[Federal Register Volume 61, Number 187 (Wednesday, September 25, 1996)]
[Notices]
[Pages 50338-50351]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: 96-24413]
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NUCLEAR REGULATORY COMMISSION
Biweekly Notice
Applications and Amendments to Facility Operating Licenses
Involving No Significant Hazards Considerations
I. Background
Pursuant to Public Law 97-415, the U.S. Nuclear Regulatory
Commission (the Commission or NRC staff) is publishing this regular
biweekly notice. Public Law 97-415 revised section 189
[[Page 50339]]
of the Atomic Energy Act of 1954, as amended (the Act), to require the
Commission to publish notice of any amendments issued, or proposed to
be issued, under a new provision of section 189 of the Act. This
provision grants the Commission the authority to issue and make
immediately effective any amendment to an operating license upon a
determination by the Commission that such amendment involves no
significant hazards consideration, notwithstanding the pendency before
the Commission of a request for a hearing from any person.
This biweekly notice includes all notices of amendments issued, or
proposed to be issued from August 30, 1996, through September 13, 1996.
The last biweekly notice was published on September 11, 1996.
Notice Of Consideration Of Issuance Of Amendments To Facility
Operating Licenses, Proposed No Significant Hazards Consideration
Determination, And Opportunity For A Hearing
The Commission has made a proposed determination that the following
amendment requests involve no significant hazards consideration. Under
the Commission's regulations in 10 CFR 50.92, this means that operation
of the facility in accordance with the proposed amendment would not (1)
involve a significant increase in the probability or consequences of an
accident previously evaluated; or (2) create the possibility of a new
or different kind of accident from any accident previously evaluated;
or (3) involve a significant reduction in a margin of safety. The basis
for this proposed determination for each amendment request is shown
below.
The Commission is seeking public comments on this proposed
determination. Any comments received within 30 days after the date of
publication of this notice will be considered in making any final
determination.
Normally, the Commission will not issue the amendment until the
expiration of the 30-day notice period. However, should circumstances
change during the notice period such that failure to act in a timely
way would result, for example, in derating or shutdown of the facility,
the Commission may issue the license amendment before the expiration of
the 30-day notice period, provided that its final determination is that
the amendment involves no significant hazards consideration. The final
determination will consider all public and State comments received
before action is taken. Should the Commission take this action, it will
publish in the Federal Register a notice of issuance and provide for
opportunity for a hearing after issuance. The Commission expects that
the need to take this action will occur very infrequently.
Written comments may be submitted by mail to the Chief, Rules
Review and Directives Branch, Division of Freedom of Information and
Publications Services, Office of Administration, U.S. Nuclear
Regulatory Commission, Washington, DC 20555-0001, and should cite the
publication date and page number of this Federal Register notice.
Written comments may also be delivered to Room 6D22, Two White Flint
North, 11545 Rockville Pike, Rockville, Maryland from 7:30 a.m. to 4:15
p.m. Federal workdays. Copies of written comments received may be
examined at the NRC Public Document Room, the Gelman Building, 2120 L
Street, NW., Washington, DC. The filing of requests for a hearing and
petitions for leave to intervene is discussed below.
By October 25, 1996, the licensee may file a request for a hearing
with respect to issuance of the amendment to the subject facility
operating license and any person whose interest may be affected by this
proceeding and who wishes to participate as a party in the proceeding
must file a written request for a hearing and a petition for leave to
intervene. Requests for a hearing and a petition for leave to intervene
shall be filed in accordance with the Commission's ``Rules of Practice
for Domestic Licensing Proceedings'' in 10 CFR Part 2. Interested
persons should consult a current copy of 10 CFR 2.714 which is
available at the Commission's Public Document Room, the Gelman
Building, 2120 L Street, NW., Washington, DC and at the local public
document room for the particular facility involved. If a request for a
hearing or petition for leave to intervene is filed by the above date,
the Commission or an Atomic Safety and Licensing Board, designated by
the Commission or by the Chairman of the Atomic Safety and Licensing
Board Panel, will rule on the request and/or petition; and the
Secretary or the designated Atomic Safety and Licensing Board will
issue a notice of a hearing or an appropriate order.
As required by 10 CFR 2.714, a petition for leave to intervene
shall set forth with particularity the interest of the petitioner in
the proceeding, and how that interest may be affected by the results of
the proceeding. The petition should specifically explain the reasons
why intervention should be permitted with particular reference to the
following factors: (1) the nature of the petitioner's right under the
Act to be made a party to the proceeding; (2) the nature and extent of
the petitioner's property, financial, or other interest in the
proceeding; and (3) the possible effect of any order which may be
entered in the proceeding on the petitioner's interest. The petition
should also identify the specific aspect(s) of the subject matter of
the proceeding as to which petitioner wishes to intervene. Any person
who has filed a petition for leave to intervene or who has been
admitted as a party may amend the petition without requesting leave of
the Board up to 15 days prior to the first prehearing conference
scheduled in the proceeding, but such an amended petition must satisfy
the specificity requirements described above.
Not later than 15 days prior to the first prehearing conference
scheduled in the proceeding, a petitioner shall file a supplement to
the petition to intervene which must include a list of the contentions
which are sought to be litigated in the matter. Each contention must
consist of a specific statement of the issue of law or fact to be
raised or controverted. In addition, the petitioner shall provide a
brief explanation of the bases of the contention and a concise
statement of the alleged facts or expert opinion which support the
contention and on which the petitioner intends to rely in proving the
contention at the hearing. The petitioner must also provide references
to those specific sources and documents of which the petitioner is
aware and on which the petitioner intends to rely to establish those
facts or expert opinion. Petitioner must provide sufficient information
to show that a genuine dispute exists with the applicant on a material
issue of law or fact. Contentions shall be limited to matters within
the scope of the amendment under consideration. The contention must be
one which, if proven, would entitle the petitioner to relief. A
petitioner who fails to file such a supplement which satisfies these
requirements with respect to at least one contention will not be
permitted to participate as a party.
Those permitted to intervene become parties to the proceeding,
subject to any limitations in the order granting leave to intervene,
and have the opportunity to participate fully in the conduct of the
hearing, including the opportunity to present evidence and cross-
examine witnesses.
If a hearing is requested, the Commission will make a final
determination on the issue of no significant hazards consideration. The
[[Page 50340]]
final determination will serve to decide when the hearing is held.
If the final determination is that the amendment request involves
no significant hazards consideration, the Commission may issue the
amendment and make it immediately effective, notwithstanding the
request for a hearing. Any hearing held would take place after issuance
of the amendment.
If the final determination is that the amendment request involves a
significant hazards consideration, any hearing held would take place
before the issuance of any amendment.
A request for a hearing or a petition for leave to intervene must
be filed with the Secretary of the Commission, U.S. Nuclear Regulatory
Commission, Washington, DC 20555-0001, Attention: Docketing and
Services Branch, or may be delivered to the Commission's Public
Document Room, the Gelman Building, 2120 L Street, NW., Washington DC,
by the above date. Where petitions are filed during the last 10 days of
the notice period, it is requested that the petitioner promptly so
inform the Commission by a toll-free telephone call to Western Union at
1-(800) 248-5100 (in Missouri 1-(800) 342-6700). The Western Union
operator should be given Datagram Identification Number N1023 and the
following message addressed to (Project Director): petitioner's name
and telephone number, date petition was mailed, plant name, and
publication date and page number of this Federal Register notice. A
copy of the petition should also be sent to the Office of the General
Counsel, U.S. Nuclear Regulatory Commission, Washington, DC 20555-0001,
and to the attorney for the licensee.
Nontimely filings of petitions for leave to intervene, amended
petitions, supplemental petitions and/or requests for a hearing will
not be entertained absent a determination by the Commission, the
presiding officer or the Atomic Safety and Licensing Board that the
petition and/or request should be granted based upon a balancing of
factors specified in 10 CFR 2.714(a)(1)(i)-(v) and 2.714(d).
For further details with respect to this action, see the
application for amendment which is available for public inspection at
the Commission's Public Document Room, the Gelman Building, 2120 L
Street, NW., Washington, DC, and at the local public document room for
the particular facility involved.
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: August 2, 1996
Description of amendment request: The proposed amendment would
eliminate from the licenses the requirement to conduct corrosion
testing for the laser welded steam generator sleeves.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. The proposed change does not involve a significant increase
in the probability or consequences of an accident previously
evaluated.
This proposed change only involves deleting laboratory testing
requirements designed to demonstrate service life of laser welded
sleeved tubes in the presence of a crevice. Current inspection
requirements ensure that premature degradation is identified and
that tubes containing degraded sleeve joints are plugged.
Operational primary-to-secondary leakage limits ensure that
appropriate action is taken if sleeve degradation results in
leakage. These actions will ensure that offsite dose will be
maintained within a small percentage of 10 CFR 100 limits. Failure
of a sleeve joint is bounded by the Steam Generator Tube Rupture
event evaluated in the [Updated Final Safety Analysis Report] UFSAR.
Therefore, the laboratory testing to determine service life of
sleeved tube joints in the presence of a crevice does not provide
any further useful data. The change does not result in the
installation of any new equipment, and no existing equipment is
modified.
Therefore, this proposed change 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.
This proposed change only addresses deleting the laboratory
testing requirements designed to demonstrate service life of sleeved
tubes in the presence of a crevice. Sleeved tubes will continue to
be inspected and plugged in accordance with existing requirements
which are sufficient to ensure detection and repair of degraded
tubes. Premature degradation of tubes is addressed through primary-
to-secondary leakage monitoring and leakage limits. No new equipment
is being installed and no existing equipment is being modified by
this proposed change. Also, no new system configurations will be
introduced as a result of this proposed change. Therefore, no new or
different failure modes are being introduced by deleting the
laboratory testing.
Thus, this proposed change 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 proposed change only involves deleting laboratory testing
requirements designed to demonstrate service life of sleeved tubes
in the presence of a crevice. Sleeve integrity will be monitored
during the operating cycle through the current primary-to-secondary
leakage monitoring program. In the event of premature degradation of
a sleeve joint that results in tube leakage, plant shutdown will
occur as required by Technical Specifications and administrative
requirements in accordance with approved plant procedures. Sleeved
tubes will be monitored for degradation in accordance with the
existing inservice inspection requirements which monitors a minimum
20 percent random sleeve sample size. Any tubes with defective
sleeve joints will be plugged as required by Technical
Specifications. Service life of sleeved tubes in the presence of a
crevice, as predicted by laboratory testing, does not affect the
margin of safety of the plant. Therefore, this proposed change does
not involve a significant reduction in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
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
Commonwealth Edison Company, Docket Nos. 50-373 and 50-374, LaSalle
County Station, Units 1 and 2, LaSalle County, Illinois
Date of amendment request: July 15, 1996
Description of amendment request: The proposed amendments would
revise Technical Specifications (TS) and associated Bases to relocate
the fire protection program elements from the TS to the Fire Protection
Program. The affected TS sections are 3/4.3.7.9, ``Fire Detection
Instrumentation;'' 3/4.7.5, ``Fire Suppression Systems;'' 3/4.7.6,
``Fire Rated Assemblies;'' and 6.1.C.4,
[[Page 50341]]
``Fire Brigade Staffing.'' In addition, the amendments revise the
Operating License to replace existing fire protection license
conditions with the NRC's standard fire protection license condition.
These changes are made in accordance with the guidance provided in
Generic Letter (GL) 86-10, ``Implementation of Fire Protection
Requirements,'' and GL 88-12, ``Removal of Fire Protection Requirements
from Technical Specifications.'' Also, the May 19, 1995, proposed
revision to remove the fire protection requirements from the TS (60 FR
35067) is withdrawn.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1) Involve a significant increase in the probability or
consequences of an accident previously evaluated because:
This amendment request does not involve any actual changes to
the fire protection systems at the station. It involves an
administrative change which relocates the control of the Fire
Protection Program from each unit's operating license and technical
specifications to the station Fire Protection Program, as suggested
in Generic Letters 86-10 and 88-12. Therefore, the relocation of
these controls does not affect the assumptions for any of the
accident analysis contained in Chapter 15 of the [Updated Final
Safety Analysis Report] UFSAR.
The Fire Protection Technical Specifications which are to be
relocated to the Fire Protection Program will be controlled by the
proposed fire protection license condition and 10CFR 50.59. These
controls ensure that the requested changes maintain the same level
of control for the Fire Protection Program as that which currently
exists in the Technical Specifications. Therefore, this change is
administrative in nature and does not involve a significant increase
in the probability or consequences of an accident previously
evaluated.
2) Create the possibility of a new or different kind of accident
from any accident previously evaluated because:
This amendment request does not involve any physical changes to
the fire protection systems or reduce the level of control of the
Fire Protection Program. It therefore does not create the
possibility of a new or different type of accident than any
previously described in the UFSAR.
3) Involve a significant reduction in the margin of safety
because:
The same level of control which is currently applied to the Fire
Protection Program by the limiting conditions for operation and the
surveillance requirements of the technical specifications will be
included in the controls applied by the unit licenses and the Fire
Protection Program. Therefore, the margin of safety as defined in
the technical specification bases will not be reduced by this
proposed amendment.
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: Jacobs Memorial Library,
Illinois Valley Community College, Oglesby, Illinois 61348.
Attorney for licensee: Michael I. Miller, Esquire; Sidley and
Austin, One First National Plaza, Chicago, Illinois 60603
NRC Project Director: Robert A. Capra
Commonwealth Edison Company, Docket Nos. 50-295 and 50-304, Zion
Nuclear Power Station, Units 1 and 2, Lake County, Illinois
Date of amendment request: July 26, 1996, and supplemented on
September 3, 1996
Description of amendment request: The proposed amendments would
allow licensee control of the reactor coolant system (RCS) pressure and
temperature (P/T) limits for heatup, cooldown, low temperature
operation and hydrostatic testing. They would also revise the reactor
vessel material surveillance program specimen withdrawal schedule such
that the Unit 2 removal of capsule X is delayed until 19 Effective Full
Power Years (EFPY). This change affects the schedule for withdrawing
surveillance capsules from the reactor vessel for testing to measure
the impact of neutron irradiation of the vessel material and is
required by Section III.B.3 of 10 CFR Part 50, Appendix H, ``Reactor
Vessel Material Surveillance Program Requirements.'' The schedule must
be approved by the Nuclear Regulator Commission (NRC) before
implementation.
Based on input from the Babcock and Wilcox Owners Group Reactor
Vessel Working Group, the data from Zion, Unit 2, capsule X would be
more useful in the overall Master Integrated Reactor Vessel
Surveillance Program (MIRVP) context if irradiated to the ASTM E185-82
maximum of twice the peak End Of Life (EOL) vessel fluence, because
data at higher fluences is needed to characterize irradiation behavior
at the higher EOL fluences characteristic of other non-Commonwealth
Edison MIRVP vessels. For this reason, the licensee is proposing
withdrawing and testing Zion, Unit 2, capsule X at 19 EFPY, which is
currently estimated to occur at refueling outage Z2R18, in the year
2002.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. Does the change involve a significant increase in the
probability or consequences of an accident previously evaluated?
The proposed change revises the 10 CFR 50, Appendix H reactor
vessel material specimen withdrawal schedule. Neither the specimens,
nor the process of withdrawal of the specimens, are considered as
initiators for any previously evaluated accident. Further, data at
all fluence levels of current interest based on ASTM E185-82 has
already been obtained from seven Zion Unit 1 and 2 capsules which
have been tested, and the existing evaluations show the reactor
vessel fracture toughness properties to be as expected, and
providing the required safety margin. Extending the time for
withdrawal of the specimen does not adversely affect the pressure
and temperature limit curves for the reactor vessel. Regulatory
Guide 1.99, Rev. 2, was used to prepare the conservative pressure
and temperature limit curves which continue to be requirements.
Additionally, Zion Station participates in the B&W Owners Group
Reactor Vessel Working Group designed to significantly increase the
amount of PWR surveillance data. Under this Working Group, Zion
Station data contributes to the overall understanding of reactor
vessel material irradiation behavior at high EOL fluences, and
obtains the benefit of data from other plants. This program
complements the Zion Station program so that postponement of the
specimen withdrawal will have minimal impact on the understanding of
the irradiation effects on the Zion Station reactor vessel.
Therefore, this change does not involve a significant increase in
the probability or consequences of an accident previously evaluated.
The proposed revision to the specimen withdrawal schedule does
not change the system operation or design, and therefore, does not
change the response of any required structures, systems or
components in the mitigation of any evaluated accident. As such,
this change does not involve a significant increase in the
consequences of an accident previously evaluated.
The proposed change relocates the RCS P/T, LTOP [low-temperature
overpressure protection] limitations, and supporting information
from the Technical Specifications to Licensee control, specifically
a Pressure Temperature Limits Report (PTLR). Compliance with these
limitations will continue to be required by the Technical
Specifications, however the limitations themselves will be relocated
to a Licensee controlled document. Changes to these limitations will
be controlled by Section 5.6.6 of the Technical Specifications.
Changes to the RCS P/T limits can only be made in accordance with
the approved methodologies listed in the Technical Specifications
which will, in combination with the limitations that continue to be
[[Page 50342]]
imposed by the Technical Specifications, continue to assure the
function of the reactor vessel as a pressure boundary. Revisions to
the LTOP limits can only be made in accordance with the approved
methodologies listed in the Technical Specifications, with any
resulting setpoint changes controlled through a process which
utilizes 10 CFR 50.59. Therefore, 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 necessitate a physical alteration
of the plant (no new or different equipment will be installed). The
proposed revision to the specimen withdrawal schedule does not
change the system operation or design, and therefore, does not
introduce any new failure mechanisms. The proposed specimen
withdrawal schedule continues to provide the required data for
subsequent reactor vessel evaluations, and previous data has
confirmed the confidence in the integrity of the reactor vessel well
beyond the completion of the evaluations following the proposed
withdrawal. Therefore, this revision to the withdrawal schedule does
not create the possibility of a new or different kind of accident
from any accident previously evaluated.
The proposed change does not necessitate a physical alteration
of the plant (no new or different equipment will be installed). The
Technical Specifications will continue to retain requirements to
maintain the RCS within acceptable operational limitations and to
assure operability of the LTOP system. As such, the Technical
Specifications will continue to require compliance with these
limitations. Thus, this change does not create the possibility of a
new or different kind of accident from any accident previously
evaluated.
3. Does this change involve a significant reduction in a margin
of safety?
The proposed change to the specimen withdrawal schedule will not
result in a significant reduction in a margin of safety because it
has no impact on any safety analysis assumptions. Additionally, data
at all fluence levels of current interest based on ASTM E185-82 has
already been obtained with the seven Zion Unit 1 and 2 capsules
which have been tested, and the existing evaluations show the
reactor vessel fracture toughness properties to be as expected, and
providing the required safety margin. The current pressure and
temperature limits are conservative and also provide sufficient
margin to ensure the integrity of the reactor vessel. The proposed
change to the withdrawal schedule does not adversely impact these
curves. Therefore, this change does not involve a significant
reduction in a margin of safety.
The proposed change will not result in a significant reduction
in a margin of safety because it has no impact on any safety
analysis assumptions. Any future changes to the RCS P/T, LTOP
limits, or supporting information must be performed in accordance
with approved NRC methodologies, and compliance with the limitations
relocated to the PTLR will continue to be required by the Technical
Specifications. Additionally, any revision to the LTOP limits which
result in setpoint changes will be controlled through a process
which utilizes 10 CFR 50.59. Therefore, this 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
requested amendments involve no significant hazards consideration.
Local Public Document Room location: Waukegan Public Library, 128
N. County Street, Waukegan, Illinois 60085
Attorney for licensee: Michael I. Miller, Esquire; Sidley and
Austin, One First National Plaza, Chicago, Illinois 60603
NRC Project Director: Robert A. Capra
Detroit Edison Company, Docket No. 50-341, Fermi-2, Monroe County,
Michigan
Date of amendment request: September 5, 1996 (NRC-96-0075)
Description of amendment request: The proposed amendment would
revise Technical Specification (TS) sections 2.1.2 and 3.4.1.1 to
incorporate cycle-specific safety limit minimum critical power ratios
(SLMCPRs) for the core that will be loaded during the upcoming
refueling outage expected to commence in November 1996.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. The proposed TS changes do not involve a significant increase
in the probability or consequences of an accident previously
evaluated.
The derivation of the revised SLMCPRs for Fermi 2 for
incorporation into the TS, and its use to determine cycle-specific
thermal limits, have been performed using NRC-approved methods.
Additionally, interim implementing procedures, which incorporate
cycle-specific parameters, have been used which result in a more
restrictive value for the SLMCPR. These calculations do not change
the method of operating the plant and have no effect on the
probability of an accident initiating event or transient. The basis
of the MCPR Safety Limit is to ensure that no mechanistic fuel
damage is calculated to occur if the limit is not violated. The new
SLMCPRs preserve the existing margin to transition boiling and the
probability of fuel damage is not increased. Therefore, the proposed
TS change does not involve an increase in the probability or
consequences of an accident previously evaluated.
2. The proposed TS changes do not create the possibility of a
new or different kind of accident from any accident previously
evaluated.
The proposed change results from analysis of the Cycle 6 core
reload using the same fuel types as previous cycles. These changes
do not involve any new method for operating the facility and do not
involve any facility modifications. No new initiating events or
transients result from these changes. Therefore, the proposed TS
change does not create the possibility of a new or different kind of
accident from any accident previously evaluated.
3. The proposed TS changes do not involve a significant
reduction in a margin of safety.
The margin of safety as defined in the TS Bases will remain the
same. The new SLMCPRs are calculated using NRC-approved methods
which are in accordance with the current fuel design and licensing
criteria. Additionally, interim implementing procedures, which
incorporate cycle-specific parameters, have been used. The MCPR
Safety Limit remains high enough to ensure that greater than 99.9%
of all fuel rods in the core will avoid transition boiling if the
limit is not violated, thereby preserving the fuel cladding
integrity. Therefore, the proposed TS change does not involve a
reduction in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: Monroe County Library System,
3700 South Custer Road, Monroe, Michigan 48161
Attorney for licensee: John Flynn, Esq., Detroit Edison Company,
2000 Second Avenue, Detroit, Michigan 48226
NRC Project Director: John Hannon
Entergy Operations, Inc., et al., Docket No. 50-416, Grand Gulf
Nuclear Station, Unit 1, Claiborne County, Mississippi
Date of amendment request: July 31, 1996, as supplemented by letter
dated September 5, 1996. These letters supersede the application
submitted in letter dated May 9, 1996, which was noticed in the Federal
Register on June 5, 1996 (61 FR 28614).
Description of amendment request: The amendment request would (1)
increase the safety limit minimum critical power ratio (MCPR) for two
loop operation and single loop operation to 1.12 and 1.14,
respectively, and (2) add a General Electric topical report to the list
of documents describing the analytical methods used to determine the
core operating limits. The proposed changes are to Section 2.1.1,
Reactor
[[Page 50343]]
Core Safety Limits, and Section 5.6.5, Core Operating Limits Report
(COLR), respectively, of the Technical Specifications (TSs). This
amendment would go into effect in Operating Cycle 9, at the end of the
upcoming Refueling Outage 8, and the plant will have a mixed core of
Siemens Power Corporation (SPS) 9x9-5 and General Electric (GE) GE11
reload fuel. The licensee also proposed changes to the Bases of the TSs
associated with the above proposed changes.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
I. The proposed change does not significantly increase the
probability or consequences of an accident previously evaluated.
The Minimum Critical Power Ratio (MCPR) safety limit is defined
in the Bases to Technical Specification 2.1.1 as that limit which
``ensures that during normal operation and during Anticipated
Operational Occurrences (AOOs), at least 99.9% of the fuel rods in
the core do not experience transition boiling.'' The MCPR safety
limit is re-evaluated for each reload and, for GGNS [Operating]
Cycle 9, the analyses have concluded that a two-loop MCPR safety
limit of 1.12 based on the application of the generic GE MCPR
methodology is necessary to ensure that this acceptance criterion is
satisfied. For single-loop operation, a MCPR safety limit of 1.14
based on the generic GE MCPR methodology was determined to be
necessary. Core MCPR operating limits are developed to support the
Technical Specification 3.2 requirements and ensure these safety
limits are maintained in the event of the worst-case transient.
Since the MCPR safety limit will be maintained at all times,
operation under the proposed changes will ensure at least 99.9% of
the fuel rods in the core do not experience transition boiling.
Therefore, The Minimum Critical Power Ratio (MCPR) safety limit
change does not affect the probability or consequences of an
accident.
The implementation of GE's GESTAR-II approved methodology has no
effect on the probability or consequences of any accidents
previously evaluated. One exception to GESTAR is that the mis-
oriented and mis-located bundle events will continue to be analyzed
as accidents subject to the acceptance criteria in the current
licensing basis. The design of the GE11 fuel bundles is such that
the bundles are not likely to be mis-oriented or mis-located and the
normal administrative controls will be in effect for assuring proper
orientation and location. Therefore, the probability of a fuel
loading error is not increased. This analysis ensures that
postulated dose releases will not exceed a small fraction (10
percent) of 10 CFR 100 limits.
Therefore, the consequences of accidents previously evaluated
are unchanged.
II. The proposed change does not create the possibility of a new
or different kind of accident from any accident previously
evaluated.
The GE11 fuel to be used in [Operating] Cycle 9 is of a design
compatible with fuel present in the core and used in the previous
cycle. Therefore, the GE11 fuel will not create the possibility of a
new or different kind of accident. The proposed changes do not
involve any new modes of operation, any changes to setpoints, or any
plant modifications. They introduce revised MCPR safety limits that
have been proved to be acceptable for Cycle 9 operation. Compliance
with the applicable criterion for incipient boiling transition
continues to be ensured. The proposed MCPR safety limits do not
result in the creation of any new precursors to an accident.
Therefore, the proposed changes do not create the possibility of
a new or different type of accident from any accident previously
evaluated.
III. The proposed change does not involve a significant
reduction in a margin of safety.
The MCPR safety limits have been evaluated to ensure that during
normal operation and during AOOs [abnormal operating occurrences],
at least 99.9% of the fuel rods in the core do not experience
transition boiling. Therefore, the implementation of the proposed
changes in the MCPR safety limit ensure there is no reduction in the
margin of safety.
As with the current SPC methodology, GGNS will implement only
the NRC-approved revisions to GE's GESTAR methodology. This GE
methodology is similar to those SPC reports currently listed in TS
5.6.5 and it will be applied in a similar, conservative fashion. One
exception to GESTAR is that the mis-oriented and mis-located bundle
events will continue to be analyzed as accidents subject to the
acceptance criteria in the current licensing basis. This analysis
ensures that postulated dose releases will not exceed a small
fraction (10 percent) of 10CFR100 [10 CFR Part 100] limits. On this
basis, the implementation of this GE methodology 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: Judge George W. Armstrong
Library, 220 S. Commerce Street, Natchez, MS 39120
Attorney for licensee: Nicholas S. Reynolds, Esquire, Winston and
Strawn, 1400 L Street, N.W., 12th Floor, Washington, DC 20005-3502
NRC Project Director: William D. Beckner
Gulf States Entergy, Cajun Electric Power Cooperative, and Entergy
Operations, Inc., Docket No. 50-458, River Bend Station, Unit 1,
West Feliciana Parish, Louisiana
Date of amendment request: August 1, 1996
Description of amendment request: The proposed amendment would
revise the Technical Specifications (TSs) to incorporate requirements
for limiting the time that the hydrogen mixing isolation valves on the
drywell are open. The requirements were contained in the old TSs and
with the conversion to the Improved Standard Technical Specifications,
the requirements were inadvertently changed. The proposed action is to
restore requirements to meet the licensing basis for the River Bend
Station. The proposed amendment would also change the time from 7 days
to 31 days to determine the cumulative time the valves are open.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
The proposed changes in this submittal put the requirements that
were in the original Technical Specifications for the Hydrogen
Mixing System back into the current Technical Specifications. The
changes reenstate into the Technical Specifications limitations that
were previously agreed to between River Bend and the Nuclear
Regulatory Commission in the FSAR Safety Evaluation Report for the
Hydrogen Mixing System.
The River Bend SER states in Supplement 2, Section 6.2.4,
``Since the applicant has not demonstrated that these valves are
capable of closing under accident conditions in the drywell, certain
restrictions apply. Technical Specification 3.6.6.2 specifies that
in Operating Modes 1 and 2, the total number of hours used should
not exceed 5 hours/365 days and in Operating Mode 3 the number of
hours should be limited to 90 hours/365 days.'' To date, the
hydrogen mixing isolation valves have not been fully demonstrated to
be capable of closing under accident conditions in the drywell. The
old Standard Technical Specifications (Attachment 2) used at River
Bend reflected this condition. When conversion to ITS was made,
these requirements were dropped but should not have been. In
addition, the requirement to operate the hydrogen mixing system
every 92 days during Modes 1, 2, and 3 was added without
consideration for the requirements in the River Bend Safety
Evaluation Report.
Consequently, for these proposed change, since the requirements
already exist and are being reenstated into the Technical
Specifications, this change is administrative in nature. The
requirements have remained in place through the SER, but were
[[Page 50344]]
inadvertently removed from the Technical Specifications. This change
places the requirements from the SER back into the Technical
Specifications.
In addition, changing the requirement from the old Technical
Specifications for determining the cumulative time that the hydrogen
mixing inlet and outlet valves are open from every 7 days to every
31 days is again administrative in nature, since this only changes
the frequency with which a given requirement is tracked
administratively. It does not change the actual requirement in any
way.
Consequently, since both of these changes are administrative in
nature and only incorporate requirements into the Technical
Specifications that already existed in the RBS FSAR Safety
Evaluation Report, the changes proposed in this amendment request do
not change the probability or consequences of an accident previously
evaluated.
This proposed change does not involve a change to the plant
design or operation. As a result, the proposed change does not
affect any of the parameters or conditions that could contribute to
the initiation of any accidents.
The changes proposed in this amendment request are
administrative in nature and merely add requirements back into the
Technical Specifications that were inadvertently deleted during the
conversion to ITS. Because of the administrative nature of the
proposed changes, it is not possible to create a new or different
kind of accident from any accident previously evaluated.
The proposed changes in this amendment request reenstate
requirements into the Technical specifications that are contained
present in the RBS FSAR Safety Evaluation Report. These requirements
were inadvertently deleted during the conversion to ITS.
Because of the administrative nature of these Technical
Specification changes, there is no change to the margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: Government Documents
Department, Louisiana State University, Baton Rouge, LA 70803
Attorney for licensee: Mark Wetterhahn, Esq., Winston & Strawn,
1400 L Street, N.W., Washington, D.C. 20005
NRC Project Director: William D. Beckner
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: August 15, 1996.
Description of amendment request: The proposed amendments would
remove a requirement for performance of a surveillance incorporating a
high toxic gas test signal.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. The proposed change does not involve a significant increase
in the probability or consequences of an accident previously
evaluated.
Analyses were performed to evaluated postulated releases of
potentially hazardous chemicals for their impact on Control Room
habitability. The latest revision of these analyses shows that none
of the potentially hazardous chemicals utilized onsite or in the
surrounding 5-mile radius around the South Texas Project pose a
credible hazard to the Control Room. Consequently, there is no need
to ensure that the Control Room Makeup and Cleanup Filtration System
can automatically switch into a recirculation mode of operation by
isolating the normal supply and exhaust flow in response to a High
Toxic Gas test signal. Therefore, elimination of the unnecessary
surveillance has no effect on the probability of an accident or its
consequences.
2. The proposed change does not create the possibility of a new
or different kind of accident from any accident previously
evaluated.
The Toxic Gas Monitoring System was provided to protect against
hazardous toxic gas releases only. Verifying automatic switch into
the recirculation mode of operation is no longer necessary since the
Toxic Gas Analyzers have been removed. This change does not affect
other tests for verification of automatic switching into the
recirculation mode of operation. Therefore, the proposed change 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.
Analyses have shown that none of the chemicals onsite and within
a 5-mile radius of the South Texas Project pose a credible hazard to
the facility. Automatic switching of the Control Room Makeup and
Cleanup Filtration System will continue to be verified using test
signals from other sources.
Based upon this evaluation, the South Texas Project has
concluded that these changes do not involve any significant hazards
considerations.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
request for amendments involves no significant hazards consideration.
Local Public Document Room location: Wharton County Junior College,
J. M. Hodges Learning Center, 911 Boling Highway, Wharton, TX 77488
Attorney for licensee: Jack R. Newman, Esq., Morgan, Lewis &
Bockius, 1800 M Street, N.W., Washington, DC 20036-5869
NRC Project Director: William D. Beckner
Illinois Power Company and Soyland Power Cooperative, Inc., Docket
No. 50-461, Clinton Power Station, Unit No. 1, DeWitt County,
Illinois
Date of amendment request: August 15, 1996
Description of amendment request: A Federal Register Notice on May
22, 1996 (61 FR 25707), stated that revisions were being proposed to
Clinton Power Station Technical Specification (TS) 3.3.6.2, ``Secondary
Containment Isolation Instrumentation;'' TS 3.3.7.1, ``Control Room
Ventilation System Instrumentation;'' TS 3.6.1.2, ``Primary Containment
Air Locks;'' TS 3.6.1.3, ``Primary Containment Isolation Valves;'' TS
3.6.4.1, ``Secondary Containment;'' TS 3.6.4.2, ``Secondary Containment
Isolation Dampers;'' TS 3.6.4.3, ``Standby Gas Treatment;'' TS 3.7.3,
``Control Room Ventilation;'' and TS 3.7.4, ``Control Room AC System.''
By letter dated August 15, 1996, the licensee revised their proposal to
consolidate the above changes under a newly proposed Special Operations
LCO (i.e., LCO 3.10.10, ``Single Control Rod Withdrawal - Refueling'').
Therefore, the Description of Amendment Request to the TSs has changed
as described herein. The Basis for No Significant Hazards Consideration
has not changed and is repeated below.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration which is presented below:
1. The proposed changes eliminate CORE ALTERATIONS as an
applicable condition requiring operability of the primary and
secondary containment and control room ventilation system. As stated
in the BASES for the associated Technical Specifications,
operability of these systems is primarily required for mitigation of
the design basis accident - fuel handling accident (DBA-FHA) and
design basis accident - loss of coolant accident (DBA-LOCA). The
performance of CORE ALTERATIONS alone is neither a
[[Page 50345]]
precursor to, nor a condition during which these DBAs are postulated
to occur. The proposed changes only delete CORE ALTERATIONS as an
applicable condition for the affected Technical Specifications. All
other applicable MODES or specified conditions, including operations
with the potential for draining the reactor vessels (OPDRVs) and the
movement of irradiated fuel assemblies within the primary or
secondary containment, remain unchanged. Further, the limitations
placed on the handling of light loads are also unchanged. The
Technical Specifications (and the separate requirements imposed on
the handling of light loads) will thus continue to require that
systems or functions designed to mitigate design-basis/previously
evaluated accidents are OPERABLE during the relevant operating MODES
or conditions. On the basis of the above, it is concluded that the
requested amendment will not increase the probability or
consequences of any accident previously evaluated.
2. The proposed changes do not involve any modification to the
plant design or to the operation of plant systems (except to
determine when certain analyzed accident-mitigating systems or
features are required to be OPERABLE). The failure modes considered
for the proposed changes are the same as those previously
considered, therefore, it can be concluded that no new failure modes
will be created. On this basis, the proposed amendment will not
create the possibility of a new or different kind of accident from
any accident previously evaluated.
3. The changes being made to eliminate CORE ALTERATIONS as an
applicable condition for which certain LCOs must be met, do not
eliminate the requirements for operability of those systems or
features assumed to mitigate design-basis or analyzed accidents
during the applicable MODES when such systems or features are
assumed to be available for performing their mitigating function.
The safety margins assumed or established by the accident analyses
for those design-basis events (as described in the accident analyses
of the Clinton Power Station Updated Final Safety Analysis Report)
therefore remain unchanged. Further, the proposed changes do not
impact the controls imposed on the handling of light loads
(including unirradiated fuel assemblies) for ensuring that such
activities cannot result in an event that yields consequences more
severe than those calculated for the DBA-FHA. With respect to
reactivity concerns during refueling operations (MODE 5), all
systems or features required to be OPERABLE for precluding
inadvertent criticality and monitoring reactivity changes will
continue to be required OPERABLE as per the current Technical
Specification requirements. The deletion of CORE ALTERATIONS as an
applicable condition only applies to the noted systems which do not
contribute to precluding reactivity events. Based on the above, the
proposed changes do not involve a significant reduction in the
margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: Vespasian Warner Public
Library, 120 West Johnson Street, Clinton, Illinois 61727
Attorney for licensee: Leah Manning Stetzner, Vice President,
General Counsel, and Corporate Secretary, 500 South 27th Street,
Decatur, Illinois 62525
NRC Project Director: Gail H. Marcus
Maine Yankee Atomic Power Company, Docket No. 50-309, Maine Yankee
Atomic Power Station, Lincoln County, Maine
Date of amendment request: August 12, 1996
Description of amendment request: The proposed amendment would add
an additional circumstance to Exception 2 of Technical Specification
(TS) 3.6, Emergency Core Cooling and Containment Spray Systems, during
which operation of a service water/component cooling pump subsystem is
permitted at reduced flow to flush the service water header or inlet
strainer. The Bases for this TS would be augmented to support the
additional circumstance of reduced service water flow.
The proposed amendment would also modify the valve surveillance
requirements of TS 4.6.A.1.b, Periodic Testing of ECCS Valves, to
provide an exception to surveillance requirements for those locked
valves that are inaccessible during power operations or located in a
locked high radiation area. The Bases for this TS would be augmented to
support the change in surveillance requirements.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration. The NRC staff's analysis is presented below.
1. The proposed amendment does not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
Invocation of the proposed addition to Exception 2 to TS 3.6 would
not alter any associated Remedial Action completion time, nor those of
TS 3.0.A, Nonconformance with a Limiting Condition for Operation. The
evolutions for which this amendment is intended (flushing a heat
exchanger inlet strainer or cleaning a service water header that has
become fouled)are administratively controlled by procedures that
require review and approval by the Plant Operation Review Committee.
The proposed change to TS 4.6.A.1.b would revise the surveillance
requirements for a very limited number of locked manual valves in the
emergency core cooling system (ECCS). The purpose of the surveillance
requirements is unchanged and is intended to verify that locked valves
remain in their correct position. The position of the valves is not
changed and the revised surveillance requirements will continue to
demonstrate ECCS valve operability.
Thus, the proposed amendment does not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
2. The proposed amendment does not create the possibility of a
new or different kind of accident from any accident previously
evaluated.
The proposed addition to Exception 2 to TS 3.6 recognizes that
service water cleaning and flushing are operations that are required to
maintain heat transfer capability and equipment reliability. The
proposed amendment does not affect the design of the plant and do not
permit operation of the plant outside the currently allowed modes of
operation.
The proposed change to TS 4.6.A.1.b maintains verification of ECCS
valve operability, while requiring no changes in system configuration
to perform surveillance testing. System functional performance is not
adversely affected.
Thus, 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.
The proposed change to TS 3.6 does not significantly alter the
availability or condition of applicable equipment and therefore does
not alter the accident analyses or the conclusions associated with
that equipment. The proposed change permits service water flow to be
reduced below that required for operation of the ECCS in the
recirculation mode, for a short time. The time during which flow is
reduced and both the mussel control and flushing evolutions are
administratively controlled by procedures reviewed and approved by
the Plant Operation Review Committee.
The proposed change to TS 4.6.A.1.b maintains verification of valve
operability. Valve position surveillances will continue to be conducted
in accordance with plant Technical Specifications to ensure valve
operational readiness.
Thus, there is no significant reduction in a margin of safety.
Based on this review, it appears that the three standards of 10
CFR 50.92(c) are satisfied. Therefore, the NRC staff proposes to
[[Page 50346]]
determine that the amendment request involves no significant hazards
consideration.
Local Public Document Room location: Wiscasset Public Library, High
Street, P.O. Box 367, Wiscasset, ME 04578
Attorney for licensee: Mary Ann Lynch, Esquire, Maine Yankee Atomic
Power Company, 329 Bath Road, Brunswick, ME 04011 NRC Deputy Director:
John A. Zwolinski
Pacific Gas and Electric Company, Docket Nos. 50-275 and 50-323,
Diablo Canyon Nuclear Power Plant, Unit Nos. 1 and 2, San Luis
Obispo County, California
Date of amendment requests: June 7, 1996
Description of amendment requests: The proposed amendments would
revise the combined Technical Specifications (TS) for the Diablo Canyon
Power Plant, Unit Nos. 1 and 2 by revising Technical Specifications
(TS) 3/4.9.14.1, ``Spent Fuel Assembly Storage - Spent Fuel Pool Region
2,'' and 3/4.9.14.3, ``Spent Fuel Assembly Storage - Spent Fuel Pool
Region 1,'' to allow storage of fuel assemblies in a checkerboard
pattern in region 2 of the spent fuel pool.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. The proposed change does not involve a significant increase
in the probability or consequences of an accident previously
evaluated.
Analysis indicates that allowing fuel storage in a checkerboard
pattern with empty storage cells in region 2 of the spent fuel
pool will not result in an inadvertent criticality event. The
keff will continue to remain below 0.95 as required to meet the
acceptance criteria in the NRC Standard Review Plan, Section 9.1.1.
Therefore, the proposed changes do not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
2. The proposed change does not create the possibility of a new
or different kind of accident from any accident previously
evaluated.
The change to allow fuel storage in a checkerboard pattern with
no minimum burnup requirements in region 2 of the spent fuel pool
would designate locations where a fuel assembly could be incorrectly
placed. However, the incorrect placement of a fuel assembly has been
analyzed and would not cause an inadvertent criticality or any other
accident.
Therefore, the proposed changes do not create the possibility of
a new or different kind of accident from any accident previously
evaluated.
3. The proposed change does not involve a significant reduction
in a margin of safety.
The NRC Standard Review Plan, Section 9.1.1, acceptance
criterion of a keff of 0.95 provides the margin to criticality.
An analysis was performed that concluded that the proposed change to
allow fuel storage in spent fuel pool region 2 in a checkerboard
pattern meets the acceptance criterion.
Therefore, the proposed changes do not involve a significant
reduction in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment requests involve no significant hazards consideration.
Local Public Document Room location: California Polytechnic State
University, Robert E. Kennedy Library, Government Documents and Maps
Department, San Luis Obispo, California 93407
Attorney for licensee: Christopher J. Warner, Esq., Pacific Gas and
Electric Company, P.O. Box 7442, San Francisco, California 94120
NRC Project Director: William H. Bateman
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 amendment request: June 6, 1996 (TS 372)
Description of amendment request: The proposed amendment revises
Section 6 of the Browns Ferry Nuclear Plant Units 1, 2, and 3 technical
specifications. Administrative controls associated with quality
assurance are relocated to the licensee's Nuclear Quality Assurance
Plan, consistent with Administrative Letter 95-06, and provides
revisions that make Section 6 more consistent with the improved
Standard Technical Specifications. Additional administrative changes
are included to ensure consistent terminology within the
specifications, and to update obsolete items such as titles and
addresses. The proposed amendment also includes minor editorial
changes.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. The proposed amendment does not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
The proposed TS change to revise items 1 through 28 above
(Section I, Description of the Proposed Change) was evaluated and
the proposed TS changes were determined to be administrative in
nature. The changes [items 2 through 9, 11, 17 through 21, 23, 26,
and 27] involve administrative title changes of TVA management
positions, the updating of an NRC mailing address and an NRC
regional office title. In addition, certain sections [items 1, 10,
12, 13, 24, and 25] are being relocated into other licensee
documents for which those provisions are adequately controlled by
regulatory requirements. [Items 14, 15, 16, 22, and 28 are editorial
changes.] These changes do not affect any of the design basis
accidents. They do not involve an increase in the probability or
consequences of an accident previously evaluated.
2. The proposed amendment does not create the possibility of a
new or different kind of accident from any accident previously
evaluated.
The proposed TS change to revise items 1 through 28 above
(Section I, Description of the Proposed Change) was evaluated and
the proposed TS changes were determined to be administrative in
nature. The changes involve administrative title changes of TVA
management positions, the updating of an NRC mailing address and an
NRC regional office title. In addition, certain sections are being
relocated into other licensee documents for which those provisions
are adequately controlled by regulatory requirements. These changes
do not affect any of the design basis accidents. No modifications to
any plant equipment are involved. There are no effects on system
interactions made by these changes. They do not create the
possibility of a new or different kind of accident from an accident
previously evaluated.
3. The proposed amendment does not involve a significant
reduction in a margin of safety.
The proposed TS change to revise items 1 through 28 above
(Section I, Description of the Proposed Change) was evaluated and
the proposed TS changes were determined to be administrative in
nature. The changes involve administrative title changes of TVA
management positions, the updating of an NRC mailing address and an
NRC regional office title. In addition, certain sections are being
relocated into other licensee documents for which those provisions
are adequately controlled by regulatory requirements. The margin of
safety as reported in the basis for the TSs is not reduced. The
proposed change is administrative and does not impact any technical
information contained in the bases of the TS.
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: Athens Public Library, South
Street, Athens, Alabama 35611
[[Page 50347]]
Attorney for licensee: General Counsel, Tennessee Valley Authority,
400 West Summit Hill Drive, ET llH, Knoxville, Tennessee 37902
NRC Project Director: Frederick J. Hebdon
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 amendment request: August 30, 1996 (TS 380)
Description of amendment request: The proposed amendment deletes
License Condition 2.C.(3) regarding thermal water quality standards
from the licenses for the Browns Ferry Nuclear Plant Units 1, 2, and 3.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. The proposed amendment does not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
The proposed License Condition change is an adminstrative change
and has no relationship to plant safety analyses. Therefore, this
change does not increase the frequency of the precursors to design
basis events or operational transients analyzed in the BFN [Browns
Ferry Nuclear Plant] Final Safety Analysis Report. Likewise, the
proposed changes will not increase the consequences of an accident
previously evaluated.
2. The proposed amendment does not create the possibility of a
new or different kind of accident from any accident previously
evaluated.
The proposed License Condition change is an administrative
change and has no relationship to plant safety analyses. Thus, the
change does not create any type of new accident sequences. Likewise,
the proposed amendment does not create the possibility of a new or
different kind of accident from any accident previously evaluated.
3. The proposed amendment does not involve a significant
reduction in a margin of safety.
The proposed License Condition change is an administrative
change and has no relationship to plant safety analyses. Therefore,
the proposed amendment does not involve a reduction in the margin of
safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: Athens Public Library, South
Street, Athens, Alabama 35611
Attorney for licensee: General Counsel, Tennessee Valley Authority,
400 West Summit Hill Drive, ET llH, Knoxville, Tennessee 37902
NRC Project Director: Frederick J. Hebdon
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.
Commonwealth Edison Company, Docket Nos. 50-295 and 50-304, Zion
Nuclear Power Station, Units 1 and 2, Lake County, Illinois
Date of amendment request: August 16, 1996
Description of amendment request: This notice relates to your
submittal to remove the uncertainty term from the specified distance
and remove the footnote which specifies the time frame it is
applicable.
Date of publication of individual notice in Federal Register:
September 11, 1996 (61 FR 47968)
Expiration date of individual notice: October 11, 1996
Local Public Document Room location: location: Waukegan Public
Library, 128 N. County Street, Waukegan, Illinois 60085.
Commonwealth Edison Company, Docket Nos. 50-295 and 50-304, Zion
Nuclear Power Station, Units 1 and 2, Lake County, Illinois
Date of amendment request: September 3, 1996
Description of amendment request: This notice relates to your
submittal to modify Technical Specification Section 4.3.1.B.4.A.10.a
which provides the acceptance criteria for steam generator tube repairs
by adding a footnote which references the cleanliness and
nondestructive examination requirements as described in CEN-629-P,
Revision 00, ``Repair of Westinghouse Series 44 and 51 Steam Generator
Tubes Using Leak Tight Sleeves.'' Date of publication of individual
notice in Federal Register: September 11, 1996 (61 FR 47966)
Expiration date of individual notice: October 11, 1996
Local Public Document Room location: location: Waukegan Public
Library, 128 N. County Street, Waukegan, Illinois 60085.
PECO Energy Company, Public Service Electric and Gas Company,
Delmarva Power and Light Company, and Atlantic City Electric
Company, Docket Nos. 50-277 and 50-278, Peach Bottom Atomic Power
Station, Unit Nos. 2 and 3, York County, Pennsylvania
Date of amendment request: March 25, 1996, as supplemented by
letter dated August 23, 1996
Brief description of amendment request: The proposed amendment
would revise the safety limit minimum critical power ratios (SLMCPRs)
to support use of GE-13 fuel at PBAPS, Units 2 and 3. Date of
publication of individual notice in Federal Register: August 30, 1996
(61 FR 45997)
Expiration date of individual notice: September 30, 1996
Local Public Document Room location: Government Publications
Section, State Library of Pennsylvania, (REGIONAL DEPOSITORY) Education
Building, Walnut Street and Commonwealth Avenue, Box 1601, Harrisburg,
PA 17105.
Pennsylvania Power and Light Company, Docket No. 50-387 Susquehanna
Steam Electric Station, Unit 1, Luzerne County, Pennsylvania
Date of amendment request: May 28, 1996, as supplemented by letter
dated July 25, 1996
Brief description of amendment request: The proposed amendment
would revise the Minimum Critical Power Ratio safety limit values,
adding two references to reflect the use of the ANF-B Critical Power
Ratio Correlation and to reflect the use of the ABB Combustion
Engineering licensing methodology, with a modification to the
associated Bases.
Date of publication of individual notice in Federal Register:
September 9, 1996 (61 FR 47529)
Expiration date of individual notice: October 9, 1996
Local Public Document Room location: Osterhout Free Library,
Reference Department, 71 South Franklin Street, Wilkes-Barre, PA 18701
[[Page 50348]]
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.
Carolina Power & Light Company, Docket No. 50-261, H. B. Robinson
Steam Electric Plant, Unit No. 2, Darlington County, South Carolina
Date of application for amendment: January 30, 1996, as
supplemented May 20, 1996
Brief description of amendment: This amendment revises the
Technical Specifications (TS) to: (1) add TS 4.6.1.5 to provide
criteria for 24-hour full-load testing of the emergency diesel
generators (EDGs) to be performed during each refueling outage; (2)
revise TS 4.6.1.2 to allow testing of the EDG protective bypasses
listed in TS 3.7.1.d to be done independent of the safety injection or
loss of offsite power testing; and (3) revise TS 4.6.1.3 to include the
EDG protective bypass inspection.
Date of issuance: September 11, 1996
Effective date: September 11, 1996
Amendment No. 174
Facility Operating License No. DPR-23. Amendment revises the
Technical Specifications.
Date of initial notice in Federal Register: February 28, 1996 (61
FR 7546) The Commission's related evaluation of the amendment is
contained in a Safety Evaluation dated September 11, 1996. The May 20,
1996, letter provided clarifying information that did not change the
initial proposed no significant hazards consideration determination. No
significant hazards consideration comments received: No
Local Public Document Room location: location: Hartsville Memorial
Library, 147 West College Avenue, Hartsville, South Carolina 29550
Carolina Power & Light Company, Docket No. 50-261, H. B. Robinson
Steam Electric Plant, Unit No. 2, Darlington County, South Carolina
Date of application for amendment: December 10, 1995, as
supplemented August 1, 1996, and September 4, 1996.
Brief description of amendment: This amendment revises Technical
Specification (TS) Section 3.5.1 and Tables 3.5-2, 3, and 4 concerning
the reactor trip system, engineering safety feature actuation system,
and isolation function.
Date of issuance: September 12, 1996Effective date: September 12,
1996
Amendment No. 175
Facility Operating License No. DPR-23. Amendment revises the
Technical Specifications.
Date of initial notice in Federal Register: February 14, 1996 (61
FR 5812). The August 1, 1996, and September 4, 1996, submittals
provided administrative changes to the TS pages that did not change the
initial proposed no significant hazards consideration determination.
The Commission's related evaluation of the amendment is contained in a
Safety Evaluation dated September 12, 1996.No significant hazards
consideration comments received: No
Local Public Document Room location: Hartsville Memorial Library,
147 West College Avenue, Hartsville, South Carolina 29550
Duke Power Company, et al., Docket No. 50-413, Catawba Nuclear
Station, Unit 1, York County, South Carolina
Date of amendment request: September 30, 1994, as supplemented
September 18, 1995, January 19, March 15, May 16, and August 27, 1996
Description of amendment: The amendment revises the Technical
Specifications to reflect the new setpoints, operational parameters,
and approved analysis methodologies associated with replacement of the
Unit 1 steam generators. The amendment also deletes references to steam
generator tube repair methods, which will no longer be applicable after
the replacement, and clarifies initial surveillances.
Date of issuance: August 29, 1996
Effective date: As of the date of issuance, to be implemented
within 30 days
Amendment No.: 151
Facility Operating License No. NPF-35: Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: April 10, 1996 (61 FR
15986) The May 16 and August 27, 1996, letters provided clarifying
information that did not change the scope of the September 30, 1994,
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 August 29, 1996.
No significant hazards consideration comments received: No
Local Public Document Room location: York County Library, 138 East
Black Street, Rock Hill, South Carolina 29730
GPU Nuclear Corporation, et al., Docket No. 50-219, Oyster Creek
Nuclear Generating Station, Ocean County, New Jersey
Date of application for amendment: July 17, 1996, as supplemented
August 28, 1996 (TSCR 242, Rev. 2). This application supersedes
applications dated February 23 (TSCR 242) and June 19, 1996 (TSCR 242,
Rev. 1).
Brief description of amendment: The amendment changes the Technical
Specifications (TS) to allow the implementation of 10 CFR Part 50,
Appendix J, Option B.
Date of Issuance: September 3, 1996
Effective date: September 3, 1996, to be implemented within 30 days
of issuance
Amendment No.: 186
Facility Operating License No. DPR-16. Amendment revises the
Technical Specifications.
Date of initial notice in Federal Register: July 31, 1996 (61 FR
40019) Supersedes notice dated March 27, 1996 (61 FR 13526). The August
28, 1996, supplement provided updated and corrected TS and bases pages.
These
[[Page 50349]]
revisions were within the scope of the original application and did not
change the staff's initial proposed no significant hazards
consideration determination. Therefore renoticing was not warranted.
The Commission's related evaluation of this amendment is contained in a
Safety Evaluation dated September 3, 1996. No significant hazards
consideration comments received: No.
Local Public Document Room location: location: Ocean County
Library, Reference Department, 101 Washington Street, Toms River, NJ
08753
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: February 22, 1996, as
supplemented by letter dated July 3, 1996
Brief description of amendment: The amendment revises the Clinton
Power Station Technical Specifications for the drywell to permit bypass
testing on a 10-year frequency with increased testing if performance
degrades, changes the drywell air lock testing and surveillance
requirements, deletes action notes for the drywell air lock and drywell
isolation valves when the bypass leakage limit is not met, and deletes
the specific leakage limits for the drywell air lock seal.
Date of issuance: September 4, 1996
Effective date: September 4, 1996
Amendment No.: 106
Facility Operating License No. NPF-62: The amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: April 24, 1996 (61 FR
18170) The July 3, 1996, submittal consisted of supporting technical
information which did not change the staff's initial proposed no
significant hazards consideration determination or expand the scope of
the original notice. The Commission's related evaluation of the
amendment is contained in a Safety Evaluation dated September 4, 1996.
No significant hazards consideration comments received: No
Local Public Document Room location: location: The Vespasian Warner
Public Library, 120 West Johnson Street, Clinton, Illinois 61727
North Atlantic Energy Service Corporation, Docket No. 50-443,
Seabrook Station, Unit No. 1, Rockingham County, New Hampshire
Date of amendment request: June 20, 1996
Description of amendment request: The proposed amendment modifies
the Seabrook Station Appendix A Technical Specifications (TSs) for the
Electrical Power Systems, Onsite Power Distribution. Specifically, the
proposed amendment changes TS 3.8.3.1, Action a., to increase from 8
hours to 7 days the allowable time that 480-volt Emergency Bus
E64 may be less than fully energized.
Date of issuance: August 30, 1996
Effective date: As of date of issuance, to be implemented within 60
days.
Amendment No.: 48
Facility Operating License No. NPF-86. Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: June 26, 1996 (61 FR
33142) The Commission's related evaluation of the amendment is
contained in a Safety Evaluation dated August 30, 1996. No significant
hazards consideration comments received: No
Local Public Document Room location: location: Exeter Public
Library, Founders Park, Exeter, NH 03833
Northeast Nuclear Energy Company, Docket No. 50-245, Millstone
Nuclear Power Station, Unit 1, New London County, Connecticut
Date of application for amendment: April 25, 1996
Brief description of amendment: The amendment modifies the
calibration requirement for the source range monitors and intermediate
range monitors by noting that the sensors are excluded.
Date of issuance: August 19, 1996
Effective date: As of the date of issuance, to be implemented
within 30 days.
Amendment No.: 96
Facility Operating License No. DPR-21. Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: June 19, 1996 (61 FR
31183) The Commission's related evaluation of the amendment is
contained in a Safety Evaluation dated August 19, 1996. No significant
hazards consideration comments received: No.
Local Public Document Room location: location: Learning Resources
Center, Three Rivers Community-Technical College, 574 New London
Turnpike, Norwich, CT 06360, and the Waterford Library, ATTN: Vince
Juliano, 49 Rope Ferry Road, Waterford, CT 06385
Northeast Nuclear Energy Company, et al., Docket No. 50-336,
Millstone Nuclear Power Station, Unit No. 2, New London County,
Connecticut
Date of application for amendment: March 28, 1996
Brief description of amendment: The amendment changes Technical
Specification 3.7.7, ``Sealed Source Contamination,'' and its Bases
that modify the criteria for testing sealed sources for contamination
and leakage. The approved changes are consistent with the testing
criteria currently used at the Millstone Nuclear Power Station, Unit
No. 3, the Haddam Neck Plant, and the Seabrook Station.
Date of issuance: September 4, 1996
Effective date: As of the date of issuance, to be implemented
within 60 days.
Amendment No.: 202
Facility Operating License No. DPR-65: Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: May 8, 1996 (61 FR
20853) The Commission's related evaluation of the amendment is
contained in a Safety Evaluation dated September 4, 1996 No significant
hazards consideration comments received: No.
Local Public Document Room location: Learning Resources Center,
Three Rivers Community-Technical College, 574 New London Turnpike,
Norwich, CT 06360, and the Waterford Library, ATTN: Vince Juliano, 49
Rope Ferry Road, Waterford, CT 06385
Pacific Gas and Electric Company, Docket No. 50-133, Humboldt Bay
Power Plant, Unit 3, Humboldt County, California
Date of application for amendment: March 13, 1996
Brief description of amendment: This amendment revised the
Technical Specification by incorporating position changes to reflect a
proposed plant staff reorganization.
Date of issuance: September 6, 1996
Effective date: This license amendment is effective as of the date
of its issuance and must be fully implemented no later than 30 days
from the date of issuance.
Amendment No.: 31Facility License No. DPR-7: This amendment revised
the Technical Specifications
Date of initial notice in Federal Register: April 24, 1996 (61 FR
18174) The Commission's related evaluation of the amendment is
contained in a Safety Evaluation dated September 6, 1996. No
significant hazards consideration comments received: No
Local Public Document Room location: Humboldt County Library, 1313
3rd Street, Eureka, California 95501
[[Page 50350]]
Pennsylvania Power and Light Company, Docket Nos. 50-387 and 50-388
Susquehanna Steam Electric Station, Units 1 and 2, Luzerne County,
Pennsylvania
Date of application for amendments: February 23, 1996, as
supplemented by letter dated June 28, 1996
Brief description of amendments: These amendments change the
Technical Specification Requirement 4.6.2.1d concerning drywell-to-
suppression chamber bypass testing interval to correspond with the
interval for Primary Containment Integrated Leak Rate Testing under 10
CFR Part 50, Appendix J, Option B.
Date of issuance: September 6, 1996
Effective date: September 6, 1996
Amendment Nos.: 160 and 131
Facility Operating License Nos. NPF-14 and NPF-22. The amendments
revised the Technical Specifications.
Date of initial notice in Federal Register: April 10, 1996 (61 FR
15992) The Commission's related evaluation of the amendments is
contained in a Safety Evaluation dated September 6, 1996. No
significant hazards consideration comments received: No
Local Public Document Room location: Osterhout Free Library,
Reference Department, 71 South Franklin Street, Wilkes-Barre, PA 18701
Local Public Document Room location: Osterhout Free Library,
Reference Department, 71 South Franklin Street, Wilkes-Barre, PA 18701
Power Authority of The State of New York, Docket No. 50-286, Indian
Point Nuclear Generating Unit No. 3, Westchester County, New York
Date of application for amendment: June 21, 1996, as supplemented
August 19, 1996, and August 21, 1996.
Brief description of amendment: The amendment extends the
surveillance interval on certain instruments from 18 to 24 months.
Date of issuance: September 5, 1996
Effective date: As of the date of issuance, to be implemented
within 30 days.
Amendment No.: 168
Facility Operating License No. DPR-64: Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: July 31, 1996 (61 FR
49027) The Commission's related evaluation of the amendment is
contained in a Safety Evaluation dated September 5, 1996. No
significant hazards consideration comments received: No
Local Public Document Room location: location: White Plains Public
Library, 100 Martine Avenue, White Plains, New York 10610.
Public Service Electric & Gas Company, Docket No. 50-354, Hope
Creek Generating Station, Salem County, New Jersey
Date of application for amendment: March 6, 1996, as supplemented
by letter dated May 30, 1996.
Brief description of amendment: The amendment changes Technical
Specification (TS) 3.8.1, ``A.C. Sources - Operating,'' to decrease the
minimum fuel oil storage capacity of the Emergency Diesel Generator
Fuel Oil Storage Tanks, from 48,800 to 44,800 gallons. In addition,
footnote ** is deleted from TS 3.8.1.1.b.2. The TS change also adds an
Action Statement to address remedial action when a fuel oil transfer
pump becomes inoperable.
Date of issuance: September 10, 1996
Effective date: As of date of issuance, to be implemented within 90
days.
Amendment No.: 96
Facility Operating License No. NPF-57: This amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: July 3, 1996 (61 FR
34897) The Commission's related evaluation of the amendment is
contained in a Safety Evaluation dated September 10, 1996. No
significant hazards consideration comments received: No
Local Public Document Room location: location: Pennsville Public
Library, 190 S. Broadway, Pennsville, New Jersey 08070
Southern California Edison Company, et al, Docket No. 50-206, San
Onofre Nuclear Generating Station, Unit No. 1, San Diego County,
California
Date of application for amendment: December 22, 1995
Brief description of amendment: The change revises the San Onofre
Unit 1 License Condition to delete a reference to License Condition
2.C(4) from License Condition 2.D. This change eliminates a reporting
requirement for violations of the physical protection plans that is
redundant to reporting requirements in 10 CFR 73.71 and 10 CFR Part 73
Appendix G.
Date of issuance: August 30, 1996
Effective date: August 30, 1996 and shall be implemented no later
than 30 days from August 30, 1996.
Amendment No.: 157
Facility Operating License No. DPR-13: The amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: July 31, 1996 (61 FR
40028) The Commission's related evaluation of the amendment is
contained in a Safety Evaluation dated August 30, 1996. No significant
hazards consideration comments received: No.
Local Public Document Room location: Science Library, University of
California, Irvine, California 92713
Southern Nuclear Operating Company, Inc., Docket Nos. 50-348 and
50-364, Joseph M. Farley Nuclear Plant, Units 1 and 2, Houston
County, Alabama
Date of amendments request: June 12, 1996
Brief description of amendments: The amendments revise the reactor
core safety limits, Overtemperature delta T (OTDT) and Overpressure
delta T (OPDT) reactor trip setpoints and allowable values, and the
power distribution limits associated with implementation of Relaxed
Axial Offset Control (RAOC) and FQ surveillance. The amendments
also include changes to the Bases associated with these specifications
and surveillances.
Date of issuance: September 3, 1996
Effective date: As of the date of issuance to be implemented within
30 days
Amendment Nos.: 121 and 113
Facility Operating License Nos. NPF-2 and NPF-8: Amendments revise
the Technical Specifications.
Date of initial notice in Federal Register: July 31, 1996 (61 FR
40029) The Commission's related evaluation of the amendments is
contained in a Safety Evaluation dated September 3, 1996. No
significant hazards consideration comments received: No
Local Public Document Room location: location: Houston-Love
Memorial Library, 212 W. Burdeshaw Street, Post Office Box 1369,
Dothan, Alabama 36302
Southern Nuclear Operating Company, Inc., Docket Nos. 50-348 and
50-364, Joseph M. Farley Nuclear Plant, Units 1 and 2, Houston
County, Alabama
Date of amendments request: June 20, 1996
Brief description of amendments: The amendments revise the
Technical Specifications to reflect the implementation of 10 CFR Part
50, Appendix J, Option B.
Date of issuance: September 3, 1996
Effective date: As of the date of issuance to be implemented within
30 days
Amendment Nos.: 122 and 114
Facility Operating License Nos. NPF-2 and NPF-8: Amendments revise
the Technical Specifications.
Date of initial notice in Federal Register: July 31, 1996 (61 FR
40030) The Commission's related evaluation of the amendments is
contained in a Safety Evaluation dated September 3, 1996. No
significant hazards consideration comments received: No
[[Page 50351]]
Local Public Document Room location: location: Houston-Love
Memorial Library, 212 W. Burdeshaw Street, Post Office Box 1369,
Dothan, Alabama 36302
Tennessee Valley Authority, Docket Nos. 50-390 Watts Bar Nuclear
Plant, Unit 1, Rhea County, Tennessee
Date of application for amendment: July 31, 1996
Brief description of amendment: The amendment revises Technical
Specification 3.6.12 to allow a one-time extension of the 3-month
surveillance requirement for the ice condenser lower inlet doors.
Date of issuance: September 9, 1996
Effective date: As of the date of issuance, to be implemented
within 30 days
Amendment No.: 3
Facility Operating License No. NPF-90: Amendment revises the
Technical Specifications.
Date of initial notice in Federal Register: August 8, 1996 (61 FR
41431) The Commission's related evaluation of the amendment is
contained in a Safety Evaluation dated September 9, 1996. No
significant hazards consideration comments received: None
Local Public Document Room location: Chattanooga-Hamilton County
Library, 1001 Broad Street, Chattanooga, TN 37402
Wisconsin Public Service Corporation, Docket No. 50-305, Kewaunee
Nuclear Power Plant, Kewaunee County, Wisconsin
Date of application for amendment: February 19, 1996, as
supplemented on July 3 and August 26, 1996
Brief description of amendment: The amendment revises Kewaunee
Nuclear Power Plant Technical Specification Section 4.2 and its
associated basis by allowing the application of a voltage-based repair
limit for the steam generator tube support plate intersections
experiencing outside diameter stress corrosion cracking. The repair
criteria are based on guidance provided in Generic Letter 95-05,
``Voltage-Based Repair Criteria for Westinghouse Steam Generator Tubes
affected by Outside Diameter Stress Corrosion Cracking,'' dated August
3, 1995, and on associated industry guidance.
Date of issuance: September 11, 1996
Effective date: September 11, 1996, and is to be implemented within
30 days of the date of issuance.
Amendment No.: 126
Facility Operating License No. DPR-43: Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: April 10, 1996 (61 FR
15999) The July 3 and August 26, 1996, submittals provided clarifying
information that did not change the initial proposed no significant
hazards consideration determination. The Commission's related
evaluation of the amendment is contained in a Safety Evaluation dated
September 11, 1996. No significant hazards consideration comments
received: No.
Local Public Document Room location: University of Wisconsin,
Cofrin Library, 2420 Nicolet Drive, Green Bay, Wisconsin 54311-7001
Wolf Creek Nuclear Operating Corporation, Docket No. 50-482, Wolf
Creek Generating Station, Coffey County, Kansas
Date of amendment request: October 24, 1995, and superseded by
letter dated May 16, 1996.
Brief description of amendment: The amendment adopts ASTM D3803-
1989 as the laboratory testing standard for charcoal samples from the
charcoal absorbers in the control room filtration system, control
building pressurization system, and the auxiliary/fuel building
emergency exhaust system. The output of the heaters in the control
building pressurization system is reduced from a nominal 15kW to a
nominal 5kW and the acceptance criterion for the testing of the
charcoal absorbers is changed.
Date of issuance: September 4, 1996
Effective date: September 4, 1996, to be implemented within 120
days of the date of issuance.
Amendment No.: 102
Facility Operating License No. NPF-42. The amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: June 5, 1996 (61 FR
28622) The Commission's related evaluation of the amendment is
contained in a Safety Evaluation dated September 4, 1996. No
significant hazards consideration comments received: No.
Local Public Document Room location: locations: Emporia State
University, William Allen White Library, 1200 Commercial Street,
Emporia, Kansas 66801 and Washburn University School of Law Library,
Topeka, Kansas 66621 Dated at Rockville, Maryland, this 18th day of
September 1996.
For the Nuclear Regulatory Commission
Steven A. Varga,
Director, Division of Reactor Projects - I/II Office of Nuclear Reactor
Regulation
[Doc. 96-24413 Filed 9-24-96; 8:45 am]
BILLING CODE 7590-01-F
