[Federal Register Volume 61, Number 80 (Wednesday, April 24, 1996)]
[Notices]
[Pages 18162-18183]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: 96-9925]



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NUCLEAR REGULATORY COMMISSION


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

I. Background

    Pursuant to Public Law 97-415, the U.S. Nuclear Regulatory 
Commission (the Commission or NRC staff) is publishing this regular 
biweekly notice. Public Law 97-415 revised section 189 of the Atomic 
Energy Act of 1954, as amended (the Act), to require the Commission to 
publish notice of any amendments issued, or proposed to be issued, 
under a new provision of section 189 of the Act. This provision grants 
the Commission the authority to issue and make immediately effective 
any amendment to an operating license upon a determination by the 
Commission that such amendment involves no significant hazards 
consideration, notwithstanding the pendency before the Commission of a 
request for a hearing from any person.
    This biweekly notice includes all notices of amendments issued, or 
proposed to be issued from March 30, 1996, through April 12, 1996. The 
last biweekly notice was published on April 10, 1996 (61 FR 15985).

[[Page 18163]]

Notice of Consideration of Issuance of Amendments to Facility Operating 
Licenses, Proposed No Significant Hazards Consideration Determination, 
and Opportunity for a Hearing

    The Commission has made a proposed determination that the following 
amendment requests involve no significant hazards consideration. Under 
the Commission's regulations in 10 CFR 50.92, this means that operation 
of the facility in accordance with the proposed amendment would not (1) 
Involve a significant increase in the probability or consequences of an 
accident previously evaluated; or (2) create the possibility of a new 
or different kind of accident from any accident previously evaluated; 
or (3) involve a significant reduction in a margin of safety. The basis 
for this proposed determination for each amendment request is shown 
below.
    The Commission is seeking public comments on this proposed 
determination. Any comments received within 30 days after the date of 
publication of this notice will be considered in making any final 
determination.
    Normally, the Commission will not issue the amendment until the 
expiration of the 30-day notice period. However, should circumstances 
change during the notice period such that failure to act in a timely 
way would result, for example, in derating or shutdown of the facility, 
the Commission may issue the license amendment before the expiration of 
the 30-day notice period, provided that its final determination is that 
the amendment involves no significant hazards consideration. The final 
determination will consider all public and State comments received 
before action is taken. Should the Commission take this action, it will 
publish in the Federal Register a notice of issuance and provide for 
opportunity for a hearing after issuance. The Commission expects that 
the need to take this action will occur very infrequently.
    Written comments may be submitted by mail to the Rules Review and 
Directives Branch, Division of Freedom of Information and Publications 
Services, Office of Administration, U.S. Nuclear Regulatory Commission, 
Washington, DC 20555, and should cite the publication date and page 
number of this Federal Register notice. Written comments may also be 
delivered to Room 6D22, Two White Flint North, 11545 Rockville Pike, 
Rockville, Maryland from 7:30 a.m. to 4:15 p.m. Federal workdays. 
Copies of written comments received may be examined at the NRC Public 
Document Room, the Gelman Building, 2120 L Street NW., Washington, DC. 
The filing of requests for a hearing and petitions for leave to 
intervene is discussed below.
    By May 24, 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.
    If a hearing is requested, the Commission will make a final 
determination on the issue of no significant hazards consideration. The 
final determination will serve to decide when the hearing is held.
    If the final determination is that the amendment request involves 
no significant hazards consideration, the Commission may issue the 
amendment and make it immediately effective, notwithstanding the 
request for a hearing. Any hearing held would take place after issuance 
of the amendment.
    If the final determination is that the amendment request involves a 
significant hazards consideration, any hearing held would take place 
before the issuance of any amendment.
    A request for a hearing or a petition for leave to intervene must 
be filed with the Secretary of the Commission, U.S. Nuclear Regulatory 
Commission, Washington, DC 20555, Attention: Docketing and Services 
Branch, or may be delivered to the Commission's Public

[[Page 18164]]

Document Room, the Gelman Building, 2120 L Street NW., Washington DC, 
by the above date. Where petitions are filed during the last 10 days of 
the notice period, it is requested that the petitioner promptly so 
inform the Commission by a toll-free telephone call to Western Union at 
1-(800) 248-5100 (in Missouri 1-(800) 342-6700). The Western Union 
operator should be given Datagram Identification Number N1023 and the 
following message addressed to (Project Director): petitioner's name 
and telephone number, date petition was mailed, plant name, and 
publication date and page number of this Federal Register notice. A 
copy of the petition should also be sent to the Office of the General 
Counsel, U.S. Nuclear Regulatory Commission, Washington, DC 20555, and 
to the attorney for the licensee.
    Nontimely filings of petitions for leave to intervene, amended 
petitions, supplemental petitions and/or requests for a hearing will 
not be entertained absent a determination by the Commission, the 
presiding officer or the Atomic Safety and Licensing Board that the 
petition and/or request should be granted based upon a balancing of 
factors specified in 10 CFR 2.714(a)(1) (i)-(v) and 2.714(d).
    For further details with respect to this action, see the 
application for amendment which is available for public inspection at 
the Commission's Public Document Room, the Gelman Building, 2120 L 
Street NW., Washington, DC, and at the local public document room for 
the particular facility involved.

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

    Date of amendment request: March 20, 1996.
    Description of amendment request: The licensee proposes to relocate 
Technical Specification (TS) 3.3.3.2, Movable Incore Detectors, to the 
Harris Nuclear Plant Core Operating Limits Report (COLR). Future 
changes to the relocated provisions will be evaluated in accordance 
with 10 CFR 50.59.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:
    1. The proposed amendment does not involve a significant increase 
in the probability or consequences of an accident previously evaluated.
    The proposed change will simplify the Technical Specifications, 
while implementing the recommendations of the Commission's Final Policy 
Statement on TS Improvements. The changes are administrative in nature 
and do not involve any modifications to plant equipment or affect plant 
operation. Since the TS provisions are being relocated to a licensee-
controlled document, any future changes will be controlled under 10 CFR 
50.59. Therefore, there would be no increase in the probability or 
consequences of an accident previously evaluated.
    2. The proposed amendment does not create the possibility of a new 
or different kind of accident from any accident previously evaluated.
    The proposed change is a relocation of existing Technical 
Specification provisions. It does not involve any physical alterations 
to plant equipment or alter the method by which any safety-related 
system performs its 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.
    The proposed change does not affect any Final Safety Analysis 
Report (FSAR) Chapter 15 accident analyses or have any impact on margin 
as defined in the Bases to the Technical Specifications. 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: W. D. Johnson, Vice President & Senior 
Counsel, Carolina Power & Light Company, Post Office Box 1551, Raleigh, 
North Carolina 27602
    NRC Project Director: Eugene V. Imbro

Connecticut Yankee Atomic Power Company, Docket No. 50-213, Haddam Neck 
Plant, Middlesex County; Northeast Nuclear Energy Company, et al., 
Docket Nos. 50-245, 50-336, 50-423, Millstone Nuclear Power Station, 
Units 1, 2, and 3, New London County, Connecticut; and North Atlantic 
Energy Service Company, Docket No. 50-443, Seabrook Station, Unit No. 
1, Rockingham County, New Hampshire

    Date of amendment request: February 1, 1996
    Description of amendment request: The amendment request would 
revise Section 6 ``Administrative Controls,'' of the Haddam Neck Plant, 
Millstone Unit Nos. 1, 2, and 3, and Seabrook Station, Unit 1 Technical 
Specifications to reflect several changes in organizational titles. The 
proposed changes are administrative title and editorial changes only.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration (SHC), which is presented below:
    * * * The proposed changes do not involve an SHC because the change 
would not:
    1. Involve a significant increase in the probability or 
consequences of an accident previously evaluated.
    No design basis accidents are affected by these proposed changes. 
The proposed changes are administrative and editorial in nature and are 
being proposed to reflect the recently announced organizational changes 
which will become effective on February 1, 1996. These changes include: 
insertion of the function Chief Nuclear Officer, in lieu of Executive 
Vice President--Nuclear; and establishment of a single point of 
operational direction for all five units in the position of the Vice 
President--Nuclear Operations. This individual is in lieu of the 
positions of Vice President--Haddam Neck, Senior Vice President--
Millstone Station, and Executive Director--Nuclear Production. These 
latter positions have been eliminated; other changes are: the 
appointment of the Haddam Neck Plant Nuclear Unit Director as chairman 
of the Haddam Neck PORC [Plant Operations Review Committee]; promotion 
of the Shift Supervisor/Shift Superintendent to the position of Shift 
Manager; revising the titles of ``additional operator'' and ``auxiliary 
operator'' to ``nuclear systems operator''; modifying the phrase 
``crewman'' to a gender neutral term ``crewperson'';

[[Page 18165]]

reassignment of the delivery of ISEG [Independent Safety Engineering 
Group] reports to the Senior Vice President--Nuclear Safety and 
Oversight; and a change to the title of the Seabrook Station Manager to 
Station Director. No safety systems are adversely affected by the 
proposed changes, and no failure modes are associated with the changes. 
Therefore, there is no impact on the probability of occurrence or the 
consequences of any accidents previously evaluated.
    2. Create the possibility of a new or different kind of accident 
from any accident previously evaluated.
    Because there are no changes in the way the plants are operated due 
to this administrative change, the potential for an unanalyzed accident 
is not created. There is no impact on plant response, and no new 
failure modes are introduced. These proposed administrative and 
editorial changes have no impact on safety limits or design basis 
accidents, and they have no potential to create a new or unanalyzed 
event.
    3. Involve a significant reduction in a margin of safety.
    The changes do not directly affect any protective boundaries nor do 
they impact the safety limits for the protective boundaries. These 
proposed changes are administrative and editorial in nature. Therefore, 
there can be no 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 locations: For the Haddam Neck Plant, 
Russell Library, 123 Broad Street, Middletown, CT 06457; for Millstone 
Nuclear Power Station, Unit Nos. 1, 2, and 3, Learning Resources 
Center, Three Rivers Community-Technical College, 574 New London 
Turnpike, Norwich, CT 06360; for Seabrook Station, Unit No. 1, Exeter 
Public Library, Founders Park, Exeter, NH 03833.
    Attorney for Licensees: 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
    Duke Power Company, et al., Docket No. 50-413, Catawba Nuclear 
Station, Unit 1, York County, South Carolina
    Date of amendment request: January 26, 1996.
    Description of amendment request: The amendment would allow a one-
time change to the Technical Specifications (TS) to allow operation of 
the containment purge ventilation system during Modes 3 and 4 during 
startup following the forthcoming Unit 1 steam generator replacement 
outage. This would alleviate respiratory hazards to personnel who would 
enter the containment to perform surveillances during Modes 4 and 3 of 
startup operations. Those hazards are expected to result from the 
thermal decomposition product gases evolving from the heatup of newly 
installed thermal insulation. Operation of the containment purge system 
to exhaust these gases would ensure that the air quality meets 
applicable standards for personnel safety.
    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 activity does not involve a significant increase in the 
probability or consequences of an accident previously evaluated.
    The VP [Containment Purge] System has no interfaces with any 
primary system, secondary system, or power transmission system. It has 
no interfaces with any reservoir of radioactive gases or liquids. None 
of the systems listed above are modified by the activity. In summary, 
no ``accident initiator'' is affected with the proposed operation of 
the VP System in Mode[s] 3 and 4. For this reason, the activity does 
not involve an increase in the probability of an accident previously 
evaluated.
    Analyses have been performed to determine upper bounds to the 
source term, the offsite doses, and the Control Room dose. The results 
of that analyses are reported above. Both the source term and the doses 
were found to be significantly lower than the results of the 
corresponding design basis analyses. No credit was taken for operation 
of the annulus ventilation system (VE) in the dose analysis. In 
addition, it has been determined that with no credit taken for any heat 
transfer from the fuel and cladding to the moderator channels, that 
sufficient time would exist for the operators to initiate recovery of 
flow from the ECCS [Emergency Core Cooling System] to the reactor core. 
The flow required from the ECCS to maintain the core in a coolable 
geometry was found to be well within the capacity of any one ECCS pump. 
Furthermore, it was determined that convective heat transfer to steam 
would be sufficient to prevent release of significant source term or a 
significant degree of fuel damage.
    For the above reasons, it is determined that operation of the VP 
System in Mode 3 or 4 immediately following the steam generator 
replacement outage does not involve a significant increase in either 
the probability or the consequences of an accident previously 
evaluated.
    (2) The activity does not create the possibility of a new or 
different type of accident from any accident previously evaluated.
    As discussed above, no ``accident initiators'' are affected by the 
proposed activity. Operation of the VP System proposed for Modes 3 and 
4 will be the same as that routinely carried in other modes of 
operation. For these reasons, the activity will not create the 
possibility of a new or different type of accident from any previously 
evaluated.
    (3) The activity does not involve a significant reduction in the 
margin of safety.
    Margin of safety is associated with confidence in the ability of 
the fission product barriers (the fuel and fuel cladding, the Reactor 
Coolant System pressure boundary, and the containment) to limit the 
level of radiation doses to the public. The proposed operation of the 
VP System will occur at the end of an extended outage. The level of 
decay heat and activity in the reactor is very low compared to the 
level of decay heat and activity associated with full power operations. 
For this reason, the likelihood of damage to the fuel following a 
DBLOCA [design basis loss-of-coolant analysis] occurring during the 
proposed purging is reduced, as determined above. Both offsite doses 
and doses to the Control Room were found to be small compared to the 
limits of 10 CFR [Part] 100 and GDC [General Design Criterion] 19. For 
these reasons, the activity 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: 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.

[[Page 18166]]

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

    Date of amendment request: December 12, 1995.
    Description of amendment request: The proposed amendments would 
correct an error in the Axial Flux Difference (AFD) Equations to more 
accurately reflect the proper AFD limit reduction, which is more 
conservative than the literal interpretation of the current Technical 
Specifications.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:
    A. The change would not involve a significant increase in the 
probability or consequences of an accident previously evaluated.
    The monitoring of core power distribution and peaking factors is to 
ensure accident analysis assumptions such as maximum local pin power at 
the initiation of an accident are satisfied, and are not involved in 
the initiation or mitigation of any previously evaluated accident.
    The proposed change is actually more conservative than the existing 
Technical Specification currently being used at McGuire.
    B. The change will not create the possibility of a new or different 
kind of accident from any accident previously evaluated.
    No plant modifications (hardware or control methods) are involved 
with this proposed change. The change is simply to correct an error in 
the Specification introduced in Amendments 130 (Unit 1) and 112 (Unit 
2). The proposed change is more restrictive than the current 
specification. No changes are proposed which could create any new 
accident scenarios.
    C. The proposed change will not involve a significant reduction in 
any margin of safety.
    The proposed change ensures the margin of safety is properly 
maintained by properly reducing (instead of increasing) the Positive 
AFD [Axial Flux Difference] limit if a peaking factor exceeds its 
surveillance limit. The change is more conservative than the existing 
Specification and will ensure the margins of safety are properly 
maintained.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: Atkins Library, University of 
North Carolina, Charlotte (UNCC Station), North Carolina 28223.
    Attorney for licensee: Mr. Albert Carr, Duke Power Company, 422 
South Church Street, Charlotte, North Carolina 28242.
    NRC Project Director: Herbert N. Berkow.

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

    Date of amendment request: March 4, 1996.
    Description of amendment request: The proposed amendments would 
delete the Flow Monitoring System from Technical Specification (TS) 
3.4.6.1 and associated surveillance requirements. The TS requires that 
either the Containment Floor and Equipment Sump Level System or the 
Flow Monitoring System be used to ensure that Reactor Coolant leakage 
is maintained within the specified limits. Duke Power does not use the 
Flow Monitoring System as a result of documented instrumentation 
inaccuracies due to the as-built piping configuration. The existing 
piping configuration does not ensure a water solid line which is 
necessary for the correct operation of any type of flow 
instrumentation. Modification to add a loop seal downstream of the flow 
element would be necessary for operability, which would create access 
difficulties as well as increase the potential for a radiological 
hazard in the form of a CRUD trap.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:
    1. This amendment will not significantly increase the probability 
or consequence of any accident previously evaluated.
    This change will not increase the probability or consequences of an 
accident since this Reactor Coolant Leakage Detection instrumentation 
is not an accident initiator or mitigator.
    This proposed Technical Specification change does not decrease the 
number of methods for Reactor Coolant leakage detection. This change 
will ensure there are still three distinctly separate methods of 
detecting NC [reactor coolant] leakage within the Containment Building. 
The first method will be detecting liquid leakage inside Containment 
via CFAE [Containment Floor and Equipment] level monitoring. The second 
method is detecting an increase in Radiation levels inside Containment 
and the third method is detecting steam leakage inside Containment. All 
three methods satisfy the diversity requirements listed in Regulatory 
Guide 1.45 for detecting a Reactor Coolant leak inside Containment.
    The sensitivity requirement listed in Regulatory Guide 1.45 is to 
detect a Reactor Coolant leak of one (1) gpm in one (1) hour. The first 
method meets this by use of the Sump level monitoring and rate of 
increase alarm from this level monitoring device. There are two sumps 
inside containment and the levels for both sumps are combined for 
detecting a one (1) gpm leak. McGuire uses the Sump Level monitoring to 
adequately address liquid leakage detection inside Containment; 
therefore, a flow monitoring system on the Sump Discharge line is not 
necessary and can be deleted.
    The Radiation Monitors are also set up to the required Regulatory 
Guide 1.45 sensitivity for detecting Reactor Coolant leakage and are 
not designed for SSE [safe-shutdown earthquake] events per the McGuire 
FSAR [Final Safety Analysis Report] (see McGuire's Request for 
Amendment: Reactor Coolant Leakage Detection Systems, dated March 4, 
1996).
    The third method for detecting Reactor Coolant leakage is to 
monitor Containment Ventilation Condensate Drain Tank (VUCDT) flow, for 
which McGuire is also using a level monitor. As in the case of the CFAE 
Unit Sump Level monitor, level monitoring for leakage detection is more 
reliable than flow monitoring.
    2. This amendment will not create the possibility of any new or 
different kind of accident not previously evaluated.
    The CFAE Flow Monitoring System has no control function, ([i.e.,] 
it is only a process monitor). Therefore, its deletion cannot create 
the pos[s]ibility of a new or different kind of accident.
    3. This amendment will not involve a significant reduction in a 
margin of safety.
    This proposed Tech Spec change does not decrease the number of 
methods for Reactor Coolant leakage detection. This change will ensure 
there are still three distinctly separate methods of detecting Reactor 
Coolant leakage within the Containment Building.
    Tech Spec 3.4.6.1 specifies two Radiation Monitors as two separate

[[Page 18167]]

required methods for Reactor Coolant Leakage Detection with the 
Containment Ventilation condensate level monitoring as a backup. The 
third method is the Containment Sump level monitoring with the flow 
monitoring as a backup.
    The new standardized Tech Spec 3.4.15, lists method one as 
Containment Sump (Level OR Discharge Flow) Monitoring Device. McGuire 
proposes to use a Sump Level monitoring device only. The second method 
listed is one Containment Radiation Monitor (either the gaseous or 
particulate monitor). McGuire will still have both available. The third 
method listed is one Containment air cooler condensate flow rate 
monitor for which McGuire plans to also use a level monitor. Liquid, 
Radiation, and Steam monitoring will still be accounted for in the Tech 
Spec, with the additional requirement of running a Reactor Coolant leak 
calculation if any of the methods are inoperable.
    Since McGuire is retaining three distinct methods of Reactor 
Coolant leakage detection per current TS [technical specification] 
requirements (and in agreement with current ISTS [improved standard 
technical specification] requirements), the proposed Technical 
Specification amendment does not cause any reduction in safety margin.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: Atkins Library, University of 
North Carolina, Charlotte (UNCC Station), North Carolina 28223.
    Attorney for licensee: Mr. Albert Carr, Duke Power Company, 422 
South Church Street, Charlotte, North Carolina 28242.
    NRC Project Director: Herbert N. Berkow.

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

    Date of amendment request: February 21, 1996.
    Description of amendment request: The licensee proposes a change to 
the Plant Hatch Unit 1 and Unit 2 Technical Specifications. The 
proposed revision would change the Drywell Air Temperature Limiting 
Condition for Operation (LCO) from less than or equal to 135 deg.F to 
less than or equal to 150 deg.F. The proposed change would provide a 
margin for the primary containment Drywell Air Temperature LCO when 
prolonged summer and high river temperatures are experienced. Also, a 
correction to a Final Safety Analysis Report (FSAR) reference would be 
made. This typographical error is strictly editorial.
    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 probability (frequency of occurrence) of previously evaluated 
accidents is not a function of the ambient drywell air temperature. 
Instrumentation setpoint calculations were assessed, and the increased 
ambient drywell air temperature does not affect any instrumentation 
setpoints or allowable values.
    The design basis accidents were reevaluated utilizing the increased 
drywell air temperature as an initial assumption. The results indicated 
that no regulatory limits or equipment design requirements will be 
exceeded as the result of the proposed change. Therefore, the change in 
drywell air temperature does not result in a significant increase in 
the probability or consequences of any previously evaluated accidents.
    2. The proposed change does not create the possibility of a new or 
different kind of accident from any accident previously analyzed. 
Revising the Drywell Air Temperature LCO does not physically modify the 
plant nor does it modify the operation of any existing equipment.
    3. The proposed change does not involve a significant reduction in 
a margin of safety. Design bases analyses performed utilizing 150 deg.F 
as the initial drywell temperature demonstrate that design and 
regulatory limits are not exceeded. Equipment in the drywell required 
to mitigate the effects of a DBA [design basis accident] is qualified 
to operate under environmental conditions expected for an accident. 
Analysis results do not affect instrumentation setpoints or 
calibration, or accident equipment qualification.
    Equipment qualified life is evaluated by an existing program which 
uses elevation-dependent drywell temperature rather than bulk average 
temperature. Therefore, the margin of safety associated with safety and 
other limits identified in the Technical Specifications are not 
significantly reduced.
    The correction to an FSAR reference is strictly editorial. 
Therefore, it meets the three criteria stated above.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: Appling County Public Library, 
301 City Hall Drive, Baxley, Georgia 31513.
    Attorney for licensee: Ernest L. Blake, Jr., Esquire, Shaw, 
Pittman, Potts and Trowbridge, 2300 N Street NW., Washington, DC 20037.
    NRC Project Director: Herbert N. Berkow.

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

    Date of amendment request: March 28, 1996 (TSCR 234).
    Description of amendment request: The proposed amendment modifies 
statements in the Technical Specifications and bases to correctly 
reflect the reference parameter for anticipatory scram signal bypass.
    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. State the basis for the determination that the proposed activity 
will or will not increase the probability of occurrence or consequences 
of an accident.
    This change modifies the terminology in a footnote to a Technical 
Specification Table and the bases. The change properly aligns the 
footnote and the bases with the FSAR [final safety analysis report] and 
the newly revised conservative setpoint which now correctly correlates 
the high pressure turbine third stage extraction steam line pressure to 
rated reactor thermal power. The change does not modify the function or 
operation of the bypass logic. Therefore, the proposed change will not 
increase the probability of occurrence or consequences of an accident.
    2. State the basis for the determination that the activity does or 
does not create the possibility of an accident or malfunction of 
equipment of

[[Page 18168]]

a different type than any previously identified in the SAR.
    The change does not involve any hardware and does not alter the 
functional intent of the pressure switches. The change of the footnote 
wording and the bases are primarily administrative and the existing 
Technical Specification Limiting Condition for Operation are preserved. 
Thus the proposed activity does not create the possibility of an 
accident or malfunction of a different type than any previously 
identified in the SAR.
    3. State the basis for the determination that the margin of safety 
as defined in the bases of any Technical Specification is not reduced.
    The revised setpoint assures that the anticipatory scram signal 
bypass is removed before reaching the Technical Specification limit of 
40 percent rated reactor thermal power (during power ascension). Thus, 
the margin of safety as stated in the bases of Technical Specification 
3.1 is preserved.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: Ocean County Library, 
Reference Department, 101 Washington Street, Toms River, NJ 08753.
    Attorney for licensee: Ernest L. Blake, Jr., Esquire. Shaw, 
Pittman, Potts & Trowbridge, 2300 N Street NW., Washington, DC 20037.
    NRC Project Director: John F. Stolz.

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

    Date of amendment request: February 22, 1996 (U-602554)
    Description of amendment request: The proposed amendment would 
modify Technical Specifications 3.3.8.1, ``Loss of Power 
Instrumentation,'' and 3.8.1, ``AC Sources-Operating.'' The proposed 
changes would delete the Surveillance Requirement (SR) 3.3.8.1.1 which 
requires a channel check for Loss of Power instrumentation and change 
Technical Specification Table 3.3.8.1-1 to change the allowable value 
for the Degraded Voltage Function (items 1.c and 2.c) from ``[greater 
than or equal to] 3762V and [less than or equal to] 3832V'' to 
``[greater than or equal to] 3876V.'' The amendment would also change 
Technical Specification Table 3.3.8-1 to modify the Division 3 degraded 
voltage logic to be the same as Divisions 1 and 2 (i.e., two-out-of-two 
rather than three-out-of-three), and increase the steady state voltage 
from [greater than or equal to] 3740V to [greater than or equal to] 
3870V for SRs 3.8.1.2, 3.8.1.7, 3.8.1.11, 3.8.1.12, 3.8.1.15, 3.8.1.19 
and 3.8.1.20.
    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) None of the proposed changes involve a significant increase in 
the probability or consequences of any accident previously evaluated. 
Each of the proposed changes is evaluated against this criteria as 
discussed below.
    The deletion of the channel check surveillance will result in 
discontinuing the recording of information that is not effective in 
assessing the capability of the degraded voltage relays to perform 
their intended function. Deletion of the channel check does not change 
the design or the expected performance of the Loss of Power (LOP) 
degraded voltage instrumentation, and therefore, the proposed change 
does not impact the intended function of this instrumentation to ensure 
adequate voltage for the ECCS equipment during DBA and other non-
accident scenarios. This surveillance provides little added assurance 
of relay operability since the relay is normally in a ``non-tripped'' 
state.
    The revision of the Allowable Values for the LOP degraded voltage 
and increase in the minimum required voltage for testing diesel 
generators will not result in any increase in the probability or 
consequences of any accident. The revised Allowable Values will 
continue to provide assurance that adequate voltage is available to run 
ECCS equipment during DBAs or any other non accident scenarios. With 
the emergency bus(es) voltage at or greater than the revised Allowable 
Values, the operability of required ECCS equipment is assured. The 
revised setpoints for the degraded voltage instrumentation, as 
controlled under 10CFR50.59 in the Clinton Power Station Operational 
Requirements Manual (ORM), are sufficiently low to assure that the 
possibility of spurious trips is minimized.
    The planned modification for Division 3 LOP degraded voltage 
sensor/relay logic will make Division 3 logic identical to the present 
designs for Division 1 and 2. The proposed design for Division 3 will 
not result in an increase in the probability of any accident because 
the proposed LOP Degraded Voltage logic for Division 3 will be 
identical to the proven design of Division 1 and 2. There will not be 
an increase in the consequences of an accident because the design of 
the LOP Degraded Voltage instrumentation will continue to ensure 
adequate voltage for ECCS equipment during any DBA and during non-
accident scenarios.
    The proposed footnotes merely assure that the proposed changes 
become effective upon installation of the corresponding plant 
modifications. Thus, these changes are purely administrative.
    Chapter 15 of the Clinton Updated Safety Analysis Report (USAR) 
discusses the effects of anticipated process disturbances to determine 
their consequences and the capability of the plant to control or 
accommodate such events. Subsection 15.2.6 discusses loss of AC power, 
including loss of grid voltage. This discussion demonstrates that fuel 
design limits and reactor coolant pressure boundary design conditions 
are not exceeded. The proposed changes do not affect the discussion nor 
the conclusion of this evaluation.
    (2) None of the proposed changes create the possibility of a new or 
different kind of accident from any accident previously evaluated. Each 
of the proposed changes is evaluated against this criterion as 
discussed below.
    The proposed changes (deletion of the channel check, the revised 
Allowable Value for the LOP degraded voltage instrumentation, revision 
of the minimum required voltage for the diesel generator (DG) 
surveillance, and change of the number of required channels for 
Division 3) do not alter the intent or purpose of the degraded voltage 
instrumentation. The instrumentation will continue to function to 
protect the loads on the emergency bus by switching automatically to 
the on site power source when the voltage has been at a degraded 
condition for greater than the Allowable Value of the time delay. The 
LOP instrumentation provides a responsive actuation (trip) to an 
accident or scenario where the protection provided by this function 
prevents damage to ECCS equipment during undervoltage (degraded 
voltage) conditions on the emergency bus(es). Because the 
instrumentation will continue to function to ensure that the emergency 
bus voltage for all three divisions is sufficient for the proper 
operation of all class 1E equipment down to the 120 volt level, the 
proposed change does not create the possibility of a new or different 
kind of accident from any accident previously evaluated. The change in 
the lower voltage for the DG

[[Page 18169]]

surveillances will not impact the way the surveillances are conducted 
because the DGs are run as close to the nominal voltage as possible. 
The lower voltage is a criterion for evaluating the surveillance and 
the revised lower voltage is adequate for its intended purpose.
    (3) None of the proposed changes involve a significant reduction in 
a margin of safety. Each of the proposed changes is evaluated against 
this criterion as discussed below.
    The proposed deletion of the channel check SR 3.3.8.1.1 will not 
result in any reduction of the margin of safety because the channel 
check is ineffective and the status of the channel will continue to be 
apparent to plant personnel because of information provided by other TS 
required surveillances. The margin of safety is provided by LOP 
instrumentation ensuring the emergency bus(es) have adequate voltage to 
support ECCS operability. The proposed revision of the Allowable Value 
for the LOP degraded voltage will provide assurance that emergency 
bus(es) voltage will be adequate for ECCS loads during DBA and other 
non-accident scenarios. These setpoints were determined based on 
revised voltage calculations and using an NRC-approved setpoint 
methodology. Thus, these changes will not involve any reduction of the 
margin of safety. The proposed revision of the number of required 
channels for Division 3 will not result in a reduction in a margin of 
safety because the proposed Division 3 LOP Degraded Voltage 
instrumentation logic will be the same as the proven design of Division 
1 and 2. This modification will improve plant maintenance and training 
by making Divisions 1, 2 and 3 similar thereby enhancing plant 
performance and safety.
    Similarly, the proposed revision of the lower voltage limit for 
voltage for the DG surveillances (SR 3.8.1.2, SR 3.8.1.7, SR 3.8.1.11, 
SR 3.8.1.12, SR 3.8.1.15, SR 3.8.1.19, and SR 3.8.1.20) will assure 
that the DGs will be capable of controlling voltage to a range that 
will be adequate for the loads on the bus. This value was determined 
using revised voltage calculations and is consistent with the proposed 
degraded voltage setpoints. None of the proposed changes will 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: Vespasian Warner Public 
Library, 120 West Johnson Street, Clinton, Illinois 61727.
    Attorney for licensee: Leah Manning Stetener, Vice President, 
General Counsel, and Corporate Secretary, 500 South 27th Street, 
Decatur, Illinois 62525.
    NRC Project Director: Gail H. Marcus.

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

    Date of amendment request: February 22, 1996 (U-602551).
    Description of amendment request: The proposed amendment would 
change Technical Specification 3.4.11, ``Reactor Coolant System (RCS) 
Pressure and Temperature (P/T) Limits,'' to incorporate specific P/T 
limits for the bottom head region of the reactor vessel, separate and 
apart from the core beltline region of the reactor vessel.
    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 results in a specific pressure and 
temperature (P/T) limit curve for the bottom head during vessel 
pressure testing evolutions, while the P/T limits for the remaining 
balance of reactor pressure vessel regions are unchanged. The limits 
for the bottom head region, which are only applicable during vessel 
system pressure or leak testing, were developed consistent with 
Regulatory Guide 1.99, Revision 2; 10CFR50, Appendix G; ASME Section 
III, Appendix G; and Welding Research Council (WRC) Bulletin 175. 
Additionally, the proposed change does not result in a change to the 
way in which the hydrostatic pressure tests are performed. That is, 
conformance to the P/T limits specified in Technical Specification 
Figure 3.4.11-1 with the proposed bottom head P/T limits incorporated, 
will continue to provide protection against brittle fracture of the 
vessel system during required testing so that vessel integrity is 
maintained. Therefore, this proposed change does not result in an 
increase in the probability or consequences of any accident previously 
evaluated.
    (2) The proposed change does not result in any change to the plant 
or the way in which the hydrostatic pressure tests are performed. As a 
result, no new failure modes are introduced. Therefore, the proposed 
change cannot create the possibility of a new or different kind of 
accident from any accident previously evaluated.
    (3) The new P/T limit curve for the bottom head has been developed 
consistent with Regulatory Guide 1.99, Revision 2; 10CFR50, Appendix G; 
ASME Section III, Appendix G; and Welding Research Council (WRC) 
Bulletin 175. All other regions of the reactor pressure vessel retain 
their applicability to appropriate and previously approved P/T limit 
curves which are based on the same methodology. Conformance to the P/T 
limit curves, with the proposed changes incorporated, will continue to 
provide adequate margins of safety against brittle fracture of the 
reactor vessel. Therefore, this proposed change does not result in 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: Vespasian Warner Public 
Library, 120 West Johnson Street, Clinton, Illinois 61727.
    Attorney for licensee: Leah Manning Stetener, Vice President, 
General Counsel, and Corporate Secretary, 500 South 27th Street, 
Decatur, Illinois 62525.
    NRC Project Director: Gail H. Marcus.

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

    Date of amendment request: February 22, 1996 (U-602522)
    Description of amendment request: The proposed amendment would 
change Technical Specification 3.3.4.1, ``End of Cycle Recirculation 
Pump Trip (EOC-RPT) Instrumentation,'' by deleting Surveillance 
Requirement (SR) 3.3.4.1.6. The SR requires the reactor recirculation 
pump trip breaker interruption time to be determined at least once per 
60 months.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration which is presented below:
    (1) End of cycle recirculation pump trip (EOC-RPT) actuation in 
response to main generator load rejection and main turbine trip events 
has previously been evaluated in Chapter 15 of Clinton Power Station 
(CPS) Updated Final

[[Page 18170]]

Safety Analysis. The proposed change does not affect the initiators of 
any of these events. In addition, the possibility of failure of the 
EOC-RPT breaker to mitigate these events has not been increased because 
there has been no change in design and no change to the plant. Deleting 
the requirement to periodically measure the breaker arc suppression 
time will not impact the EOC-RPT breakers' capability of performing 
their intended function because CPS will continue to perform 
inspections, testing and maintenance that supports breaker operation as 
intended and provides assurance that breaker interruption time will be 
within limits. Thus, the EOC-RPT breaker trip may be expected to 
operate as before to mitigate pressurization transient effects.
    The EOC-RPT breaker trip is also assumed to occur in the analyses 
for the loss of feedwater heating, feedwater controller failure, 
pressure regulator failure, recirculation flow control failure, and 
recirculation pump seizure events. However, the EOC-RPT breaker trip is 
not an initiator or mitigating feature for these events. The proposed 
change cannot therefore impact the probability or consequences for 
these events. Nonetheless, the EOC-RPT breaker trip may be assumed to 
function as before for these scenarios.
    For scenarios where the EOC-RPT breaker trip could initiate an 
event (i.e., inadvertent recirculation pump trip events), the 
probability of occurrence is not increased. The design and operation of 
the EOC-RPT system has not been changed, and therefore, the 
consequences resulting from the EOC-RPT breaker trip are unchanged.
    Based on the above, neither the probability nor the consequences of 
any accident previously evaluated have been increased.
    (2) As noted above, the EOC-RPT breakers will continue to function 
as before. The proposed change involves no design change or physical 
change in the plant. Therefore, previous accident analyses are 
unchanged. Further, no new operations or testing is involved. On this 
basis, no new failure modes are introduced. Therefore, this proposed 
change does not create the possibility of a new or different kind of 
accident from any accident previously evaluated.
    (3) This proposed change does not involve a significant reduction 
in a margin of safety. The capability of the EOC-RPT breaker trip to 
provide additional insertion of negative reactivity for mitigating 
design-basis events remains unchanged. That is, the EOC-RPT will 
continue to be capable of reducing the peak reactor pressure and power 
resulting from turbine trip or generator load rejection transients, 
thus providing additional margin to core thermal MCPR Safety Limits.
    The margin of safety is assured by the EOC-RPT breaker trip 
occurring within established limits such that the overall system 
performs its intended safety function within the time analyzed for the 
system safety response. No system time limit change is proposed. The 
robust design of the breakers, combined with continued performance of 
vendor-recommended testing and maintenance that ensures proper 
mechanical and electrical performance of the breakers, will continue to 
provide assurance that breaker interruption time is within the 
acceptable limit. Therefore, there is no significant reduction in the 
margin of safety.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 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: Vespasian Warner Public 
Library, 120 West Johnson Street, Clinton, Illinois 61727.
    Attorney for licensee: Leah Manning Stetener, Vice President, 
General Counsel, and Corporate Secretary, 500 South 27th Street, 
Decatur, Illinois 62525.
    NRC Project Director: Gail H. Marcus.

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

    Date of amendment request: February 22, 1996 (U-602549).
    Description of amendment request: The proposed amendment would 
revise Technical Specification (TS) 3.6.5.1, ``Drywell,'' to allow 
drywell bypass leakage tests to be performed at intervals of up to ten 
years based, in part, on the demonstrated performance of the drywell 
barrier with respect to leak tightness. The proposed amendment would 
also revise TS 3.6.5.2, ``Drywell Air Lock,'' to extend the testing 
intervals for the surveillances on drywell air lock overall leakage and 
interlock operability, relocate the specific leakage limits on the air 
lock barrel and door seals to the TS Bases, relocate the requirement to 
pressurize the drywell air lock to 19.7 psid prior to performance of 
the overall drywell air lock leakage test to the TS Bases, and other 
administrative changes.
    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 change to the plant 
design or operation. As a result, the proposed changes do not affect 
any of the parameters or conditions that contribute to initiation of 
any accidents previously evaluated. Therefore, the proposed changes 
cannot increase the probability of any accidents previously evaluated.
    The proposed changes do potentially affect the leaktight integrity 
of the drywell, a structure used to mitigate the consequences of a loss 
of coolant accident (LOCA). The function of the drywell is to force the 
steam released from a LOCA through the suppression pool, limiting the 
amount of steam released to the primary containment atmosphere. This 
serves to limit the containment pressurization due to the LOCA. The 
leakage of the drywell is limited to ensure that the primary 
containment does not exceed its design limits of 185 deg.F and 15 psig. 
Because the proposed change to replace the current 18-month frequency 
for performing drywell bypass leakage tests (DBLRTs) with a 
performance-based frequency does not alter the plant design, the 
proposed change does not directly result in an increase in the drywell 
leakage. However, decreasing the test frequency can increase the 
probability that a large increase in drywell bypass leakage could go 
undetected for an extended period of time. This potential has been 
evaluated, and Illinois Power has determined that the proposed change 
to the DBLRT frequency will not result in the potential for undetected, 
large increases in leakage, as further discussed below.
    There are several potential drywell bypass leakage paths. These 
include potential cracks in drywell concrete structure, the drywell 
vacuum breakers, and various penetrations through the drywell 
structure. Based on the results of the structural integrity test 
conducted at the design pressure of 30 psig as part of the 
preoperational test program, additional cracking of the drywell is not 
expected during the remaining life of the plant. Ventilation and piping 
penetrations (including the drywell vacuum breaker penetrations) are 
designed to ASME Code Class 2 and Seismic Category 1 requirements. 
These penetrations are typically designed with two isolation valves in 
series with one valve in the drywell and another either outside primary 
containment or in the wetwell. Technical Specification (TS) 
Surveillance Requirements (SRs) require, as applicable, periodic 
verification of drywell isolation valve

[[Page 18171]]

position, stroke time, and automatic isolation capability. High energy 
lines that extend into the wetwell, such as the main steam lines and 
feedwater lines, are encapsulated by guard pipes to direct energy back 
into the drywell in case of a piping rupture. Electrical penetrations 
are sealed with a high strength/density material that will prevent 
leakage, as well as provide radiation shielding.
    The proposed changes for the drywell air lock involve relocation of 
the separate limits on the drywell air lock barrel and seal leakage 
rates to the TS Bases, relocation of the requirement to pressurize the 
air lock to 19.7 psid prior to performance of the air lock overall 
(barrel) leakage test, and changing the frequency for these tests from 
18 months to 24 months. While the proposed changes will eliminate 
separate TS limits on leakage of the drywell air lock, the overall 
drywell bypass leakage TS limit (which includes leakage through the air 
lock) is not affected by this proposed change. The limiting scenario 
for drywell bypass leakage is a small break LOCA which results in 
drywell pressures of approximately 3 psid. Only a large break LOCA can 
create drywell pressures of 19.7 psid. For this event, the allowable 
drywell bypass leakage rate is over eight times larger than for a small 
break LOCA. Thus, relocation of these requirements to the TS Bases will 
continue to provide adequate control of these requirements. The 
proposed air lock overall leakage rate testing frequency is consistent 
with the guidance for testing primary containment air locks in Nuclear 
Energy Institute (NEI) 94-01, ``Industry Guideline for Implementing 
Performance-Based Option of 10CFR50, Appendix J.'' The drywell air lock 
is tested in a manner similar to the primary containment air locks, 
even though the drywell air lock is not a direct leakage path from 
primary containment and, therefore, 10CFR50, Appendix J test 
requirements do not apply. The drywell air lock's use is limited during 
plant operation due to radiation and temperature in the drywell. Since 
sufficient confidence in the door's sealing capability is assured 
considering past performance and the air lock door usage is very low 
throughout an operating cycle, it is justified to allow performance of 
these tests at refueling-outage intervals, whether the unit is on a 18-
month or a 24-month refueling cycle.
    Operational experience has shown that the leak tightness of the 
drywell has been maintained well below the allowable leakage limits at 
Clinton Power Station. The TS limit of 10% of the design [maximum 
allowable leakage path area] provides a large margin for degradation. 
Drywell performance to date suggests that drywell degradation, even 
with a ten-year interval between tests, will not exceed this margin. 
The most recent DBLRT performed during the fourth refueling outage (RF-
4) measured a drywell bypass leakage rate of 0.07% of the design limit.
    An analysis was also conducted to determine the potential risk to 
the public from unacceptable drywell bypass leakage going undetected as 
a result of the proposed change. Based on this probabilistic risk 
analysis, for several different accident scenarios, the risk of 
radioactivity release from containment was found to be insignificant.
    Based on the above, Illinois Power has concluded that the proposed 
changes will not result in a significant increase in the consequences 
of any accident previously evaluated.
    (2) The proposed change does not involve a change to the plant 
design or operation. As a result, the proposed change does not affect 
any of the parameters or conditions that could contribute to initiation 
of any accidents. Drywell bypass leakage cannot, of itself, create an 
accident. Thus, it has been concluded that the proposed change cannot 
create the possibility of an accident not previously evaluated.
    (3) The NRC has provided standards for determining whether a no 
significant hazards consideration exists as stated in 10CFR50.92(c). 
These proposed changes involve the withdrawal of operating restrictions 
previously imposed because acceptable operation of the Mark III primary 
containment design had not been demonstrated at the time of initial 
licensing. As published in the Federal Register (FR) regarding no 
significant hazards consideration criteria, granting of a relief based 
upon demonstration of acceptable operation from an operating 
restriction that was imposed because acceptable operation had not yet 
been demonstrated does not involve a significant hazards consideration 
(reference 48 FR 14870).
    The proposed change only affects the frequency of measuring the 
drywell bypass leakage rate and does not change the bypass leakage rate 
limit. The proposed change could potentially increase the probability 
that a large increase in drywell bypass leakage could go undetected for 
an extended period of time. However, operational experience has shown 
that the leaktightness of the drywell has been maintained well below 
the allowable leakage limits. In addition, there are TS surveillances 
which require, as applicable, periodic verification of drywell 
isolation valve position, stroke time, and automatic isolation 
capability. Further, qualitative methods (such as periodic verification 
that the drywell pressurizes, which ensures that the drywell leak rate 
is less than the instrument air leak and usage rates) are available to 
provide assurance that the drywell leakage rate is being maintained 
within limits. The Clinton Power Station TS require the drywell leakage 
rate measured during DBLRTs to be less than or equal to 10% of the 
design limit. This request does not affect this required margin. Nor 
does it affect the existing margin between the primary containment 
design pressure and the actual pressure at which primary containment 
would fail.
    With respect to proposed changes to the drywell air lock overall 
leakage testing and interlock testing requirements, the proposed leak 
test frequencies are consistent with the guidance for testing primary 
containment air locks in NEI 94-01. Due to the limited use of the 
drywell air locks during plant operation, it is justified to allow 
performance of interlock operability testing on a refueling outage 
basis, whether the unit is on an 18-month or a 24-month refueling 
cycle. The separate limits on the drywell air lock and barrel are being 
relocated from the TS, these limits are being controlled under 
10CFR50.59 and the TS Bases Control program of TS 5.5.11. Leakage 
through these pathways will continue to be a part of the overall 
drywell bypass leakage limited by LCO 3.6.5.1.
    An analysis was also conducted to determine the potential risk to 
the public from the proposed change. Based on this probabilistic risk 
analysis, for several different accident scenarios, the risk of 
radioactivity release from containment was found to be insignificant.
    As a result, Illinois Power has concluded that the proposed changes 
will continue to assure that the drywell bypass leakage will be within 
design limits if challenged and therefore, will not result in 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: Vespasian Warner Public 
Library, 120 West Johnson Street, Clinton, Illinois 61727.

[[Page 18172]]

    Attorney for licensee: Leah Manning Stetener, Vice President, 
General Counsel, and Corporate Secretary, 500 South 27th Street, 
Decatur, Illinois 62525.
    NRC Project Director: Gail H. Marcus.

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

    Date of amendment requests: February 26, 1996 (AEP:NRC:1071U).
    Description of amendment requests: The proposed amendments would 
modify the technical specifications (TS) to increase the current limit 
on nominal fuel assembly enrichment for new, Westinghouse-fabricated, 
fuel stored in the new fuel storage racks from 4.55 weight percent 
uranium-235 isotope to 4.95 weight percent uranium-235 isotope with 
certain provisions. Also, TS 5.6.2 would be reformatted similar to that 
used in the standard TS (NUREG-1431, Rev. 1).
    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:
    Per 10 CFR 50.92, a proposed amendment will not involve a 
significant hazards consideration if the proposed amendment does not:
    (1) involve a significant increase in the probability or 
consequences of an accident previously evaluated,
    (2) create the possibility of a new or different kind of accident 
from any accident previously evaluated, or
    (3) involve a significant reduction in a margin of safety.

Criterion 1

    The proposed changes will not involve a significant increase in the 
probability of an accident previously evaluated because similar 
administrative controls to those presently used to identify new fuel 
storage rack inventory and compliance with T/S limits will be used. 
There are no physical changes to the plant associated with this T/S 
change. The consequences of an accident previously evaluated will not 
be increased because the reactivity of the fuel stored in the new fuel 
storage racks under the proposed T/S limits will be no greater than the 
reactivity of fuel stored in the new fuel storage racks presently 
allowable under the current T/S limits.

Criterion 2

    The proposed changes will not create the possibility of a new or 
different kind of accident from any accident previously evaluated 
because the changes will involve no physical changes to the plant nor 
any changes in plant operations. Furthermore, the reactivity of the 
fuel stored in the new fuel storage racks under the proposed T/S limits 
will be no greater than the reactivity of fuel stored in the new fuel 
storage rack presently allowable under the current T/S limits.

Criterion 3

    The proposed amendment will not involve a significant reduction in 
a margin of safety because the reactivity of the fuel stored in the new 
fuel storage racks under the proposed T/S limits will be no greater 
than the reactivity of fuel stored in the new fuel storage racks 
presently allowable under the current T/S limits.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. In addition, the reformatting of TS 5.6.2 is a purely 
administratiave change having no effect on the physical plant or its 
operation. Therefore, the NRC staff proposes to determine that the 
amendment requests involve no significant hazards consideration.
    Local Public Document Room location: Maud Preston Palenske Memorial 
Library, 500 Market Street, St. Joseph, Michigan 49085.
    Attorney for licensee: Gerald Charnoff, Esq., Shaw, Pittman, Potts 
and Trowbridge, 2300 N Street, NW, Washington, DC 20037.
    NRC Project Director: Mark Reinhart, Acting.

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

    Date of amendment requests: February 29, 1996 (AEP:NRC:1232).
    Description of amendment requests: The proposed amendments would 
revise the technical specifications to reduce the boric acid 
concentration in the boric acid storage system from approximately 12 
percent to approximately 4 percent by weight. Related changes are also 
proposed to increase the minimum required flow rate in action 
statements for certain affected TS and add an additional surveillance 
requirement for this flow rate, and decrease the minimum temperature 
requirement in certain affected TS to 63  deg.F. The bases section is 
also updated to reflect these proposed changes.
    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:
    Per 10 CFR 50.92, a proposed change does not involve a significant 
hazards consideration if the change does not:
    1. involve a significant increase in the probability or 
consequences of an accident previously evaluated,
    2. create the possibility of a new or different kind of accident 
from any accident previously evaluated, and
    3. involve a significant reduction in a margin of safety.

Criterion 1

    Does the change involve a significant increase in the probability 
or consequences of an accident previously evaluated?
    NO. The BAST [boric acid storage tank] water volume and boron 
concentration were not credited in any Chapter 14 safety analysis. 
Therefore, no change in the probabilities of the accident analysis will 
result from the BAST water volume and boron concentration change. In 
addition, since the BAST water volume and boron concentration are not 
taken into consideration in any safety analysis, the consequences of an 
accident previously evaluated in the FSAR [final safety analysis 
report] are not increased. The heat tracing system is currently only 
necessary to prevent precipitation of existing high boric acid 
concentration in the plant systems. The reduction in boron 
concentration in this proposal eliminates the need for the heat tracing 
system. The existence of the heat tracing system was not part of any 
safety analysis and disabling of the heat tracing system will not 
result in a significant increase in the probability or consequences of 
an accident previously evaluated.

Criterion 2

    Does the change create the possibility of a new or different kind 
of accident from any accident previously evaluated?
    NO. Since the minimum required water flow from the boric acid 
storage system to the reactor coolant system was increased to 
counteract any possible operational transients, as shown in Attachment 
4 [of the application], the change in BAST water volume and boron 
concentration and disabling of the heat tracing system do not create 
the possibility of an accident which is different from any already 
evaluated in the FSAR. No new or different failure modes have been 
defined for any system or component nor has any new limiting single 
failure been identified.

[[Page 18173]]

Criterion 3

    Does the change involve a significant reduction in a margin of 
safety?
    NO. The margin of safety requirements are not affected by the 
removal of the heat tracing system and the reduction of the boric acid 
concentration in the boric acid storage system. The required flow paths 
and borated water sources are unaffected by this proposal. The required 
quantity of borated water is still available based upon the performed 
evaluation, and appropriate surveillance requirements ensure the 
ability to deliver this borated water. The reduction of the boric acid 
concentration in the BASTs will ensure that the boric acid remains in 
solution at the normal room temperature in the auxiliary building. With 
the above changes, there will be a net improvement in system 
reliability and accordingly the proposed changes do not affect 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 requests involve no significant hazards consideration.
    Local Public Document Room location: Maud Preston Palenske Memorial 
Library, 500 Market Street, St. Joseph, Michigan 49085.
    Attorney for licensee: Gerald Charnoff, Esq., Shaw, Pittman, Potts 
and Trowbridge, 2300 N Street, NW, Washington, DC 20037.
    NRC Project Director: Mark Reinhart, Acting.

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

    Date of amendment requests: March 13, 1996.
    Description of amendment requests: The proposed amendments would 
revise the combined Technical Specifications (TS) for the Diablo Canyon 
Power Plant, Unit Nos. 1 and 2 to revise TS 4.0.5, ``Surveillance 
Requirements,'' to delete reference to prior NRC approval for written 
relief from the Inservice Inspection (ISI) and Inservice Testing 
Program (IST) requirements and to add ASME Section XI definition of 
``Biennially or every 2 years--At least once per 731 days'' in TS 
4.0.5b.
    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.
    The proposed changes implement the NRC's recommendation contained 
in NUREG-1482, ``Guidelines for Inservice Testing Programs at Nuclear 
Power Plants,'' endorsed by Generic Letter
    89-04, Supplement 1, ``Guidance on Developing Acceptable Inservice 
Testing Programs.'' The changes are consistent with 10 CFR 50.55a, 
``Codes and Standards,'' which does not prohibit the implementation of 
relief from ASME Section XI requirements prior to specific written 
approval when those changes are found acceptable by change process 
specified in 10 CFR 50.59, ``Changes, Tests and Experiments.'' The 
proposed changes are administrative in nature and do not involve any 
modifications to any plant equipment or affect plant operation.
    Therefore, the proposed changes do not involve a significant 
increase in the probability or consequences of an accident previously 
evaluated.
    2. The proposed changes do not create the possibility of a new or 
different kind of accident from any accident previously evaluated.
    The proposed changes are administrative in nature, do not involve 
any physical alterations to any plant equipment, and cause no change in 
the method by which any safety-related system performs its 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 changes do not involve a significant reduction in a 
margin of safety.
    The proposed changes do not alter the basic regulatory requirements 
and do not affect any safety analyses.
    Therefore, the proposed changes do not involve a significant 
reduction in a margin of safety.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment requests involve no significant hazards consideration.
    Local Public Document Room location: California Polytechnic State 
University, Robert E. Kennedy Library, Government Documents and Maps 
Department, San Luis Obispo, California 93407.
    Attorney for licensee: Christopher J. Warner, Esq., Pacific Gas and 
Electric Company, P.O. Box 7442, San Francisco, California 94120.
    NRC Project Director: William H. Bateman.

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

    Date of amendment request: April 3, 1996.
    Description of amendment request: The proposed amendments would 
revise the combined Technical Specifications (TS) for the Diablo Canyon 
Power Plant, Unit Nos. 1 and 2 to revise Technical Specifications 3/
4.7.5, ``Control Room Ventilation System,'' 3/4.7.6, ``Auxiliary 
Building Safeguards Air Filtration System,'' and 3/4.9.12, ``Fuel 
Handling Building Ventilation System,'' to clarify the testing 
methodology utilized by PG&E to determine the operability of the 
charcoal and high-efficiency particulate air (HEPA) filters in the 
engineering safeguards features (ESF) air handling units.
    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 charcoal testing protocol changes will not affect system 
operation or performance, nor do they affect the probability of any 
event initiators. These changes do not affect any engineered safety 
features actuation setpoints or accident mitigation capabilities. The 
new charcoal adsorber sample laboratory testing protocol more 
accurately demonstrates the required performance of the adsorbers in 
the control room ventilation system and auxiliary building safeguards 
air filtration system following a design basis loss of coolant accident 
or in the fuel handling building ventilation system following a fuel 
handling accident outside containment. The decontamination efficiencies 
used in the offsite and control room dose analyses are not affected by 
these changes. Therefore, offsite and control room dose analyses are 
not affected by this change, and all offsite and control room doses 
will remain within the limits of 10 CFR 100 and 10 CFR 50, Appendix A, 
General Design Criterion (GDC) 19.

[[Page 18174]]

    The requirements of ANSI N510-1980 encompass the requirements of 
ANSI N510-1975, which is referenced in Regulatory Guide (RG) 1.52, as 
it applies to testing at Diablo Canyon Power Plant (DCPP). 
Consequently, revising the Technical Specifications (TS) to reference 
ANSI N510-1980 will have no effect on filter testing.
    The proposed changes are consistent with the new Standard Technical 
Specifications (NUREG-1431, Rev. 1).
    Therefore, the proposed changes do 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 changes to the charcoal sample testing protocol will not affect 
the method of operation of the system. The proposed changes only affect 
the testing parameters for the charcoal samples. No new or different 
accident scenarios, transient precursors, failure mechanisms, or 
limiting single failures will be introduced as a result of these 
changes.
    The requirements of ANSI N510-1980 encompass the requirements of 
ANSI N510-1975, which is referenced in RG 1.52, as it applies to 
testing at DCPP. Consequently, revising the TSs to reference ANSI N510-
1980 will have no effect on filter testing.
    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 change does not involve a significant reduction in 
a margin of safety.
    The changes in charcoal sample testing protocol will not affect 
system performance or operation. The decontamination efficiencies used 
in the offsite and control room dose analyses are not affected by these 
changes. Therefore, offsite and control room dose analyses are not 
affected by this change, and all offsite and control room doses will 
remain within the limits of 10 CFR 100 and 10 CFR 50, Appendix A, GDC 
19.
    The requirements of ANSI N510-1980 encompass the requirements of 
ANSI N510-1975, which is referenced in RG 1.52, as it applies to 
testing at DCPP. Consequently, revising the TSs to reference ANSI N510-
1980 will have no effect on filter testing.
    Therefore, the proposed changes do not involve a significant 
reduction in a margin of safety.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment requests involve no significant hazards consideration.
    Local Public Document Room location: California Polytechnic State 
University, Robert E. Kennedy Library, Government Documents and Maps 
Department, San Luis Obispo, California 93407.
    Attorney for licensee: Christopher J. Warner, Esq., Pacific Gas and 
Electric Company, P.O. Box 7442, San Francisco, California 94120.
    NRC Project Director: William H. Bateman.

Pacific Gas and Electric Company, Docket No. 50-133, Humboldt Bay Power 
Plant, Unit 3, Humboldt County, California

    Date of amendment request: March 13, 1996.
    Description of amendment request: The proposed amendment would 
revise the Humboldt Bay Power Plant (HBPP), Unit 3, Technical 
Specifications (TS) by incorporating position changes to reflect a 
proposed plant staff reorganization. The TS changes proposed are as 
follows:
    (1) TS Section VII.C.2.c and VII.D.1.b--change the position title 
from ``Power Plant Engineer'' to ``Senior Power Production Engineer.''
    (2) TS Section VII.C.2.d--change the position title from ``Senior 
Chemical and Radiological Engineer'' to ``Senior Radiation Protection 
Engineer.''
    (3) TS Section VII.C.2.e and VII.D.1.b--change the position title 
from ``Maintenance Planner'' to ``Supervisor of Maintenance.''
    (4) TS Section VII.C.2.g and VII.D.1.b--add the position of 
``Assistant Plant Manager/Power Plant Engineer.''
    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 administrative and organizational changes provide 
editorial corrections and reflect the proposed HBPP and current NRC 
organizations. These changes do not affect the operating methodology of 
HBPP, and they are not related to the probability or consequences of an 
accident previously evaluated.
    Therefore, the proposed changes do 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 revisions to the HBPP TS are organizational and 
administrative in nature, and do not change the method by which any 
safety-related system performs its function.
    Therefore, the proposed changes do not create the possibility of a 
new of different kind of accident from any accident previously 
evaluated.
    3. Does the change involve a significant reduction in a margin of 
safety?
    The proposed changes have no effect on the current operating 
methodologies or actions that govern plant performance. In addition, 
the proposed changes do not affect the margin of safety associated with 
parameters for any accident analysis.
    Therefore, the proposed changes do not involve a significant 
reduction in a margin of safety.
    The NRC staff has reviewed the analysis of the licensee and, based 
on this review, it appears that the three standards of 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: Humboldt County Library, 1313 
3rd Street, Eureka, California 95501.
    Attorney for licensee: Christopher J. Warner, Esquire, Pacific Gas 
& Electric Company, P.O. Box 7442, San Francisco, California 94120.
    NRC Project Director: Seymour H. Weiss.

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

    Date of application request: February 9, 1996, as superseded by 
letter dated March 22, 1996.
    Description of amendment request: The amendment would revise 
Technical Specification (TS) Definition 1.7, TS 3/4.6, TS 6.8, and 
their associated bases to directly reference Regulatory Guide 1.163 as 
required by 10 CFR 50, Appendix J, Option B, for the Type A containment 
integrated leak rate tests (ILRTs) and the Type B and C local leak rate 
tests (LLRTs).
    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

[[Page 18175]]

consideration, which is presented below:
    1. The proposed change does not involve a significant increase in 
the probability or consequences of an accident previously evaluated.
    The proposed changes to TS 1.7e, 4.6.1.1, 3/4.6.1.3, Bases 3/
4.6.1.1 and the program addition to TS 6.8.4g have no effect on plant 
operation. The proposed changes only provide mechanisms within TS for 
implementing a performance-based methodology for determining the 
frequency of leak rate testing, as allowed by the NRC. The test type, 
method, and acceptance criteria will not be changed. Containment 
leakage will continue to be maintained within the required limits. 
Based on industry and NRC evaluations performed in support of 
developing Option B, these changes potentially result in a minor 
increase in the consequences of an accident previously evaluated due to 
the increased testing intervals. However, the proposed changes do not 
result in an increase in the core damage frequency since the 
containment system is used for mitigation purposes only.
    Directly referencing the Containment Leakage Rate Testing Program 
for Containment ILRT and LLRT requirements does not involve any 
modification to plant equipment or affect the operation or design basis 
of the containment. Leakage rate testing is not a precursor to or an 
initiating event for any accident.
    Therefore, these changes do 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 changes only allow for implementation of 10 CFR 50, 
Appendix J, Option B and do not involve any modifications to any plant 
equipment or affect the operation or design basis of the containment. 
The proposed changes do not affect the response of the containment 
during a design basis accident.
    3. The proposed change does not involve a significant reduction in 
a margin of safety.
    The proposed changes do not affect or change a safety limit, any 
limiting condition for operation or affect plant operations. The 
changes only implement the Appendix J, Option B test frequencies that 
have been determined by NRC not to involve a safety concern. The 
testing methods, acceptance criteria and bases are not changed and 
still provide assurance that the containment will provide its intended 
function.
    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: Callaway County Public 
Library, 710 Court Street, Fulton, Missouri 65251.
    Attorney for licensee: Gerald Charnoff, Esq., Shaw, Pittman, Potts 
& Trowbridge, 2300 N Street, N.W., Washington, D.C. 20037.
    NRC Project Director: William H. Bateman.

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

    Date of amendment request: March 21, 1996.
    Description of amendment request: The proposed changes to the 
Technical Specifications (TS) for the North Anna Power Station, Units 
1&2 (NA-1&2) would clarify the requirements for testing charcoal 
adsorbent in the Waste Gas Charcoal Filter System, the Control Room 
Emergency Habitability System, and the Safeguards Area Ventilation 
System. No change in the testing is being proposed, only clarification 
of the description of the required testing in TS 3/4.6.4.3, 3/4.7.7.1, 
and 3/4.7.8.1.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:
    The proposed Technical Specifications changes will revise 
Surveillance Requirements for the charcoal adsorbent in the Waste Gas 
Charcoal Filter System (TS 3/4.6.3.), Control Room Emergency 
Habitability System (TS 3/4.7.7.2), and the Safeguards Area Ventilation 
System (TS 3/4.7.8.1) to reflect the current testing methodology for 
new and used carbon adsorbent. These proposed changes specify ASTM D 
3803-1979 as the laboratory testing standard for both new and used 
charcoal adsorbent for the ventilation system identified above.
    Virginia Electric and Power has evaluated the proposed Technical 
Specification changes to the North Anna Units 1 and 2 Technical 
Specifications against the Significant Hazards Criteria of 10 CFR 50.92 
and determined that the changes do not involve any significant hazard 
for the following reasons:
    1. The probability or consequences of an accident previously 
evaluated is not significantly increased.
    The proposed changes are administrative in nature in that the 
changes only explicitly specify the current testing methodology for 
charcoal adsorbent. The proposed changes will not affect system 
operation or performance, nor do they affect the probability of any 
event initiators. These changes do not affect any Engineered Safety 
Features actuation setpoints or accident mitigation capabilities. 
Therefore, the proposed changes will not significantly increase the 
consequences of an accident or malfunction of equipment important to 
safety previously evaluated in the UFSAR.
    2. The possibility of an accident or a malfunction of a different 
type than any previously evaluated is not created.
    The proposed changes only clarify the requirements for charcoal 
testing and will not affect the method of operation of the ventilation 
systems. Furthermore, the proposed changes are only intended to clarify 
the existing requirements to explicitly specify the current test 
methodology. No new or different accident scenarios, transient 
precursors, failure mechanisms, or limiting single failures will be 
introduced as a result of these changes. Therefore, the possibility of 
a new or different kind of accident other than those already evaluated 
will not be created by this change.
    3. The margin of safety has not been significantly reduced.
    The proposed changes which represent the current laboratory testing 
methodology for charcoal adsorber samples, demonstrates the required 
performance of the adsorbent following a design basis LOCA or Fuel 
Handling Accident. Changing the Technical Specification to clarify the 
methodology for charcoal sample testing will not affect system 
performance or operation.
    Therefore, these changes will not result in a significant reduction 
in any margin of safety.
    Based on the above discussions, it has been determined that the 
requested Technical Specification changes do not involve a significant 
increase in the probability or consequences of an accident or other 
adverse condition over previous evaluations; or create the possibility 
of a new or different kind of accident or condition over previous 
evaluation; or involve a significant reduction in a margin of safety. 
Therefore, the requested license amendment does not involve a 
significant hazards consideration.

[[Page 18176]]

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: The Alderman Library, Special 
Collections Department, University of Virginia, Charlottesville, 
Virginia 22903-2498.
    Attorney for licensee: Michael W. Maupin, Esq., Hunton and 
Williams, Riverfront Plaza, East Tower, 951 E. Byrd Street, Richmond, 
Virginia 23219
    NRC Project Director: Eugene V. Imbro.

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.

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

    Date of amendment request: March 21, 1996.
    Brief description of amendments: The amendments provide changes to 
Technical Specifications (TS) for CR3 relating to the Once Through 
Steam Generator's (OTSG's) tube inspection acceptance criteria, and 
repair limit for removing steam generator tubes from service. The 
proposed TS change would be applicable for one cycle duration, and only 
to Inter-Granular-Attack (IGA) degradation mechanism in a limited 
region of the OTSG.
    Date of publication of individual notice in Federal Register: March 
28, 1996 (61 FR 13888)
    Expiration date of individual notice: April 29, 1996.
    Local Public Document Room location: Coastal Region Library, 8619 
W. Crystal Street, Crystal River, Florida 32629.

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

    Date of amendment request: May 30, 1995, as supplemented by letter 
dated February 8, 1996.
    Description of amendment request: The proposed amendment would 
increase the spent fuel pool heat load licensing basis to provide 
greater flexibility for normal refueling practices.
    Date of individual notice in the Federal Register: April 3, 1996 
(61 FR 14832)
    Expiration date of individual notice: May 3, 1996.
    Local Public Document Room location: Wharton County Junior College, 
J. M. Hodges Learning Center, 911 Boling Highway, Wharton, TX 77488.

Notice of Issuance of Amendments to Facility Operating Licenses

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

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

    Date of application for amendments: December 20, 1995.
    Brief description of amendments: These amendments change the 
instrument setpoint for the reactor trip and main steam isolation 
signal actuation on low steam generator pressure from greater than or 
equal to 919 psia with an allowable value of 911 psia to 895 psia with 
an allowable value of greater than or equal to 890 psia.
    Date of issuance: April 5, 1996.
    Effective date: April 5, 1996, to be implemented within 45 days of 
issuance.
    Amendment Nos.: Unit 1-105; Unit 2-97; Unit 3-77.
    Facility Operating License Nos. NPF-41, NPF-51, and NPF-74: The 
amendments revised the Technical Specifications.
    Date of initial notice in Federal Register: February 28, 1996 (61 
FR 7544) The Commission's related evaluation of the amendments is 
contained in a Safety Evaluation dated April 5, 1996.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: Phoenix Public Library, 1221 
N. Central Avenue, Phoenix, Arizona 85004.

Baltimore Gas and Electric Company, Docket Nos. 50-317 and 50-318, 
Calvert Cliffs Nuclear Power Plant, Unit Nos. 1 and 2, Calvert County, 
Maryland

    Date of application for amendments: November 1, 1995 as 
supplemented on December 1, 1995.
    Brief description of amendments: The amendments reflect the new 
plant electrical distribution configuration, surveillance and limiting 
condition for operation of the new safety-related (SR) emergency diesel 
generator (EDG), the increased electrical capacities for the two of the 
three existing SR EDGs, the increased EDG fuel oil storage capacity, 
and the fire protection system for the

[[Page 18177]]

new EDG building. The remaining existing SR EDG will be upgraded during 
the Unit No. 2 refueling outage scheduled for the spring of 1997.
    Date of issuance: April 2, 1996.
    Effective date: As of the date of issuance to be implemented within 
30 days.
    Amendment Nos.: 214 and 191.
    Facility Operating License Nos. DPR-53 and DPR-69: Amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: January 3, 1996 (61 FR 
175) The Commission's related evaluation of these amendments is 
contained in a Safety Evaluation dated April 2, 1996.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: Calvert County Library, Prince 
Frederick, Maryland 20678.

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

Docket Nos. STN 50-456 and STN 50-457, Braidwood Station, Unit Nos. 1 
and 2, Will County, Illinois

    Date of application for amendments: December 6, 1995, as 
supplemented February 27, 1996, and March 28, 1996.
    Brief description of amendments: The amendments modify the 
technical specifications to replace the existing scheduling 
requirements for overall integrated and local containment leakage rate 
testing with a requirement to perform the testing in accordance with 10 
CFR Part 50, Appendix J, Option B. Option B allows test scheduling to 
be adjusted based on past performance.
    Date of issuance: April 4, 1996.
    Effective date: April 4, 1996.
    Amendment Nos.: 81, 81, 73, and 73.
    Facility Operating License Nos. NPF-37, NPF-66, NPF-72 and NPF-77: 
The amendments revised the Technical Specifications.
    Date of initial notice in Federal Register: February 28, 1996 (61 
FR 7547) The February 27, 1996, and March 28, 1996, supplements 
modified the Technical Specification pages to be more consistent with 
the published guidance, were within this scope of the initial notice, 
and did not affect the initial proposed no significant hazards 
consideration determination. The Commission's related evaluation of the 
amendments is contained in a Safety Evaluation dated April 4, 1996.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: For Byron, the Byron Public 
Library District, 109 N. Franklin, P.O. Box 434, Byron, Illinois 61010; 
for Braidwood, the Wilmington Public Library, 201 S. Kankakee Street, 
Wilmington, Illinois 60481.

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

Docket Nos. STN 50-456 and STN 50-457, Braidwood Station, Unit Nos. 1 
and 2, Will County, Illinois

    Date of application for amendments: October 3, 1995, as 
supplemented on February 21, 1996, and April 2, 1996.
    Brief description of amendments: The amendments revise the 
Technical Specifications (TS) to implement ten of the line-item TS 
improvements recommended in Generic Letter (GL) 93-05, ``Line-Item 
Technical Specifications Improvements to Reduce Surveillance 
Requirements for Testing During Power Operation,'' dated September 27, 
1993. The amendments also include editorial changes on the affected TS 
pages.
    Date of issuance: April 10, 1996.
    Effective date: April 10, 1996.
    Amendment Nos.: 82, 82 and 74, 74.
    Facility Operating License Nos. NPF-37, NPF-66, NPF-72 and NPF-77: 
The amendments revised the Technical Specifications.
    Date of initial notice in Federal Register: November 27, 1995 (60 
FR 58397). The February 21, 1996, and April 2, 1996, submittals did not 
change the initial proposed no significant hazards consideration 
determination.
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated April 10, 1996.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: For Byron, the Byron Public 
Library District, 109 N. Franklin, P.O. Box 434, Byron, Illinois 61010; 
for Braidwood, the Wilmington Public Library, 201 S. Kankakee Street, 
Wilmington, Illinois 60481.

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

Docket Nos. STN 50-456 and STN 50-457, Braidwood Station, Unit Nos. 1 
and 2, Will County, Illinois

    Date of application for amendments: May 17, 1995, as supplemented 
by letters dated January 17, March 8, March 18, April 4 and April 9, 
1996.
    Brief description of amendments: The amendments revised the 
Facility Operating Licenses and the technical specifications to permit 
the steam generator tubes to be repaired using the tungsten inert gas 
welded sleeve process developed by ABB-Combustion Engineering and 
remove references to the kinetically welded sleeving process.
    Date of issuance: April 12, 1996.
    Effective date: April 12, 1996.
    Amendment Nos.: 83, 83, 75, and 75.
    Facility Operating License Nos. NPF-37, NPF-66, NPF-72 and NPF-77: 
The amendments revised licenses and the Technical Specifications.
    Date of initial notice in Federal Register: July 5, 1995 (60 FR 
35064) The additional submittals provided information that did not 
change the initial proposed no significant hazards consideration 
determination. The Commission's related evaluation of the amendments is 
contained in a Safety Evaluation dated April 12, 1996.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: For Byron, the Byron Public 
Library District, 109 N. Franklin, P.O. Box 434, Byron, Illinois 61010; 
for Braidwood, the Wilmington Public Library, 201 S. Kankakee Street, 
Wilmington, Illinois 60481.

Commonwealth Edison Company, Docket Nos. 50-237 and 50-249, Dresden 
Nuclear Power Station, Units 2 and 3, Grundy County, Illinois

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

    Date of application for amendments: September 1, 1995, for Dresden 
and September 20, 1995, for Quad Cities.
    Brief description of amendments: This application upgrades the 
current custom Technical Specifications (TS) for Dresden and Quad 
Cities to the Standard Technical Specifications contained in NUREG-
0123, ``Standard Technical Specification General Electric Plants BWR/
4.'' This application upgrades only Section 6.0, ``Administrative 
Controls.''
    Date of issuance: April 2, 1996.
    Effective date: Immediately, to be implemented no later than June 
30, 1996.
    Amendment Nos.: 149, 143, 170, and 166.
    Facility Operating License Nos. DPR-19, DPR-25, DPR-29 and DPR-30: 
The amendments revised the Technical Specifications.
    Date of initial notice in Federal Register: September 20, 1995 (60 
FR 48728) for Dresden and October 5, 1995 (60 FR 52226) for Quad 
Cities. The Commission's related evaluation of the amendments is 
contained in a Safety Evaluation dated April 2, 1996.

[[Page 18178]]

    No significant hazards consideration comments received: No.
    Local Public Document Room location: for Dresden, Morris Area 
Public Library District, 604 Liberty Street, Morris, Illinois 60450; 
for Quad Cities, Dixon Public Library, 221 Hennepin Avenue, Dixon, 
Illinois 61021.

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

    Date of application for amendments: January 18, 1996, as 
supplemented on March 1, March 22, March 26, and April 3, 1996.
    Brief description of amendments: The amendments change the 
setpoints for the automatic primary containment isolation signal upon 
detection of a high main steamline tunnel differential temperature and 
delete the automatic isolation function upon detection of a high main 
steamline tunnel temperature. Additionally, the amendments provide a 12 
hour allowed outage time for the Main Steam Line Tunnel Differential 
Temperature--High isolation signal upon loss of the Reactor Building 
Ventilation System.
    Date of issuance: April 4, 1996.
    Effective date: Immediately, to be implemented prior to restart 
from refueling outage L1R07 (Unit 1) and L2R07 (Unit 2).
    Amendment Nos.: 111 and 96.
    Facility Operating License Nos. NPF-11 and NPF-18: The amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: February 27, 1996 (61 
FR 7281). The March 1, March 22, March 26 and April 3, 1996, submittals 
provided additional clarifying information that did not change the 
initial proposed no significant hazards consideration determination. 
The Commission's related evaluation of the amendments is contained in a 
Safety Evaluation dated April 4, 1996.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: Jacobs Memorial Library, 
Illinois Valley Community College, Oglesby, Illinois 61348.

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

    Date of application for amendments: August 25, 1995 as supplemented 
on December 15, 1995, February 5, February 9, February 28, March 4, 
March 28 and April 3, 1996.
    Brief description of amendments: These amendments revise the 
LaSalle Facility Operating Licenses and Technical Specifications (TSs) 
to reflect the deletion of the leakage control system (LCS) presently 
installed to control and contain the leakage past the main steamline 
isolation valves (MSIVs) on each of the four main steamlines. The TSs 
are also revised to raise the allowable leakage rates from 25 standard 
cubic feet per hour (scfh) for each set of MSIVs and a total of 100 
scfh from all four main steamlines to values of 100 scfh per steamline 
and 400 scfh for all four steamlines.
    Date of issuance: April 5, 1996.
    Effective date: Immediately, to be implemented by startup from 
refueling outage L1R07 (Unit 1) and L2R07 (Unit 2).
    Amendment Nos.: 112 and 97.
    Facility Operating License Nos. NPF-11 and NPF-18: The amendments 
revised the licenses and technical specifications.
    Date of initial notice in Federal Register: October 25, 1995 (60 FR 
54717). The December 15, 1995, February 5, February 9, February 28, 
March 4, March 28 and April 3, 1996, submittals provided additional 
information that did not change the initial proposed no significant 
hazards consideration determination. The Commission's related 
evaluation of the amendments is contained in a Safety Evaluation dated 
April 5, 1996.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: Jacobs Memorial Library, 
Illinois Valley Community College, Oglesby, Illinois 61348.

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

    Date of application for amendment: June 16, 1994, as supplemented 
February 6, 1995.
    Brief description of amendment: The amendment revises License 
Condition 2.K and relocates the Indian Point Nuclear Generating Unit 
No. 2 (IP2) fire protection requirements from the IP2 Technical 
Specifications to the IP2 fire protection program plan in accordance 
with the guidance provided in Generic Letter (GL) 86-10, 
``Implementation of Fire Protection Requirements,'' April 24, 1986, and 
GL 88-12, ``Removal of Fire Protection Requirements from Technical 
Specifications,'' August 2, 1988.
    Date of issuance: March 26, 1996.
    Effective date: As of the date of issuance to be implemented within 
9 months.
    Amendment No.: 186.
    Facility Operating License No. DPR-26: Amendment revised the 
Technical Specifications and the Facility Operating License.
    Date of initial notice in Federal Register: August 17, 1994 (59 FR 
42335) The February 6, 1995, submittal provided clarifying information 
and did not expand the scope of the original application, and 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 March 26, 1996.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: White Plains Public Library, 
100 Martine Avenue, White Plains, New York 10610.

Consumers Power Company, Docket No. 50-255, Palisades Plant, Van Buren 
County, Michigan

    Date of application for amendment: October 17, 1995.
    Brief description of amendment: This amendment revises the 
Palisades Facility Operating License to reference 10 CFR Part 40, allow 
the use of source materials as reactor fuel, delete references to 
specific amendments and specific revisions in the listed titles of the 
Physical Security Plan, Suitability Training and Qualification Plan, 
and the Safeguards Contingency Plan and make minor editorial changes to 
the license. In addition, the Technical Specifications (TS) are 
modified as follows: (1) TS 3.1.2 is modified to change the pressurizer 
cooldown limit from 100 deg.F to 200 deg.F/hour; (2) the shield cooling 
system requirements are relocated to the Final Safety Analysis Report; 
(3) several minor editorial changes and corrections are made, including 
corrections requested in the licensee's letter of March 24, 1995; and 
(4) several TS bases pages have been revised. The portion of the 
amendment request deleting license paragraph 2.F on reporting 
requirements was denied.
    Date of issuance: April 5, 1996.
    Effective date: April 5, 1996.
    Amendment No.: 171.
    Facility Operating License No. DPR-20: Amendment revised the 
Facility Operating License and the Technical Specifications.
    Date of initial notice in Federal Register: November 27, 1995 (60 
FR 58399).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated April 5, 1996, and an Environmental 
Assessment dated March 11, 1996 (61 FR 10811).

[[Page 18179]]

    No significant hazards consideration comments received: No.
    Local Public Document Room location: Van Wylen Library, Hope 
College, Holland, Michigan 49423.

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

    Date of application for amendment: December 7, 1995, as 
supplemented January 4, March 1, March 5, March 7, March 11, March 27, 
and March 29, 1996.
    Brief description of amendment: The amendment revises Technical 
Specifications 3/4.4.5 and 3/4.4.6.2 and their Bases to maintain 
voltage-based steam generator tube repair criteria for the tube support 
plate elevations for future cycles of operation. The amendment replaces 
a 1.0 volt repair limit which had been approved on an interim basis by 
License Amendment No. 184 (issued February 3, 1995) with a 2.0 volt 
repair limit. The amendment also includes additional changes to reflect 
the guidance provided in NRC Generic Letter 95-05, ``Voltage-Based 
Repair Criteria for Westinghouse Steam Generator Tubes Affected by 
Outside Diameter Stress Corrosion Cracking.''
    Date of issuance: April 1, 1996.
    Effective date: As of the date of issuance, to be implemented 
within 60 days.
    Amendment No: 198.
    Facility Operating License No. DPR-66: Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: January 3, 1996 (61 FR 
178) The January 4, March 1, March 5, March 7, March 11, March 27, and 
March 29, 1996, letters provided clarifying information that did not 
change the initial proposed no significant hazards consideration 
determination or expand the amendment request beyond the scope of the 
January 3, 1996 notice.
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated April 1, 1996.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: B. F. Jones Memorial Library, 
663 Franklin Avenue, Aliquippa, PA 15001.

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: December 14, 1995.
    Brief description of amendment: The amendment consists of several 
changes to the instrumentation sections of the Clinton Power Station 
Technical Specifications. These changes were required due to 
engineering reanalyses or plant modifications. The affected 
instrumentation includes: (1) steam line flow high channels for the 
reactor core isolation cooling (RCIC) system, (2) ambient temperature 
channels in the residual heat removal (RHR) system heat exchanger 
rooms, (3) reactor vessel pressure channels that provide a permissive 
for operation of the shutdown cooling mode of the RHR system, and (4) 
RCIC storage tank water level instrument channels.
    Date of issuance: April 10, 1996.
    Effective date: April 10, 1996.
    Amendment No.: 104.
    Facility Operating License No. NPF-62: The amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: January 22, 1996 (61 FR 
1631) The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated April 10, 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.
    No significant hazards consideration comments received: No.

Northern States Power Company, Docket No. 50-263, Monticello Nuclear 
Generating Plant, Wright County, Minnesota

    Date of application for amendment: August 15, 1995, as supplemented 
November 14, and December 20, 1995.
    Brief description of amendment: The amendment modifies the 
Monticello Technical Specifications (TS) to: (1) revise the main steam 
line isolation valve leak rate test acceptance criterion to be based 
upon the combined maximum flow path leakage for all four main steam 
lines of 46 standard cubic feet per hour (scfh) in lieu of the current 
limit of 11.5 scfh per valve; (2) revise the operability test interval 
for the drywell spray header and nozzles from 5 years to 10 years; and 
(3) revise TS 3/4.7.a.2, Primary Containment Integrity, to remove 
information specific to the primary containment leakage rate testing 
program and adopt the requirements of 10 CFR Part 50, Appendix J, 
Option B, for Type A testing, while remaining under Appendix J, Option 
A, for Type B and C testing.
    Date of issuance: April 3, 1996.
    Effective date: April 3, 1996.
    Amendment No.: 95.
    Facility Operating License No. DPR-22: Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: January 22, 1996 (61 FR 
1632) The December 20, 1995, letter provided clarifying information 
that was within the scope of the initial notice and did not change the 
staff's initial proposed no significant hazards considerations 
determination. The Commission's related evaluation of the amendment is 
contained in a Safety Evaluation dated April 3, 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.

Northern States Power Company, Docket No. 50-263, Monticello Nuclear 
Generating Plant, Wright County, Minnesota

    Date of application for amendment: March 1, 1996 (supersedes 
December 11, 1995, application).
    Brief description of amendment: The amendment modifies Technical 
Specification Section 4.7, Surveillance Requirements for Primary 
Containment Automatic Isolation Valves, by revising Surveillance 
Requirement 4.7.D.4 to require that the seat seals of the drywell and 
suppression chamber purge and vent valves be replaced every six 
operating cycles.
    Date of issuance: April 9, 1996.
    Effective date: April 9, 1996.
    Amendment No.: 96.
    Facility Operating License No. DPR-22: Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: March 8, 1996 (61 FR 
9504). The Commission's related evaluation of the amendment is 
contained in a Safety Evaluation dated April 9, 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.

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

    Date of application for amendments: December 22, 1995.
    Brief description of amendments: The amendments change Technical 
Specification 3.6.1.8, ``Drywell and Suppression Chamber Purge 
System,'' increasing the drywell and suppression

[[Page 18180]]

chamber purge system operating time limit from 90 hours each 365 days 
to 180 hours each 365 days.
    Date of issuance: March 29, 1996.
    Effective date: As of date of issuance, to be implemented within 30 
days.
    Amendment Nos.: 115 and 77.
    Facility Operating License Nos. NPF-39 and NPF-85. The amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: February 28, 1996 (61 
FR 7555).
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated March 29, 1996.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: Pottstown Public Library, 500 
High Street, Pottstown, Pennsylvania 19464.

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

    Date of application for amendment: February 9, 1996, as 
supplemented March 20, 1996.
    Brief description of amendment: The proposed amendment would revise 
the Technical Specifications (TSs) to use an installed retractable 
overhead door assembly and change TS 3.9.3 to satisfy closure 
requirements for the containment equipment hatch during core 
alterations or fuel movement in the containment building. The 
retractable door is to be used as a functionally equivalent closure 
plate currently required by TS 3.9.3.
    Date of issuance: April 1, 1996.
    Effective date: April 1, 1996.
    Amendment No.: 62.
    Facility Operating License No. DPR-18: Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: February 28, 1996 (61 
FR 7557). The Commission's related evaluation of the amendment is 
contained in a Safety Evaluation dated April 1, 1996.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: Rochester Public Library, 115 
South Avenue, Rochester, New York 14610.

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: August 18, 1995, as supplemented 
on November 1, 1995, February 14, March 14 (there are two supplemental 
letters with this date), and March 25, 1996.
    Brief description of amendment: The amendment revises the Operating 
License (OL) to increase the authorized core power level from 2775 
Megawatts thermal (MWt) to 2900 MWt. The amendment also approves 
changes to the technical specifications (TS) to implement uprated power 
operation.
    Date of issuance: April 12, 1996.
    Effective date: April 12, 1996.
    Amendment No.: 133.
    Facility Operating License No. NPF-12: Amendment revises the OL and 
TS.
    Date of initial notice in Federal Register: December 6, 1995 (60 FR 
62495). The original Federal Register notice included information from 
the licensee's November 1, 1995 supplemental letter. The February 14, 
March 14, and March 25, 1996 supplemental letters provided 
clarification and amplification of the analysis in the November 1, 1995 
letter and were not outside the scope of the initial Federal Register 
notice. The Commission's related evaluation of the amendment is 
contained in an Environmental Assessment dated April 12, 1996 and in a 
Safety Evaluation dated April 12, 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 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 30, 1992, as 
supplemented by letters dated September 7, 1993, August 17, 1994, and 
March 7, 1996.
    Brief description of amendments: These amendments add a new 
technical specification (TS) 3/4.7.3.1, ``Component Cooling Water (CCW) 
Safety Related Makeup System,'' and its associated Bases. The new TS 
will ensure that sufficient CCW capacity is available for continued 
operation of safety-related equipment during normal conditions and 
design-basis events.
    Date of issuance: April 11, 1996.
    Effective date: April 11, 1996.
    Amendment Nos.: Unit 2-129; Unit 3-118.
    Facility Operating License Nos. NPF-10 and NPF-15: The amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: March 3, 1993 (58 FR 
12268). The September 7, 1993, August 17, 1994, and March 7, 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 April 11, 1996.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: Main Library, University of 
California, P. O. Box 19557, Irvine, California 92713.

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

    Date of application for amendment: December 12, 1995, and 
supplemented March 4, 1996 (TS 95-23).
    Brief description of amendment: The amendment revises the TS 
surveillance requirements and bases to incorporate alternate S/G tube 
plugging criteria at tube support plate (TSP) intersections. The 
approach taken is based on guidance given in Generic Letter (GL) 95-05, 
``Voltage-Based Repair Criteria for Westinghouse Steam Generator Tubes 
Affected by Outside Diameter Stress Corrosion Cracking.'' The amendment 
is applicable for Cycle 8 operation only.
    Date of issuance: April 3, 1996.
    Effective date: April 3, 1996.
    Amendment No.: 211.
    Facility Operating License Nos. DPR-77: Amendment revises the 
technical specifications.
    Date of initial notice in Federal Register: January 3, 1996 (61 FR 
183) The March 6, 1996 supplemental letter provided clarifying 
information which did not change the proposed no significant hazards 
consideration.
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated April 3, 1996.
    No significant hazards consideration comments received: None
    Local Public Document Room location: Chattanooga-Hamilton County 
Library, 1101 Broad Street, Chattanooga, Tennessee 37402

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

    Date of application for amendment: February 27, 1996, as 
supplemented by letter dated March 1, 1996.
    Brief description of amendment: The amendment allows the drywell 
personnel air lock shield doors to be open during Operational 
Conditions 1, 2, and 3 until the end of Operating Cycle 6.

[[Page 18181]]

    Date of issuance: March 22, 1996.
    Effective date: March 22, 1996.
    Amendment No.: 84.
    Facility Operating License No. NPF-58: This amendment approved a 
change to the design basis as described in the Updated Safety Analysis 
Report. Public comments requested as to proposed no significant hazards 
consideration: Yes (61 FR 8982 dated March 8, 1996). 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 for an opportunity to request a 
hearing BiWeekly Notice by March 18, 1996, corrected to April 5, 1996 
(61 FR 10600 dated March 14, 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 March 1, 1996, supplemental letter 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 and final no 
significant hazards consideration determination is contained in a 
Safety Evaluation dated March 22, 1996.
    Local Public Document Room location: Perry Public Library, 3753 
Main Street, Perry, Ohio 44081.

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 requests: November 21, 1995 (TXX-95288) as 
supplemented by letters dated December 15, 1995 (TXX-95306), and 
February 2, 1996 (TXX-96040).
    Brief description of amendments: The amendments revised the core 
safety limit curves and revised N-16 Overtemperature reactor trip 
setpoints as a result of the reload analyses for CPSES Unit 2, Cycle 3. 
In addition, the minimum required Reactor Coolant System (RCS) flow was 
increased and an administrative enhancement was included in the 
footnotes of the RCS flow-low reactor trip function setpoint for both 
Units 1 and 2.
    Date of issuance: April 1, 1996.
    Effective date: April 1, 1996.
    Amendment Nos.: Unit 1-49; Unit 2-35.
    Facility Operating License Nos. NPF-87 and NPF-89. The amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: January 3, 1996 (61 FR 
185) The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated April 1, 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, Texas 76019

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

    Date of application for amendments: July 26, 1995.
    Brief description of amendments: The amendments revise the 
Technical Specifications to increase the pressurizer safety valve lift 
setpoint tolerance and reduce the pressurizer high pressure reactor 
trip setpoint and allowable value.
    Date of issuance: April 1, 1996.
    Effective date: April 1, 1996.
    Amendment Nos.: 200 and 181.
    Facility Operating License Nos. NPF-4 and NPF-7: Amendments revised 
the Technical Specifications.
    Date of initial notice in Federal Register: August 30, 1995 (60 FR 
45189) The Commission's related evaluation of the amendments is 
contained in a Safety Evaluation dated April 1, 1996.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: The Alderman Library, Special 
Collections Department, University of Virginia, Charlottesville, 
Virginia 22903-2498.

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

    Date of amendment request: March 8, 1996, as supplemented by letter 
dated March 26, 1996.
    Brief description of amendment: This amendment reduces the 
calculated thermal design flow of the reactor coolant system and 
increases the trip setpoint of the low pressurizer pressure.
    Date of issuance: April 4, 1996.
    Effective date: April 4, 1996.
    Amendment No.: 99.
    Facility Operating License No. NPF-42: The amendment revised the 
Technical Specifications.
    Public comments requested as to proposed no significant hazards 
consideration: Yes (61 FR 10389 dated March 13, 1996). The 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 for an opportunity to request a 
hearing by April 12, 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 is contained 
in a Safety Evaluation dated April 4, 1996.
    Local Public Document Room locations: Emporia State University, 
William Allen White Library, 1200 Commercial Street, Emporia, Kansas 
66801 and Washburn University School of Law Library, Topeka, Kansas 
66621.

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.

[[Page 18182]]

    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 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 May 24, 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, 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

[[Page 18183]]

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

Arizona Public Service Company, et al., Docket No. STN 50-529, Palo 
Verde Nuclear Generating Station, Unit 2, Maricopa County, Arizona

    Date of application for amendment: April 1, 1996, as supplemented 
by letter dated April 3, 1996.
    Brief description of amendment: The amendment modifies Technical 
Specification (TS) 3/4.9.6 to temporarily allow the use of a hoist 
instead of the refueling machine for the movement of the fuel assembly 
at core location A-07.
    Date of issuance: April 3, 1996.
    Effective date: April 3, 1996.
    Amendment No.: Unit 2--96.
    Facility Operating License No. NPF-51: The amendment revised the 
Technical Specifications.
    Public comments requested as to proposed no significant hazards 
consideration: No.
    The Commission's related evaluation of the amendment, finding of 
emergency circumstances, and final determination of no significant 
hazards consideration are contained in a Safety Evaluation dated April 
3, 1996.
    Local Public Document Room location: Phoenix Public Library, 1221 
N. Central Avenue, Phoenix, Arizona 85004.
    Attorney for licensee: Nancy C. Loftin, Esq., Corporate Secretary 
and Counsel, Arizona Public Service Company, P.O. Box 53999, Mail 
Station 9068, Phoenix, Arizona 85072-3999.
    NRC Project Director: William H. Bateman.

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

    Date of application of amendments: April 2, 1996.
    Brief description of amendments: The amendments revise Technical 
Specification (TS) Section 4.5.4, ``Penetration Room Ventilation 
System'' and TS Section 4.14, ``Reactor Building Purge Filters and 
Spent Fuel Pool Ventilation System.'' The change updates the industry 
guidance reference for testing charcoal absorber units for the system 
covered by those TS.
    Date of Issuance: April 2, 1996.
    Effective date: April 2, 1996, to be implemented within 30 days.
    Amendment Nos.: 215, 215, and 212.
    Facility Operating License Nos. DPR-38, DPR-47, and DPR-55: The 
amendments revised the Technical Specifications.
    Public comments requested as to proposed no significant hazards 
consideration: No.
    The Commission's related evaluation of the amendments, finding of 
emergency circumstances, and final determination of no significant 
hazards consideration are contained in a Safety Evaluation dated April 
2, 1996.
    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.

    Dated at Rockville, Maryland, this 17th day of April 1996.

    For the Nuclear Regulatory Commission.
Steven A. Varga,
Director, Division of Reactor Projects--I/II, Office of Nuclear Reactor 
Regulation.
[FR Doc. 96-9925 Filed 4-23-96; 8:45 am]
BILLING CODE 7590-01-P