[Federal Register Volume 61, Number 50 (Wednesday, March 13, 1996)]
[Notices]
[Pages 10391-10406]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: X96-10313]



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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 February 16, 1996, through March 1, 1996. 
The last biweekly notice was published on February 28, 1996 (61 FR 
7542).

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 April 12, 1996, the licensee may file a request for a hearing 
with respect to issuance of the amendment to the subject facility 
operating license and any person whose interest may be affected by this 
proceeding and who wishes to participate as a party in the proceeding 
must file a written request for a hearing and a petition for leave to 
intervene. Requests for a hearing and a petition for leave to intervene 
shall be filed in accordance with the Commission's ``Rules of Practice 
for Domestic Licensing Proceedings'' in 10 CFR Part 2. Interested 
persons should consult a current copy of 10 CFR 2.714 which is 
available at the Commission's Public Document Room, the Gelman 
Building, 2120 L Street, NW., Washington, DC and at the local public 
document room for the particular facility involved. If a request for a 
hearing or petition for leave to intervene is filed by the above date, 
the Commission or an Atomic Safety and Licensing Board, designated by 
the Commission or by the Chairman of the Atomic Safety and Licensing 
Board Panel, will rule on the request and/or petition; and the 
Secretary or the designated Atomic Safety and Licensing Board will 
issue a notice of a hearing or an appropriate order.
    As required by 10 CFR 2.714, a petition for leave to intervene 
shall set forth with particularity the interest of the petitioner in 
the proceeding, and how that interest may be affected by the results of 
the proceeding. The petition should specifically explain the reasons 
why intervention should be permitted with particular reference to the 
following factors: (1) the nature of the petitioner's right under the 
Act to be made a party to the proceeding; (2) the nature and extent of 
the petitioner's property, financial, or other interest in the 
proceeding; and (3) the possible effect of any order which may be 
entered in the proceeding on the petitioner's interest. The petition 
should also identify the specific aspect(s) of the subject matter of 
the proceeding as to which petitioner wishes to intervene. Any person 
who has filed a petition for leave to intervene or who has been 
admitted as a party may amend the petition without requesting leave of 
the Board up to 15 days prior to the first prehearing conference 
scheduled in the proceeding, but such an amended petition must satisfy 
the specificity requirements described above.
    Not later than 15 days prior to the first prehearing conference 
scheduled in the proceeding, a petitioner shall file a supplement to 
the petition to intervene which must include a list of the contentions 
which are sought to be litigated in the matter. Each contention must 
consist of a specific statement of the issue of law or fact to be 
raised or controverted. In addition, the petitioner shall provide a 
brief explanation of the bases of the contention and a concise 
statement of the alleged facts or expert opinion which support the 
contention and on which the petitioner intends to rely in proving the 
contention at the hearing. The petitioner must also provide references 
to those specific sources and documents of which the petitioner is 
aware and on which the petitioner intends to rely to establish those 
facts or expert opinion. Petitioner must provide sufficient information 
to show that a genuine dispute exists with the applicant on a material 
issue of law or fact. Contentions shall be limited to matters within 
the scope of the amendment under consideration. The contention must be 
one which, if proven, would entitle the petitioner to relief. A 
petitioner who fails to file such a supplement which satisfies these 
requirements with respect to at least one

[[Page 10393]]

contention will not be permitted to participate as a party.
    Those permitted to intervene become parties to the proceeding, 
subject to any limitations in the order granting leave to intervene, 
and have the opportunity to participate fully in the conduct of the 
hearing, including the opportunity to present evidence and cross-
examine witnesses.
    If a hearing is requested, the Commission will make a final 
determination on the issue of no significant hazards consideration. The 
final determination will serve to decide when the hearing is held.
    If the final determination is that the amendment request involves 
no significant hazards consideration, the Commission may issue the 
amendment and make it immediately effective, notwithstanding the 
request for a hearing. Any hearing held would take place after issuance 
of the amendment.
    If the final determination is that the amendment request involves a 
significant hazards consideration, any hearing held would take place 
before the issuance of any amendment.
    A request for a hearing or a petition for leave to intervene must 
be filed with the Secretary of the Commission, U.S. Nuclear Regulatory 
Commission, Washington, DC 20555, Attention: Docketing and Services 
Branch, or may be delivered to the Commission's Public Document Room, 
the Gelman Building, 2120 L Street, NW., Washington DC, by the above 
date. Where petitions are filed during the last 10 days of the notice 
period, it is requested that the petitioner promptly so inform the 
Commission by a toll-free telephone call to Western Union at 1-(800) 
248-5100 (in Missouri 1-(800) 342-6700). The Western Union operator 
should be given Datagram Identification Number N1023 and the following 
message addressed to (Project Director): petitioner's name and 
telephone number, date petition was mailed, plant name, and publication 
date and page number of this Federal Register notice. A copy of the 
petition 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.

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 amendment request: September 16, 1994, as supplemented on 
January 31, 1996.
    Description of amendment request: The proposed amendment would 
revise the technical specifications to eliminate periodic response time 
testing requirements for selected pressure and differential pressure 
sensors in the reactor trip system and engineered safety features 
actuation instrumentation channels.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 50.91(a), the licensee has provided 
its analysis of the issue of no significant hazards consideration, 
which is presented below:
    1. The proposed change does not involve a significant increase 
in the probability or consequences of an accident previously 
evaluated.
    This change to the Technical Specifications does not result in a 
condition where the design, material, and construction standards 
that were applicable prior to the change are altered. The same RTS 
and ESFAS instrumentation is being used; the time response 
allocations/modeling assumptions in the Updated Final Safety 
Analysis Report (UFSAR), Chapter 15, Accident Analyses, are still 
the same; only the method of verifying time response is changed. The 
proposed change will not modify any system interface and could not 
increase the likelihood of an accident since these events are 
independent of this change. The proposed activity will not change, 
degrade or prevent actions or alter any assumptions previously made 
in evaluating the radiological consequences of an accident described 
in the UFSAR. Therefore, the proposed amendment does not result in 
any 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.
    This change does not alter the performance of the identified 
pressure and differential pressure transmitters and switches used in 
the plant protection systems. All sensors will still have response 
time verified by test before placing the sensor in operational 
service, and after any maintenance that could affect response time. 
Changing the method of periodically verifying instrument response 
for these sensors (assuring equipment operability) from time 
response testing to calibration and channel checks does not result 
in any design, installation, or operational changes and thus will 
not create any new accident initiators or scenarios. Periodic 
surveillance of these instruments will detect significant 
degradation in the sensor response characteristics. Implementation 
of the proposed amendment does 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.
    This change does not affect the total system response time 
assumed in the safety analyses. The periodic system response time 
verification method for the identified pressure and differential 
pressure sensors and switches is modified to allow use of (1) 
historical records based on acceptable response time tests 
(hydraulic, noise, or power interrupt tests), (2) inplace, onsite or 
offsite (e.g. vendor) test measurements, or (3) using vendor 
engineering specifications.
    The method of verification still provides assurance that the 
total system response is within that defined in the safety analyses, 
since calibration tests will detect any degradation which might 
significantly affect sensor response time. Based on the above, it is 
concluded that the proposed license amendment request does not 
result in a reduction in margin with respect to plant safety.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
requested amendments involve no significant hazards consideration.
    Local Public Document Room location: 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.
    Attorney for licensee: Michael I. Miller, Esquire; Sidley and 
Austin, One First National Plaza, Chicago, Illinois 60603
    NRC Project Director: Robert A. Capra

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

    Date of amendment request: February 12, 1996
    Description of amendment request: The proposed amendment would 
revise Technical Specification (TS) 4.6.2.2.d to delete the reference 
to the specific test acceptance criteria for the Containment 
Recirculation Spray Pumps and replace the specific test acceptance 
criteria with

[[Page 10394]]
reference to the requirements of the Inservice Testing (IST) Program. 
In addition, the 18-month test frequency would be replaced with the 
test frequency requirements specified in the IST Program. The proposed 
amendment would make this TS the same as Beaver Valley Power Station, 
Unit No. 2 TS 4.6.2.2.d which was revised by License Amendment No. 68 
on May 3, 1995.
    The proposed amendment would also revise the Bases of TS 4.6.2.2.d 
for both Unit Nos. 1 and 2 to describe the proposed revision to TS 
4.6.2.2.d.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 50.91(a), the licensee has provided 
its analysis of the issue of no significant hazards consideration, 
which is presented below:
    1. Does the change involve a significant increase in the 
probability or consequences of an accident previously evaluated?
    The change does not result in a modification to plant equipment 
nor does if affect the manner in which the plant is operated. The 
Recirculation Spray System (RSS) pumps are normally in a standby 
condition and only operate during accident mitigation. Since the 
physical plant equipment and operating practices are not changed, as 
noted above, there is no change in the probability of an accident 
previously evaluated.
    The proposed change, for Beaver Valley Power Station (BVPS) Unit 
No. 1 only, will not lower the pump performance operability criteria 
for the RSS pumps. The required values for developed pump head and 
flow will continue to satisfy accident mitigation requirements and 
will be maintained and controlled in the BVPS Unit No. 1 Inservice 
Testing (IST) Program.
    Since the proposed change does not lower the RSS pump 
performance acceptance criteria, the containment depressurization 
system will continue to meet its design basis requirements. The 
proposed change will not impose additional challenges to the 
containment structure in terms of peak pressure. The calculated 
offsite does consequences of a design basis accident (DBA) will 
remain unchanged since the one hour release duration remains 
unchanged. Future changes to the RSS pump head and flow requirements 
will be made under the 10 CFR 50.59 process to ensure that the 
containment performance requirements continue to be met.
    The proposed change in the RSS pump surveillance interval from 
18 months to every refueling, will not affect the ability of the 
pumps to perform as assumed in the Safety Analyses. The proposed 
change to the Bases section, for BVPS Unit Nos. 1 and 2, will ensure 
that safety analyses assumptions for assumed pump performance 
continue to be met. The words ``required developed head'' will be 
clearly defined to reflect that they refer to the value assumed in 
the safety analysis for the recirculation spray pump's developed 
head at a specific point. The proposed changes to the Index pages 
are administrative in nature and do not affect plant safety. 
Therefore, the proposed change does not involve a significant 
increase in the consequences of an accident previously evaluated.
    Based on the above discussion, it is concluded that this change 
does not involve a significant increase in the probability or 
consequences of an accident previously evaluated.
    2. Does the change create the possibility of a new or different 
kind of accident from any accident previously evaluated?
    The proposed change does not alter the method of operating the 
plant. The recirculation spray system is an accident mitigation 
system and is normally in standby. System operation would be 
initiated following a containment pressure increase resulting from a 
DBA. The RSS pumps will continue to provide sufficient flow to 
mitigate the consequences of a DBA. RSS operation continues to 
fulfill the safety function for which it was designed and no changes 
to plant equipment will occur. As a result, an accident which is new 
or different than any already evaluated in the Updated Final Safety 
Analysis Report will not be created due to this change.
    Therefore, the proposed change does not create the possibility 
of a new or different kind of accident from any accident previously 
evaluated.
    3. Does the change involve a significant reduction in a margin 
of safety?
    The surveillance requirements for demonstrating that the RSS 
pumps are operable will continue to assure the ability of the system 
to satisfy its design function. Therefore, the proposed change will 
not affect the ability of the RSS to perform its safety function.
    The containment spray system design requirement to restore the 
containment to subatmospheric condition within one hour will 
continue to be satisfied. This proposed change does not have any 
affect on the containment peak pressure since the containment peak 
pressure occurs prior to the initiation of any of the two 
containment spray systems. There is no resultant change in dose 
consequences since the containment will continue to reach a 
subatmospheric pressure within the first hour following a DBA.
    The RSS pumps' performance requirements will continue to be 
controlled in a manner to ensure safety analysis assumptions are 
met.
    Therefore, based on the above discussion, it can be concluded 
that the proposed change does not involve a significant reduction in 
a margin of safety.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: B. F. Jones Memorial Library, 
663 Franklin Avenue, Aliquippa, PA 15001
    Attorney for licensee: Jay E. Silberg, Esquire, Shaw, Pittman, 
Potts & Trowbridge, 2300 N Street, NW., Washington, DC 20037
    NRC Project Director: John F. Stolz

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

    Date of amendment request: November 30, 1995
    Description of amendment request: The proposed amendment would 
implement the Option I-D long-term stability solution and remove the 
existing SIL-380 Rev. 1-based specifications. In addition, the proposed 
change would require a plant scram be initiated should the plant enter 
natural circulation conditions and would prohibit restarting a 
recirculation pump while in natural circulation. The proposed change 
would define natural circulation. Finally, this change would delete 
Technical Specification (TS) actions and surveillance requirements 
related to core plate differential pressure noise while in single 
recirculation pump operation (SLO).
    Basis for proposed no significant hazards consideration 
determination:
    As required by 10 50.91(a), the licensee has provided its analysis 
of the issue of no significant hazards consideration, which is 
presented below:
    1) The proposed license amendment does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated. The implementation of the [boiling water 
reactor] BWR Owner's Group long term solution Option I-D does not 
modify the assumptions in the existing accident analysis. The use of 
an exclusion region and the operator actions required to avoid and 
minimize operation inside the region do not increase the possibility 
of an accident. Licensing Topical Report, 'Evaluation of the 
``Regional Exclusion with Flow-Biased APRM [average power range 
monitor] Neutron Flux Scram'' Stability Solution', GENE-A000-04021-
01 (attachment 1) demonstrates that the APRM flow-biased scram 
function provides a high degree of assurance that the fuel safety 
limit will not be exceeded should power oscillations occur during 
plant operation within the restricted region. Regional mode core 
oscillations are not predicted to occur at the [Duane Arnold Energy 
Center] DAEC because of its small core size and tight core inlet 
orifices. Conditions for operation outside of the exclusion region 
are within the assumptions of the existing accident analysis. The 
operator action requirement to exit the exclusion region upon entry 
minimizes the probability of an instability event occurring. 
Inserting control rods or increasing recirculation flow, the 
evolutions to be used to exit the region, are normal plant 
maneuvers.
    The proposed clarifications to explicitly direct the operator to 
initiate a reactor scram

[[Page 10395]]
in the event of operation in natural circulation are conservative 
and consistent with current plant operating practices. Likewise, the 
proposed prohibition from starting a recirculation pump as a means 
of exiting the natural circulation mode of operation is also 
conservative. Therefore, the proposed license amendment does not 
involve a significant increase in the probability or consequences of 
an accident previously evaluated.
    The core plate differential pressure noise surveillances that 
are performed while in single recirculation pump operation were 
included in TS Amendment <greek-i>119 due to NRC concerns at the 
time that high core plate noise observed during [single-loop 
operation] SLO at Brown's Ferry in 1985 could be an indication of 
thermal hydraulic instability. [General Electric] GE has since 
determined that core plate differential pressure noise is not a 
cause of thermal hydraulic instability and that the noise does not 
pose a safety concern. Therefore, the proposed license amendment 
does not involve a significant increase in the probability or 
consequences of an accident previously evaluated.
    2) The proposed license amendment does not create the 
possibility of a new or different kind of accident from any 
previously evaluated. As stated above, the proposed changes either 
mandate operation within the envelope of previously analyzed plant 
operating conditions or direct the operator to immediately return 
the plant to within these analyzed conditions using normal plant 
maneuvers. In addition, analysis has demonstrated that the APRM 
flow-biased scram function provides a high degree of assurance that 
the fuel safety limit will not be exceeded should power oscillations 
occur during plant operation within the restricted region. 
Therefore, the potential for a new or different type of accident 
from those previously evaluated is not created.
    The proposed clarifications to explicitly direct the operator to 
initiate a reactor scram in the event of operation in natural 
circulation are conservative and consistent with current plant 
operating practices. Likewise, the proposed prohibition from 
starting a recirculation pump as a means of exiting the natural 
circulation mode of operation is also conservative. Therefore, the 
potential for a new or different type of accident from those 
previously evaluated is not created.
    The core plate differential pressure noise surveillances that 
are performed while in single recirculation pump operation were 
included in TS Amendment <greek-i>119 due to NRC concerns at the 
time that high core plate noise observed during SLO at Brown's Ferry 
in 1985 could be an indication of thermal hydraulic instability. GE 
has since determined that core plate differential pressure noise is 
not a cause of thermal hydraulic instability and that the noise does 
not pose a safety concern. Therefore, the potential for a new or 
different type of accident from those previously evaluated is not 
created.
    3) The proposed amendment will not reduce the margin of safety. 
The combination of the proposed requirements to avoid possible 
unstable conditions and the automatic flow biased high reactor flux 
scram provide defense in depth to provide fuel protection. Therefore 
the individual or combination of means to detect and suppress 
thermal hydraulic instability supplements the margin of safety.
    The proposed specification related to initiating a reactor scram 
while in natural circulation is conservative. Likewise, the proposed 
prohibition from starting a recirculation pump as a means of exiting 
the natural circulation mode of operation is also conservative and 
therefore does not constitute a reduction in the margin of safety.
    The core plate differential pressure noise surveillances that 
are performed while in single recirculation pump operation were 
included in TS Amendment <greek-i>119 due to NRC concerns at the 
time that high core plate noise observed during SLO at Brown's Ferry 
in 1985 could be an indication of thermal hydraulic instability. GE 
has since determined that core plate differential pressure noise is 
not a cause of thermal hydraulic instability and that the noise does 
not pose a safety concern. Therefore, the elimination of these 
surveillance tests does not constitute a reduction in the margin of 
safety.
    The NRC staff has reviewed the licensee's analysis and, based on 
thisreview, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: Cedar Rapids Public Library, 
500 First Street, S.E., Cedar Rapids, Iowa 52401
    Attorney for licensee: Jack Newman, Kathleen H. Shea, Morgan, 
Lewis, & Bockius, 1800 M Street, NW., Washington, DC 20036-5869
    NRC Project Director: Gail H. Marcus

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

    Date of amendment request: November 16, 1995
    Description of amendment request: The proposed amendment would 
revise the technical specifications (TS) to add a Limiting Condition 
for Operation and surveillance test for safety related inverters and 
deletes requirements for non-safety related instrument buses.
    Basis for proposed no significant hazards consideration 
determination:
    As required by 10 50.91(a), the licensee has provided its analysis 
of the issue of no significant hazards consideration, which is 
presented below:
    1. The proposed change does not involve a significant increase 
in the probability or consequences of an accident previously 
evaluated.
    The proposed changes will delete requirements from the Technical 
Specifications (TS) for non-safety related 120 Volt a-c instrument 
panels AI-42A and AI-42B, and incorporate new requirements for the 
safety-related 125 Volt d-c to 120 Volt a-c inverters (A, B, C, and 
D) similar to the Standard Technical Specification for Combustion 
Engineering plants as contained in NUREG-1432.
    TS 2.7 requires that 120 Volt instrument panels AI-42A and AI-
42B be operable whenever the reactor coolant temperature is above 
300 F. Either of these instrument panels may be inoperable for up 
to 8 hours or a plant shutdown is required. These instrument panels 
are non-safety related and do not receive or actuate any Engineered 
Safeguards Features (ESF) or Reactor Protection System (RPS) and the 
panels are not required for, nor do they indicate the status of, 
containment integrity. The FCS plant specific Probabilistic Risk 
Assessment (PRA) model was reviewed to determine the effect of 
unavailability of these instrument panels on the core damage 
frequency. The results of the review show that the unavailability of 
these panels is not a contributor to risk. Therefore these 
instrument panels do not meet any of the four criteria contained in 
10 CFR 50.36 for inclusion into TS. The operation of these panels 
are controlled by plant procedures that are governed by 10 CFR 
50.59.
    Therefore, deletion of the requirements for AI-42A and AI-42B 
from the TS would not significantly increase the probability or 
consequences of an accident previously evaluated.
    It is also proposed to incorporate new requirements for the 
safety-related 125 Volt d-c to 120 Volt a-c inverters (A, B, C, and 
D). Currently, there are no TS requirements for inoperability of the 
safety-related inverters. However, if an inverter is inoperable and 
its associated 120 Volt a-c instrument bus is powered by its safety-
related bypass transformer, the a-c instrument bus is considered 
inoperable and an 8 hour Limiting Condition for Operation is 
applied. The bus is declared inoperable even though it is being 
powered from a safety related power source because this source is 
not an uninterruptible power supply. Operating experience has shown 
that, in many instances, 8 hours is insufficient time to 
troubleshoot and conduct repairs on an inverter. FCS initiated a TS 
required plant shutdown in November 1994, and again in January 1995, 
due to inoperable inverters that could not be repaired in the 8 
hours allowed by TS. If FCS had 24 hours to conduct repairs, a power 
reduction, and the potential to challenge plant systems, would not 
have been necessary.
    The proposed change does not increase the probability of an 
accident since loss of power to a vital bus is not an initiator of 
any analyzed accident. The proposed change does not increase the 
consequences of any accident since the TS currently allow one 120 V 
instrument bus to be inoperable and de-energized. The proposed 
change would only allow one 120 V instrument bus to be energized 
from a safety related bypass source. The proposed changes do not 
reduce the number of RPS or ESF actuation channels that are required 
to be operable. Should a

[[Page 10396]]
loss of offsite power event occur, power to the instrument bus would 
only be interrupted during the time required for the emergency 
diesel generator to start and load.
    The FCS plant specific PRA model was reviewed to determine the 
effect of unavailability of the 120 V instrument panels supplied by 
inserters A, B, C, and D on the core damage frequency. The results 
of the review show that the loss of one of the panels has an 
insignificant effect on the PRA model. Therefore, the proposed 
change of allowing a 24 hour period with one instrument panel 
powered from a interruptible power supply has a insignificant effect 
on the PRA results.
    Therefore, the proposed change to include specific operability 
requirements for safety related inverters does not significantly 
increase 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.
    There will be no physical alterations to the plant 
configuration, changes to setpoint values, or changes to the 
implementation of setpoints or limits as a result of these proposed 
changes. The proposed changes do not reduce the number of RPS or ESF 
actuation channels that are required to be operable. Therefore, the 
proposed changes do not create the possibility of a new or different 
kind of accident from any previously evaluated.
    3. The proposed change does not involve a significant reduction 
in a margin of safety.
    The proposed changes delete TS requirements for nonsafety 
related instrument panels and incorporate additional operability 
requirements for safety related inverters. The proposed changes do 
not revise any setpoints or limits monitored by the instrument 
panels or buses. In addition, a review of the FCS plant specific PRA 
shows that these proposed changes are insignificant to core damage 
frequency. 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: W. Dale Clark Library, 215 
South 15th Street, Omaha, Nebraska 68102
    Attorney for licensee: Perry D. Robinson, Winston & Strawn, 1400 L 
Street, N.W., Washington, DC 20005-3502
    NRC Project Director: William H. Bateman
Omaha Public Power District, Docket No. 50-285, Fort Calhoun 
Station, Unit No. 1, Washington County, Nebraska
    Date of amendment request: February 1, 1996
    Description of amendment request: The proposed amendment would 
revise the Technical Specifications (TS) to allow an increase in the 
initial nominal enrichment limit of fuel assemblies to be stored in the 
spent fuel pool.
    Basis for proposed no significant hazards consideration 
determination:
    As required by 10 50.91(a), the licensee has provided its analysis 
of the issue of no significant hazards consideration, which is 
presented below:
    1. The proposed change does not involve a significant increase 
in the probability or consequences of an accident previously 
evaluated.
    The proposed change to the Technical Specifications to increase 
the enrichment limit for fuel assembly storage requirements does not 
involve a significant increase in the probability of an accident. 
The enrichment limit is not a precursor to any analyzed event and 
therefore cannot impact probability.
    The safety evaluation for the existing Spent Fuel Pool (SFP) 
storage racks was approved by the NRC in Amendment 155 (TAC M85116). 
This amendment approved the current limit on fuel enrichment, and 
the mechanical, structural, and thermal/hydraulic design of the fuel 
racks. This amendment also evaluated the radiological consequences 
of a fuel handling accident with fuel enrichments equivalent to the 
proposed change. The proposed change will not impact this previously 
approved evaluation with the exception of the nuclear criticality 
analysis. The nuclear criticality analysis supporting the proposed 
change used calculational methods conforming to NRC guidance, 
industry codes, standards, and specifications. In meeting the 
acceptance criteria for criticality in the SFP, such that k<INF>eff 
is always less than or equal to O.95 at a 95%/95% probability 
tolerance level, the proposed change from 4.2 weight percent (w/o) 
to 4.5 w/o Uranium-235 (U<SUP>235) does not involve an increase in 
the consequences of an accident previously evaluated.
    Therefore, it is concluded that the proposed change to increase 
the enrichment limit for fuel storage does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    2. The proposed change does not create the possibility of a new 
or different kind of accident from any accident previously 
evaluated.
    The proposed change was evaluated in accordance with the 
guidance of the NRC Position Paper entitled, ``OT Position for 
Review and Acceptance of Spent Fuel Storage and Handling 
Applications'', appropriate sections of the NRC Standard Review 
Plan, Regulatory Guides, industry codes, and standards. In addition, 
the NRC Safety Evaluation Report for Amendment 155 was also reviewed 
with respect to the proposed change.
    No new or different mode of operation is proposed. No unproven 
technology was utilized in the analytical techniques necessary to 
justify the planned fuel storage change. The analytical techniques 
used have been developed and used in over 15 applications previously 
approved by the NRC. Based upon the reviews, it is concluded that 
the proposed change does not create the possibility of a new or 
different type accident from any accident previously evaluated.
    3. The proposed change does not involve a significant reduction 
in a margin of safety.
    The only margin of safety potentially impacted by the proposed 
change is related to nuclear criticality considerations. The 
established acceptance criterion for criticality is that the neutron 
multiplication factor in spent fuel pools shall be less than or 
equal to 0.95, including all uncertainties, under all conditions. 
This margin of safety has been adhered to in the criticality 
analysis methods for the proposed change. Therefore the proposed 
change does not involve a significant reduction in a margin of 
safety.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: W. Dale Clark Library, 215 
South 15th Street, Omaha, Nebraska 68102
    Attorney for licensee: Perry D. Robinson, Winston & Strawn, 1400 L 
Street, N.W., Washington, DC 20005-3502
    NRC Project Director: William H. Bateman
PECO Energy Co., Public Service Electric and Gas Co., Delmarva 
Power and Light Co., and Atlantic City Electric Co., Dockets Nos. 
50-277 and 50-278, Peach Bottom Atomic Power Station, Units Nos. 2 
and 3, York County, Pennsylvania
    Date of application for amendments: December 21, 1995
    Description of amendment request: The proposed amendments would 
modify the Peach Bottom Atomic Power Station (PBAPS), Units 2 and 3 
Facility Operating Licenses (FOLs) to provide for elimination of 
outdated or superseded material regarding, among other things, 
environmental monitoring and modifications to the low pressure coolant 
injection system, and for making the FOL of Unit 2 consistent with the 
FOL of Unit 3.
    Basis for proposed no significant hazards consideration 
determination:
    As required by 10 50.91(a), the licensee has provided its analysis 
of the issue of no significant hazards consideration, which is 
presented below:
    The changes proposed in the Application do not constitute a 
Significant Hazards Consideration in that:
    i) The proposed changes do not involve a significant increase in 
the probability or

[[Page 10397]]

consequences of an accident previously evaluated because the changes 
are purely administrative and do not involve any physical changes to 
plant SSC [structures, systems, and components]. Therefore, these 
changes will not involve a significant increase in the probability 
or consequences of an accident previously evaluated.
    ii) The proposed changes do not create the possibility of a new 
or different kind of accident from any accident previously evaluated 
because the changes will not alter the plant or the manner in which 
the plant is operated. The changes do not allow plant operation in 
any mode that is not already evaluated in the safety analysis. The 
changes will not alter assumptions made in the safety analysis and 
licensing bases. Therefore, these changes will not create the 
possibility of a new or different kind of accident from any accident 
previously evaluated.
    iii) The proposed changes do not involve a significant reduction 
in a margin of safety because they are purely administrative and 
have no impact on any safety analysis assumptions.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: Government Publications 
Section, State Library of Pennsylvania, (REGIONAL DEPOSITORY) Education 
Building, Walnut Street and Commonwealth Avenue, Box 1601, Harrisburg, 
Pennsylvania 17105.
    Attorney for licensee: J. W. Durham, Sr., Esquire, Sr. V. P. and 
General Counsel, PECO Energy Company, 2301 Market Street, Philadelphia, 
Pennsylvania 19101
    NRC Project Director: John F. Stolz

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

    Date of amendment request: January 11, 1996
    Description of amendment request: The proposed amendment adds a new 
action statement to Section 3.8.3.1. of the Technical Specifications 
which precludes the need for entry into Limiting Condition for 
Operation (LCO) 3.0.3 to allow the performance of certain Emergency 
Diesel Generator testing.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 50.91(a), the licensee has provided 
its analysis of the issue of no significant hazards consideration, 
which is presented below:
    The proposed changes do not:
    I. Involve a significant increase in the probability or 
consequences of an accident previously evaluated.
    The proposed change to allow 8 hours to perform Emergency Diesel 
Generator testing and eliminate the need to enter LCO 3.0.3 to 
perform this testing does not increase the chances for a previously 
analyzed accident to occur. The 8 hour time limit before requiring a 
unit shutdown balances the benefit of performing the required test 
with the low probability of a LOCA/LOOP [loss-of-coolant accident/
loss of offsite power] while being in the degraded condition for the 
duration of the test. To ensure that this risk is minimized, a 
significant amount of precautions are taken prior to test 
initiation. The governing surveillance procedures have a very 
restrictive list of test prerequisites and limitations, which ensure 
the availability of remaining ac [alternating current] electrical 
power distribution systems and reduce the potential for any single 
failure. The allowance of 8 hours to complete the required test 
prior to initiating shutdown actions ensures operator attention is 
focused on minimizing the potential loss of power to the remaining 
division, and restoring power to the effected division upon test 
completion; thus, not redirecting operator attention towards a plant 
shutdown per 3.0.3. Therefore, the proposed change will not involve 
a significant increase in the probability or consequences of an 
accident previously evaluated.
    II. Create the possibility of a new or different kind of 
accident from any accident previously evaluated.
    Inhibiting the ESS [electronic switching system] Buses in Unit 1 
requires that an LCO be entered in Unit 2 due to the common loads 
shared between the Units. However, performance of the LOCA/LOOP or 
LOOP surveillance procedures does not cause any diesel generator to 
become inoperable as a result of inhibiting an ESS Bus. The time 
frame the diesels are fully loaded in the testing evolution is for a 
five-minute period to fulfill a Technical Specification requirement. 
If at that precise moment a LOCA/LOOP occurs in the operating unit, 
the ESS Buses in Unit 1 and 2 will de-energize except for the ESS 
Buses that are already connected to the diesels. In the first few 
minutes of a postulated LOCA/LOOP occurring in the operating Unit 
while performing a LOCA/LOOP test, the operator would have to take 
immediate action to shed non-essential loads from the diesels in the 
Unit under test to prepare the diesels for the shutdown loads via 
the load sequence timers in the operating unit. Existing emergency 
procedures require that these actions will be taken. Therefore, the 
incorporation of this change will not create the possibility of a 
new or different kind of accident from any accident previously 
evaluated.
    III. Involve a significant reduction in a margin of safety.
    With one or more required ac buses, (two load groups) de-
energized, the remaining ac electrical power distribution subsystems 
are capable of supporting the minimum safety functions necessary to 
shutdown the reactor and maintain it in a safe shutdown condition, 
assuming no single failure. The overall reliability is reduced, 
however, because a single failure in the remaining power 
distribution subsystems could result in the minimum required ESF 
[engineered safety feature] functions not being supported. 
Therefore, the required ac buses must be restored to OPERABLE status 
within a relatively short period of time. Eight hours has been 
accepted by the NRC as documented in NUREG-1433, Revision 1, 
``Standard Technical Specifications.'' Therefore, the incorporation 
of this change will not involve a significant reduction in the 
margin to 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) 
aresatisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: Osterhout Free Library, 
Reference Department, 71 South Franklin Street, Wilkes-Barre, 
Pennsylvania 18701
    Attorney for licensee: Jay Silberg, Esquire, Shaw, Pittman, Potts 
and Trowbridge, 2300 N Street NW., Washington, DC 20037
    NRC Project Director: John F. Stolz

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

    Date of amendment request: February 5, 1996
    Description of amendment request: The proposed amendment would 
revise Davis-Besse Nuclear Power Station (DBNPS) Technical 
Specification (TS) 3/4.3.2.1 - Safety Features Actuation System 
Instrumentation and its associated Bases. The revision changes the 
following items in the Sequence Logic Channels portion of Table 3.3-3: 
Functional Unit 4.a, Sequencer; Functional Unit 4.b, Essential Bus 
Feeder Breaker Trip (90%); Functional Unit 4.c, Diesel Generator Start, 
Load Shed on Essential Bus (59%); and the associated Bases, to clarify 
the design and actuation logic and to specify actions to take if 
instrumentation channels become inoperable.
    Basis for proposed no significant hazards consideration 
determination:
    As required by 10 50.91(a), the licensee has provided its analysis 
of the issue of no significant hazards consideration, which is 
presented below:
    Toledo Edison has reviewed the proposed changes and determined 
that a significant hazards consideration does not exist because 
operation of the Davis-Besse Nuclear Power Station, Unit No. 1 in 
accordance with these changes would:
    1a. Not involve a significant increase in the probability of an 
accident previously

[[Page 10398]]
evaluated because the proposed change to accurately reflect the 
design and actuation logic of the sequencers and essential bus 
undervoltage relays, and provide TS actions for two inoperable 
functional units does not make a change to any accident initiator, 
initiating condition or assumption. The accident previously 
evaluated in the DBNPS Updated Safety Analysis Report (USAR) Section 
15.2.9, Loss of All AC Power to the Station Auxiliaries (Station 
Blackout), is not affected by this proposed change. The proposed 
action statements maintain the USAR requirement for starting and 
loading of one [emergency diesel generator] EDG to meet the minimum 
[engineered safety features] ESF requirements. The proposed change 
accurately reflects the plant design, therefore, the change does not 
involve a significant change to the plant design or operation.
    1b. Not involve a significant increase in the consequences of an 
accident previously evaluated because the proposed changes do not 
invalidate assumptions used in evaluating the radiological 
consequences of an accident, do not alter the source term or 
containment isolation and do not provide a new radiation release 
path or alter potential radiological releases.
    2. Not create the possibility of a new or different kind of 
accident from any accident previously evaluated because the proposed 
changes do not introduce a new or different accident initiator or 
introduce a new or different equipment failure mode or mechanism.
    3. Not involve a significant reduction in a margin of safety 
because the proposed changes do not reduce the margin to safety 
which exists in the present TS or USAR. The proposed changes permit 
continued operation with one unit of the sequencer, 59% or 90% 
undervoltage protection inoperable provided the unit is placed in 
the tripped condition which is consistent with the current TS. With 
two units of the same function inoperable the associated EDG is 
declared inoperable and the requirements of the TS for an inoperable 
EDG entered, including verification that the requirements of TS 
3.0.5 are met to assure that the minimum ESF requirement is met. The 
operability requirements of the proposed TS are consistent with the 
initial condition assumptions of the safety analyses.
    The NRC staff has reviewed the licensees' analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: University of Toledo, William 
Carlson Library, Government Documents Collection, 2801 West Bancroft 
Avenue, Toledo, Ohio 43606
    Attorney for licensee: Jay E. Silberg, Esquire, Shaw, Pittman, 
Potts and Trowbridge, 2300 N Street, NW., Washington, DC 20037
    NRC Project Director: Gail H. Marcus

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

    Date of amendment request: February 5, 1996
    Description of amendment request: The proposed amendment would 
correct typographical errors, textual inconsistencies, and minor 
errors. In addition, equipment identification numbers would be added to 
the tables.
    Basis for proposed no significant hazards consideration 
determination:
    As required by 10 50.91(a), the licensee has provided its analysis 
of the issue of no significant hazards consideration, which is 
presented below:
    1. The administrative changes proposed herein will have no 
effect on plant hardware, plant design, safety limit setting, or 
plant system operation and therefore do not modify or add any 
initiating parameters that would significantly increase the 
probability or consequences of any previously analyzed accident.
    2. These changes do not affect any equipment nor do they involve 
any potential initiating events that would create any new or 
different kind of accident. Therefore, the proposed change does not 
create the possibility of a new or different kind of accident from 
any accident previously evaluated.
    3. These changes do not affect any equipment involved in 
potential initiating events or safety limits. Therefore, it is 
concluded that the proposed change does not involve a significant 
reduction in a margin of safety.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 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: Brooks Memorial Library, 224 
Main Street, Brattleboro, VT 05301
    Attorney for licensee: R. K. Gad, III, Ropes and Gray, One 
International Place, Boston, MA 02110-2624
    NRC Project Director: Ledyard B. Marsh

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

    Date of amendment request: February 8, 1996
    Description of amendment request: The proposed amendments will 
modify Technical Specification Section 15.3.10, ``Control Rod and Power 
Distribution Limits,'' and Section 15.4.1, ``Operational Safety 
Review.'' Changes and additions are proposed to clarify the 
specifications and to more closely conform to current staff guidance.
    Basis for proposed no significant hazards consideration 
determination:
    As required by 10 50.91(a), the licensee has provided its analysis 
of the issue of no significant hazards consideration which is presented 
below:
    1. Operation of this facility under the proposed Technical 
Specifications change will not create a significant increase in the 
probability or consequences of an accident previously evaluated.
    The probabilities of accidents previously evaluated are based on 
the probability of initiating events for these accidents. Initiating 
events for accidents previously evaluated for Point Beach include: 
control rod withdrawal and drop, CVCS [chemical and volume control 
system] malfunction (boron dilution), startup of an inactive reactor 
coolant loop, reduction in feedwater enthalpy, excessive load 
increase, losses of reactor coolant flow, loss of external 
electrical load, loss of normal feedwater, loss of all AC power to 
the auxiliaries, turbine overspeed, fuel handling accidents, 
accidental releases of waste liquid or gas, steam generator tube 
rupture, steam pipe rupture, control rod ejection, and primary 
coolant system ruptures.
    The consequences of the accidents previously evaluated in the 
PBNP [Point Beach Nuclear Plant] FSAR [Final Safety Analysis Report] 
are determined by the results of analyses that are based on initial 
conditions of the plant, the type of accident, transient response of 
the plant, and the operation and failure of equipment and systems.
    This change request proposes to improve the clarity of the 
requirements concerning shutdown margin, rod group alignment limits, 
rod position indication, bank insertion limits, power distribution 
limits, at-power physics tests exceptions, and low power physics 
tests exceptions. The proposed changes do not affect the probability 
of any accident initiating event, because these Technical 
Specification requirements do not control any factors that could be 
accident initiators. These Technical Specifications establish the 
requirements that provide the limitations on the initial conditions, 
transient response of the plant, and operation and failure of 
equipment and systems. The proposed changes establish the 
appropriate limiting conditions for operation, action statements, 
and allowable outage times that will continue to ensure that the 
results of the accident analyses are not changed. Additionally, 
there is no physical change to the facility or its systems. 
Therefore, the probability and consequences of any accident 
previously evaluated is not increased.
    2. Operation of this facility under the proposed Technical 
Specifications change will not create the possibility of a new or 
different kind of accident from any accident previously evaluated.
    New or different kinds of accidents can only be created by new 
or different accident initiators or sequences. This change request 
proposes to improve the clarity of the

[[Page 10399]]
Technical Specifications requirements contained in Technical 
Specification Section 15.3.10. The proposed specifications will 
clarify the existing Technical Specifications where identified by 
rewording, supplementing, or replacing existing requirements. There 
is no physical change to the facility or its systems. Therefore, a 
new or different kind of accident cannot occur, because no factors 
have been introduced that could create a new or different accident 
initiator.
    3. Operation of this facility under the proposed Technical 
Specifications change will not create a significant reduction in a 
margin of safety.
    The margins of safety for Point Beach are based on the design 
and operation of the reactor and containment and the safety systems 
that provide their protection.
    This change request proposes to improve the clarity of the 
Technical Specifications requirements contained in Technical 
Specification Section 15.3.10. The proposed specifications will 
clarify the existing Technical Specifications where identified by 
rewording, supplementing, or replacing existing requirements. There 
is no physical change to the facility or its systems. Section 
15.3.10 of the Technical Specifications provides the requirements 
that limit the operation of the reactor and establish the 
operability requirements for reactivity control by the control rod 
system. The proposed Technical Specifications changes continue to 
provide the appropriate limiting conditions for operation, action 
statements, and allowable outage times that ensure the applicable 
margins of safety to protect the reactor are preserved. Therefore, 
no reduction in any margin of safety has been introduced.
    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: Joseph P. Mann Library, 1516 
Sixteenth Street, Two Rivers, Wisconsin 54241
    Attorney for licensee: Gerald Charnoff, Esq., Shaw, Pittman, Potts, 
and Trowbridge, 2300 N Street, NW., Washington, DC 20037
    NRC Project Director: Gail H. Marcus

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.

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

    Date of application for amendments: February 16, 1996
    Brief description of amendments: The amendments provide a one-time 
surveillance requirement extension for the performance of the trip 
actuating device operational test for one of the safety injection 
manual initiation switches.
    Date of publication of individual notice in Federal Register: 
February 26, 1996 (61 FR 7125)
    Expiration date of individual notice: March 27, 1996
    Local Public Document Room location: Cameron Village Regional 
Library, 1930 Clark Avenue, Raleigh, North Carolina 27605.

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 19, 1995, as 
supplemented by letter dated February 9, 1996.
    Brief description of amendments: These amendments allow the 
implementation of the recently approved Option B to 10 CFR Part 50, 
Appendix J, Option B, by referring to Regulatory Guide 1.163, 
``Performance Based Containment Leakage - Test Program.'' This new rule 
allows a performance-based option for determining the test frequency 
for containment leakage rate testing. The amendment would modify 
Technical Specifications (TS) 1.7, 3/4.6.1.1, 3/4.6.1.2, 3/4.6.1.3, and 
3/4.6.3, and the Bases of TS 3/4.6.1.2, and would add a new TS 6.16.
    Date of issuance: February 23, 1996
    Effective date: February 23, 1996, to be implemented within 15 days 
of issuance.
    Amendment Nos.: Unit 1 - Amendment No. 103; Unit 2 - Amendment No. 
92; Unit 3 - Amendment No. 75.
    Facility Operating License Nos. NPF-41, NPF-51, and NPF-74: The 
amendments revised the Technical Specifications.
    Date of initial notice in Federal Register: January 22, 1996 (61 FR 
1627) The February 9, 1996, supplemental letter provided 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 February 23, 
1996.No significant hazards consideration comments received: No.
    Local Public Document Room location: Phoenix Public Library, 1221 
N. Central Avenue, Phoenix, Arizona 85004

[[Page 10400]]


Baltimore Gas and Electric Company, Docket No. 50-317, Calvert 
Cliffs Nuclear Power Plant, Unit No. 1, Calvert County, Maryland

    Date of application for amendment: December 21, 1995
    Brief description of amendment: The amendment allows the use of 
cladding material other than Zircaloy or ZIRLO. The Safety Evaluation 
addresses the safety significance of loading four (4) lead fuel 
assemblies (LFAs) into the Calvert Cliffs Nuclear Power Plant, Unit No. 
1, reactor vessel during cycles 13, 14, and 15. A Temporary Exemption 
was issued on November 28, 1995, (60 FR 62483) approving the loading of 
the 4 LFAs into the Unit 1 reactor vessel for the cycles noted above. 
The technical basis for the Temporary Exemption, which is the same 
basis for the requested TS amendment, was provided in the Baltimore Gas 
and Electric Company submittal dated July 13, 1995.
    Date of issuance: February 21, 1996
    Effective date: As of the date of issuance to be implemented within 
30 days.
    Amendment No.: 211
    Facility Operating License No. DPR-53: Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: January 22, 1996 (61 FR 
1627) The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated February 21, 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, Units 1 and 2, Ogle County, IllinoisDocket Nos. STN 
50-456 and STN 50-457, Braidwood Station, Units 1 and 2, Will 
County, Illinois

    Date of application for amendments: June 8, 1995
    Brief description of amendments: The amendments revise Technical 
Specification (TS) 3/4.8.1 by (1) replacing Table 4.8-1, ``Diesel 
Generator Test Schedule,'' with a single surveillance interval of at 
least once per 31 days, and (2) deleting TS 4.8.1.1.3, ``Reports.'' The 
amendments also revise ACTION statements and surveillances in TS 
3.8.1.1 related to certain diesel generator testing and startup 
requirements.Date of issuance: February 16, 1996Effective date: 
Immediately, to be implemented within 90 days.
    Amendment Nos.: 79 and 71
    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: August 30, 1995 (60 FR 
45176) The Commission's related evaluation of the amendments is 
contained in a Safety Evaluation dated February 16, 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. 50-373 and 50-374, LaSalle 
County Station, Units 1 and 2, LaSalle County, Illinois

    Date of application for amendments: June 8, 1995
    Brief description of amendments: The amendments revise Technical 
Specification (TS) 3/4.8.1 by (1) replacing Table 4.8.1.1.2-1, ``Diesel 
Generator Test Schedule,'' with a single surveillance interval of at 
least once per 31 days, and (2) deleting TS 4.8.1.1.3, ``Reports.'' The 
amendments also revise ACTION statements and surveillances in TS 
3.8.1.1 related to certain diesel generator testing and startup 
requirements.
    Date of issuance: February 16, 1996
    Effective date: Immediately, to be implemented within 90 days.
    Amendment Nos.: 109 and 94
    Facility Operating License Nos. NPF-11 and NPF-18: The amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: August 30, 1995 (60 FR 
45176) The Commission's related evaluation of the amendments is 
contained in a Safety Evaluation dated February 16, 1996.No significant 
hazards consideration comments received: No
    Local Public Document Room location: Jacobs Memorial Library, 
Illinois alley Community College, Oglesby, Illinois 61348.

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

    Date of application for amendment: November 3, 1995
    Brief description of amendment: This amendment allows deferral of 
the Reactor Coolant Pump flywheel inspection until outage 11, scheduled 
for the spring of 1998.
    Date of issuance: February 15, 1996
    Effective date: February 15, 1996
    Amendment No.: 153
    Facility Operating License No. DPR-72. Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: December 20, 1995 (60 
FR 65679) The Commission's related evaluation of the amendment is 
contained in a Safety Evaluation dated February 15, 1996.No significant 
hazards consideration comments received: No.
    Local Public Document Room location: Coastal Region Library, 8619 
W. Crystal Street, Crystal River, Florida 32629

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

    Date of application for amendments: May 5, 1995, as supplemented by 
letter dated September 28, 1995
    Brief description of amendments: The amendments consist of changes 
to the Technical Specifications (TS) relating to implementation of a 
revised thermal design procedure and steam generator water level low-
low setpoint
    Date of issuance: February 20, 1996
    Effective date: February 20, 1996
    Amendment Nos.: 183 and 177Facility Operating Licenses Nos. DPR-31 
and DPR-41: Amendments revised the Technical Specifications.
    Date of initial notice in Federal Register: October 25, 1995 (60 FR 
54719) The Commission's related evaluation of the amendments is 
contained in a Safety Evaluation dated February 20, 1996.No significant 
hazards consideration comments received: No
    Local Public Document Room location: Florida International 
University, University Park, Miami, Florida 33199.

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 22, 1995, as supplemented by letter 
dated October 9, 1995.
    Brief description of amendments: The amendments revised Technical 
Specification 4.8.1.1.2.e.7 to allow the performance of the 24-hour 
surveillance test of the diesel generators during power operation.Date 
of issuance: February 21, 1996Effective date: February 21, 1996, to be 
implemented within 30 days of issuance.

[[Page 10401]]

    Amendment Nos.: Unit 1 - Amendment No. 81; Unit 2 - Amendment No. 
70
    Facility Operating License Nos. NPF-76 and NPF-80. The amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: July 19, 1995 (60 FR 
37091) The October 9, 1995, supplement provided clarifying information 
and did not change the original no significant hazards consideration 
determination. The Commission's related evaluation of the amendments is 
contained in a Safety Evaluation dated February 21, 1996.No significant 
hazards consideration comments received: No
    Local Public Document Room location: Wharton County Junior College, 
J. M. Hodges Learning Center, 911 Boling Highway, Wharton, TX 77488

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: October 27, 1995
    Brief description of amendment: The amendment revises Technical 
Specification (TS) 3.1.3, ``Control Rod OPERABILITY,'' to include the 
25% surveillance overrun allowed by Limiting Condition for Operation 
(LCO) 3.0.2 into the allowances of the surveillance Notes for control 
rod ``notch'' testing per Surveillance Requirement (SR) 3.1.3.2 and SR 
3.1.3.3. The amendment also includes a clarification to the description 
of TS Table 3.3.3.1-1, ``Post Accident Monitoring Instrumentation,'' 
Function 7, to indicate that the Function's requirements apply to the 
position indication for only automatic primary containment isolation 
valves, rather than all primary containment isolation valves. Finally, 
the amendment includes changes to correct a number of editorial and 
typographical errors inadvertently contained in TS 3.3.4.1, ``End of 
Cycle Recirculation Pump Trip (EOC-RPT) Instrumentation,'' TS 3.3.6.1, 
``Primary Containment and Drywell Isolation Instrumentation,'' TS 
3.3.8.2, ``Reactor Protection System (RPS) Electric Power Monitoring,'' 
and TS 3.6.5.2, ``Drywell Air Lock.''
    Date of issuance: February 29, 1996
    Effective date: February 29, 1996
    Amendment No.: 102
    Facility Operating License No. NPF-62: The amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: December 20, 1995 (60 
FR 65680) The Commission's related evaluation of the amendment is 
contained in a Safety Evaluation dated February 29, 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

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

    Date of application for amendment: August 30, 1995, as supplemented 
by letter dated January 15, 1996.
    Brief description of amendment: The amendment revises Technical 
Specification 1.3, ``Reactor'', to (1) allow the use of fuel rods clad 
with Zircaloy or ZIRLO, rather than restrict use to fuel rods clad with 
Zircaloy-4, and (2) replace the specified enrichment limit with a 
limitation similar to that found in NUREG-1432, ``Standard Technical 
Specifications for Combustion Engineering Plants.''
    Date of issuance: February 29, 1996
    Effective date: As of the date of issuance, to be implemented 
concurrent with Amendment No. 144.
    Amendment No.: 155
    Facility Operating License No. DPR-36: Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: October 11, 1995 (60 FR 
52932) The January 15, 1996, submittal provided clarifying information 
and did not change the initial proposed no significant hazards 
determination.The Commission's related evaluation of the amendment is 
contained in Safety Evaluation dated February 29, 1996. No significant 
hazards consideration comments received: No
    Local Public Document Room location: Wiscasset Public Library, High 
Street, P.O. Box 367, Wiscasset, ME 04578.

Niagara Mohawk Power Corporation, Docket Nos. 50-220, and 50-410, 
Nine Mile Point Nuclear Station, Unit Nos. 1 and 2, Oswego County, 
New York

    Date of application for amendments: October 25, 1995, as 
supplemented February 7, 1996.
    Brief description of amendments: The amendments revise portions of 
Chapter 6 of the Technical Specifications to reflect management 
position title and responsibility changes.Date of issuance: February 
20, 1996
    Effective date: As of the date of issuance to be implemented within 
60 days.
    Amendment Nos.: 157 and 71
    Facility Operating License Nos. DPR-63 and NPF-69: Amendments 
revise the Technical Specifications.
    Date of initial notice in Federal Register: November 16, 1995 (60 
FR 57605) The Commission's related evaluation of the amendments is 
contained in a Safety Evaluation dated February 20, 1996.No significant 
hazards consideration comments received: No
    Local Public Document Room location: Reference and Documents 
Department, Penfield Library, State University of New York, Oswego, New 
York 13126.

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

    Date of application for amendments: August 1, 1995
    Brief description of amendments: These amendments revise the 
Technical Specifications Section 3/4.9.1, ``Reactor Mode Switch,'' in 
order to provide alternate actions to allow the continuation of core 
alterations in the event certain Reactor Manual Control System (RMCS) 
and refueling interlocks are inoperable, while preserving the intended 
function of the inoperable interlocks.
    Date of issuance: February 23, 1996
    Effective date: As of date of issuance, to be implemented within 30 
days.
    Amendment Nos.: 114 and 76
    Facility Operating License Nos. NPF-39 and NPF-85. The amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: September 27, 1995 (60 
FR 49944) The Commission's related evaluation of the amendments is 
contained in a Safety Evaluation dated February 23, 1996.No significant 
hazards consideration comments received: No
    Local Public Document Room location: Pottstown Public Library, 500 
High Street, Pottstown, Pennsylvania 19464.

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

    Date of application for amendment: October 31, 1994
    Brief description of amendment: This amendment deletes certain 
valves from Technical Specification Table 3.6.3-1, ``Primary 
Containment Isolation Valves,'' that no longer need to be tested in 
accordance with 10 CFR Part 50, Appendix J.
    Date of issuance: February 22, 1996
    
[[Page 10402]]

    Effective date: As of the date of issuance to be implemented within 
60 days from the date of issuance.
    Amendment No.:  93
    Facility Operating License No. NPF-57: This amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: March 29, 1995 (60 FR 
16198) The Commission's related evaluation of the amendment is 
contained in a Safety Evaluation dated February 22, 1996.No significant 
hazards consideration comments received: No
    Local Public Document Room location:  Pennsville Public Library, 
190 S. Broadway, Pennsville, New Jersey 08070

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

    Date of application for amendment: February 5, 1996, as 
supplemented by letter dated February 14, 1996.
    Brief description of amendment: The amendment changes Technical 
Specifications 4.6.2.2b, ``Suppression Pool Spray,'' and 4.6.2.3b, 
``Suppression Pool Cooling,'' to include flow through the RHR heat 
exchanger bypass line (in addition to the RHR heat exchanger) in the 
Suppression Pool Cooling and Suppression Pool Spray flow path used 
during RHR pump testing.
    Date of issuance: February 26, 1996
    Effective date: As of date of issuance, to be implemented within 3 
days.
    Amendment No.: 94
    Facility Operating License No. NPF-57: This amendment revised the 
Technical Specifications. Public comments requested as to proposed no 
significant hazards consideration: Yes (61 FR 5040) February 9, 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 by March 11, 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 February 26, 1996.
    Local Public Document Room location: Pennsville Public Library, 190 
S. Broadway, Pennsville, New Jersey 08070

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

    Date of application for amendments: September 28, 1995
    Brief description of amendments: The changes relocate ``Reactor 
Coolant System - Chemistry'' Technical Specification 3/4.4.7 for Salem 
Unit 1 and 3/4.4.8 for Salem Unit 2 and their associated Bases to the 
Salem Updated Final Safety Analysis Report and the Surveillance 
Requirements and Limiting Conditions for Operations to applicable plant 
procedures controlled by the 10 CFR 50.59 process. Also, the 
applicability will be changed from ``At all times'' to ``Modes 1, 2, 3, 
4, 5, and 6.''
    Date of issuance: February 22, 1996
    Effective date: Units 1 and 2, as of date of issuance and shall be 
implemented within 60 days of date of issuance.
    Amendment Nos.: 180 and 161
    Facility Operating License Nos. DPR-70 and DPR-75. The amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: November 8, 1995 (60 FR 
56369) The Commission's related evaluation of the amendments is 
contained in a Safety Evaluation dated February 22, 1996.No significant 
hazards consideration comments received: No
    Local Public Document Room location: Salem Free Public Library, 112 
West Broadway, Salem, New Jersey 08079

Sacramento Municipal Utility District, Docket No. 50-312, Rancho 
Seco Nuclear Generating Station, Sacramento County, California

    Date of application for amendment: June 20, 1995, as supplemented 
on December 19, 1995 and February 7, 1996.
    Brief description of amendment: This amendment modifies the 
technical specification requirements on qualifications for reviewers of 
facility modifications, programs, and documents affecting nuclear 
safety and changes the required schedule for reporting changes 
requested to environmental permits.
    Date of issuance: February 26, 1996
    Effective date: February 26, 1996
    Amendment No.: 124
    Facility Operating License No. NPF-1: The amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: July 19, 1995 (60 FR 
37099) The Commission's related evaluation of the amendment is 
contained in a Safety Evaluation dated February 26, 1996.No significant 
hazards consideration comments received: No.
    Local Public Document Room location: Central Library, Government 
Documents, 828 I Street, Sacramento, California 95814

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: November 20, 1995
    Brief description of amendment: The amendment adds the following 
footnote to Technical Specification (TS) 3/4.5.2: ``The allowable 
outage time for each RHR train may be extended to 7 days for the 
purpose of maintenance and modification. This exception may only be 
used one time per RHR train and is not valid after December 31, 1997.''
    Date of issuance: February 21, 1996
    Effective date: February 21, 1996
    Amendment No.: 132
    Facility Operating License No. NPF-12: Amendment revises the TS.
    Date of initial notice in Federal Register: December 20, 1995 (60 
FR 65684) The Commission's related evaluation of the amendment is 
contained in a Safety Evaluation dated February 21, 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: October 14, 1992, as 
supplemented by letter dated December 18, 1995
    Brief description of amendments: These amendments revise TS 3/
4.7.5, ``Control Room Emergency Air Cleanup System,'' by reducing the 
test duration for the control room emergency air cleanup system and 
deleting requirements for duct heaters and diverting valves. The 
associated Bases are also revised to reflect these changes.
    Date of issuance: February 28, 1996
    Effective date: February 28, 1996, to be implemented within 30 days 
of issuance.
    Amendment Nos.: Unit 1 - Amendment No. 128; Unit 2 - Amendment No. 
117
    Facility Operating License Nos. NPF-10 and NPF-15: The amendments 
revised the Technical Specifications.

[[Page 10403]]

    Date of initial notice in Federal Register: March 3, 1993 (58 FR 
12267) The December 18, 1995, supplemental letter 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 
February 28, 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 Nos. 50-259, 50-260, and 50-296, 
Browns Ferry Nuclear Plant, Units 1, 2, and 3, Limestone County, 
Alabama

    Date of application for amendments:  December 8, 1995 supplemented 
January 10, 1996 (TS 364)
    Brief description of amendment: The amendments implement recent 
changes to 10 CFR 50 Appendix J for performance-based testing of 
containment leakage.
    Date of issuance: February 22, 1996
    Effective Date: February 22, 1996
    Amendment Nos.: 228, 243 and 203
    Facility Operating License Nos. DPR-33, DPR-52 and DPR-68: 
Amendments revised the Technical Specifications.
    Date of initial notice in Federal Register: January 22, 1996 (61 FR 
1637) The letter dated January 10, 1996 provided information that did 
not change the initial proposed finding of no significant hazards 
consideration. The Commission's related evaluation of the amendment is 
contained in a Safety Evaluation dated February 22, 1996.No significant 
hazards consideration comments received: None
    Local Public Document Room location:  Athens Public library, South 
Street, Athens, Alabama 35611

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: November 22, 1995
    Brief description of amendment: The amendment added OES Nuclear, 
Inc. as an owner.
    Date of issuance: February 27, 1996
    Effective date: February 27, 1996
    Amendment No.: 81
    Facility Operating License No. NPF-58: This amendment revised the 
license.
    Date of initial notice in Federal Register: December 20, 1995 (60 
FR 65685) The Commission's related evaluation of the amendment is 
contained in a Safety Evaluation dated February 27, 1996. No 
significant hazards consideration comments received: No
    Local Public Document Room location: Perry Public Library, 3753 
Main Street, Perry, Ohio 44081

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

    Date of application for amendment: December 12, 1995, supplemented 
by facsimile transmission dated January 26, 1996
    Brief description of amendment: This amendment revises TS 3/
4.6.1.1, Containment Systems - Primary Containment -Containment 
Integrity; TS 3/4.6.1.2, Containment Systems - Containment Leakage; TS 
3/4.6.1.6, Containment Systems - Containment Vessel Structural 
Integrity; TS 3/4.6.5.3, Containment Systems - Shield Building 
Structural Integrity; and associated Bases. The revisions incorporate 
changes to the TS to adopt the provisions of Appendix J, Option B for 
Type A containment leakage testing as modified by approved exemptions 
and in accordance with Regulatory Guide 1.163, to provide consistency 
with these new requirements, and to make administrative changes.
    Date of issuance: February 22, 1996
    Effective date: February 22, 1996, and implemented not later than 
90 days after issuance.
    Amendment No.: 205
    Facility Operating License No. NPF-3. Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: January 22, 1996 (61 FR 
1637) The January 26, 1996, facsimile transmission was clarifying in 
nature and did not affect the initial no significant hazards 
consideration determination. The Commission's related evaluation of the 
amendment is contained in a Safety Evaluation dated February 22, 
1996.No significant hazards consideration comments received: No.
    Local Public Document Room location:  University of Toledo, William 
Carlson Library, Government Documents Collection, 2801 West Bancroft 
Avenue, Toledo, Ohio 43606.

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

    Date of application for amendment: June 1, 1995, as supplemented on 
October 20, 1995, December 13, 1995, and January 26, 1996.
    Brief description of amendment: The amendment revised the allowed 
outage time for one unavailable emergency diesel generator from 72 
hours to 7 days.
    Date of issuance: February 26, 1996
    Effective date: February 26, 1996
    Amendment No.: 206
    Facility Operating License No. NPF-3. Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: August 2, 1995 (60 FR 
39453) Supplemental information submitted on October 20, 1995, December 
13, 1995, and January 26, 1996, provided clarification only and was not 
outside the scope of the original no significant hazards determination. 
The Commission's related evaluation of the amendment is contained in a 
Safety Evaluation dated February 26, 1996. No significant hazards 
consideration comments received: No
    Local Public Document Room location: University of Toledo Library, 
Documents Department, 2801 Bancroft Avenue, Toledo, Ohio 43606.

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

    Date of application for amendment: September 29, 1995
    Brief description of amendment: The amendment increases the minimum 
available borated water volume requirement for the boric acid addition 
system, the minimum and maximum boron concentration requirements for 
the borated water storage tank, the minimum boron concentration 
requirement for the core flood tanks; modifies the surveillance 
requirements for trisodium phosphate dodecahydrate; and modifies the 
refueling boron concentration and the associated Action statement.
    Date of issuance: February 27, 1996
    Effective date: February 27, 1996
    Amendment No.: 207
    Facility Operating License No. NPF-3. Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: November 8, 1995 (60 FR

[[Page 10404]]
56371) The Commission's related evaluation of the amendment is 
contained in a Safety Evaluation dated February 27, 1996.No significant 
hazards consideration comments received: No.
    Local Public Document Room location: University of Toledo, William 
Carlson Library, Government Documents Collection, 2801 West Bancroft 
Avenue, Toledo, Ohio 43606.

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: November 29, 1994
    Brief description of amendments: The amendments revise and update 
the North Anna Units 1 and 2 Environmental Protection Plan (EPP) to 
reflect current obligations to the Commonwealth of Virginia, revise 
portions of the transmission corridor rights-of-way erosion control 
program for clarification and to be consistent with the state 
regulations, eliminate inconsistencies, and delete obsolete material.
    Date of issuance: February 20, 1996
    Effective date: February 20, 1996
    Amendment Nos.: 197 and 198
    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 
45188) The Commission's related evaluation of the amendments is 
contained in a Safety Evaluation dated February 20, 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.

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:  October 17, 1995, as 
supplemented by facsimile dated February 26, 1996.
    Brief description of amendments: The amendments revise the North 
Anna Units 1 and 2 Technical Specifications (TS) to allow both of the 
containment personnel airlock doors to remain open during refueling 
operations, delete License Condition 2.G for Unit 1 and 2.I for Unit 2, 
which reference the analyses for limiting doses to control room 
operators, and modify the TS Bases to clarify the emergency power 
system requirements relative to mitigation of the consequences of a 
Fuel Handling Accident.
    Date of issuance: February 27, 1996
    Effective date: As of the date of issuance to be implemented within 
30 days from the date of issuance.
    Amendment Nos.: Unit 1 - 198; Unit 2 -179
    Facility Operating License Nos. NPF-4 and NPF-7. Amendments revised 
the Technical Specifications and License Conditions.
    Date of initial notice in Federal Register: January 3, 1996 (61 FR 
187) The February 26, 1996, facsimile provided clarifying information 
that did not change the scope of the October 17, 1995, application and 
the initial proposed no significant hazards consideration 
determination.The Commission's related evaluation of the amendments is 
contained in a Safety Evaluation dated February 27, 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: November 22, 1995, as supplemented by 
letter dated February 8, 1996.
    Brief description of amendment: This amendment allows the personnel 
airlock doors to be open during core alterations and movement of 
irradiated fuel in containment. The surveillance requirements for 
containment penetrations have also been revised to require that each be 
in its ``required condition'' instead of ``closed/isolated condition.'' 
The Bases section has been updated.
    Date of issuance: February 28, 1996
    Effective date: February 28, 1996, to be implemented within 30 days 
of issuance.
    Amendment No.: Amendment No. 95
    Facility Operating License No. NPF-42. The amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: December 20, 1995 (60 
FR 65687) The February 8, 1996, supplemental letter provided additional 
clarifying information and did not change the original no significant 
hazards consideration determination. The Commission's related 
evaluation of the amendment is contained in a Safety Evaluation dated 
February 28, 1996.No significant hazards consideration comments 
received: No.
    Local Public Document Room locations: Emporia State University, 
William Allen White Library, 1200 Commercial Street, Emporia, Kansas 
66801 and Washburn University School of Law Library, Topeka, Kansas 
66621

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

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

[[Page 10405]]
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 April 12, 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 2.714, a petition for leave to intervene shall 
set forth with particularity the interest of the petitioner in the 
proceeding, and how that interest may be affected by the results of the 
proceeding. The petition should specifically explain the reasons why 
intervention should be permitted with particular reference to the 
following factors: (1) the nature of the petitioner's right under the 
Act to be made a party to the proceeding; (2) the nature and extent of 
the petitioner's property, financial, or other interest in the 
proceeding; and (3) the possible effect of any order which may be 
entered in the proceeding on the petitioner's interest. The petition 
should also identify the specific aspect(s) of the subject matter of 
the proceeding as to which petitioner wishes to intervene. Any person 
who has filed a petition for leave to intervene or who has been 
admitted as a party may amend the petition without requesting leave of 
the Board up to 15 days prior to the first prehearing conference 
scheduled in the proceeding, but such an amended petition must satisfy 
the specificity requirements described above.
    Not later than 15 days prior to the first prehearing conference 
scheduled in the proceeding, a petitioner shall file a supplement to 
the petition to intervene which must include a list of the contentions 
which are sought to be litigated in the matter. Each contention must 
consist of a specific statement of the issue of law or fact to be 
raised or controverted. In addition, the petitioner shall provide a 
brief explanation of the bases of the contention and a concise 
statement of the alleged facts or expert opinion which support the 
contention and on which the petitioner intends to rely in proving the 
contention at the hearing. The petitioner must also provide references 
to those specific sources and documents of which the petitioner is 
aware and on which the petitioner intends to rely to establish those 
facts or expert opinion. Petitioner must provide sufficient information 
to show that a genuine dispute exists with the applicant on a material 
issue of law or fact. Contentions shall be limited to matters within 
the scope of the amendment under consideration. The contention must be 
one which, if proven, would entitle the petitioner to relief. A 
petitioner who fails to file such a supplement which satisfies these 
requirements with respect to at least one contention will not be 
permitted to participate as a party.
    Those permitted to intervene become parties to the proceeding, 
subject to any limitations in the order granting leave to intervene, 
and have the opportunity to participate fully in the conduct of the 
hearing, including the opportunity to present evidence and cross-
examine witnesses. Since the Commission has made a final determination 
that the amendment involves no significant hazards consideration, if a 
hearing is requested, it will not stay the effectiveness of the 
amendment. Any hearing held would take place while the amendment is in 
effect.
    A request for a hearing or a petition for leave to intervene must 
be filed with the Secretary of the Commission, U.S. Nuclear Regulatory 
Commission, Washington, DC 20555, Attention: Docketing and Services 
Branch, or may be delivered to the Commission's Public Document Room, 
the Gelman Building, 2120 L Street, NW., Washington, DC, by the above 
date. Where petitions are filed during the last 10 days of the notice 
period, it is requested that the petitioner promptly so inform the 
Commission by a toll-free telephone call to Western Union at 1-(800) 
248-5100 (in Missouri 1-(800) 342-6700). The Western Union operator 
should be given Datagram Identification Number N1023 and the following 
message addressed to (Project Director): petitioner's name and 
telephone number, date petition was mailed, plant name, and publication 
date and page number of this Federal Register notice. A copy of the 
petition should also be sent to the Office of the General Counsel, U.S. 
Nuclear 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

[[Page 10406]]
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).

Tennesse Valley Authority, Docket No. 50-390, Watts Bar Nuclear 
Plant, Unit No. 1, Rhea County, Tennessee

    Date of application for amendment: February 26, 1996
    Brief description of amendment: The proposed amendment revises 
Technical Specifications (TS) to allow implementation of a proposed 
plant modification to preclude inadvertent transfer of the turbine-
driven auxiliary feedwater pump suction from the condensate storage 
tank to the emergency raw cooling water system.
    Date of issuance: February 28, 1996
    Effective date: February 28, 1996
    Amendment No.: 1
    Facility Operating License No. NPF-90: Amendment revises the TS. 
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 
February 28, 1996.Public comments requested as to proposed no 
significant hazards consideration: No
    Local Public Document Room location: Chattanooga-Hamilton County 
Library, 1001 Broad Street, Chattanooga, Tennessee 37402.
    Attorney for licensee: General Counsel, Tennessee Valley Authority, 
400 West Summit Hill Drive, ET 11H, Knoxville, Tennessee 37902
    NRC Project Director: Frederick J. Hebdon
    Dated at Rockville, Maryland, this 6th day of March 1996.
    For the Nuclear Regulatory Commission
Steven A. Varga,
Director, Division of Reactor Projects - I/II, Office of Nuclear 
Reactor Regulation
[Doc. 96-5817 Filed 3-12-96; 8:45 am]
BILLING CODE 7590-01-F