[Federal Register Volume 61, Number 138 (Wednesday, July 17, 1996)]
[Notices]
[Pages 37295-37307]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: X96-20717]
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NUCLEAR REGULATORY COMMISSION
Biweekly Notice
Applications and Amendments to Facility Operating Licenses
Involving No Significant Hazards Considerations
I. Background
Pursuant to Public Law 97-415, the U.S. Nuclear Regulatory
Commission (the Commission or NRC staff) is publishing this regular
biweekly notice. Public Law 97-415 revised section 189 of the Atomic
Energy Act of 1954, as amended (the Act), to require the Commission to
publish notice of any amendments issued, or proposed to be issued,
under a new provision of section 189 of the Act. This provision grants
the
[[Page 37296]]
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 June 22, 1996, through July 5, 1996. The
last biweekly notice was published on July 3, 1996 (61 FR 34884).
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 August 16, 1996, the licensee may file a request for a hearing
with respect to issuance of the amendment to the subject facility
operating license and any person whose interest may be affected by this
proceeding and who wishes to participate as a party in the proceeding
must file a written request for a hearing and a petition for leave to
intervene. Requests for a hearing and a petition for leave to intervene
shall be filed in accordance with the Commission's ``Rules of Practice
for Domestic Licensing Proceedings'' in 10 CFR Part 2. Interested
persons should consult a current copy of 10 CFR 2.714 which is
available at the Commission's Public Document Room, the Gelman
Building, 2120 L Street, NW., Washington, DC and at the local public
document room for the particular facility involved. If a request for a
hearing or petition for leave to intervene is filed by the above date,
the Commission or an Atomic Safety and Licensing Board, designated by
the Commission or by the Chairman of the Atomic Safety and Licensing
Board Panel, will rule on the request and/or petition; and the
Secretary or the designated Atomic Safety and Licensing Board will
issue a notice of a hearing or an appropriate order.
As required by 10 CFR 2.714, a petition for leave to intervene
shall set forth with particularity the interest of the petitioner in
the proceeding, and how that interest may be affected by the results of
the proceeding. The petition should specifically explain the reasons
why intervention should be permitted with particular reference to the
following factors: (1) the nature of the petitioner's right under the
Act to be made a party to the proceeding; (2) the nature and extent of
the petitioner's property, financial, or other interest in the
proceeding; and (3) the possible effect of any order which may be
entered in the proceeding on the petitioner's interest. The petition
should also identify the specific aspect(s) of the subject matter of
the proceeding as to which petitioner wishes to intervene. Any person
who has filed a petition for leave to intervene or who has been
admitted as a party may amend the petition without requesting leave of
the Board up to 15 days prior to the first prehearing conference
scheduled in the proceeding, but such an amended petition must satisfy
the specificity requirements described above.
Not later than 15 days prior to the first prehearing conference
scheduled in the proceeding, a petitioner shall file a supplement to
the petition to intervene which must include a list of the contentions
which are sought to be litigated in the matter. Each contention must
consist of a specific statement of the issue of law or fact to be
raised or controverted. In addition, the petitioner shall provide a
brief explanation of the bases of the contention and a concise
statement of the alleged facts or expert opinion which support the
contention and on which the petitioner intends to rely in proving the
contention at the hearing. The petitioner must also provide references
to those specific sources and documents of which the petitioner is
aware and on which the petitioner intends to rely to establish those
facts or expert opinion. Petitioner must provide sufficient information
to show that a genuine dispute exists with the applicant on a material
issue of law or fact. Contentions shall be limited to matters within
the scope of the amendment under consideration. The contention must be
one which, if proven, would entitle the petitioner to relief. A
petitioner who fails to file such a supplement which satisfies these
requirements with respect to at least one contention will not be
permitted to participate as a party.
Those permitted to intervene become parties to the proceeding,
subject to any limitations in the order granting leave to intervene,
and have the opportunity to participate fully in the conduct of the
hearing, including the opportunity to present evidence and cross-
examine witnesses.
If a hearing is requested, the Commission will make a final
determination on the issue of no significant hazards consideration. The
final determination will serve to decide when the hearing is held.
If the final determination is that the amendment request involves
no significant hazards consideration, the Commission may issue the
amendment and make it immediately effective,
[[Page 37297]]
notwithstanding the request for a hearing. Any hearing held would take
place after issuance of the amendment.
If the final determination is that the amendment request involves a
significant hazards consideration, any hearing held would take place
before the issuance of any amendment.
A request for a hearing or a petition for leave to intervene must
be filed with the Secretary of the Commission, U.S. Nuclear Regulatory
Commission, Washington, DC 20555-0001, Attention: Docketing and
Services Branch, or may be delivered to the Commission's Public
Document Room, the Gelman Building, 2120 L Street, NW., Washington DC,
by the above date. Where petitions are filed during the last 10 days of
the notice period, it is requested that the petitioner promptly so
inform the Commission by a toll-free telephone call to Western Union at
1-(800) 248-5100 (in Missouri 1-(800) 342-6700). The Western Union
operator should be given Datagram Identification Number N1023 and the
following message addressed to (Project Director): petitioner's name
and telephone number, date petition was mailed, plant name, and
publication date and page number of this Federal Register notice. A
copy of the petition should also be sent to the Office of the General
Counsel, U.S. Nuclear Regulatory Commission, Washington, DC 20555-0001,
and to the attorney for the licensee.
Nontimely filings of petitions for leave to intervene, amended
petitions, supplemental petitions and/or requests for a hearing will
not be entertained absent a determination by the Commission, the
presiding officer or the Atomic Safety and Licensing Board that the
petition and/or request should be granted based upon a balancing of
factors specified in 10 CFR 2.714(a)(1)(i)-(v) and 2.714(d).
For further details with respect to this action, see the
application for amendment which is available for public inspection at
the Commission's Public Document Room, the Gelman Building, 2120 L
Street, NW., Washington, DC, and at the local public document room for
the particular facility involved.
Carolina Power & Light Company, et al., Docket Nos. 50-325 and 50-
324, Brunswick Steam Electric Plant, Units 1 and 2, Brunswick
County, North Carolina
Date of amendments request: April 4, 1996
Description of amendments request: The proposed amendments would
revise the Technical Specifications (TS) to add an allowance to
complete a TS-required surveillance within 24 hours of discovery of a
missed surveillance in accordance with the guidance of Generic Letter
(GL) 87-09, ``Sections 3.0 and 4.0 of the Standard Technical
Specifications (STS) on the Applicability of Limiting Conditions for
Operation and Surveillance Requirements'' and NUREG-1433, ``Standard
Technical Specifications, General Electric Plants, BWR/4,'' Revision 1,
April 1995. Typographical errors are being corrected and wording
adjustments are being incorporated for consistency between plant TS
terminology and the associated Bases.
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 amendments do not involve a significant increase
in the probability or consequences of an accident previously
evaluated. The operational flexibility resulting from the proposed
revision to Technical Specification 3.0.4 is consistent with that
allowed by the existing individual LCO [limiting condition for
operation] and their associated ACTION requirements, which provide
an acceptable level of safety for continued operation. A delay of up
to 24 hours or the time of the surveillance interval, whichever is
less, provided by Technical Specification 4.0.3 to complete a missed
surveillance reduces the probability of a transient occurring when
the affected system or component is either out of service to allow
performance of the surveillance test, or there is a lower level of
confidence in the operability because the normal surveillance was
exceeded. The revision to Technical Specification 4.0.4 makes it
clear that Technical Specification 4.0.4 does not prevent passage
through or to OPERATIONAL CONDITIONS as required to comply with
ACTION requirements. The revision to the wording in Unit 2 Technical
Specification Table 3.12.1-1, Notation (h), revisions to the Bases
of the Technical Specifications, and the elimination of specific
exemptions to Technical Specifications 3.0.4 are administrative in
nature.
Based on the above, the proposed license amendments do not
involve a significant increase in the probability or consequences of
an accident previously evaluated.
2. The proposed amendments do not create the possibility of a
new or different kind of accident from any accident previously
evaluated. The proposed license amendments do not introduce any new
equipment nor do they require any existing equipment or systems to
perform a different type of function than they are presently
designed to perform. The proposed changes result in improved
Technical Specifications by removing unnecessary restrictions on
changes in OPERATIONAL CONDITIONS and facility operation, removing
unnecessary shutdowns caused by inadvertently exceeding surveillance
intervals, and removing conflicts between various Technical
Specifications. The revision to the wording in Unit 2 Technical
Specification Table 3.12.1-1, Notation (h), revisions to the Bases
of the Technical Specifications, and the elimination of specific
exemptions to Technical Specification 3.0.4 are administrative in
nature.
Based on the above, the proposed license amendments do not
create a new or different kind of accident from any previously
evaluated.
3. The proposed license amendments do not involve a significant
reduction in a margin of safety. The operational flexibility that
results from the proposed revision to Technical Specification 3.0.4
is consistent with that allowed by the existing individual LCO and
associated ACTION requirements, which provide an acceptable level of
safety for continued operation. Therefore, there is no change in the
margin of safety associated with this change. A delay of up to 24
hours or the length of the surveillance interval, whichever is less,
provided by Technical Specification 4.0.3 to complete a missed
surveillance reduces the probability of a transient occurring when
the affected system or component is either out of service to allow
performance of the surveillance test, or there is a lower level of
confidence in the operability because the normal surveillance was
exceeded. In addition, the proposed change acknowledges that the
most common outcome of the performance of a surveillance is the
successful demonstration that acceptance criteria are met. The
proposed change provides the potential benefit of avoiding a
shutdown transient when required equipment is still capable of
performing its function, and variables are still within limits. The
revision to Technical Specification 4.0.4 makes it clear that
Technical Specification 4.0.4 does not prevent passage through or to
OPERATIONAL CONDITIONS as required to comply with ACTION
requirements. This change is considered to be a clarification to
achieve consistency with existing Technical Specification
requirements. The revision to the wording in Unit 2 Technical
Specification Table 3.12.1-1, Notation (h), revisions to the Bases
of the Technical Specifications, and the elimination of specific
exemptions to Technical Specification 3.0.4 are administrative in
nature.
The proposed changes would result in improved Technical
Specifications and eliminate unnecessary plant challenges. Based on
the above, the proposed license amendments 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: University of North Carolina
at Wilmington, William Madison Randall Library, 601 S. College Road,
Wilmington, North Carolina 28403-3297
[[Page 37298]]
Attorney for licensee: William D. Johnson, Vice President and
Senior Counsel, Carolina Power & Light Company, Post Office Box 1551,
Raleigh, North Carolina 27602
NRC Project Director: Eugene V. Imbro
Consumers Power Company, Docket No. 50-255, Palisades Plant, Van
Buren County, Michigan
Date of amendment request: December 6, 1995
Description of amendment request: The proposed amendment would
relocate the crane operation and movement of heavy loads requirements
and their bases from the Technical Specifications (TS) to other plant
documents.
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 change moves the requirements from TS to other plant
documents controlled under 10 CFR 50.59 without affecting their
technical content. Since this change does not alter the technical
content of any requirements, the operation of the facility in
accordance with the proposed change cannot involve a significant
increase in the probability or consequences of an accident
previously evaluated, create the possibility of a new or different
kind of accident from any previous evaluated, or 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: Van Wylen Library, Hope
College, Holland, Michigan 49423.
Attorney for licensee: Judd L. Bacon, Esquire, Consumers Power
Company, 212 West Michigan Avenue, Jackson, Michigan 49201
NRC Project Director: Mark Reinhart
Duke Power Company, Docket Nos. 50-269, 270 and 50-287, Oconee
Nuclear Station, Units 1, 2 and 3, Oconee County, South Carolina
Date of amendment request: June 6, 1996
Description of amendment request: The proposed change would remove
the Engineered Safeguard (ES) signals that presently open the outlet
valves on the Low Pressure Service Water (LPSW) System coolers, LPSW-4
and LPSW-5, on high reactor coolant system pressure or high reactor
building pressure. The valves will continue to be operable from the
control room when needed. The proposed change to Technical
Specification (TS) 4.5.1.1.2.a.(2) would require that the refueling
outage test signal be applied to the LPSW pumps, but no longer to LPSW-
4 and LPSW-5, and that the operability of the valves be verified by
cycling them from the control room. A note would be added to reflect
that the refueling outage test of LPSW-4 and LPSW-5 response to the ES
signal will continue to be verified until the signal is removed from
the ES system for each unit during the specified refueling outages. In
addition, TS 4.5.1.1.2.b would be clarified to differentiate between
test acceptance criteria for automatic actuation of the appropriate
LPSW pumps and valves in response to the ES signal, and completion of
travel of LPSW-4 and LPSW-5 in response to manual operation of the
valves. A proposed change to the Bases would also reflect these
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:
Pursuant to 10CFR50.91, Duke Power Company (Duke) has made the
determination that this amendment involves a No Significant Hazards
Consideration by applying the standards established by NRC
regulations in 10CFR50.92. The following discusses the basis for our
analysis:
Will operation of the facility in accordance with the proposed
amendment:
A. Involve a significant increase in the probability or
consequences of an accident previously evaluated?
No. Eliminating the automatic signal that opens Low Pressure
Service Water (LPSW) System valves, LPSW-4 and LPSW-5, upon an
Engineered Safeguards (ES) actuation does not increase the
probability of any accident previously evaluated. The proposed
change would involve a delay in providing cooling water to the Low
Pressure Injection (LPI) System coolers after a design basis
accident. Cooling water flow to the LPI coolers is isolated during
normal power operation. During normal cold shutdown conditions,
cooling water flow to the LPI coolers is normally open without
relying on the ES actuation signal. This cooling water flow is
needed to mitigate certain accidents, but a delay in providing this
cooling water flow after a design basis accident does not
significantly increase the probability of any accident previously
evaluated.
Eliminating the ES actuation signal for LPSW-4 and LPSW-5 will
not increase the consequences of an accident previously evaluated.
After a loss of coolant accident (LOCA), operators will operate the
appropriate valves from the control room in sufficient time to
provide adequate cooling water flow to maintain containment
temperature and pressure within acceptable limits. Duke has also
evaluated the delay of LPSW cooling flow's impact on core cooling
and concluded that there are no adverse impacts on the capability to
maintain core cooling. Since the containment temperature and
pressure limits after a LOCA will not be exceeded, this change will
not increase any potential off-site dose consequences after a LOCA.
Due to the time available for operator action (approximately one
hour), there is no significant increase in operator burden during
this accident scenario.
B. Create the possibility of a new or different kind of accident
from the accidents previously evaluated?
No. As stated above, due to the time available for operator
action (approximately 1 hour), there is no significant increase in
operator burden during this accident scenario. Eliminating the ES
signal that automatically opens valves LPSW-4 and LPSW-5 results in
significantly lower flow demand on the LPSW pumps. If all LPSW pumps
are successfully started, this could result in a stronger pump
causing deadhead conditions on a weaker pump since the pumps feed
into the same piping system. To prevent any potential adverse
effects on the LPSW pumps due to inadequate flow during the initial
stages of a LOCA, minimum flow piping will be installed for the LPSW
pumps to provide adequate flowpaths for pump minimum flow. Testing
will be performed to validate that the LPSW pumps can operate at the
chosen design value for pump minimum flow. In addition, Duke
conducted an evaluation, based on manufacturer input, of the thermal
effects on the LPI coolers due to delaying LPSW cooling flow. This
evaluation concluded that the 30 minute delay of LPSW cooling flow
has no adverse thermal effects on the LPI coolers. Therefore,
because there is no significant increase in operator burden and
because there will be no adverse effects on the LPSW pumps, LPI
coolers, and associated piping caused by the delayed LPSW cooling
flow, the proposed change will not create the possibility of a new
or different kind of accident from the accidents previously
evaluated.
C. Involve a significant reduction in a margin of safety?
No. There are no safety limits or limiting safety system
settings associated with the LPSW System in the Oconee Nuclear
Station Technical Specifications. The proposed change will not
affect any existing safety limits or limiting safety system
settings. The proposed change will not affect any existing Limiting
Conditions for Operation in the Technical Specifications. The
proposed change involves an alternative method of initiating cooling
water flow to the LPI coolers after a LOCA. This alternative method
will achieve the required results since there will be no significant
change in the containment temperature and pressure after a LOCA.
Duke has concluded based on the above that there are no
significant hazards considerations involved in this amendment
request.
The NRC has reviewed the licensee's analysis and, based on this
review, it
[[Page 37299]]
appears that the three standards of 10 CFR 50.92(c) are satisfied.
Therefore, the NRC staff proposes to determine that the amendment
request involves no significant hazards consideration.
Local Public Document Room location: Oconee County Library, 501
West South Broad Street, Walhalla, South Carolina 29691
Attorney for licensee: J. Michael McGarry, III, Winston and Strawn,
1200 17th Street, NW., Washington, DC 20036
NRC Project Director: Herbert N. Berkow
Entergy Operations, Inc., et al., Docket No. 50-416, Grand Gulf
Nuclear Station (GGNS), Unit 1, Claiborne County, Mississippi
Date of amendment request: June 20, 1996
Description of amendment request: The amendment would redefine the
secondary containment boundary to allow the enclosure building to be
inoperable during the upcoming refueling outage 8 (RFO 8) scheduled to
begin in October 1996. The amendment would add a condition to the
license that the enclosure building may be inoperable during core
alterations and movement of non-recently irradiated fuel (i.e., fuel
that has not occupied part of a critical reactor core for 12 days)
during RFO 8 and the standby gas treatment (SGT) system may be unable
to automatically start or achieve and maintain the required vacuum,
provided the following conditions exist:
a. All dampers communicating between the auxiliary building and the
enclosure building are closed.
b. The access door between the auxiliary building and the enclosure
building is closed, except when the access opening is being used for
entry and exit.
c. The SGT system is blocked from automatic initiation.
d. SGT system is available for manual initiation or the actions for
Limiting Condition for Operation 3.6.4.3 in the Technical
Specifications for GGNS are complied with.
The non-recently irradiated fuel is spent fuel that has decayed at
least 12 days after the reactor was shut down for refueling.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. The proposed changes do not significantly increase the
probability or consequences of an accident previously evaluated.
The equipment affected by the proposed change is not considered
an initiator to any previously analyzed accident, therefore,
inoperability of the equipment does not increase the probability of
any previously evaluated accident.
As described in Updated Final Safety Analysis Report [for GGNS,]
Chapter 15, the accidents postulated to occur during core
alterations in addition to fuel handling accidents are [the
following]: inadvertent criticality due to a control rod removal
error or continuous control rod withdrawal error during refueling
and the inadvertent loading of a fuel assembly in an improper
location. These events are not postulated to result in fuel cladding
integrity damage. The only accident postulated to occur during core
alterations that results in a significant radioactive release is the
fuel handling accident. The proposed requirements in conjunction
with existing administrative controls on light loads, bounds the
conditions of the current design basis fuel handling accident
analysis which concludes that the radiological consequences are
within the acceptance criteria of NUREG 0800, Section 15.7.4 and
General Design Criteria [GDC] 19 [of Appendix A to 10 CFR Part 50].
Therefore, the proposed changes do not significantly increase
consequences of any previously evaluated accident.
Based on the above, the proposed changes do not significantly
increase the probability or consequences of any accident previously
evaluated.
2. The proposed changes would not create the possibility of a
new or different kind of accident from any previous analyzed.
The leaktightness of the enclosure building does not affect the
function of any plant system other than the ability of the SGT
System to ensure the secondary containment is at the specified
pressure. The proposed change in [the] normal SGT System
alignment[,] by defeating the automatic start feature of the SGT
System and the inability to ensure secondary containment is at the
specified pressure[,] does not affect the operation of any [other]
plant system or component. The SGT System is not relied upon to
provide normal or accident cooling to plant systems or components.
The function of the enclosure building and the SGT System is only to
mitigate the release of radioactivity to the environment in the
event of an accident.
Therefore, the proposed changes do not create the possibility of
a new or different kind of accident from any accident previously
analyzed.
3. The proposed changes do not involve a significant reduction
in a margin of safety.
The proposed changes continue to ensure that the radiological
consequences are at or below the current GGNS licensing limit.
Safety margins and analytical conservatisms have been evaluated and
are well understood. Substantial margins are retained to ensure that
the analysis adequately bounds all postulated event scenarios. The
current margin of safety is retained.
Specifically, the margin of safety for the fuel handling
accident is the difference between the 10CFR100 [dose consequence
guidelines of 300 rem thyroid and 25 rem whole- body] and the
licensing limit defined by NUREG-0800, Section 15.7.4. With respect
to the control room personnel doses, the margin of safety is the
difference between the 10CFR100 [guidelines] and the licensing limit
defined by 10CFR50 [10 CFR Part 50], Appendix A, Criterion 19 (GDC
19). The proposed applicability continues to ensure that the whole-
body and thyroid doses at the exclusion area and low population zone
boundaries[,] as well as control room doses[,] are at or below the
corresponding licensing limit. The margin of safety is unchanged;
therefore, the proposed changes do not involve a significant
reduction in a margin of safety.
In excess to the margin of safety supplied by the licensing
limits of NUREG-0800 and GDC 19, the proposed change incorporates an
additional layer of conservative requirements. The proposed change
leaves in effect a redefined secondary containment boundary which
will provide a low leakage boundary (consisting of the primary
containment and the auxiliary building) by automatically isolating
in the event of the design basis fuel handling accident and requires
that the SGT System be available for manual initiation when desired.
These requirements will ensure that doses will be even lower than
those calculated.
Therefore, the proposed changes do not result in a significant
reduction in a margin of safety.
Based on the above evaluation, operation in accordance with the
proposed amendment involves no significant hazards considerations.
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
Georgia Power Company, Oglethorpe Power Corporation, Municipal
Electric Authority of Georgia, City of Dalton, Georgia, Docket Nos.
50-424 and 50-425, Vogtle Electric Generating Plant, Units 1 and 2,
Burke County, Georgia
Date of amendment request: June 17, 1996
Description of amendment request: The proposed amendments would
revise Technical Specification Section 5.3.1 to allow use of fuel
assemblies containing fuel rods clad with ZIRLOTM.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the
[[Page 37300]]
licensee has provided its analysis of the issue of no significant
hazards consideration, which is presented below:
The proposed change does not involve a significant increase in
the probability or consequences of an accident previously evaluated
because:
The methodologies used in the accident analyses remain
unchanged. The proposed change does not change or alter the design
assumptions for the systems or components used to mitigate the
consequences of an accident. Use of ZIRLOTM fuel cladding does
not adversely affect fuel performance or impact nuclear design
methodology. Therefore, accident analysis results are not
significantly impacted.
The operating limits will not be changed and the analysis
methods to demonstrate operation within the limits will remain in
accordance with NRC-approved methodologies. Other than the changes
to the fuel assemblies cladding, there are no physical changes to
the plant associated with this Technical Specification change. A
safety analysis will continue to be performed for each specific
reload cycle to demonstrate compliance with all fuel safety design
bases.
The 10 CFR 50.46 criteria are applied to the ZIRLOTM clad
fuel rods. The use of these fuel assemblies will not result in a
change to the reload design and safety analysis limits. Since the
original design criteria are met, the ZIRLOTM clad fuel rods
will not be an initiator for any new accident. The clad material is
similar in chemical composition and has similar physical and
mechanical properties as Zircaloy-4. Thus, the cladding integrity is
maintained and the structural integrity of the fuel assembly is not
affected. ZIRLOTM cladding improves corrosion performance and
dimensional stability. Since the dose predictions in the safety
analyses are not sensitive to the fuel rod cladding material used,
the radiological consequences of accidents previously evaluated in
the safety analysis remain valid.
The proposed change does not create the possibility of a new or
different kind of accident from any accident previously evaluated
because:
The possibility for a new or different kind of accident from any
accident previously evaluated is not created since the fuel
assemblies containing ZIRLOTM clad fuel rods will satisfy the
same design bases as that currently used for Zircaloy-4 clad fuel
assemblies. All design and performance criteria will continue to be
met and no new single failure mechanisms have been defined. In
addition, the use of ZIRLOTM fuel assemblies does not involve
any alterations to plant equipment or procedures which would
introduce any new or unique operational mode or accident precursor.
Therefore, the possibility for a new or different kind of accident
from any accident previously evaluated is not created.
The proposed change does not involve a significant reduction in
a margin of safety because:
The margin of safety is not significantly reduced since the
ZIRLOTM clad fuel assemblies will not change the reload design
and safety analysis limits. Their use will take into consideration
the normal core operating conditions allowed for in the Technical
Specifications. Each specific cycle's reload core will continue to
be specifically evaluated using NRC approved reload design methods
and approved fuel rod design models. This will include consideration
of the core physics analysis peaking factor and core average linear
heat rate effects.
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: BurkeCounty Public Library,
412 Fourth Street, Waynesboro, Georgia 30830
Attorney for licensee: Mr. Arthur H. Domby, Troutman Sanders,
NationsBank Plaza, Suite 5200, 600 Peachtree Street, NE., Atlanta,
Georgia 30308
NRC Project Director: Herbert N. Berkow
Georgia Power Company, Oglethorpe Power Corporation, Municipal
Electric Authority of Georgia, City of Dalton, Georgia, Docket Nos.
50-424 and 50-425, Vogtle Electric Generating Plant, Units 1 and 2,
Burke County, Georgia
Date of amendment request: June 17, 1996
Description of amendment request: The proposed amendments would
clarify the requirement of Technical Specification Surveillance
Requirement 4.8.1.1.2.j(2) that requires a pressure test of those
portions of the diesel fuel-oil system that are designed to Section
III, Subsection ND of the American Society of Mechanical Engineers
(ASME) Code. The system pressure test would be performed at a pressure
of 110% of the design pressure, at least once per 10 years and only on
those sections of piping that are isolable.
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 clarification of T/S [Technical Specification]
4.8.1.1.2.j(2) does not involve a significant hazards consideration
because operation of [the Vogtle Electric Generating Plant] with
this change would not:
1. Involve a significant increase in the probability or
consequences of an accident previously evaluated. The configuration
of the diesel fuel-oil system as currently installed and operated is
such that a pressure test of 110% of design pressure would be
impractical to perform. The system contains tanks designed for
atmospheric pressure and isolation of them and their vent lines from
the specified pressure test is not practical. The ASME Code, Section
XI, provides alternate test methods to use when storage tanks are
involved in a system pressure test. By clarifying this T/S
requirement, the requirements set forth in ASME Section XI can be
utilized as guidance for testing requirements to ensure the
integrity of the diesel fuel-oil system to perform its intended
safety function.
2. Create the possibility of a new or different kind of accident
from any accident previously evaluated. There are no design changes
being made that would create a new type of accident or malfunction
and the method and manner of plant operation remain unchanged. Using
ASME Section XI as guidance for pressure testing the isolable
sections of piping provides assurance that the fuel oil supply
system will perform its intended function.
3. Involve a significant reduction in a margin of safety. There
are no changes being made to the safety limits or safety system
settings that would adversely impact plant safety. Utilizing ASME
Section XI as guidance for determining those sections of piping that
should be pressure-tested and atmospheric-tested will ensure proper
operation of the diesel generator fuel oil supply system.
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: Burke County Public Library,
412 Fourth Street, Waynesboro, Georgia 30830
Attorney for licensee: Mr. Arthur H. Domby, Troutman Sanders,
NationsBank Plaza, Suite 5200, 600 Peachtree Street, NE., Atlanta,
Georgia 30308
NRC Project Director: Herbert N. Berkow
GPU Nuclear Corporation, et al., Docket No. 50-289, Three Mile
Island Nuclear Station, Unit No. 1, Dauphin County, Pennsylvania
Date of amendment request: April 10, 1996
Description of amendment request: The proposed changes bring the
surveillance requirements to conformance with Amendment No. 196 issued
September 19, 1995. Additionally, this request changes frequency
notation for a group of 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
[[Page 37301]]
consideration (SHC), which is presented below:
1. Operation of the facility in accordance with the proposed
amendment would not involve a significant increase in the
probability of occurrence of the consequences of an accident
previously evaluated.
The proposed amendment extends the interval between successive
refueling interval surveillances to once every 24 months for those
surveillances evaluated herein, and to make administrative changes
serving to conform the Technical Specifications to Amendment No.
196. Except for the administrative changes, the proposed
surveillance interval changes do not involve any change to the
actual surveillance requirements, nor does it involve any
change to the limits and restrictions on plant operations. The
reliability of systems and components relied upon to prevent or
mitigate the consequences of accidents previously evaluated is not
degraded by the proposed change to the surveillance interval.
Assurance of system and equipment availability is maintained. This
change does not involve any change to system or equipment
configuration. Therefore, this change does not increase the
probability of occurrence or the consequences of an accident
previously evaluated.
2. Operation of the facility in accordance with the proposed
amendment would not create the possibility of a new or different
kind of accident from any accident previously evaluated.
The proposed amendment extends the interval between successive
refueling interval surveillances to once every 24 months for those
surveillances evaluated herein, and to make administrative changes
serving to conform the Technical Specifications to Amendment No.
196. Except for the administrative changes the proposed surveillance
interval changes do not involve any change to the limits and
restrictions in plant operation. This change does not involve any
change to system or equipment configuration. Therefore, this change
is unrelated to the possibility of creating a new or different kind
of accident from any previously evaluated.
3. Operation of the facility in accordance with the proposed
amendment would not involve a significant reduction in a margin of
safety.
The proposed amendment extends the interval between successive
refueling interval surveillances to once every 24 months for the
surveillances evaluated herein, and to make administrative changes
serving to conform the Technical Specifications to Amendment No.
196. Except for the administrative changes the proposed surveillance
interval changes do not involve any change to the actual
surveillance requirements, nor does it involve any change to the
limits and restrictions on plant operation. The reliability of
systems and components is not degraded by the proposed change to the
surveillance interval. Assurance of system and equipment
availability is maintained. Therefore, it is concluded that
operation of the facility in accordance with the proposed amendment
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: Law/Government Publications
Section, State Library of Pennsylvania, (REGIONAL DEPOSITORY) Walnut
Street and Commonwealth Avenue, Box 1601, Harrisburg, PA 17105.
Attorney for licensee: Ernest L. Blake, Jr., Esquire, Shaw,
Pittman, Potts & Trowbridge, 2300 N Street, NW., Washington, DC 20037.
NRC Project Director: John F. Stolz
Northeast Nuclear Energy Company (NNECO), Docket No. 50-245,
Millstone Nuclear Power Station, Unit 1, New London County,
Connecticut
Date of amendment request: May 2, 1996
Description of amendment request: The proposed change would remove
Technical Specification Figure 5.1, which is used in maintaining
Keff values, and substitute in its place a defined requirement for
maximum K-infinity for any fuel placed in the Millstone Unit 1 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:
Pursuant to 10CFR50.92, NNECO [Northeast Nuclear Energy Company]
has reviewed the proposed change and concludes that the change does
not involve a significant hazards consideration (SHC) since the
proposed change satisfies the criteria in 10CFR50.92(c). That is,
the proposed change does not:
1. Involve a significant increase in the probability or
consequences of an accident previously evaluated.
There are no spent fuel pool accident conditions discussed in
Chapter 15 of the FSAR [Final Safety Analysis Report]. FSAR section
15.8 discusses a fuel handling accident which drops a fuel assembly
into the core during refueling. Changing the maximum allowed fuel
reactivity or allowing gaps in the Boraflex
panels will have no effect on the probability or consequences of
a fuel assembly drop onto the core.
Therefore, based on the above, the proposed change to the
Technical Specifications does not involve a significant increase in
the probability or consequences of any previously analyzed accident.
2. Create the possibility of a new or different kind of accident
from any accident previously evaluated.
The reduction in the allowable fuel reactivity in the SFP [spent
fuel pool] is conservative and does not create the possibility of a
new or different type of accident. Allowing boraflex gaps does not
create the possibility of a new or different kind of accident from
any accident previously evaluated.
3. Involve a significant reduction in a margin of safety.
The margin to safety, for this proposed technical specification
change, is to maintain the SFP Keff to be less than or equal to
0.90. As described in the HOLTEC analysis, gaps in the Boraflex of
up to 5 inches can exist in every boraflex panel of every rack with
Boraflex in the SFP, with Keff still less than 0.90. This is
true even if all of the gaps are uniformly lined up at the same
elevation. These calculations conservatively assumed 4% Boraflex
width shrinkage as well as the axial Boraflex gaps. Older fuel
designs were also considered to ensure that they had not become
limiting with the reduced allowable K-infinity limit of 1.24. With
no boraflex gaps, the maximum Keff is less than .844. With 5
inch Boraflex gaps in every panel at the same elevation, the maximum
Keff is 0.896, which is less than 0.90. NNECO has implemented a
1 year decay time requirement to minimize gamma irradiation damage
to the Boraflex, and will continue to measure via ``blackness
testing'' the actual gap size to ensure the margin of safety in
maintained.
Therefore, this change has no impact on the margin to safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: Learning Resources Center,
Three Rivers Community-Technical College, 574 New London Turnpike,
Norwich, CT 06360, and the Waterford Library, ATTN: Vince Juliano, 49
Rope Ferry Road, Waterford, CT 06385
Attorney for licensee: Lillian M. Cuoco, Esq., Senior Nuclear
Counsel, Northeast Utilities Service Company, P.O. Box 270, Hartford,
CT 06141-0270
NRC Project Director: Phillip F. McKee
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 amendment request: March 29, 1996
Description of amendment request: The proposed amendment would add
limits associated with Departure from Nucleate Boiling (DNB) to the
Indian Point 3 (IP3) Technical Specifications.
Basis for proposed no significant hazards consideration
determination:
[[Page 37302]]
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:
Consistent with the criteria of 10 CFR 50.92, the enclosed
application is judged to involve no significant hazards based on the
following information:
(1) Does the proposed license amendment involve a significant
increase in the probability or consequences of an accident
previously analyzed?
Response:
The proposed amendment makes no changes to the way in which the
plant is operated and has no effect on accident initiators
associated with analyzed transients. The probability of previously
analyzed accidents is not increased. The proposed amendment
clarifies the relationship between measurable parameters (RCS
[reactor coolant system] temperature, pressure, and flow rate) and
the resulting heat transfer regime in the reactor core, as
characterized by the Departure from Nucleate Boiling (DNB) ratio.
This clarification ensures that safety analysis initial conditions
regarding heat transfer remain valid, so that the consequences of
previously analyzed accidents are not increased. The changes ensure
that RCS pressure, temperature, and flow are within analytical
bounds. This ensures that the plant is operated in a manner that
will not increase the probabilities of previously analyzed accidents
nor the consequences of previously analyzed accidents.
(2) Does the proposed license amendment create the possibility
of a new or different kind of accident from any accident previously
evaluated?
Response:
The proposed amendment does not involve any modifications to
plant systems, structures, or components. The proposed change
clarifies existing limits on RCS parameters and makes no changes to
plant setpoints or operating limits. The amendment does not involve
any physical mechanism which could contribute to a new or different
kind of accident. The changes ensure that RCS pressure, temperature,
and flow are within analytical bounds. This ensures that the plant
is operated in a manner that will not create the possibility of a
new [or] different kind of accident from any previously evaluated.
(3) Does the proposed amendment involve a significant reduction
in a margin of safety?
Response:
The proposed amendment clarifies existing limits on the
measurable parameters (RCS temperature, pressure, and flow rate) so
that the resulting DNB value is consistent with initial condition
assumptions used in existing safety analyses. Maintaining these
limits during normal plant operation ensures that the existing
margins of safety remain valid. The proposed amendment 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 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: White Plains Public Library,
100 Martine Avenue, White Plains, New York 10601.
Attorney for licensee: Mr. Charles M. Pratt, 10 Columbus Circle,
New York, New York 10019.
NRC Project Director: Jocelyn A. Mitchell, Acting
Public Service Electric & Gas Company, Docket Nos. 50-272 and 50-
311, Salem Nuclear Generating Station, Unit Nos. 1 and 2, Salem
County, New Jersey
Date of amendment request: June 18, 1996
Description of amendment request: The proposed amendment would
change Technical Specification (TS) 5.2.2, ``Design Pressure and
Temperature,'' by adding design parameters for Main Steam Line Break
(MSLB). The MSLB analysis results in a higher containment air
temperature than the current value in TS 5.2.2.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. The proposed change does not involve a significant increase
in the probability or consequences of an accident previously
evaluated.
The accidents considered for this change are the Loss of Coolant
Accident (LOCA) and the Main Steam Line Break (MSLB). The proposed
change ensures the design limiting containment pressure and
temperature data specified in the TS is consistent with the [Updated
Final Safety Analysis Report] UFSAR. Since no physical changes to
the containment are being made there will be no change in the
probability of either accident occurring.
Detailed structural analysis presented in Supplement 1 of
Licensee Event Report (LER) 272/95-016 shows that the Design Basis
LOCA combination of pressure and temperature result in more severe
loading for the containment concrete structure and, therefore,
bounds the temperature and pressure scenario associated with a MSLB
accident. The pressure retaining capability of the liner is governed
by the loads generated in the MSLB. Since containment leakage is
maintained within the limits assumed in the Accident Analysis for
either scenario there is no change in the consequences of either
accident.
Therefore the 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.
The changes proposed affect the post-accident condition of the
containment, and have no impact on the pre-accident condition. Since
there is no physical change proposed the containment and all systems
in the containment will continue to perform as designed. With no
physical changes being proposed and no change to the pre-accident
condition of the containment it can be concluded that there will be
no change in the probability of a new or different accident being
created.
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.
Although calculations indicate that some yielding of the liner
plate could occur during a MSLB, loading is transferred to the
containment concrete structure and leakage from the containment is
maintained within the limits assumed in the Accident Analysis. Since
containment leakage is maintained within the limits assumed in the
Accident Analysis the proposed change does not involve a significant
change the margin of safety provided by the containment for the
MSLB.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: Salem Free Public library, 112
West Broadway, Salem, NJ 08079
Attorney for licensee: Mark J. Wetterhahn, Esquire, Winston and
Strawn, 1400 L Street, NW, Washington, DC 20005-3502
NRC Project Director: John F. Stolz
Tennessee Valley Authority, Docket Nos. 50-327 and 50-328, Sequoyah
Nuclear Plant, Units 1 and 2, Hamilton County, Tennessee
Date of amendment request: June 7, 1996 (TSC 95-19)
Description of amendment request: The proposed change would revise
Section 6 of the plant Technical Specifications to be more closely
aligned with the Revised Standard Technical Specifications for
Westinghouse-designed nuclear plants (NUREG-1431). Additionally, the
proposed changes would be consistent with the guidance provided in
Administrative Letter 95-06, ``Relocation of Technical Specification
Administrative Controls Related to Quality Assurance.''
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the
[[Page 37303]]
licensee has provided its analysis of the issue of no significant
hazards consideration, which is presented below:
TVA [Tennessee Valley Authority] has concluded that operation of
SQN [Sequoyah Nuclear Plant] Units 1 and 2 in accordance with the
proposed changes to the TS [Technical Specification] does not
involve a significant hazards consideration. TVA's conclusion is
based on its evaluation, in accordance with 10 CFR 50.91(a)(1), of
the three standards set forth in 10 CFR 50.92(c).
A. The proposed amendment does not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
The proposed TS change is administrative. TVA has evaluated the
proposed TS changes and has determined that the proposed changes are
administrative in nature. 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. They do not involve an increase
in the probability or consequences of an accident previously
evaluated.
B. The proposed amendment does not create the possibility of a
new or different kind of accident from any previously evaluated.
The proposed TS change is administrative. TVA has evaluated the
proposed TS changes and has determined that the proposed changes are
administrative in nature. 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.
C. The proposed amendment does not involve a significant
reduction in a margin of safety.
The proposed TS change is administrative. TVA has evaluated the
proposed TS changes and has determined that the proposed changes are
administrative in nature. 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 has reviewed the licensee's analysis and, based on
thisreview, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: Chattanooga-Hamilton County
Library, 1001 Broad Street, Chattanooga, Tennessee 37402
Attorney for licensee: General Counsel, Tennessee Valley Authority,
400 West Summit Hill Drive, ET 11H, Knoxville, Tennessee 37902
NRC Project Director: Frederick J. Hebdon
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.
North Atlantic Energy Service Company, Docket No. 50-443, Seabrook
Plant Unit No. 1, Rockingham County, New Hampshire
Date of amendment request: June 20, 1996
Description of amendment request: The proposed amendment would
increase the allowed time for an inoperable service water cooling tower
loop electrical supply to be the same as the allowed outage time for an
operable service water cooling tower loop.
Date of publication of individual notice in Federal Register: June
26, 1996 (61 FR 33142)
Expiration date of individual notice: July 26, 1996
Local Public Document Room location: Exeter Public Library,
Founders Park, Exeter, New Hampshire
Northeast Utilities Service Company, Docket No. 50-336, Millstone
Nuclear Power Station, Unit No. 2, New London County, Connecticut
Date of amendment request: June 3, 1996
Description of amendment request: The proposed amendments would
provide a one-time change to Technical Specification 3.9.1, ``Refueling
Operations, Boron Concentration.'' This change would remove the
requirement that the boron concentration in all filled portions of the
Reactor Coolant System be ``uniform'' and would only be applicable
during Millstone 2 Cycle 13 mid-cycle core offload.
Date of publication of individual notice in Federal Register: June
12, 1996 (61 FR 29771)
Expiration date of individual notice: July 12, 1996
Local Public Document Room location: Learning Resources Center,
Three Rivers Community-Technical College, 574 New London Turnpike,
Norwich, Connecticut, and the Waterford Library, ATTN: Vince Juliano,
49 Rope Ferry Road, Waterford, Connecticut
Northeast Nuclear Energy Company, et al., Docket No. 50-423,
Millstone Nuclear Power Station, Unit No. 3, New London County,
Connecticut
Date of amendment request: May 23, 1996
Description of amendment request: The proposed amendment would
revise the Technical Specifications (TS) for the Overtemperature delta
T time constants in TS Table 2.2-1 and the Steam Line Pressure Negative
Rate High Steam Line Isolation time constant in TS Table 3.3-4. Date of
publication of individual notice in Federal Register: June 17, 1996 (61
FR 30639)
Expiration date of individual notice: July 17, 1997
Local Public Document Room location: Learning Resources Center,
Three Rivers Community-Technical College, 574 New London Turnpike,
Norwich, Connecticut, and the Waterford Library, ATTN: Vince Juliano,
49 Rope Ferry Road, Waterford, Connecticut
Public Service Electric & Gas Company, Docket Nos. 50-272 and 50-
311, Salem Nuclear Generating Station, Unit Nos. 1 and 2, Salem
County, New Jersey
Date of amendment request: June 10, 1996
Brief description of amendment request: The amendment proposes
changes to Technical Specification 3/4.7.6, ``Control Room Emergency
Air Conditioning System,'' to reflect a control room design in which
the common Salem Unit 1 and 2 control room envelope is supplied by 2
one hundred percent capable Control Room Emergency Air Conditioning
System trains. Date of publication of individual notice in Federal
Register: June 24, 1996 (61 FR 32468)
Expiration date of individual notice: July 24, 1996
Local Public Document Room location: Salem Free Public Library, 112
West Broadway, Salem, NJ 08079
[[Page 37304]]
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 amendment request: June 24, 1996
Description of amendment request: The proposed amendments would
revise Technical Specification Table 4.3.1 to delete the requirement
for surveillance of the manual safety injection to the reactor trip
circuitry until the next unit shutdown, following which, this testing
will be performed prior to Mode 2 entry. This change is applicable only
to Unit 1, Cycle 14 and Unit 2, Cycle 11. Date of publication of
individual notice in Federal Register: July 3, 1996 (61 FR 34880)
Expiration date of individual notice: August 2, 1996
Local Public Document Room location: Houston-Love Memorial Library,
212 W. Burdeshaw Street, P. O. Box 1369, Dothan, Alabama
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.
Commonwealth Edison Company, Docket Nos. STN 50-454 and STN 50-455,
Byron Station, Unit Nos. 1 and 2, Ogle County, Illinois Docket Nos.
STN 50-456 and STN 50-457, Braidwood Station, Unit Nos. 1 and 2,
Will County, Illinois
Date of application for amendments: September 16, 1994, as
supplemented January 31, 1996.
Brief description of amendments: The amendments revise the
technical specifications to eliminate periodic response time testing
requirements for selected pressure and differential pressure sensors in
the reactor trip system and engineered safety features actuation
instrumentation channels.
Date of issuance: June 26, 1996
Effective date: Immediately, to be implemented within 30 days.
Amendment Nos.: 84, 84, 76 and 76
Facility Operating License Nos. NPF-37, NPF-66, NPF-72 and NPF-77:
The amendments revised the Technical Specifications.
Date of initial notice in Federal Register: March 13, 1996 (61 FR
10393). The Commission's related evaluation of the amendments is
contained in a Safety Evaluation dated June 26, 1996. No significant
hazards consideration comments received: No
Local Public Document Room location: For Byron, the Byron Public
Library District, 109 N. Franklin, P.O. Box 434, Byron, Illinois 61010;
for Braidwood, the Wilmington Public Library, 201 S. Kankakee Street,
Wilmington, Illinois 60481.
Commonwealth Edison Company, Docket Nos. 50-237 and 50-249, Dresden
Nuclear Power Station, Units 2 and 3, Grundy County, Illinois
Docket Nos. 50-254 and 50-265, Quad Cities Nuclear Power Station,
Units 1 and 2, Rock Island County, Illinois
Date of application for amendments: November 14, 1995, as
supplemented by letters dated February 23, March 1, March 13, March 25,
March 26, May 10, June 10, June 14, two letters dated June 25 and a
letter dated June 26, 1996.
Brief description of amendments: The proposed amendments closed out
additional open items identified in the NRC staff's review of the
upgrade of the Dresden and Quad Cities Technical Specifications (TS) to
the Standard Technical Specifications (STS) contained in NUREG-0123.
The Technical Specification Upgrade Program (TSUP) is not a complete
adaptation of the STS. The TS upgrade focuses on (1) integrating
additional information such as equipment operability requirements
during shutdown conditions, (2) clarifying requirements such as
limiting conditions for operation and action statements utilizing STS
terminology, (3) deleting superseded requirements and modifications to
the TS based on the licensee's responses to Generic Letter (GL), and
(4) relocating specific items to more appropriate TS locations.
Date of issuance: June 28, 1996
Effective date: June 28, 1996
Amendment Nos.: 150, 145, 171, and 167
Facility Operating License Nos. DPR-19, DPR-25, DPR-29 and DPR-30.
The amendments revised the Technical Specifications and operating
licenses.
Date of initial notice in Federal Register: November 29, 1995 (60
FR 61272) The Commission's related evaluation of the amendments is
contained in a Safety Evaluation dated June 28, 1996. No significant
hazards consideration comments received: No
Local Public Document Room location: for Dresden, Morris Area
Public Library District, 604 Liberty Street, Morris, Illinois 60450;
for Quad Cities, Dixon Public Library, 221 Hennepin Avenue, Dixon,
Illinois 61021.
Detroit Edison Company, Docket No. 50-341, Fermi-2, Monroe County,
Michigan
Date of application for amendment: November 22, 1995 (NRC-95-0124)
Brief description of amendment: The amendment revises the Technical
Specifications to remove accelerated testing frequencies and special
reporting requirements for Fermi 2 emergency diesel generators (EDGs)
in accordance with guidance contained in Generic Letter 94-01, dated
May 31, 1994. NRC will issue a separate safety evaluation on extending
the allowed outage time for the EDGs at a later date.
Date of issuance: June 20, 1996
Effective date: June 20, 1996, with full implementation within 60
days
Amendment No.: 107
Facility Operating License No. NPF-43. Amendment revises the
Technical Specifications.
Date of initial notice in Federal Register: February 28, 1996 (61
FR 7550) The Commission's related evaluation of the amendment is
contained in a Safety Evaluation dated
[[Page 37305]]
June 20, 1996. No significant hazards consideration comments received:
No
Local Public Document Room location: Monroe County Library System,
3700 South Custer Road, Monroe, Michigan 48161
Duke Power Company, Docket Nos. 50-369 and 50-370, McGuire Nuclear
Station, Units 1 and 2, Mecklenburg County, North Carolina
Date of application for amendments: December 12, 1995, as
supplemented by letter dated June 10, 1996
Description of amendment request: The amendments revise the
absolute values in the Axial Flux Difference (AFD) Equations to reflect
the proper AFD limit reduction in the current Technical Specifications.
Date of issuance: July 2, 1996
Effective date: As of the date of issuance to be implemented within
30 days.
Amendment Nos.: 167 and 149
Facility Operating License Nos. NPF-9 and NPF-17: Amendments
revised the Technical Specifications.
Date of initial notice in Federal Register: April 24, 1996 (61 FR
18166) The June 10, 1996, letter provided clarifying information that
did not change the scope of the December 12, 1995, application and the
initial proposed no significant hazards consideration determination.
The Commission's related evaluation of the amendments is contained in a
Safety Evaluation dated July 2, 1996. No significant hazards
consideration comments received: No
Local Public Document Room location: Atkins Library, University of
North Carolina, Charlotte (UNCC Station), North Carolina 28223
Entergy Operations, Inc., Docket No. 50-382, Waterford Steam
Electric Station, Unit 3, St. Charles Parish, Louisiana
Date of amendment request: August 11, 1995, as supplemented by
letter dated February 12, 1996
Brief description of amendment: The amendment reduced the minimum
reactor coolant cold leg temperature to 541 deg.F from 544 deg.F in
Technical Specification Section 3.2.6, ``Reactor Coolant Cold Leg
Temperature.''
Date of issuance: June 24, 1996
Effective date: June 24, 1996
Amendment No.: 120
Facility Operating License No. NPF-38. Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: June 5, 1996 (61 FR
25706) The Commission's related evaluation of the amendment is
contained in a Safety Evaluation dated June 24, 1996. No significant
hazards consideration comments received: No.
Local Public Document Room location: University of New Orleans
Library, Louisiana Collection, Lakefront, New Orleans, LA 70122.
Florida Power and Light Company, Docket Nos. 50-250 and 50-251,
Turkey Point Plant Units 3 and 4, Dade County, Florida
Date of application for amendments: March 20, 1996, as supplemented
by letter date April 23, 1996.
Brief description of amendments: These amendments relocate the
requirements for surveillance testing of the water level and pressure
channel instrumentation for the reactor coolant system accumulators.
These amendments also modify the existing action statements of TS 3.5.1
for accumulators to reflect the requirements of NUREG-1431 by requiring
a 72- hour period to restore boron concentration if it is not within
the limits, and a 1-hour period to restore any other condition
rendering the accumulators inoperable.
Date of issuance: June 24, 1996
Effective date: June 24, 1996
Amendment Nos. 185 and 179Facility Operating Licenses Nos. DPR-31
and DPR-41: Amendments revised the Technical Specifications.
Date of initial notice in Federal Register: May 22, 1996 (61 FR
25707) The Commission's related evaluation of the amendments is
contained in a Safety Evaluation dated June 24, 1996. No significant
hazards consideration comments received: No
Local Public Document Room location: Florida International
University, University Park, Miami, Florida 33199.
Georgia Power Company, Oglethorpe Power Corporation, Municipal
Electric Authority of Georgia, City of Dalton, Georgia, Docket Nos.
50-321 and 50-366, Edwin I. Hatch Nuclear Plant, Units 1 and 2,
Appling County, Georgia
Date of application for amendments: February 21, 1996, as
supplemented by letters dated May 1 and June 4, 1996.
Brief description of amendments: The amendments revise the
Technical Specifications to change the Drywell Air Temperature Limiting
Condition for Operation (LCO) from less than or equal to 135 deg.F to
less than or equal to 150 deg.F. The proposed change would provide a
margin for the primary containment Drywell Air Temperature LCO when
prolonged summer and high river temperatures are experienced. Also, a
strictly editorial correction to a Final Safety Analysis Report
reference would be made.
Date of issuance: 201 and 142
Effective date: As of the date of issuance to be implemented within
30 days.
Amendment Nos.: 201 and 142
Facility Operating License Nos. DPR-57 and NPF-5: Amendments
revised the Technical Specifications.
Date of initial notice in Federal Register: April 24, 1996 (61 FR
18167) The May 1 and June 4, 1996, letters provided clarifying
information that did not change the scope of the February 21, 1996,
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 June 27, 1996. No significant hazards
consideration comments received: No
Local Public Document Room location: Appling County Public Library,
301 City Hall Drive, Baxley, Georgia 31513
Indiana Michigan Power Company, Docket Nos. 50-315 and 50-316,
Donald C. Cook Nuclear Plant, Unit Nos. 1 and 2, Berrien County,
Michigan
Date of application for amendments: May 19, 1995, and supplemented
October 20, 1995, and April 8, 1996 (AEP:NRC:1213A)
Brief description of amendments: The amendments modify the neutron
flux high setpoints for one or more main steam safety valves inoperable
in response to Westinghouse Nuclear Safety Advisory Letter 94-001. The
associated action statements are also revised and an exemption to TS
4.0.4 is added to support the operability surveillance.
Date of issuance: June 28, 1996
Effective date: June 28, 1996, with full implementation within 45
days.
Amendment Nos.: Unit 1 - 210, Unit 2 - 195
Facility Operating License Nos. DPR-58 and DPR-74. Amendments
revised the Technical Specifications.
Date of initial notice in Federal Register: December 20, 1995 (60
FR 65681) The April 8, 1996, submittal provided information clarifying
the location of the TS 4.0.4 exemption statement. This information was
within the scope of the original application and did not alter the
staff's no significant hazards considerations determination. Therefore
renoticing was not warranted. The Commission's related evaluation of
the amendments is contained in a Safety Evaluation dated June 28, 1996.
No significant hazards consideration comments received: No.
[[Page 37306]]
Local Public Document Room location: Maud Preston Palenske Memorial
Library, 500 Market Street, St. Joseph, Michigan 49085
Niagara Mohawk Power Corporation, Docket No. 50-410, Nine Mile
Point Nuclear Station, Unit 2, Oswego County, New York
Date of application for amendment: January 17, 1996
Brief description of amendment: The amendment revises the Technical
Specifications (TSs) and associated Bases by relocating certain
response time limit tables from the TSs to the Updated Safety Analysis
Report in accordance with the guidance of NRC Generic Letter 93-08. The
relocated tables are for instrumentation for the Reactor Protection
System, Isolation Actuation System, Emergency Core Cooling System, and
the Recirculation Pump Trip System.
Date of issuance: June 25, 1996
Effective date: As of the date of issuance to be implemented within
30 days.
Amendment No.: 73
Facility Operating License No. NPF-69: Amendment revises the
Technical Specifications.
Date of initial notice in Federal Register: May 8, 1996 (61 FR
20850) The Commission's related evaluation of the amendment is
contained in a Safety Evaluation dated June 25, 1996. No significant
hazards consideration comments received: No
Local Public Document Room location: Reference and Documents
Department, Penfield Library, State University of New York, Oswego, New
York 13126.
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: December 18, 1995
Brief description of amendment: The amendment changes the Reactor
Coolant Flow - Low Flow in Technical Specification Table 2.2-1,
``Reactor Instrumentation Protective Trip Setpoint Limits.'' The
proposed change increases the allowable value from greater than or
equal to 90.1% to greater than or equal to 90.9% of the reactor coolant
flow with four pumps operating. As an editorial change for
clarification, the word ``flow'' is added after ``reactor coolant'' in
the above sentence.
Date of issuance: July 2, 1996
Effective date: As of the date of issuance, to be implemented
within 60 days.
Amendment No.: 199
Facility Operating License No. DPR-65: Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: February 14, 1996 (61
FR 5815) The Commission's related evaluation of the amendment is
contained in a Safety Evaluation dated July 2, 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
Northeast Nuclear Energy Company, et al., Docket No. 50-423,
Millstone Nuclear Power Station, Unit No. 3, New London County,
Connecticut
Date of application for amendment: June 27, 1995, as supplemented
July 21, 1995
Brief description of amendment: The amendment revises the Technical
Specifications (TS) to relocate TS requirements for the containment
purge exhaust and supply valves, and to remove a duplicate testing
requirement for the safety injection input from engineered safety
features from the TS.
Date of issuance: June 27, 1996
Effective date: As of the date of issuance, to be implemented
within 60 days.
Amendment No.: 129
Facility Operating License No. NPF-49. Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: December 6, 1995 (60 FR
62494) The Commission's related evaluation of the amendment is
contained in a Safety Evaluation dated June 27, 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
Pennsylvania Power and Light Company, Docket No. 50-387,
Susquehanna Steam Electric Station, Unit 1, Luzerne County,
Pennsylvania
Date of application for amendment: January 26, 1996
Brief description of amendment: The amendment deletes three
residual heat removal (RHR) system relief valves from Technical
Specification (TS) Table 3.6.3-1, ``Primary Containment Isolation
Valves.'' These valves are no longer needed to support the steam
condensing mode of RHR and are being removed from the plant during the
Unit 1 ninth refueling outage.
Date of issuance: June 24, 1996
Effective date: As of date of issuance to be implemented within 60
days.
Amendment No.: 157
Facility Operating License No. NPF-14: This amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: March 27, 1996 (61 FR
13531) The Commission's related evaluation of the amendment is
contained in a Safety Evaluation dated June 24, 1996. No significant
hazards consideration comments received: No
Local Public Document Room location: Osterhout Free Library,
Reference Department, 71 South Franklin Street, Wilkes-Barre,
Pennsylvania 18701.
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 29, 1996
Brief description of amendments: These amendments relocate
Specification 3/4.9.6, ``Refueling Platform,'' to the Susquehanna Steam
Electric Station Technical Requirements Manual, a document which is
controlled under the requirements of 10 CFR 50.59.
Date of issuance: July 2, 1996
Effective date: July 2, 1996
Amendment Nos.: 158 and 129
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 July 2, 1996. No significant
hazards consideration comments received: No
Local Public Document Room location: Osterhout Free Library,
Reference Department, 71 South Franklin Street, Wilkes-Barre,
Pennsylvania 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: March 12, 1996
Brief description of amendment: The proposed changes would remove a
requirement to cross tie safety injection accumulators.
Date of issuance: July 3, 1996
Effective date: As of the date of issuance to be implemented within
30 days.
[[Page 37307]]
Amendment No.: 167
Facility Operating License No. DPR-64: 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 July 3, 1996. No significant
hazards consideration comments received: No
Local Public Document Room location: White Plains Public Library,
100 Martine Avenue, White Plains, New York 10610.
Power Authority of the State of New York, Docket No. 50-333, James
A. FitzPatrick Nuclear Power Plant, Oswego County, New York
Date of application for amendment: April 24, 1996
Brief description of amendment: The amendment proposes to relocate
Specification 3.11.B/4.11.B ``Crescent Area Ventilation'' and
associated Bases from the TS to an Authority controlled procedure.
Date of issuance: June 28, 1996
Effective date: As of the date of issuance to be implemented within
30 days.
Amendment No.: 231
Facility Operating License No. DPR-59: Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: May 22, 1996 (61 FR
25710) The Commission's related evaluation of the amendment is
contained in a Safety Evaluation dated June 28, 1996. No significant
hazards consideration comments received: No
Local Public Document Room location: Reference and Documents
Department, Penfield Library, State University of New York, Oswego, New
York 13126.
Public Service Electric & Gas Company, Docket Nos. 50-272 and 50-
311, Salem Nuclear Generating Station, Unit Nos. 1 and 2, Salem
County, New Jersey
Date of application for amendments: February 6, 1996
Brief description of amendments: The amendments change the
Technical Specifications to lower the 125 Volt Battery Charger
surveillance amperage from at least 200 amps to at least 170 amps.
Date of issuance: June 27, 1996
Effective date: As of date of issuance, to be implemented within 30
days.
Amendment Nos. 183 and 164
Facility Operating License Nos. DPR-70 and DPR-75. The amendments
revised the Technical Specifications.
Date of initial notice in Federal Register: February 28, 1996 (61
FR 7556) The Commission's related evaluation of the amendments is
contained in a Safety Evaluation dated June 27, 1996. No significant
hazards consideration comments received: No
Local Public Document Room location: Salem Free Public Library, 112
West Broadway, Salem, New Jersey 08079
Union Electric Company, Docket No. 50-483, Callaway Plant, Unit 1,
Callaway County, Missouri
Date of application for amendment: June 26, 1995, as supplemented
by letter dated February 2, 1996.
Brief description of amendment: The amendment revised the allowed
outage time for component cooling water motor operated containment
isolation valves, moved the list of containment isolation valves from
the technical specifications to the final safety analysis report, and
allowed containment penetration check valves to be used as isolation
devices.
Date of issuance: June 28, 1996
Effective date: June 28, 1996, to be implemented within 30 days of
the date of issuance.
Amendment No.: 113
Facility Operating License No. NPF-30: The amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: August 30, 1995 (60 FR
45187) The February 2, 1996, supplemental letter provided additional
clarifying information and did not change the staff's original no
significant hazards consideration determination. The Commission's
related evaluation of the amendment is contained in a Safety Evaluation
dated June 28, 1996. No significant hazards consideration comments
received: No.
Local Public Document Room location: Callaway County Public
Library, 710 Court Street, Fulton, Missouri 65251.
Dated at Rockville, Maryland, this 10th day of July 1996.
For the Nuclear Regulatory Commission
Steven A. Varga,
Director, Division of Reactor Projects - I/II, Office of Nuclear
Reactor Regulation.
[Doc. 96-18007 Filed 7-16-96; 8:45 am]
BILLING CODE 7590-O1-F