2, 1996.No significant hazards consideration comments 
received: No
    Local Public Document Room location: Fairfield County Library, 300 
Washington Street, Winnsboro, SC 29180

Southern California Edison Company, et al, Docket No. 50-206, San 
Onofre Nuclear Generating Station, Unit No. 1, San Diego County, 
California

    Date of application for amendment: March 13, 1996
    Brief description of amendment: The change revises the San Onofre 
Unit 1 License Condition 2.D. This change eliminates a reporting 
requirement that is redundant to reporting requirements in 10 CFR 50.72 
and 50.73. Additionally, the amendment makes administrative and 
editorial changes to the Permanently Defueled Technical Specifications.
    Date of issuance: October 3, 1996

[[Page 55046]]

    Effective date: October 3, 1996
    Amendment No.: 158
    Facility Operating License No. DPR-13: The amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: July 31, 1996 (61 FR 
40028) The Commission's related evaluation of the amendment is 
contained in a Safety Evaluation dated October 3, 1996.No significant 
hazards consideration comments received: No.
    Local Public Document Room location: Science Library, University of 
California, Irvine, California 92713

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: December 6, 1995, as 
supplemented by letters dated August 30, 1996, and September 20, 1996
    Brief description of amendments: These amendments revise Technical 
Specifications (TS) Section 4.3 ``Fuel Storage'' to allow fuel 
assemblies having a maximum U-235 enrichment of 4.8 weight percent (w/
o) to be stored in both the spent fuel racks and the new fuel racks. 
Additionally, TS Section 3.7.18 ``Spent Fuel Assembly Storage,'' 
Figures 3.7.18-1 ``Unit 1 Fuel Minimum Burnup vs. Initial Enrichment 
for Region II Racks,'' and 3.7.18-2 ``Units 2 and 3 Fuel Minimum Burnup 
vs. Initial Enrichment for Region II Racks,'' are being revised and 
relabeled.
    Date of issuance: October 3, 1996
    Effective date: October 3, 1996, to be implemented within 30 days 
as of the date of issuance.
    Amendment Nos.: Unit 2 - 131; Unit 3 - 120
    Facility Operating License Nos. NPF-10 and NPF-15: The amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: April 10, 1996 (61 FR 
15997) The August 30, 1996, and September 20, 1996, letters provided 
additional clarifying information and did not change the initial no 
significant hazards consideration determination.The Commission's 
related evaluation of the amendments is contained in a Safety 
Evaluation dated October 3, 1996.No significant hazards consideration 
comments received: No.
    Temporary Local Public Document Room location: Science Library, 
University of California, P. O. Box 19557, Irvine, California 92713

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

    Date of amendment request: July 31, 1996 (TXX-96433)
    Brief description of amendments: The amendments revised core safety 
limit curves (Technical Specification (TS) Figure 2.1-1a) and new N-16 
setpoint values and parameters (TS Table 2.1-1) for Unit 1, and 
reference to topical report RXE-95-001-P as an approved methodology for 
small break loss of coolant accident analysis for Units 1 and 2.
    Date of issuance: September 30, 1996
    Effective date: September 30, 1996, to be implemented within 30 
days
    Amendment Nos.: 52 and 38
    Facility Operating License Nos. NPF-87 and NPF-89. The amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: August 28, 1996 (61 FR 
44362) The Commission's related evaluation of the amendments is 
contained in a Safety Evaluation dated September 30, 1996.No 
significant hazards consideration comments received: No.
    Local Public Document Room location: University of Texas at 
Arlington Library, Government Publications/Maps, 702 College, P.O. Box 
19497, Arlington, TX 76019

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

    Date of application for amendment: April 12, 1996, as supplemented 
by letters dated August 2, 1996, August 19, 1996, and September 5, 
1996.
    Brief description of amendment: The amendment revises the Technical 
Specifications to address the installation of laser welded tube sleeves 
in the Callaway Plant steam generators.
    Date of issuance: October 1, 1996
    Effective date: October 1, 1996, and will be implemented within 30 
days of the date of issuance.
    Amendment No.: 116
    Facility Operating License No. NPF-30: The amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: May 8, 1996 (61 FR 
20857) The August 2, 1996, August 19, 1996, and September 5, 1996, 
supplemental letters provided clarifying information and did not change 
the original no significant hazards consideration determination. The 
Commission's related evaluation of the amendment is contained in a 
Safety Evaluationdated October 1, 1996.No significant hazards 
consideration comments received: No.
    Local Public Document Room location: Callaway County Public 
Library, 710 Court Street, Fulton, Missouri 65251.

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

    Date of application for amendment: April 17, 1996, as supplemented 
by letters dated July 15, 1996, July 31, 1996, and August 28, 1996.
    Brief description of amendment: The amendment would change 
Technical Specification (TS) 3/4.3 to support a future modification to 
replace existing digital portions of the main steam and feedwater 
isolation system (MSFIS) with digital processor equipment and would 
authorize revision of the FSAR to include a description of the MSFIS 
modification.
    Date of issuance: October 1, 1996
    Effective date: October 1, 1996, to be implemented prior to startup 
from the Callaway Plant Refuel 8.
    Amendment No.: 117
    Facility Operating License No. NPF-30: The amendment revised the 
Technical Specifications and the Final Safety Analysis Report.
    Date of initial notice in Federal Register: June 5, 1996 (61 FR 
28619) The July 15, 1996, July 31, 1996 and August 28, 1996 
supplemental letters provided additional clarifying information and did 
not change the staff's original no significant hazards consideration 
determination. The Commission's related evaluation of the amendment is 
contained in a Safety Evaluation dated October 1, 1996.No significant 
hazards consideration comments received: No.
    Local Public Document Room location: Callaway County Public 
Library, 710 Court Street, Fulton, Missouri 65251

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

    Date of application for amendment: April 4, 1996
    Brief description of amendment: The amendment revises the Technical 
Specifications regarding the surveillance requirement for control rod 
over-travel by moving the specific testing methodology to licensee 
administratively controlled documents. Specifically, the amendment 
removes the requirement in Specification 4.3.B.1(b) to verify prior to 
coupling that the over-travel indicating light is working properly by 
withdrawing an uncoupled control rod drive to the over-travel position.

[[Page 55047]]

    Date of issuance: September 30, 1996
    Effective date: As of the date of issuance, to be implemented 
within 30 days.
    Amendment No.: 149
    Facility Operating License No. DPR-28: Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: May 8, 1996 (61 FR 
20860) The Commission's related evaluation of the amendment is 
contained in a Safety Evaluation dated September 30, 1996.No 
significant hazards consideration comments received: No
    Local Public Document Room location: Brooks Memorial Library, 224 
Main Street, Brattleboro, VT 05301

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

    Date of application for amendment: August 9, 1996
    Brief description of amendment: The amendment changes the 
operations manager qualification requirements to allow either of two 
alternatives (having held a senior reactor operator's license or having 
been certified for equivalent senior reactor operator knowledge) to the 
requirement for the operations manager to hold a senior reactor 
operator's license.
    Date of issuance: October 1, 1996
    Effective date: October 1, 1996
    Amendment No.: 148
    Facility Operating License No. NPF-21: The amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: August 28, 1996 (61 FR 
44350) The Commission's related evaluation of the amendment is 
contained in a Safety Evaluation dated October 1, 1996.No significant 
hazards consideration comments received: No.
    Local Public Document Room location: Richland Public Library, 955 
Northgate Street, Richland, Washington 99352

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

    Date of application for amendment: July 3, 1996, as supplemented on 
July 23, August 28, and September 16, 1996
    Brief description of amendment: The amendment revises Kewaunee 
Nuclear Power Plant Technical Specification 4.2.b, ``Steam Generator 
Tubes,'' and its associated basis, by revising the acceptance criteria 
for indications of tube degradation occurring in the tubesheet crevice 
region.
    Date of issuance: October 2, 1996
    Effective date: October 2, 1996
    Amendment No.: 129
    Facility Operating License No. DPR-43: Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: July 31, 1996 (61 FR 
40031) The July 23, August 28, and September 16, 1996, submittals 
provided clarifying information that did not change the initial 
proposed no significant hazards consideration determination.The 
Commission's related evaluation of the amendment is contained in a 
Safety Evaluation dated October 2, 1996.No significant hazards 
consideration comments received: No.
    Local Public Document Room location: University of Wisconsin, 
Cofrin Library, 2420 Nicolet Drive, Green Bay, Wisconsin 54311-7001

Wisconsin Electric Power Company, Docket Nos. 50-266 and 50-301, 
Point Beach Nuclear Plant, Unit Nos. 1 and 2, Town of Two Creeks, 
Manitowoc County, Wisconsin

    Date of application for amendments: May 29, 1996, as supplemented 
August 20, 1996
    Brief description of amendments: These amendments revise Technical 
Specification (TS) Section 15.4.4, ``Containment Tests,'' to 
incorporate the provisions of 10 CFR Part 50, Appendix J, ``Primary 
Reactor Containment Leakage Testing for Water-Cooled Power Reactors,'' 
Option B. Revisions have also been made to TS Sections 15.1, 
``Definitions,'' 15.3.6, ``Containment System,'' and 15.6, 
``Administrative Controls,'' to support the proposed changes to Section 
15.4.4.
    Date of issuance: October 9, 1996
    Effective date: October 9, 1996, to be implemented within 45 days.
    Amendment Nos.: Unit 1 - 169 and Unit 2 - 173
    Facility Operating License Nos. DPR-24 and DPR-27: Amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: July 3, 1996 (61 FR 
34901) The supplemental information did not affect the staff's initial 
no significant hazards consideration determination.The Commission's 
related evaluation of the amendments is contained in a Safety 
Evaluation dated October 9, 1996.No significant hazards consideration 
comments received: No
    Local Public Document Room location: Joseph P. Mann Library, 1516 
Sixteenth Street, Two Rivers, Wisconsin 54241

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

    During the period since publication of the last biweekly notice, 
the Commission has issued the following amendments. The Commission has 
determined for each of these amendments that the application for the 
amendment complies with the standards and requirements of the Atomic 
Energy Act of 1954, as amended (the Act), and the Commission's rules 
and regulations. The Commission has made appropriate findings as 
required by the Act and the Commission's rules and regulations in 10 
CFR Chapter I, which are set forth in the license amendment.
    Because of exigent or emergency circumstances associated with the 
date the amendment was needed, there was not time for the Commission to 
publish, for public comment before issuance, its usual 30-day Notice of 
Consideration of Issuance of Amendment, Proposed No Significant Hazards 
Consideration Determination, and Opportunity for a Hearing.
    For exigent circumstances, the Commission has either issued a 
Federal Register notice providing opportunity for public comment or has 
used local media to provide notice to the public in the area 
surrounding a licensee's facility of the licensee's application and of 
the Commission's proposed determination of no significant hazards 
consideration. The Commission has provided a reasonable opportunity for 
the public to comment, using its best efforts to make available to the 
public means of communication for the public to respond quickly, and in 
the case of telephone comments, the comments have been recorded or 
transcribed as appropriate and the licensee has been informed of the 
public comments.
    In circumstances where failure to act in a timely way would have 
resulted, for example, in derating or shutdown of a nuclear power plant 
or in prevention of either resumption of operation or of increase in 
power output up to the plant's licensed power level, the Commission may 
not have had an opportunity to provide for public comment on its no 
significant hazards consideration determination. In such case, the 
license amendment has been issued without opportunity for comment. If 
there has been some time for public comment but less than 30 days, the 
Commission may provide an

[[Page 55048]]

opportunity for public comment. If comments have been requested, it is 
so stated. In either event, the State has been consulted by telephone 
whenever possible.
    Under its regulations, the Commission may issue and make an 
amendment immediately effective, notwithstanding the pendency before it 
of a request for a hearing from any person, in advance of the holding 
and completion of any required hearing, where it has determined that no 
significant hazards consideration is involved.
    The Commission has applied the standards of 10 CFR 50.92 and has 
made a final determination that the amendment involves no significant 
hazards consideration. The basis for this determination is contained in 
the documents related to this action. Accordingly, the amendments have 
been issued and made effective as indicated.
    Unless otherwise indicated, the Commission has determined that 
these amendments satisfy the criteria for categorical exclusion in 
accordance with 10 CFR 51.22. Therefore, pursuant to 10 CFR 51.22(b), 
no environmental impact statement or environmental assessment need be 
prepared for these amendments. If the Commission has prepared an 
environmental assessment under the special circumstances provision in 
10 CFR 51.12(b) and has made a determination based on that assessment, 
it is so indicated.
    For further details with respect to the action see (1) the 
application for amendment, (2) the amendment to Facility Operating 
License, and (3) the Commission's related letter, Safety Evaluation 
and/or Environmental Assessment, as indicated. All of these items are 
available for public inspection at the Commission's Public Document 
Room, the Gelman Building, 2120 L Street, NW., Washington, DC, and at 
the local public document room for the particular facility involved.
    The Commission is also offering an opportunity for a hearing with 
respect to the issuance of the amendment. By November 22, 1996, the 
licensee may file a request for a hearing with respect to issuance of 
the amendment to the subject facility operating license and any person 
whose interest may be affected by this proceeding and who wishes to 
participate as a party in the proceeding must file a written request 
for a hearing and a petition for leave to intervene. Requests for a 
hearing and a petition for leave to intervene shall be filed in 
accordance with the Commission's ``Rules of Practice for Domestic 
Licensing Proceedings'' in 10 CFR Part 2. Interested persons should 
consult a current copy of 10 CFR 2.714 which is available at the 
Commission's Public Document Room, the Gelman Building, 2120 L Street, 
NW., Washington, DC and at the local public document room for the 
particular facility involved. If a request for a hearing or petition 
for leave to intervene is filed by the above date, the Commission or an 
Atomic Safety and Licensing Board, designated by the Commission or by 
the Chairman of the Atomic Safety and Licensing Board Panel, will rule 
on the request and/or petition; and the Secretary or the designated 
Atomic Safety and Licensing Board will issue a notice of a hearing or 
an appropriate order.
    As required by 10 CFR 2.714, a petition for leave to intervene 
shall set forth with particularity the interest of the petitioner in 
the proceeding, and how that interest may be affected by the results of 
the proceeding. The petition should specifically explain the reasons 
why intervention should be permitted with particular reference to the 
following factors: (1) the nature of the petitioner's right under the 
Act to be made a party to the proceeding; (2) the nature and extent of 
the petitioner's property, financial, or other interest in the 
proceeding; and (3) the possible effect of any order which may be 
entered in the proceeding on the petitioner's interest. The petition 
should also identify the specific aspect(s) of the subject matter of 
the proceeding as to which petitioner wishes to intervene. Any person 
who has filed a petition for leave to intervene or who has been 
admitted as a party may amend the petition without requesting leave of 
the Board up to 15 days prior to the first prehearing conference 
scheduled in the proceeding, but such an amended petition must satisfy 
the specificity requirements described above.
    Not later than 15 days prior to the first prehearing conference 
scheduled in the proceeding, a petitioner shall file a supplement to 
the petition to intervene which must include a list of the contentions 
which are sought to be litigated in the matter. Each contention must 
consist of a specific statement of the issue of law or fact to be 
raised or controverted. In addition, the petitioner shall provide a 
brief explanation of the bases of the contention and a concise 
statement of the alleged facts or expert opinion which support the 
contention and on which the petitioner intends to rely in proving the 
contention at the hearing. The petitioner must also provide references 
to those specific sources and documents of which the petitioner is 
aware and on which the petitioner intends to rely to establish those 
facts or expert opinion. Petitioner must provide sufficient information 
to show that a genuine dispute exists with the applicant on a material 
issue of law or fact. Contentions shall be limited to matters within 
the scope of the amendment under consideration. The contention must be 
one which, if proven, would entitle the petitioner to relief. A 
petitioner who fails to file such a supplement which satisfies these 
requirements with respect to at least one contention will not be 
permitted to participate as a party.
    Those permitted to intervene become parties to the proceeding, 
subject to any limitations in the order granting leave to intervene, 
and have the opportunity to participate fully in the conduct of the 
hearing, including the opportunity to present evidence and cross-
examine witnesses. Since the Commission has made a final determination 
that the amendment involves no significant hazards consideration, if a 
hearing is requested, it will not stay the effectiveness of the 
amendment. Any hearing held would take place while the amendment is in 
effect.
    A request for a hearing or a petition for leave to intervene must 
be filed with the Secretary of the Commission, U.S. Nuclear Regulatory 
Commission, Washington, DC 20555-001, Attention: Docketing and Services 
Branch, or may be delivered to the Commission's Public Document Room, 
the Gelman Building, 2120 L Street, NW., Washington, DC, by the above 
date. Where petitions are filed during the last 10 days of the notice 
period, it is requested that the petitioner promptly so inform the 
Commission by a toll-free telephone call to Western Union at 1-(800) 
248-5100 (in Missouri 1-(800) 342-6700). The Western Union operator 
should be given Datagram Identification Number N1023 and the following 
message addressed to (Project Director): petitioner's name and 
telephone number, date petition was mailed, plant name, and publication 
date and page number of this Federal Register notice. A copy of the 
petition should also be sent to the Office of the General Counsel, U.S. 
Nuclear Regulatory Commission, Washington, DC 20555-001, and to the 
attorney for the licensee.
    Nontimely filings of petitions for leave to intervene, amended 
petitions, supplemental petitions and/or requests for a hearing will 
not be entertained absent a determination by the Commission, the 
presiding officer or the Atomic Safety and Licensing Board that the 
petition and/or request should be granted based upon a balancing of the 
factors specified in 10 CFR 2.714(a)(1)(i)-(v) and 2.714(d).

[[Page 55049]]

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

    Date of application for amendments: September 21, 1996
    Brief description of amendments: The amendments approve changes to 
the Updated Final Analysis Report (UFSAR), and require that the changes 
be submitted with the next update of the UFSAR pursuant to 10 CFR 
50.71(e). The associated Safety Evaluation delineates the staff's 
review and findings, including finding that the as-built condition of 
the subject power system protective devices is acceptable as-is.
    Date of issuance: September 28, 1996
    Effective date: September 28, 1996
    Amendment Nos.: 153 and 145
    Facility Operating License Nos. NPF-35 and NPF-52: The amendments 
revised the Updated Final Safety Analysis Report. Public comments 
requested as to proposed no significant hazards consideration: Yes. The 
NRC staff published a public notice of the proposed amendments, issued 
a proposed finding of no significant hazards consideration, and 
requested that any comments on the proposed no significant hazards 
consideration be provided to the staff no later than 5:00 p.m., 
September 28, 1996. The notice was published in ``The Herald'' of Rock 
Hill, South Carolina, from September 25 through 27, 1996. No comments 
have been received.
    The Commission's related evaluation of the amendments, finding of 
exigent circumstances, consultation with the State of South Carolina, 
and final determination of no significant hazards consideration are 
contained in a Safety Evaluation dated September 28, 1996.
    Local Public Document Room location: York County Library, 138 East 
Black Street, Rock Hill, South Carolina 29730
    Attorney for licensee: Mr. Albert Carr, Duke Power Company, 422 
South Church Street, Charlotte, North Carolina 28242
    NRC Project Director: Herbert N. Berkow

PECO Energy Company, Public Service Electric and Gas Company, 
Delmarva Power and Light Company, and Atlantic City Electric 
Company, Docket No. 50-277, Peach Bottom Atomic Power Station, Unit 
No. 2, York County, Pennsylvania

    Date of application for amendment: March 25, 1996 as supplemented 
by letters dated August 23, 1996 and September 27, 1996.
    Brief description of amendment: The amendment revises Peach Bottom 
Technical Specification 2.1.1.2 safety limit minimum critical power 
ratios to be consistent with the use of GE-13 fuel in the Unit 2 core 
for operating cycle 12.
    Date of issuance: September 27, 1996
    Effective date: As of date of issuance
    Amendment No.: 217
    Facility Operating License No. DPR-44: Amendment revised the 
Technical Specifications.Public comments requested as to proposed no 
significant hazards consideration: Yes (61 FR 45997). That notice 
provided an opportunity to submit comments on the Commission's proposed 
no significant hazards consideration determination. No comments have 
been received. The notice also provided an opportunity to request a 
hearing by September 30, 1996, but indicated that if the Commission 
makes a final no significant hazards consideration determination any 
such hearing would take place after issuance of the amendment.The 
Commission's related evaluation of the amendment, finding of exigent 
circumstances, and final no significant hazards consideration 
determination are contained in a Safety Evaluation dated September 27, 
1996.
    Attorney for licensee: J. W. Durham, Sr., Esquire, Sr. Vice 
President and General Counsel, PECO Energy Company, 2301 Market Street, 
Philadelphia, Pennsylvania 19101
    Local Public Document Room location: Government Publications 
Section, State Library of Pennsylvania, (REGIONAL DEPOSITORY) Education 
Building, Walnut Street and Commonwealth Avenue, Box 1601, Harrisburg, 
PA 17105.
    Dated at Rockville, Maryland, this 16th day of October 1996.
    For the Nuclear Regulatory Commission
John A. Zwolinski,
Acting Director, Division of Reactor Projects - I/II, Office of Nuclear 
Reactor Regulation
[FR Doc. 96-27025 Filed 10-22-96; 8:45 am]
BILLING CODE 7590-O1-F