[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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[[Page 10392]]
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 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 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 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 keff
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 (U235) 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
