[Federal Register Volume 64, Number 76 (Wednesday, April 21, 1999)]
[Notices]
[Pages 19554-19570]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: 99-9839]
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NUCLEAR REGULATORY COMMISSION
Biweekly Notice; Applications and Amendments to Facility
Operating Licenses Involving No Significant Hazards Considerations
I. Background
Pursuant to Public Law 97-415, the U.S. Nuclear Regulatory
Commission (the Commission or NRC staff) is publishing this regular
biweekly notice. Public Law 97-415 revised section 189 of the Atomic
Energy Act of 1954, as amended (the Act), to require the Commission to
publish notice of any amendments issued, or proposed to be issued,
under a new provision of section 189 of the Act. This provision grants
the Commission the authority to issue and make immediately effective
any amendment to an operating license upon a determination by the
Commission that such amendment involves no significant hazards
consideration, notwithstanding the pendency before the Commission of a
request for a hearing from any person.
This biweekly notice includes all notices of amendments issued, or
proposed to be issued from March 27 through April 9, 1999. The last
biweekly notice was published on April 7, 1999 (64 FR 17021).
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 and
Directives Branch, Division of Administration Services, Office of
[[Page 19555]]
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 May 21, 1999, the licensee may file a request for a hearing with
respect to issuance of the amendment to the subject facility operating
license and any person whose interest may be affected by this
proceeding and who wishes to participate as a party in the proceeding
must file a written request for a hearing and a petition for leave to
intervene. Requests for a hearing and a petition for leave to intervene
shall be filed in accordance with the Commission's ``Rules of Practice
for Domestic Licensing Proceedings'' in 10 CFR Part 2. Interested
persons should consult a current copy of 10 CFR 2.714 which is
available at the Commission's Public Document Room, the Gelman
Building, 2120 L Street, NW., Washington, DC and at the local public
document room for the particular facility involved. If a request for a
hearing or petition for leave to intervene is filed by the above date,
the Commission or an Atomic Safety and Licensing Board, designated by
the Commission or by the Chairman of the Atomic Safety and Licensing
Board Panel, will rule on the request and/or petition; and the
Secretary or the designated Atomic Safety and Licensing Board will
issue a notice of a hearing or an appropriate order.
As required by 10 CFR 2.714, a petition for leave to intervene
shall set forth with particularity the interest of the petitioner in
the proceeding, and how that interest may be affected by the results of
the proceeding. The petition should specifically explain the reasons
why intervention should be permitted with particular reference to the
following factors: (1) the nature of the petitioner's right under the
Act to be made a party to the proceeding; (2) the nature and extent of
the petitioner's property, financial, or other interest in the
proceeding; and (3) the possible effect of any order which may be
entered in the proceeding on the petitioner's interest. The petition
should also identify the specific aspect(s) of the subject matter of
the proceeding as to which petitioner wishes to intervene. Any person
who has filed a petition for leave to intervene or who has been
admitted as a party may amend the petition without requesting leave of
the Board up to 15 days prior to the first prehearing conference
scheduled in the proceeding, but such an amended petition must satisfy
the specificity requirements described above.
Not later than 15 days prior to the first prehearing conference
scheduled in the proceeding, a petitioner shall file a supplement to
the petition to intervene which must include a list of the contentions
which are sought to be litigated in the matter. Each contention must
consist of a specific statement of the issue of law or fact to be
raised or controverted. In addition, the petitioner shall provide a
brief explanation of the bases of the contention and a concise
statement of the alleged facts or expert opinion which support the
contention and on which the petitioner intends to rely in proving the
contention at the hearing. The petitioner must also provide references
to those specific sources and documents of which the petitioner is
aware and on which the petitioner intends to rely to establish those
facts or expert opinion. Petitioner must provide sufficient information
to show that a genuine dispute exists with the applicant on a material
issue of law or fact. Contentions shall be limited to matters within
the scope of the amendment under consideration. The contention must be
one which, if proven, would entitle the petitioner to relief. A
petitioner who fails to file such a supplement which satisfies these
requirements with respect to at least one contention will not be
permitted to participate as a party.
Those permitted to intervene become parties to the proceeding,
subject to any limitations in the order granting leave to intervene,
and have the opportunity to participate fully in the conduct of the
hearing, including the opportunity to present evidence and cross-
examine witnesses.
If a hearing is requested, the Commission will make a final
determination on the issue of no significant hazards consideration. The
final determination will serve to decide when the hearing is held.
If the final determination is that the amendment request involves
no significant hazards consideration, the Commission may issue the
amendment and make it immediately effective, notwithstanding the
request for a hearing. Any hearing held would take place after issuance
of the amendment.
If the final determination is that the amendment request involves a
significant hazards consideration, any hearing held would take place
before the issuance of any amendment.
A request for a hearing or a petition for leave to intervene must
be filed with the Secretary of the Commission, U.S. Nuclear Regulatory
Commission, Washington, DC 20555-0001, Attention: Rulemakings and
Adjudications Staff, or may be delivered to the Commission's Public
Document Room, the Gelman Building, 2120 L Street, NW, Washington DC,
by the above date. 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.
Detroit Edison Company, Docket No. 50-341, Fermi 2, Monroe County,
Michigan
Date of amendment request: March 23, 1999.
Description of amendment request: The proposed amendment would
modify Technical Specification Surveillance Requirement 4.4.1.1.1 to
require each recirculation pump discharge valve to be demonstrated
OPERABLE at least once every 18 months and will delete footnote * that
applies to Technical Specification 4.4.1.1.1.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. The change does not involve a significant increase in the
probability or consequences of an accident previously evaluated.
The proposed changes to the Technical Specifications (TS) would
modify the frequency of cycling the recirculation pump discharge
valves from ``each STARTUP*
[[Page 19556]]
prior to THERMAL POWER exceeding 25% of RATED THERMAL POWER'' to
``at least once per 18 months;'' and replace the footnote applicable
to TS 4.4.1.1, ``*If not performed in the previous 31 days'' with
``*Not Used.'' The change in testing frequency does not affect the
probability of an accident since the valve testing is not related to
accident initiation sequences. Consequences of accidents are not
significantly increased because the proposed testing interval
provides reasonable assurance that the valves will function. Testing
of the valves will still be performed on a frequency that is allowed
by TS if no events occur that require entry into Mode 3 or Mode 4.
Therefore, the change will not involve a significant increase in the
probability or consequences of an accident previously evaluated.
Testing the valves in accordance with the inservice testing (IST)
program on the same testing frequency as testing performed for the
low pressure coolant injection system, provides adequate assurance
that the valves can perform their safety function and will not
increase the consequences of an accident previously evaluated. The
change to the footnote is administrative in nature and will have no
effect on the probability of an accident and will not increase any
safety consequences.
2. The change does not create the possibility of a new or
different kind of accident from any accident previously evaluated.
The proposed changes revise performing the testing of the
recirculation pump discharge valves from ``prior to Startup* not to
exceed 25% of rated thermal power.'' to ``at least once per 18
months'' and replace the footnote applicable to TS 4.4.1.1'' *If not
performed in the previous 31 days'' with ``*Not Used'' does not
result in a new accident precursor since the test only verifies that
the valve can close which is its safety function. Deleting the
information contained in footnote ``*'' that applies to TS 4.4.1.1.1
and designating it as ``* Not Used.'' is administrative in nature
with no safety significance. Therefore, no different type of
accident from any previously evaluated is introduced.
3. The change does not involve a significant reduction in the
margin of safety.
The proposed changes revise the frequency of cycling the
recirculation pump discharge valves from ``each STARTUP* prior to
THERMAL POWER exceeding 25% of RATED THERMAL POWER'' to ``at least
once per 18 months'' and replace the footnote applicable to TS
4.4.1.1 ``*If not performed in the previous 31 days'' with ``*Not
Used.'' Altering the test frequency does not change valve stroke
time or other performance or design characteristics related to the
safety function of the valves. The potential for failure of the
valve to close is not changed as a result of the proposed change
since the same frequency is allowed by the current TS if no events
occur that require entry into Mode 3 or Mode 4. Performing stroke
time testing on a refueling outage basis and MOV testing on a
periodic basis does not decrease the margin of safety associated
with the valve performing its safety function. Revising footnote *
is an administrative change and has no safety consequence.
Therefore, the proposed changes do not involve a significant
reduction in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: Monroe County Library System,
Ellis Reference and Information Center, 3700 South Custer Road, Monroe,
Michigan 48161.
Attorney for licensee: John Flynn, Esq., Detroit Edison Company,
2000 Second Avenue, Detroit, Michigan 48226.
NRC Section Chief: George F. Dick, Acting.
Duquesne Light Company, et al., Docket Nos. 50-334 and 50-412, Beaver
Valley Power Station, Unit Nos. 1 and 2, Shippingport, Pennsylvania
Date of amendment request: March 3, 1999.
Description of amendment request: The proposed amendments would
change the required qualifications for operations management specified
in the Technical Specifications (TSs) for the Beaver Valley Power
Station, Units 1 and 2 (BVPS-1 and BVPS-2). The requirement that the
operations manager hold a Senior Reactor Operator (SRO) license at the
time of appointment would be changed in the TSs to require that the
assistant operations managers, one for each unit, hold an SRO license
on their assigned unit. The TSs would not then require the operations
manager hold an SRO license. Additionally, the Updated Final Safety
Analysis Report (UFSAR) for each unit would be changed to require the
operations manager to hold, or have held, an SRO license rather than
presently hold a license. The UFSAR would require the same as the TS;
that the assistant operations managers hold an SRO license on the unit
to which they are assigned. Finally, the proposed amendments would
substitute generic personnel titles for plant-specific personnel titles
in the BVPS-1 and BVPS-2 TSs. The correlation between generic titles
and plant-specific titles would be provided in the BVPS-2 UFSAR.
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 changes are administrative in nature. The revised
requirements for who must hold a current senior reactor operator
(SRO) License does not involve any change to the configuration or
method of operation of any plant equipment that is used to mitigate
the consequences of an accident nor alter the conditions or
assumptions in any of the Updated Final Safety Analysis Report
[UFSAR] accident analyses. The requirement that the operations
manager hold or have held an SRO License is included in the revised
Position Qualifications in the Unit 2 UFSAR, Table 13.1-2, sheet 30
of 35. The title changes are being made, consistent with TSTF-65,
Rev 1 and help avoid the need for future Technical Specification
changes. Therefore, it can be concluded that the proposed changes do
not involve any 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?
No new failure modes are defined for any plant system or
component important to safety nor has any new limiting failure been
identified as a result of the proposed changes. Therefore, it can be
concluded that the proposed change does not create the possibility
of a new or different kind of accident from those previously
evaluated.
3. Does the change involve a significant reduction in a margin
of safety?
The proposed changes are administrative in nature. One of the
proposed changes requires that the manager who directly supervises
the licensed operators at each unit be the holder of a current SRO
license. The other change modifies personnel titles. Therefore, it
can be concluded that the proposed changes do not involve any
reduction in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: B.F. Jones Memorial Library,
663 Franklin Avenue, Aliquippa, PA 15001.
Attorney for licensee: Jay E. Silberg, Esquire, Shaw, Pittman,
Potts & Trowbridge, 2300 N Street, NW, Washington, DC 20037.
NRC Section Chief: Singh Bajwa.
Duquesne Light Company, et al., Docket No. 50-412, Beaver Valley Power
Station, Unit No. 2, Shippingport, Pennsylvania
Date of amendment request: March 16, 1999.
Description of amendment request: The proposed amendment would
revise Technical Specification (TS) 3/4.7.1.3
[[Page 19557]]
and associated Bases for the Primary Plant Demineralized Water (PPDW)
System to clarify that the minimum specified volume of water in the
PPDW Storage Tank is a usable volume. Additionally, the minimum usable
volume of water in the PPDW Storage tank is increased, and a clarifying
footnote that the specified value is an analysis value is added.
Finally, several editorial and administrative changes, such as revision
of action statement wording, addition of license number to the TS page,
and addition of clarifying information to the TS Bases regarding
analysis assumptions are made.
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 failure of the primary plant demineralized water (PPDW)
storage tank to provide a sufficient source of water to the
Auxiliary Feedwater (AFW) System is not an accident initiating
event. Therefore, the probability of an accident previously
evaluated is not increased by this proposed amendment.
Limiting Condition for Operation (LCO) 3.7.1.3 titled ``Primary
Plant Demineralized Water (PPDW)'' will be revised to specify the
required value for PPDW storage tank volume as a usable volume. To
reflect the value currently assumed in the analysis, the value
stated in the LCO, for minimum required PPDW storage tank volume,
would be slightly increased. The addition of proposed Footnote (1)
to LCO 3.7.1.3 will ensure that plant operators recognize that the
specified volume is an analysis value and that the value does not
include measurement uncertainties. This footnote will require plant
procedures to specify an increased required volume in the PPDW
storage tank to account for measurement uncertainties. The proposed
revisions to LCO 3.7.1.3 will assure that the PPDW storage tank
minimum usable volume is maintained consistent with the design basis
for the PPDW storage tank. The PPDW storage tank will continue to
provide a sufficient source of water to the AFW pumps. Maintaining a
sufficient source of water will ensure that the AFW System is
capable of mitigating the consequences of Design Basis Accidents
(DBAs) that could result in overpressurization of the RCS pressure
boundary. The AFW system will continue to be capable of providing an
emergency source of feedwater to the steam generators to act as heat
sinks for sensible and decay heat removal from the reactor core. A
sufficient volume of water will continue to be maintained in the
PPDW storage tank to satisfy the Safe Shutdown evaluation.
The proposed changes to the Action statements will remove the
required water volume value and add wording pertaining to the water
volume not being within the limit. The LCO clearly states the value
for the minimum required volume in the PPDW storage tank. Therefore,
the proposed modification to the Action statements is administrative
in nature and does not affect plant safety. The additional Bases
wording pertaining to reactor coolant pump operation is
administrative in nature and does not affect plant safety. The
remaining change, which consists of the addition of plant operating
license number, is editorial in nature and does not affect plant
safety.
Therefore, operation of the facility in accordance with the
proposed amendment does not involve a significant increase in the
probability or consequence 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 amendment will not change the physical plant or the
modes of plant operation defined in the operating license. This
change does not involve the addition or modification of plant
equipment nor does it alter the design or operation of plant
systems. The proposed amendment will require that the minimum volume
in the PPDW storage tank be maintained consistent with analysis
assumptions.
Therefore, operation of the facility in accordance with the
proposed amendment will not create the possibility of a new or
different kind of accident from any accident previously evaluated.
3. Does the change involve a significant reduction in a margin
of safety?
The minimum required volume in the PPDW storage tank would be
slightly increased over the currently required value. This increase
in the required volume will ensure that an adequate volume of water
is maintained in the PPDW storage tank. The proposed addition of the
term ``usable,'' along with the addition of Footnote (1), will
ensure that the water volume specified in LCO 3.7.1.3 is
appropriately increased in plant procedures to account for unusable
volume in the tank and for measurement uncertainties. A sufficient
volume of water will continue to be maintained in the PPDW storage
tank to satisfy the Safe Shutdown evaluation.
The PPDW storage tank will continue to provide a sufficient
source of water to the AFW pumps to ensure that the AFW System is
capable of mitigating the consequences of DBAs that could result in
overpressurization of the RCS pressure boundary. The AFW system will
continue to be capable of providing an emergency source of feedwater
to the steam generators to act as heat sinks for sensible and decay
heat removal from the reactor core.
The proposed changes to the Action statements will remove the
required water volume value and add wording pertaining to the water
volume not being within the limit. The LCO clearly states the value
for the minimum required volume in the PPDW storage tank. Therefore,
the proposed modification to the Action statements is administrative
in nature and does not affect plant safety. The additional Bases
wording pertaining to reactor coolant pump operation is
administrative in nature and does not affect plant safety. The
remaining change, which consists of the addition of plant operating
license number, is editorial in nature and does not affect plant
safety.
Therefore, 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: B.F. Jones Memorial Library,
663 Franklin Avenue, Aliquippa, PA 15001.
Attorney for licensee: Jay E. Silberg, Esquire, Shaw, Pittman,
Potts & Trowbridge, 2300 N Street, NW., Washington, DC 20037.
NRC Section Chief: S. Singh Bajwa.
Florida Power Corporation, et al., Docket No. 50-302, Crystal River
Nuclear Generating Plant, Unit No. 3, Citrus County, Florida
Date of amendment request: August 31, 1998, and revised March 18,
1999.
Description of amendment request: The proposed amendment would
revise Improved Technical Specification (ITS) 5.6.2.10, ``Steam
Generator (OTSG [once-through steam generator]) Tube Surveillance
Program,'' to include a new repair process, called a ``repair roll'' or
``re-roll.'' The process would be used to repair steam generator tubes
with defects within the upper tubesheet. Changes to inservice
inspection and reporting requirements are proposed for tubes which are
repaired using this process. In addition, several format and editorial
changes are proposed to ITS 5.6.2.10 and to ITS 5.7.2, ``Special
Reports,'' for clarification purposes. The March 18, 1999 revision
superceded the August 31, 1998 request, and includes the results of
recent accident analyses conducted to identify the maximum OTSG tube
tensile loads. As a result of the increased tube tensile loads, some
tubes will require a double repair roll. The double repair roll
methodology was not included in the original amendment request.
Therefore, this notice revises the previous Notice of Consideration of
Issuance of Amendment (63 FR 56249).
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.
[[Page 19558]]
(1) Involve a significant increase in the probability or
consequences of an accident previously evaluated?
The repair roll process is a method to create a new primary-to-
secondary pressure boundary joint in the upper tubesheet of Babcock
& Wilcox (B&W) Once Through Steam Generators (OTSGs) manufactured
with Inconel Alloy 600 tubes. The repair roll process creates a new
roll joint in the OTSG tubes at a point closer to the secondary face
of the tubesheet than the existing roll joint. The new pressure
boundary is established by the repair roll to remove degradation of
the existing roll joint from pressure boundary service. The repair
roll process has been qualified as an acceptable repair methodology
for use in the upper tubesheet of the Crystal River Unit 3 (CR-3)
OTSGs. The proposed License Amendment Request (LAR) proposes to
implement the qualified OTSG tube repair roll process, and also
addresses several editorial and format changes which do not impact
the current CR-3 accident analyses.
The qualification of the OTSG tube repair roll methodology is
based on establishing a mechanical joint length that will carry all
structural loads imposed on the OTSG tubes while maintaining the
required margins during normal and accident conditions. A series of
tests and analyses were performed to establish the minimum
acceptable length of the OTSG tube repair roll. Tests performed
included leak, tensile, fatigue, ultimate load and eddy-current
measurement uncertainty. The analyses evaluated plant operating and
faulted load conditions, in addition to OTSG tubesheet bow effects.
OTSG tube leakage remains bounded by the evaluation presented in the
CR-3 Final Safety Analysis Report (FSAR) for a main steam line break
(MSLB). The proposed change also includes a description of the
required inspection program for the OTSG tube repair rolls. The
additional inspection requirements do not change any accident
initiators. The proposed inspections following OTSG tube repair roll
installation, and during future inservice inspections, assure
continuous monitoring of these tubes such that inservice degradation
of tubes repaired by the repair roll process will be detected. Based
on the qualification testing and analyses performed, as well as the
industry experience with the use of OTSG tube repair roll processes,
there are no new safety issues associated with the use of repair
roll methodology. Therefore, this change does not involve a
significant increase in the probability or consequences of any
accident previously evaluated.
(2) Create the possibility of a new or different kind of
accident from previously evaluated accidents?
The repair roll creates no new failure modes or accident
scenarios. The new pressure boundary joint created by the repair
roll process has been demonstrated, by testing and analysis, to
provide structural and leakage integrity equivalent to the original
design and construction for all normal operating and accident
conditions. Furthermore, the testing and analysis demonstrate the
repair roll process creates no new adverse effects for the repaired
tube and does not change the design or operating characteristics of
the OTSGs. In the unlikely event that a tube with a repair roll
should fail and sever completely at the transition of the repair
roll region, the tube would remain engaged in the tubesheet bore,
preventing interaction with other surrounding tubes. In this case,
leakage is bounded by the steam generator tube rupture (SGTR)
accident analysis. Therefore, this change does not create the
possibility of a new or different kind of accident from any
previously evaluated.
(3) Involve a significant reduction in a margin of safety?
The repair roll process effectively removes the defective/
degraded area of the tube from service. The repair roll interface
created with the tubesheet satisfies the necessary structural,
leakage and heat transfer requirements. The mechanical joint is
constrained within the tubesheet bore; thus, there is no additional
risk associated with tube rupture. The accident leakage is shown to
be less than one gallon per minute primary-to-secondary leakage.
Therefore, the FSAR analyzed accident scenarios remain bounding, and
the use of the repair roll process does not reduce 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 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: Coastal Region Library, 8619
W. Crystal Street, Crystal River, Florida 34428.
Attorney for licensee: R. Alexander Glenn, General Counsel, Florida
Power Corporation, MAC-A5A, P. O. Box 14042, St. Petersburg, Florida
33733-4042.
NRC Section Chief: Sheri R. Peterson.
Illinois Power Company, Docket No. 50-461, Clinton Power Station, Unit
1, DeWitt County, Illinois
Date of amendment request: March 1, 1999.
Description of amendment request: The proposed amendment would
approve changes to the Updated Safety Analysis Report (USAR) concerning
design requirements for physical protection from tornado missiles for
safety-related equipment.
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 activity does not involve a significant
increase in the probability or consequences of any accident
previously evaluated.
The associated USAR changes reflect use of the Electric Power
Research Institute (EPRI) Topical Report, ``Tornado Missile Risk
Evaluation Methodology, (EPRI NP-2005),'' Volumes I and II. This
methodology has been reviewed, accepted and documented in an NRC
Safety Evaluation dated October 26, 1983. The NRC concluded that:
``the EPRI methodology can be utilized when assessing the need for
positive tornado missile protection for specific safety-related
plant features in accordance with the criteria of SRP Section
3.5.1.4.''
The EPRI methodology has been previously applied at CPS to
resolve previously identified missile protection issues during the
initial licensing of the plant. The NRC documented their acceptance
of this methodology in Supplement 6 to the CPS Safety Evaluation
Report (NUREG-0853, July 1986).
As permitted in the Standard Review Plan (NUREG-0800), the total
probability of damage to plant systems or components initiated from
tornado missiles leading to consequences in excess of 10 CFR Part
100 guidelines will be maintained below an acceptable level. The
results of the current tornado missile hazards analysis are such
that the calculated total tornado missile hazard probability is
approximately 3.4 x 10-7 per year. This is lower than the value
determined to be acceptable, i.e., 1 x 10-6 per year.
Although it has been calculated that these targets have a higher
total probability of being exposed to tornado missiles than that
described to be acceptable in SER Supplement 6, Section 3.5.1.3, the
revised tornado missile hazards analysis for CPS has determined that
this probability is acceptably low.
With respect to the probability of occurrence or the
consequences of an accident previously analyzed in the USAR, the
possibility of a tornado reaching CPS and causing damage to plant
systems, structures and components is a design basis event
considered in the USAR. The changes being proposed herein do not
affect the probability that a tornado will reach the plant, but they
do, from a licensing basis perspective, reflect a slightly
increased, calculated probability that missiles generated by the
winds of a tornado might strike certain plant systems or components.
The tornado missile analysis determined that there are a limited
number of safety-related components that theoretically could be
struck. The probability of tornado-generated missile strikes on
important systems and components (as discussed in Regulatory Guide
1.117) was analyzed using the probability methods described above.
Based on the low, calculated probability, the total (cumulative)
probability of strikes will be maintained below an adequately low
acceptance criterion to ensure overall plant safety. On this basis,
the proposed change is not considered to constitute a significant
increase in the probability of occurrence or the consequences of an
accident, due to the low probability of a tornado missile striking
safety-related systems or components.
Therefore, the proposed changes do not involve a significant
increase in the probability or consequences of previously evaluated
accidents.
[[Page 19559]]
(2) The proposed activity does not create the possibility of a
new or different kind of accident from any accident previously
evaluated.
The proposed changes involve evaluation of whether any physical
protection of safety-related equipment from tornado missiles is
required relative to the probability of such damage without physical
protection. A tornado at CPS is a design basis event considered in
the USAR, however, a tornado is not postulated to act as an
initiator for any new or different kind of accident, or to occur
coincident with any of the design basis accidents in the USAR. The
low probability threshold established for missile damage to plant
systems is consistent with these assumptions.
Therefore, the proposed changes do not create the possibility of
a new or different kind of accident.
(3) The proposed activity does not involve a significant
reduction in a margin of safety.
Under the proposed change, physical protection of safety-related
equipment from tornado missiles must be considered if it has been
determined that the calculated total tornado missile hazard
probability is greater than 1 x 10-6 per year. The proposed change
to the USAR to specifically identify this threshold may slightly
increase the probability of a malfunction of equipment important to
safety previously evaluated in the safety analysis report (i.e.,
changing the requirements from protecting all safety-related systems
and components to not requiring protection if there is an extremely
low probability that a tornado missile could strike portions of
safety related systems and components). However, the changes are
consistent with the minimum acceptable requirements as documented in
the NRC's Safety Evaluation Report dated October 23, 1983.
Therefore, there will be no significant reduction to the margin of
safety that may be associated with the potential for safety-related
equipment to be damaged from tornado-generated missiles.
Therefore, the proposed changes do not involve a significant
reduction in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: Vespasian Warner Public
Library, 120 West Johnson Street, Clinton, IL 61727.
Attorney for licensee: Leah Manning Stetzner, Vice President,
General Counsel, and Corporate Secretary, 500 South 27th Street,
Decatur, IL 62525.
NRC Section Chief: Anthony J. Mendiola.
Northeast Nuclear Energy Company (NNECO), et al., Docket No. 50-336,
Millstone Nuclear Power Station, Unit No. 2, New London County,
Connecticut
Date of amendment request: March 17, 1999.
Description of amendment request: The licensee is proposing to
change Technical Specifications 3.5.2, ``Emergency Core Cooling
Systems--ECCS Subsystems--Tavg greater than or equal to 300 deg.F;''
3.7.1.7, ``Plant Systems--Atmospheric Steam Dump Valves;'' and 3.7.6.1,
``Plant Systems--Control Room Emergency Ventilation System.'' The
proposed Technical Specification changes will revise (1) surveillance
requirements for Emergency Core Cooling System valves, (2) the
atmospheric steam dump valve requirements to focus on the steam release
path instead of the individual valves, and (3) the allowed outage times
for the atmospheric steam dump valves and Control Room Emergency
Ventilation System.
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:
In accordance with 10 CFR 50.92, NNECO has reviewed the proposed
changes and has concluded that they do not involve a Significant
Hazards Consideration (SHC). The basis for this conclusion is that
the three criteria of 10 CFR 50.92(c) are not compromised. The
proposed changes do not involve an SHC because the changes would
not:
1. Involve a significant increase in the probability or
consequences of an accident previously evaluated.
Technical Specification 3.5.2
The removal of 2-CH-434, a manual valve, from the list of valves
to be checked every 31 days by Surveillance Requirement (SR)
4.5.2.a.10 will not change the requirement for this containment
isolation valve to be locked closed. The position of valve 2-CH-434,
and the associated locking device, will be verified by SR 4.6.1.1.a.
Although this change will result in the position of 2-CH-434 being
checked less often, there are sufficient Technical Specification and
administrative requirements to ensure that 2-CH-434 will be
maintained in the proper position. An additional benefit of this
proposed change will be a reduction in personnel exposure since 2-
CH-434 is located inside containment. This proposed change will not
result in any modification to Emergency Core Cooling System (ECCS)
alignment or operation.
The addition of the footnote to SR 4.5.2.a.10 will clarify that
2-SI-306 is pinned and locked open to the required throttle
position. 2-SI-306, which is the Shutdown Cooling (SDC) System
throttle valve in the discharge piping of the SDC pumps, is required
to be left in a throttled position after SDC has been secured to
ensure sufficient low pressure safety injection (LPSI) flow will be
available. This proposed change will not result in any modification
to ECCS alignment or operation.
The change in the valve nomenclature used in SR 4.5.2.e and
Table 4.5-1 from throttle valve to injection valve will eliminate
any confusion between valve description and valve operation. This
proposed change will not result in any modification to ECCS
alignment or operation.
The addition of the License Amendment Number to the bottom of
Page 3/4 5-6a will not result in a technical change to this
Technical Specification.
Technical Specification 3.7.1.7
The proposed changes will expand the scope of Technical
Specification 3.7.1.7 to include the steam release path, instead of
just the individual atmospheric dump valves (ADVs). The allowed
outage times will be modified to address inoperable ADV lines and
the impact inoperable ADV lines will have on the ability of
Millstone Unit No. 2 to mitigate a loss of coolant accident (LOCA).
If one ADV line is inoperable, a plant shutdown will be required if
the ADV line is not restored to operable status within 48 hours. An
allowed outage time of 48 hours to restore the ADV line to operable
status is acceptable based on the low probability of a LOCA
occurring during this time period, and the subsequent loss of
offsite power and the failure of one train of high pressure safety
injection (HPSI). This is also consistent with the allowed outage
time for one ECCS train (Technical Specification 3.5.2).
If two ADV lines are inoperable, a plant shutdown will be
required if at least one ADV line is not restored to operable status
within one hour. The plant will be required to be in Mode 3 within
the following 6 hours. These time requirements are based on
Technical Specification 3.0.3. However, the time to reach Mode 4
will remain at the ``following 24 hours'' to reflect the impact
inoperable ADV lines may have on the time to cool down the plant.
The proposed change to the surveillance requirement will ensure
operation of the ADV lines, consistent with the accident analysis,
is verified.
The proposed change in component nomenclature is consistent with
current Millstone Unit No. 2 terminology. This is not a technical
change.
The proposed changes to the Bases of Technical Specification
3.7.1.7 are consistent with the changes just described.
Technical Specification 3.7.6.1
The action requirements for the Control Room Emergency
Ventilation System will be modified to address the situation when
both Control Room Emergency Ventilation Trains are inoperable in
Modes 1, 2, 3, and 4. This situation is expected to occur during
normal plant operation when the air filters in the common supply
header to both trains are cleaned/replaced. Since this is a common
supply header, both trains are affected and would be inoperable. The
proposed action requirements will address this situation so
[[Page 19560]]
that Technical Specification 3.0.3 will not be entered as a result
of an expected plant activity. However, since the proposed action
requirements are the same as the requirements of Technical
Specification 3.0.3, the time the plant is allowed to operate in
this situation will not change.
The proposed changes to the Technical Specifications and
associated Bases will have no adverse effect on plant operation or
accident mitigation equipment. The proposed changes will ensure that
the necessary equipment to mitigate the design basis accidents will
be available, or a plant shutdown will be required. In addition, the
proposed changes can not cause an accident, and they will ensure the
accident mitigation equipment will continue to operate as assumed in
the analyses to mitigate the design basis accidents. Therefore,
there will be no 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.
The proposed changes to the Technical Specifications and
associated Bases will have no adverse effect on plant operation or
accident mitigation equipment. The proposed changes will ensure that
the necessary equipment to mitigate the design basis accidents will
be available, or a plant shutdown will be required. Therefore, the
proposed changes will 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 proposed changes to the Technical Specifications and
associated Bases will ensure that the necessary equipment to
mitigate the design basis accidents will be available, or a plant
shutdown will be required. The proposed changes will not result in
any plant configuration changes. There will be no adverse effect on
plant operation or accident mitigation equipment. The plant response
to the design basis accidents will not change. Therefore, there will
be no significant reduction in the margin of safety as defined in
the Bases for the Technical Specifications affected by these
proposed changes.
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, Connecticut, and the Waterford Library, ATTN: Vince Juliano,
49 Rope Ferry Road, Waterford, Connecticut.
Attorney for licensee: Lillian M. Cuoco, Esq., Senior Nuclear
Counsel, Northeast Utilities Service Company, P.O. Box 270, Hartford,
Connecticut.
NRC Section Chief: James W. Clifford.
Northeast Nuclear Energy Company (NNECO), et al., Docket No. 50-336,
Millstone Nuclear Power Station, Unit No. 2, New London County,
Connecticut
Date of amendment request: March 19, 1999.
Description of amendment request: The proposed changes will
relocate Technical Specifications (TSs) 3.3.3.2, ``Instrumentation,
Incore Detectors,'' 3.3.3.3, ``Instrumentation, Meteorological
Instrumentation,'' to the Millstone, Unit No. 2 Technical Review Manual
(TRM). Index Page V will be revised by eliminating the sections
corresponding to incore detectors (Page \3/4\ 3-0), seismic
instrumentation (Page \3/4\ 3-32), and meteorological instrumentation
(Page \3/4\ 3-36). These sections, as well as changes to the associated
Bases, will be relocated to the TRM.
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:
In accordance with 10 CFR 50.92, NNECO has reviewed the proposed
changes and has concluded that they do not involve a Significant
Hazards Consideration (SHC). The basis for this conclusion is that
the three criteria of 10 CFR 50.92(c) are not compromised. The
proposed changes do not involve an SHC because the changes would
not:
1. Involve a significant increase in the probability or
consequences of an accident previously evaluated.
Technical Specification 3.3.3.2, Instrumentation, ``Incore
Detectors,'' is proposed to be relocated to the TRM where future
changes will be controlled in accordance with 10 CFR 50.59.
Relocation of this Technical Specification to the TRM does not imply
any reduction in its importance in confirming that core power
distribution are bounded by safety analysis limits. These
instruments are neither used for, nor capable of, detecting a
significant abnormal degradation of the reactor coolant pressure
boundary before a design basis accident, nor do they function as a
primary success path to mitigate events which assume a failure of,
or a challenge to, the integrity of fission product barriers.
Although the core power distribution (measured by the incore
detectors) constitutes an important initial condition to design
basis accidents and therefore needs to be addressed by Technical
Specifications, the detectors themselves are not an active design
feature needed to preclude analyzed accidents or transients. The
proposed change will not alter the way core power distribution is
measured by the incore detectors, nor will it alter any of the power
distribution assumptions used in the accident analysis. Therefore,
this change will not significantly increase the probability or
consequences of an accident previously evaluated.
Technical Specification 3.3.3.3, Instrumentation, ``Seismic
Instrumentation,'' is proposed to be relocated to the TRM where
future changes will be controlled in accordance with 10 CFR 50.59.
Relocation of Technical Specification 3.3.3.3 to the TRM does not
imply any reduction in its importance in determining the response of
those nuclear power plant features important to safety in the event
of an earthquake. Seismic instrumentation does not actuate any
protective equipment or serve any direct role in the mitigation of
an accident. The capability of the plant to withstand a seismic
event or other design basis accident is determined by the initial
design and construction of systems, structures, and components. The
instrumentation is used to alert operators to the seismic event and
evaluate the plant response. The seismic instrumentation does not
serve as a protective design feature or part of a primary success
path for events which challenge fission product barriers. The
proposed change will not alter the way these instruments are used in
determining the response of those nuclear power plant features
important to safety in the event of an earthquake, nor will it alter
the capability of the plant to withstand a seismic event. Therefore,
this change will not significantly increase the probability or
consequences of an accident previously evaluated.
Technical Specification 3.3.3.4, Instrumentation,
``Meteorological Instrumentation,'' is proposed to be relocated to
the TRM where future changes will be controlled in accordance with
10 CFR 50.59. Relocation of Technical Specification 3.3.3.4 to the
TRM does not imply any reduction in its importance in providing a
basis for estimating annual radiation doses resulting from
radioactive materials released in airborne effluents. The
instrumentation does not serve to ensure that the plant is operated
within the bounds of initial conditions assumed in design basis
accident and transient analyses or that the plant will be operated
to preclude transients or accidents. Likewise, the meteorological
instrumentation does not serve as part of the primary success path
of a safety sequence analysis used to demonstrate that the
consequences of these events are within the appropriate acceptance
criteria. The proposed change will not alter the way these
instruments are used in providing a basis for estimating annual
radiation doses resulting from radioactive materials released in
airborne effluents. Therefore, this change will not significantly
increase the probability or consequences of an accident previously
evaluated.
Revision of Index page V and the proposed changes to the
associated Bases sections are administrative changes. Therefore,
these changes will not significantly increase the probability or
consequences of an accident previously evaluated.
The proposed changes do not alter how any structure, system, or
component functions. There will be no effect on
[[Page 19561]]
equipment important to safety. The proposed changes have no effect
on any of the design basis accidents previously evaluated.
Therefore, this License Amendment Request does not impact the
probability of an accident previously evaluated, nor does it involve
a significant increase in the consequences of an accident previously
evaluated.
2. Create the possibility of a new or different kind of accident
from any accident previously evaluated.
The proposed changes do not alter the plant configuration (no
new or different type of equipment will be installed) or require any
new or unusual operator actions. They do not alter the way any
structure, system, or component functions and do not alter the
manner in which the plant is operated. The proposed changes do not
introduce any new failure modes. Therefore, the proposed changes
will 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 proposed relocation of incore detector instrumentation
requirements to the TRM does not imply any reduction in their
importance in confirming that core power distribution is bounded by
safety analysis limits. The incore detectors will still be used to
measure core power distribution and the assumptions used in the
accident analysis will be verified. The proposed relocation of
seismic instrumentation requirements to the TRM does not imply any
reduction in their importance in determining the response of those
nuclear power plant features important to safety in the event of an
earthquake. The seismic instrumentation will still be used to
determine the response of those nuclear power plant features
important to safety in the event of an earthquake. The capability of
the plant to withstand a seismic or other design basis accident,
which is determined by the initial design and construction of
systems, structures, and components will not be altered. The
relocation of meteorological instrumentation requirements to the TRM
does not change the way these instruments are used in providing a
basis for estimating annual radiation doses resulting from
radioactive materials released in airborne effluents. The
meteorological instrumentation will continue to perform their
function in exactly the same way.
The proposed changes do not affect any of the assumptions used
in the accident analysis, nor do they affect any operability
requirements for equipment important to plant safety. Therefore, the
proposed changes will not result in a significant reduction in the
margin of safety as defined in the Bases for Technical
Specifications covered in this License Amendment Request.
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, Connecticut, and the Waterford Library, ATTN: Vince Juliano,
49 Rope Ferry Road, Waterford, Connecticut.
Attorney for licensee: Lillian M. Cuoco, Esq., Senior Nuclear
Counsel, Northeast Utilities Service Company, P.O. Box 270, Hartford,
Connecticut.
NRC Section Chief: James W. Clifford.
Pacific Gas and Electric Company, Docket Nos. 50-275 and 50-323, Diablo
Canyon Nuclear Power Plant, Unit Nos. 1 and 2, San Luis Obispo County,
California
Date of amendment request: September 11, 1998, as supplemented by
letter dated January 14, 1999.
Description of amendment request: The proposed amendments would
change the combined Technical Specifications (TS) for the Diablo Canyon
Power Plant, Unit Nos. 1 and 2 to revise TS 6.8.4f., ``Containment
Polar and Turbine Building Cranes,'' to control the operation of the
containment polar cranes in jet impingement zones.
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 Technical Specification (TS) 6.8.4f requirement to have a
program that will ensure the position of the polar cranes precludes
jet impingement from a postulated pipe rupture was previously
evaluated in the NRC staff's safety evaluation for License
Amendments (LA) 20 and 21. The proposed change is to control the
operation of the containment polar cranes in jet impingement zones.
PG&E evaluated a high energy line break (HELB) scenario for core
damage frequency (CDF) considering operation of a polar crane. A
postulated HELB would have to damage the crane or cause its load to
drop in a manner that damages a component that exacerbates the HELB
event and leads to core damage. The PRA evaluation for this scenario
concluded the CDF is 1.6E-9 per year. It is not a significant
increase in CDF compared to never operating the polar crane in jet
impingement zones. The CDF for this scenario is nonrisk significant
when compared to the industry standard threshold for risk
significance for an operational evolution, which is 1E-6 per year.
Several factors that further lower the risk of CDF include: 1) the
movement of heavy loads is done in accordance with the DCPP Heavy
Loads Program, which provides assurance that a dropped load would
not lead to core damage, 2) the polar crane had been evaluated to
withstand jet impingement loads without the seismic loads, and 3)
the probability of simultaneous seismic and HELB events is low.
Therefore, based on probabilistic considerations, the risk
associated with 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.
Deterministic engineering methods required combining both the
seismic and jet impingement loads to qualify Design Class I
structures. The polar cranes were not originally qualified for these
combined loads. This resulted in administrative controls that
prohibited parking the polar cranes in jet impingement zones to
preclude jet impingement loads from a postulated pipe rupture. The
proposed change does not involve a physical change to the plant, but
it does involve a change to the TS required program for containment
polar crane operation.
The proposed change is to control the operation of the
containment polar cranes in jet impingement zones. It recognizes
that there are jet (HELB) and target (polar crane) interactions.
They were previously not considered for postulated jet impingement
analyses because administrative controls prohibited parking the
polar cranes in jet impingement zones. PG&E has evaluated jet
impingement loads on the polar crane and determined it is able to
withstand these loads without seismic loads. Based on this
evaluation, the polar crane would not fail due to a HELB event. The
movement of a heavy load would be done in accordance with the DCPP
Heavy Loads Program. Thus, there would be no consequential failures
that would lead to core damage.
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.
The current TS 6.8.4f. requirement to have a program that will
ensure the position of the polar cranes precludes jet impingement
from a postulated pipe rupture was previously evaluated in the NRC
staff's safety evaluation for LAs 20 and 21.
The credible HELB sources that could impinge on the polar crane
were identified and evaluated. The feedwater and main steam line
steam generator nozzles are the only credible HELBs that could
impinge upon the polar crane. The structural integrity of these
lines was evaluated and determined to be of robust design.
The margin of safety affected by the proposed change involves a
comparison between the margin of safety afforded by no operation of
the polar crane and operation that is controlled by procedures. The
margin of safety in this case is the increase in risk for CDF caused
by a scenario that postulates that operation of the polar crane
would lead
[[Page 19562]]
to core damage. The risk for CDF has been evaluated and determined
to be nonrisk significant. The CDF value is well below the industry
standard threshold for acceptable risk for an operational evolution,
which is 1E-6 per year.
Therefore, the proposed change does not involve a significant
reduction in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 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 &
Electric Company, P.O. Box 7442, San Francisco, California 94120.
NRC Project Director: Stuart A. Richards.
Pacific Gas and Electric Company, Docket Nos. 50-275 and 50-323, Diablo
Canyon Nuclear Power Plant, Unit Nos. 1 and 2, San Luis Obispo County,
California
Date of amendment request: December 12, 1998.
Description of amendment request: The proposed amendments would
change the combined Technical Specifications (TS) for the Diablo Canyon
Power Plant, Unit Nos. 1 and 2 to revise TS 6.9.1.8, ``Core Operating
Limits Report,'' to allow use of NRC approved addenda to WCAP-10054-P-
A, ``Westinghouse Small Break ECCS Evaluation Model Using NOTRUMP
Code,'' August 1985, to determine core operating limits.
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 change is administrative in nature in that it revises the
Technical Specification (TS) Administrative Controls for the Core
Operating Limits Report to include reference to NRC approved addenda
to WCAP-10054-P-A, ``Westinghouse Small Break ECCS Evaluation Model
Using the NOTRUMP Code,'' August 1985. The proposed change would
allow the use of the analytical methods in WCAP-10054-P-A, Addendum
2, Revision 1, Addendum to the Westinghouse Small Break ECCS.
Evaluation Model Using the NOTRUMP Code: Safety Injection Into
the Broken Loop and COSI Condensation Model,'' July 1997, and other
NRC approved addenda to WCAP-10054-P-A to determine core operating
limits for Diablo Canyon Power Plant (DCPP). Because plant operation
will continue to be limited in accordance with cycle specific core
operating limits that are established using an NRC approved
methodology, NRC approved addenda to WCAP-10054-P-A are acceptable
for use in determining DCPP Unit 1 and 2 cycle specific core
operating limits.
The change does not affect plant operation, or physically alter
or change the function of structures, systems, or components
required to mitigate the consequences of a design basis accident. In
addition, it cannot initiate a transient or affect the probability
of occurrence of any previously analyzed 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 proposed change revises the TS to allow the use of NRC
approved analytical methods in WCAP-10054-P-A, Addendum 2, Revision
1, and other NRC approved addenda to WCAP-10054-P-A, to determine
core operation limits. The change is consistent with the
requirements of the TS, and does not affect plant operation, or
physically alter or change the function of structures, systems, or
components required to mitigate the consequences of a design basis
accident.
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.
The proposed change revises the TS to allow the use of the NRC
approved analytical methods in WCAP-10054-P-A, Addendum 2, Revision
1 and other NRC approved addenda to WCAP-10054-P-A, to determine
core operating limits. The change is consistent with the
requirements of the TS, and does not affect plant operation, or
physically alter or change the function of structures, systems, or
components required to mitigate the consequences of a design basis
accident. The acceptance limits for the small break loss-of-coolant
accident are not affected by this change and will continue to be
met.
Therefore, the proposed change does not involve a significant
reduction in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 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 &
Electric Company, P.O. Box 7442, San Francisco, California 94120.
NRC Project Director: Stuart A. Richards.
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: January 25, 1999.
Description of amendment request: This application for amendment to
the Indian Point 3 (IP3) Technical Specifications (TSs) proposes to
relocate the time restriction for movement of irradiated fuel and its
related basis page from the TSs to the IP3 Final Safety Analysis Report
(FSAR).
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 proposed license amendment involve a significant
increase in the probability or consequences of an accident
previously [evaluated]?
Response
Relocation (i.e., removal from TS) of TS 3.8.A.9 and its basis
for the minimum time prior to movement of more than 76 irradiated
fuel assemblies (267 hour limit) will not involve a significant
increase in the probability or consequences of an accident since the
relocation of the TS to administrative controls governed by 10 CFR
50.59 (FSAR) does not affect the availability or function of fuel
storage and handling equipment or the SFP [spent fuel pool] cooling
system. The waiting time of 267 hours following plant shutdown
before unloading more than 76 assemblies from the reactor is to
ensure that the maximum SFP water temperature will be within design
objectives as stated in the FSAR.
The waiting time of 267 hours is not an initiator of an accident
and the proposed change does not alter overall system operation,
physical design, system configuration, or operational setpoints.
There will be no significant increase in the consequences of an
accident because the restricted movement time for irradiated fuel
will continue to be administratively controlled under 10 CFR 50.59.
The other TS of section 3.8.A (such as the remaining portion of
3.8.A.9, and 3.8.A. 10) and the other controls ensure that doses
from a postulated FHA are within 10 CFR 100 limits.
[[Page 19563]]
(2) Does the proposed license amendment create the possibility
of a new or different kind of accident from any accident previously
evaluated?
Response
The basis for the waiting time of 267 hours following plant
shutdown before unloading more than 76 assemblies from the reactor
is to ensure that the maximum pool water temperature will be within
design objectives as stated in the FSAR. Relocation of this waiting
time of 267 hours for irradiated fuel will not create the
possibility of a new or different kind of accident from any
previously evaluated. The TS change will not create the possibility
of a new or different kind of accident from any previously evaluated
since it does not alter the administrative controls for fuel
handling or the operation, physical design, system configuration, or
operational setpoints for fuel handling and SFP cooling. The plant
systems for fuel storage and handling, and SFP cooling are operated
in the same manner as before and, consequently, the relocation does
not introduce any new accident initiators or failure mechanisms and
does not invalidate the existing FHA response. The minimum waiting
time for movement of more than 76 irradiated fuel assemblies is not
an accident initiator. The minimum waiting time will continue to be
controlled under 10 CFR 50.59.
(3) Does the proposed amendment involve a significant reduction
in a margin of safety?
Response
Relocation (i.e., removal from TS) of TS 3.8.A.9 and its basis
for the waiting time of 267 hours following plant shutdown for
irradiated fuel will not involve a significant reduction in margin
of safety. The waiting time of 267 hours following plant shutdown
before unloading more than 76 assemblies from the reactor is to
ensure that the maximum SFP water temperature will be within design
objectives as stated in the FSAR. The relocation is a change to the
administrative controls that are used to limit the heat load on the
SFP cooling system, and those administrative controls will be
governed by 10 CFR 50.59. The manner in which fuel storage and
handling is performed, and how the SFP cooling system is operated
does not change and there is no change to physical design, system
configuration, or operational setpoints. The other controls and the
existing TS assure that dose from a postulated FHA are within 10 CFR
100 limits. Previous analyses remain unchanged. The current TS does
not meet the criteria in 10 CFR 50.36(c)(2)(ii) for inclusion in the
Technical Specifications.
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. David E. Blabey, 10 Columbus Circle, New
York, New York 10019.
NRC Section Chief: S. Singh Bajwa.
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: January 28, 1999.
Description of amendment request: This application for amendment to
the Indian Point 3 (IP3) Technical Specifications (TSs) proposes to
change the setpoint of the automatic reactor trip on turbine trip to at
or below the P-8 setpoint.
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 proposed license amendment involve a significant
increase in the probability or consequences of an accident
previously evaluated?
The addition of reactor trip on turbine trip at [greater than or
equal to] 50% to the P-8 Permissive function versus its current
setting of [greater than or equal to] 10%, as revised in TS section
2.3.1.C.(3), 2.3.2.A, 2.3.2.B, Table 3.5-2, item 12, Table 4.1-1,
item 21 and associated bases, does not significantly increase the
probability or consequences of an accident previously evaluated.
This additional function, change in reactor trip on turbine trip
setpoint, does not cause the initiation of any accident, nor create
any new credible limiting single failure, nor result in any event
previously deemed incredible being made credible. The existing
separation of the reactor and protection functions are not adversely
impacted. In addition, the safety functions of safety related
systems and component, which are related to accident mitigation,
have not been altered. The change in the P-7 or P-8 circuitry does
not directly initiate an accident. The consequences of accidents
previously [evaluated] in the IP3 FSAR [final safety analysis
report] are unaffected by this change because no change to any
equipment response or accident mitigation scenario has resulted.
There are no additional challenges to fission product barrier
integrity. Therefore, the probability or consequences of an accident
previously evaluated will not be increased.
(2) Does the proposed license amendment create the possibility
of a new or different kind of accident from any accident previously
evaluated?
By adding the reactor trip on turbine trip at [greater than or
equal to] 50% to the P-8 Permissive function and setpoint, versus
its current setting of [greater than or equal to] 10% and revising
TS sections 2.3. l.C.(3), 2.3.2.A, 2.3.2.B, Table 3.5-2, item 12,
Table 4.1-1, item 21 and associated bases, does not create the
possibility of a new or different kind of accident than any accident
already evaluated. The additional function added to the P-8
Permissive does not result in any event previously deemed incredible
being made credible. No new accident scenarios, failure mechanisms,
or limiting single failures are introduced as a result of this
change. In addition, the safety functions of safety related systems
and components, which are related to accident mitigation, have not
been altered. Therefore, the possibility of a new or different kind
of accident is not created.
(3) Does the proposed amendment involve a significant reduction
in a margin of safety?
The addition of the reactor trip on turbine trip at [greater
than or equal to] 50% to the P-8 Permissive function, versus its
current setting of [greater than or equal to] 10% and associated
changes to TS Sections 2.3. l.C.(3), 2.3.2.A, 2.3.2.B, Table 3.5-2,
item 12, Table 4.1-1, item 21 and the associated bases, will have no
effect on the availability, operability or performance of the
safety-related systems and components and does not affect the plant
TS requirements. The current licensing basis safety analyses for IP3
remain bounding with the modification to the P-8 Permissive
function; therefore, the margin of safety as defined in the TS is
not reduced. The change to the IP3 TS does not involve a significant
reduction in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: White Plains Public Library,
100 Martine Avenue, White Plains, New York 10601.
Attorney for licensee: Mr. David E. Blabey, 10 Columbus Circle, New
York, New York 10019.
NRC Section Chief: S. Singh Bajwa.
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: January 28, 1999.
Description of amendment request: This application for amendment to
the Indian Point 3 Technical Specifications (TSs) proposes to reduce
the number of Emergency Diesel Generators (EDGs) required to be
operable during cold shutdown from 2 to 1 under certain conditions.
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 proposed license amendment involve a significant
increase in the
[[Page 19564]]
probability or consequences of an accident previously [evaluated]?
Response
No. The equipment, which is affected by the proposed Technical
Specification change, is not an initiator to those accidents
postulated to occur during Cold Shutdown or Refueling operating
conditions. A comprehensive systems review and EDG loading
electrical analysis has demonstrated the ability of those shutdown
support systems, necessary to provide safe shutdown needs, to
perform their accident mitigation functions for the postulated
accidents during Cold Shutdown and Refueling conditions. One EDG can
support the necessary electrical loads required in Cold Shutdown and
Refueling in the event of postulated accidents along with a LOOP
[loss of offsite power] in the time frame required to prevent
reactor core/cavity/SFP [spent fuel pool] heatup concerns. This EDG
support relies upon existing plant designed manual closure of 480VAC
EDS [electrical distribution system] bus tie breakers to allow a
single EDG to pick up other 480VAC EDS bus loads, such as supplying
an RHR [residual heat removal] pump and SFP cooling pump, located on
480VAC EDS buses 3A, 5A, or 6A. Together, operability of the
required offsite circuit(s) and one EDG ensures the availability of
sufficient AC sources to operate the unit in a safe manner and to
mitigate the consequences of postulated accidents during shutdown
(e.g., Fuel Handling Accidents). Action statements provide prompt,
specific guidance to ensure sufficiently conservative plant response
should the expected EDG power supply not be available. These Action
Statements are similar to those in the STS [Standard Technical
Specifications]. Therefore, the proposed license amendment (i.e.,
changes to 3.7.F.4 and the added sections of 3.7.F.5 & 3.7.F.6) does
not involve a significant increase in the probability or
consequences of an accident previously analyzed.
(2) Does the proposed license amendment create the possibility
of a new or different kind of accident from any accident previously
evaluated?
Response
No. The proposed license amendment does not involve any physical
changes to plant systems or component set points. The use of 480VAC
EDS bus tie breakers to power loads from an energized 480VAC bus is
part of present plant design and included within the present LOOP
Off-Normal operating procedures when the reactor is in Cold Shutdown
operating conditions. As discussed in the Standard Technical
Specifications, NUREG 1431, during plant shutdown with one EDG, it
is not required to assume a single failure and concurrent loss of
all offsite or all onsite power. Worst case bounding events are
deemed not credible in Cold Shutdown and Refueling conditions
because the energy contained within the reactor pressure boundary,
reactor coolant temperature and pressure, and the corresponding
stresses result in the probabilities of occurrence being
significantly reduced or eliminated, and ultimately result in
minimal consequences. The lone EDG is capable of accepting and
starting required loads within the assumed loading sequence
intervals and continue to operate until offsite power can be
provided to the 480VAC EDS buses. Action statements provide prompt,
specific guidance to ensure sufficiently conservative plant response
should the expected EDG power supply not be available. These action
statements are similar to those in the STS. Therefore, the proposed
license amendment (i.e., changes to 3.7.F.4 and added sections
3.71.5 & 3.7.F.6) does not create the possibility of a new or
different kind of accident from any accident previously evaluated.
(3) Does the proposed amendment involve a significant reduction
in a margin of safety?
Response
No. The electrical power system specifications support the
equipment required to be operable, commensurate with the current
level of safety, including the equipment requiring an EDG backed
power source. The design review results demonstrate that operation
in the conditions of Cold Shutdown and Refueling, in accordance with
the proposed Technical Specification change, is acceptable from an
accident mitigation standpoint. The basic system functions in Cold
Shutdown and Refueling operating conditions are not changed. One EDG
can supply the necessary electrical power needs during these plant
operating conditions, and in the time frame required to prevent
reactor core/cavity/SFP heatup concerns, with sufficient ``kw
loading'' to spare. The analysis conducted shows that the systems
are capable of performing their design basis functions. Applicable
safety analysis in the Standard Technical Specifications, NUREG
1431, discusses these system requirements as well (i.e., it is not
required to assume a single failure and concurrent loss of all
offsite or all onsite power). Action statements, similar to those in
the Standard Technical Specifications, provide prompt, specific
guidance to ensure sufficiently conservative plant response should
the expected EDG power supply not be available. On this basis, the
proposed license amendment (i.e., changes to 3.7.F.4 and added
sections 3.7.F.5 & 3.7.F.6) does not involve a significant reduction
in the margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 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. David E. Blabey, 10 Columbus Circle, New
York, New York 10019.
NRC Section Chief: S. Singh Bajwa.
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: January 29, 1999.
Description of amendment request: This application for amendment to
the Indian Point 3 (IP3) Technical Specifications (TSs) proposes to
change the allowable indicated control rod misalignment.
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 proposed license amendment involve a significant
increase in the probability or consequences of an accident
previously evaluated?
Response
No. Based on the Westinghouse evaluation in WCAP-14668, the
Authority has determined that all pertinent licensing basis
acceptance criteria have been met, and the margin of safety as
defined in the TS Bases is not reduced in any of the IP3 licensing
basis accident analysis. Increasing the magnitude of allowed control
rod indicated misalignment (in section 3.10.5) is not a contributor
to the mechanistic cause of an accident evaluated in the FSAR [Final
Safety Analysis Report]. Neither the rod control system nor the rod
position indicator function is being altered. Therefore, the
probability of an accident previously evaluated has not
significantly increased. Because design limitations continue to be
met, and the integrity of the reactor coolant system pressure
boundary is not challenged, the assumptions employed in the
calculation of the offsite radiological doses remain valid.
Therefore, the consequences of an accident previously evaluated
will not be significantly increased.
(2) Does the proposed license amendment create the possibility
of a new or different kind of accident from any accident previously
evaluated?
Response
No. Based on the Westinghouse evaluation in WCAP-14668, the
Authority has determined that all pertinent licensing basis
acceptance criteria have been met, and the margin of safety as
defined in the TS Bases is not reduced in any of the IP3 licensing
basis accident analysis. Increasing the magnitude of allowed control
rod indicated misalignment is not a contributor to the mechanistic
cause of any accident. Neither the rod control system nor the rod
position indicator function is being altered. Therefore, an accident
which is new or different than any previously evaluated will not be
created.
(3) Does the proposed amendment involve a significant reduction
in a margin of safety?
Response
No. Based on the Westinghouse evaluation in WCAP-14668, the
Authority has
[[Page 19565]]
determined that all pertinent licensing basis acceptance criteria
have been met, and the margin of safety as defined in the TS Bases
is not reduced in any of the IP3 licensing basis accident analysis
based on the changes to safety analyses input parameter values as
discussed in WCAP-14668. Since the evaluations in Section 3.0 of
WCAP-14668 demonstrate that all applicable acceptance criteria
continue to be met, the proposed change will not involve a
significant reduction in 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. David E. Blabey, 10 Columbus Circle, New
York, New York 10019.
NRC Section Chief: S. Singh Bajwa.
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: January 29, 1999.
Description of amendment request: This application for amendment to
the Indian Point 3 (IP3) Technical Specifications (TSs) proposes to
change the allowable indicated control rod misalignment.
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 proposed license amendment involve a significant
increase in the probability or consequences of an accident
previously evaluated?
Response
No. Based on the Westinghouse evaluation in WCAP-14668, the
Authority has determined that all pertinent licensing basis
acceptance criteria have been met, and the margin of safety as
defined in the TS Bases is not reduced in any of the IP3 licensing
basis accident analysis. Increasing the magnitude of allowed control
rod indicated misalignment (in Section 3.10.5) is not a contributor
to the mechanistic cause of an accident evaluated in the FSAR [Final
Safety Analysis Report]. Neither the rod control system nor the rod
position indicator function is being altered. Therefore, the
probability of an accident previously evaluated has not
significantly increased. Because design limitations continue to be
met, and the integrity of the reactor coolant system pressure
boundary is not challenged, the assumptions employed in the
calculation of the offsite radiological doses remain valid.
Therefore, the consequences of an accident previously evaluated
will not be significantly increased.
(2) Does the proposed license amendment create the possibility
of a new or different kind of accident from any accident previously
evaluated?
Response
No. Based on the Westinghouse evaluation in WCAP-14668, the
Authority has determined that all pertinent licensing basis
acceptance criteria have been met, and the margin of safety as
defined in the TS Bases is not reduced in any of the IP3 licensing
basis accident analysis. Increasing the magnitude of allowed control
rod indicated misalignment is not a contributor to the mechanistic
cause of any accident. Neither the rod control system nor the rod
position indicator function is being altered. Therefore, an accident
which is new or different than any previously evaluated will not be
created.
(3) Does the proposed amendment involve a significant reduction
in a margin of safety?
Response
No. Based on the Westinghouse evaluation in WCAP-14668, the
Authority has determined that all pertinent licensing basis
acceptance criteria have been met, and the margin of safety as
defined in the TS Bases is not reduced in any of the IP3 licensing
basis accident analysis based on the changes to safety analyses
input parameter values as discussed in WCAP-14668. Since the
evaluations in Section 3.0 of WCAP-14668 demonstrate that all
applicable acceptance criteria continue to be met, the proposed
change will not involve a significant reduction in 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. David E. Blabey, 10 Columbus Circle, New
York, New York 10019.
NRC Section Chief: S. Singh Bajwa.
STP Nuclear Operating Company, Docket Nos. 50-498 and 50-499, South
Texas Project, Units 1 and 2, Matagorda County, Texas
Date of amendment request: March 22, 1999.
Description of amendment request: The proposed amendments would
revise Technical Specification (TS) 3.7.1.6, ``Atmospheric Steam Relief
Valves,'' and add a new TS for atmospheric steam relief valve
instrumentation, to ensure that the automatic feature of the steam
generator power-operated relief valve (i.e., atmospheric steam relief
valves) remains operable during Modes 1 and 2. In addition, the
proposed change would add an associated surveillance requiring that a
channel calibration on the steam generator power-operated relief valve
be performed every 18 months.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. Does the change involve a significant increase in the
probability or consequences of an accident previously evaluated?
The methodologies used in the accident analyses remain
unchanged. The automatic actuation of the Steam Generator Power
Operated Relief Valves is not a new design feature. The effects of
the inadvertent opening of a Steam Generator Power Operated Relief
Valve are currently analyzed as described in Section 15.1.4 of the
Updated Final Safety Analysis Report. The radiological consequences
for the Small Break Loss of Coolant Accident (SBLOCA) event
presented in the Updated Final Safety Analysis Report remain
unchanged. The calculated Peak Clad Temperature is 1849 deg.F
remaining substantially below the 2200 deg.F acceptance limit of 10
CFR 50.46. Although the manual control specification is relocated
from Specification 3.7.1.6 to the new instrumentation specification,
the limiting condition for operation, applicability and action
statements for manual controls remain unchanged. Therefore no
increase in the probability or consequences of any accident
previously evaluated will occur.
2. Does the proposed change create the possibility of a new or
different kind of accident from any accident previously evaluated?
The automatic actuation of the Steam Generator Power Operated
Relief Valves is not an accident initiator for the SBLOCA event. The
automatic actuation of the Steam Generator Power Operated Relief
Valves currently exists at the South Texas Project and is not a new
design feature. The description of the Steam Generator Power
Operated Relief Valves currently exists in the Updated Final Safety
Analysis Report. This change does not represent a change to the
facility and does not affect the safety functions and reliability of
systems, structures, or components in any new manner. Operating
procedures have a temporary administrative control to ensure the
automatic actuation of the Steam Generator Power Operated Relief
Valves remains operable in Modes 1 and 2. This condition will become
permanent with the approval of this Technical Specification
Amendment proposal. Although the manual control specification is
relocated from Specification 3.7.1.6 to the new instrumentation
specification, the limiting condition for operation, applicability
and
[[Page 19566]]
action statements for manual controls remain unchanged. Since the
automatic actuation of the Steam Generator Power Operated Relief
Valves is not an accident initiator and is not a new design feature
to the facility, no possibility exists for a new or different kind
of accident from those previously evaluated.
3. Does this change involve a significant reduction in a margin
of safety?
The proposed change results in the calculated Peak Clad
Temperature of 1849 deg.F remaining well below the acceptance limit
of 10 CFR 50.46 and comparable to the results currently described in
the Updated Final Safety Analysis Report. Therefore, the change does
not involve a significant reduction in a margin of safety.
Based on the above, the South Texas Project has evaluated the
proposed Technical Specification change and determined it does not
represent a significant hazards consideration.
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, Texas 77488.
Attorney for licensee: Jack R. Newman, Esq., Morgan, Lewis &
Bockius, 1800 M Street, N.W., Washington, DC 20036-5869.
NRC Section Chief: Robert A. Gramm.
Tennessee Valley Authority (TVA), Docket Nos. 50-327 and 50-328,
Sequoyah Nuclear Plant, (SQN), Units 1 and 2, Hamilton County,
Tennessee
Date of application for amendments: March 19, 1999 (TS 99-01).
Brief description of amendments: The proposed amendments would
change the SQN Technical Specifications (TS) for Operating Licenses
DPR-77 (Unit 1) and DPR-79 (Unit 2) by relocating TS Sections 3.8.3.1,
3.8.3.2, and 3.8.3.3 to the SQN Technical Requirements Manual. These
sections provide requirements for electrical overcurrent isolation
devices.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
A. The proposed amendment does not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
The proposed revision to the TS relocates the requirements for
SQN's electrical equipment protective devices without changing the
current requirements. TVA does not consider these devices to be the
source of any accident; therefore, this administrative relocation of
the requirements will not increase the possibility of an accident.
SQN's electrical equipment protective devices will continue to
provide fault protection for circuits and equipment. Changes to the
relocated requirements will be processed, in accordance with 10 CFR
50.59, to ensure changes are not implemented that would reduce the
functionality or introduce an unreviewed safety question to SQN's
electrical equipment devices. Therefore, the proposed relocation of
the TS requirements for electrical equipment protective devices will
not increase the consequences of an accident.
B. The proposed amendment does not create the possibility of a
new or different kind of accident from any accident previously
evaluated.
SQN's electrical equipment protective devices ensure proper
operation of plant equipment. These devices are not associated with
accident mitigation or previously evaluated accidents and would not
be the initiator of any new or different kind of accident. The
proposed change does not alter the current functions of these
devices, therefore, this proposed change will not create the
possibility of a new or different kind of accident.
C. The proposed amendment does not involve a significant
reduction in a margin of safety.
The requirements for SQN's electrical equipment protective
devices are unchanged by the proposed relocation of the requirements
to the SQN Technical Requirements Manual. The function of these
devices and the surveillance testing to ensure operability of these
devices remains unchanged. Any future changes to these requirements
will be evaluated, in accordance with 10 CFR 50.59, to ensure
acceptability and NRC review as required. Accordingly, the proposed
change will not result in a reduction in a margin of safety.
The NRC 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: Chattanooga-Hamilton County
Library, 1001 Broad Street, Chattanooga, Tennessee 37402.
Attorney for licensee: General Counsel, Tennessee Valley Authority,
400 West Summit Hill Drive, ET 10H, Knoxville, Tennessee 37902.
NRC Section Chief: Sheri R. Peterson.
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, et al., Docket No. 50-400, Shearon
Harris Nuclear Power Plant, Unit 1, Wake and Chatham Counties, North
Carolina
Date of application for amendment: December 16, 1997, as
supplemented August 31, and December 7, 1998.
Brief description of amendment: This amendment changes Technical
Specification 4.7.1.2.1.a.2.a, Auxiliary Feedwater (AFW) System
Surveillance Requirements, by changing the differential pressure and
flow requirements of the steam turbine-driven AFW pump to allow testing
of the pump at a lower speed.
Date of issuance: April 1, 1999.
Effective date: April 1, 1999.
Amendment No.: 87.
Facility Operating License No. NPF-63: Amendment revises the
Technical Specifications.
Date of initial notice in Federal Register: February 11, 1998 (63
FR 6981).
[[Page 19567]]
The August 31, and December 7, 1998, submittals contained
clarifying information only, and did not change the initial no
significant hazards consideration determination.
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated April 1, 1999.
No significant hazards consideration comments received: No.
Local Public Document Room location: Cameron Village Regional
Library, 1930 Clark Avenue, Raleigh, North Carolina 27605.
Carolina Power & Light Company, et al., Docket No. 50-400, Shearon
Harris Nuclear Power Plant, Unit 1, Wake and Chatham Counties, North
Carolina
Date of application for amendment: September 1, 1998, as
supplemented on March 19, 1999.
Brief description of amendment: This amendment changes Technical
Specification (TS) \3/4\.9.11, ``Water Level--New and Spent Fuel
Pools,'' and its associated Bases by requiring 23 feet of water above
the top of fuel rods within irradiated fuel assemblies seated in the
storage racks.
Date of issuance: April 8, 1999.
Effective date: April 8, 1999.
Amendment No.: 88.
Facility Operating License No. NPF-63: Amendment revises the
Technical Specifications.
Date of initial notice in Federal Register: September 23, 1998 (63
FR 50935).
The March 19, 1999, submittal contained clarifying information
only, and did not change the initial no significant hazards
consideration determination.
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated April 8, 1999.
No significant hazards consideration comments received: No.
Local Public Document Room location: Cameron Village Regional
Library, 1930 Clark Avenue, Raleigh, North Carolina 27605.
Commonwealth Edison Company, Docket Nos. STN 50-456 and STN 50-457,
Braidwood Station, Unit Nos. 1 and 2, Will County, Illinois
Date of application for amendments: November 25, 1998.
Brief description of amendments: The amendments revised the
Technical Specifications (TS) to support on-line replacement of the
Braidwood, Unit 2, vital batteries.
Date of issuance: March 26, 1999.
Effective date: Immediately, to be implemented within 30 days.
Amendment Nos.: 99 and 99.
Facility Operating License Nos. NPF-72 and NPF-77: The amendments
revised the Technical Specifications.
Date of initial notice in Federal Register: February 24, 1999 (64
FR 9185).
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated March 26, 1999.
No significant hazards consideration comments received: No.
Local Public Document Room location: Wilmington Public Library, 201
S. Kankakee Street, Wilmington, Illinois 60481.
Commonwealth Edison Company, 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: January 21, 1999.
Brief description of amendments: The amendments revised the
Technical Specifications (TSs) by relocating TS Section \3/4\.6.I,
``Primary System Boundary-Chemistry'' and associated bases to the
Updated Final Safety Analysis Report (UFSAR) and to applicable plant
procedures.
Date of issuance: March 31, 1999.
Effective date: Immediately, to be implemented within 30 days.
Amendment Nos.: 187 and 184.
Facility Operating License Nos. DPR-29 and DPR-30: The amendments
revised the Technical Specifications.
Date of initial notice in Federal Register: February 24, 1999 (64
FR 9186).
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated March 31, 1999.
No significant hazards consideration comments received: No.
Local Public Document Room location: Dixon Public Library, 221
Hennepin Avenue, Dixon, Illinois 61021.
Duke Energy Corporation, et al., Docket Nos. 50-413 and 50-414, Catawba
Nuclear Station, Units 1 and 2, York County, South Carolina
Date of application for amendments: January 28, 1999.
Brief description of amendments: The amendments revised Technical
Specifications Section 3.7.13, ``Fuel Handling Ventilation Exhaust
System,'' and associated Bases to correct discrepancies between the
current design and this section.
Date of issuance: March 26, 1999.
Effective date: As of the date of issuance to be implemented within
30 days from the date of issuance.
Amendment Nos.: Unit 1-176; Unit 2-168.
Facility Operating License Nos. NPF-35 and NPF-52: Amendments
revised the Technical Specifications.
Date of initial notice in Federal Register: February 24, 1999 (64
FR 9187).
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated March 26, 1999.
No significant hazards consideration comments received: No.
Local Public Document Room location: York County Library, 138 East
Black Street, Rock Hill, South Carolina.
Duke Energy Corporation, Docket Nos. 50-269, 50-270, and 50-287, Oconee
Nuclear Station, Units 1, 2, and 3, Oconee County, South Carolina
Date of application of amendments: October 15, 1998, as
supplemented December 15, 1998, and January 11 and 21, 1999.
Brief description of amendments: The amendments revised the
Technical Specifications (TSs) to change the heatup, cooldown, and
inservice test limitations for the reactor coolant system of each unit
to a maximum of 26 effective full-power years. The amendments also
revise the TSs for low temperature overpressure protection to reflect
the revised pressure-temperature limits of the reactor vessels.
Date of Issuance: March 30, 1999.
Effective date: As of the date of issuance to be implemented within
90 days from the date of issuance.
Amendment Nos.: Unit 1-302; Unit 2-302; Unit 3-302.
Facility Operating License Nos. DPR-38, DPR-47, and DPR-55:
Amendments revised the Technical Specifications.
Date of initial notice in Federal Register: December 2, 1998 (63 FR
66592).
The December 15, 1998, and January 11 and 21, 1999, letters
provided clarifying information that did not change the scope of the
original Federal Register notice and the initial proposed no
significant hazards consideration determination.
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated March 30, 1999.
No significant hazards consideration comments received: No.
Local Public Document Room location: Oconee County Library, 501
West South Broad Street, Walhalla, South Carolina.
Duquesne Light Company, et al., Docket Nos. 50-334 and 50-412, Beaver
Valley Power Station, Unit Nos. 1 and 2, Shippingport, Pennsylvania
Date of application for amendments: October 15, 1998, as
supplemented
[[Page 19568]]
December 14, 1998, February 18, 1999, and February 23, 1999.
Brief description of amendments: These amendments made several
changes that are administrative in nature. The changes (1) made
editorial changes that delete obsolete material or material adequately
described elsewhere, changed action statement numbers, updated
technical specification (TSs) index pages, and made changes to be
consistent with the guidance provided in the improved standard
technical specifications for Westinghouse reactors (NUREG-1431,
Revision 1); (2) deleted reporting requirements that are duplicated in
various sections of Title 10 of the Code of Federal Regulations; and
(3) relocated the requirement for meteorological monitoring
instrumentation from the TSs to the Licensing Requirements Manual.
The February 18, 1999, and February 23, 1999, letters withdrew a
portion of the amendment request that would have deleted the
description of the site exclusion boundary from the TSs. The
description of the site exclusion boundary will remain in the TS.
Date of issuance: March 26, 1999.
Effective date: Units 1 and 2, as of date of issuance, to be
implemented within 60 days.
Amendment Nos.: 220 and 97.
Facility Operating License Nos. DPR-66 and NPF-73: Amendments
revised the Technical Specifications and licenses.
Date of initial notice in Federal Register: November 18, 1998 (63
FR 64111).
The December 14, 1998, February 18, 1999, and February 23, 1999,
letters did not change the initial proposed no significant hazards
consideration determination or expand the amendment request beyond the
scope of the initial notice.
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated March 26, 1999.
No significant hazards consideration comments received: No.
Local Public Document Room location: B. F. Jones Memorial Library,
663 Franklin Avenue, Aliquippa, PA 15001
Duquesne Light Company, et al., Docket No. 50-412, Beaver Valley Power
Station, Unit 2, Shippingport, Pennsylvania
Date of application for amendment: March 10, 1997, as supplemented
July 28, 1997, September 17, 1997, April 30, 1998, January 29, 1999,
and February 26, 1999.
Brief description of amendment: The amendment modifies Technical
Specification 3/4.4.5, ``Steam Generators,'' and its associated Bases
and adds a new license condition to Appendix D to allow repair of steam
generator tubes by installation of sleeves developed by ABB Combustion
Engineering. In addition, the amendment deletes the option for using
the kinetic sleeving methodology previously approved for use at Beaver
Valley Power Station, Unit 2.
Date of issuance: March 26, 1999.
Effective date: As of date of issuance, to be implemented within 60
days.
Amendment No: 98.
Facility Operating License No. NPF-73. Amendment revised the
Technical Specifications and License.
Date of initial notice in Federal Register: April 23, 1997 (62 FR
19829).
The July 28, 1997, September 17, 1997, April 30, 1998, January 29,
1999, and February 26, 1999, letters provided clarifying information
that did not change the initial proposed no significant hazards
consideration determination or expand the amendment request beyond the
scope of the April 23, 1997, Federal Register notice.
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated March 26, 1999.
No significant hazards consideration comments received: No.
Local Public Document Room location: B.F. Jones Memorial Library,
663 Franklin Avenue, Aliquippa, PA 15001.
Entergy Gulf States, Inc., and Entergy Operations, Inc., Docket No. 50-
458, River Bend Station, Unit 1, West Feliciana Parish, Louisiana
Date of amendment request: January 12, 1999, supersedes application
dated May 31, 1996.
Brief description of amendment: The amendment adds an additional
required action to the Limiting Condition for Operation (LCO) 3.9.1,
``Refueling Equipment Interlocks,'' of the RBS Technical
Specifications. The additional action will allow an alternative to the
current action for one or more inoperable refueling equipment
interlocks. The current action is to ``suspend in-vessel fuel movement
with equipment associated with the inoperable interlock(s).'' The
alternative action will be to (1) insert a control rod withdrawal
block, and (2) verify all control rods are fully inserted in core cells
containing one or more fuel assemblies. The amendment also revised the
Bases for LCO 3.9.1 actions to describe the alternative action.
Date of issuance: March 26, 1999.
Effective date: March 26, 1999.
Amendment No.: 104.
Facility Operating License No. NPF-47: The amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: February 10, 1999 (64
FR 6695).
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated March 26, 1999.
No significant hazards consideration comments received: No.
Local Public Document Room location: Government Documents
Department, Louisiana State University, Baton Rouge, LA 70803.
Niagara Mohawk Power Corporation, Docket No. 50-220, Nine Mile Point
Nuclear Station Unit No. 1, Oswego County, New York
Date of application for amendment: November 30, 1998.
Brief description of amendment: The amendment changes Technical
Specification 3.1.2, ``Liquid Poison System,'' and its associated Bases
to correct the required concentration and volume of boron solution.
Date of issuance: April 2, 1999.
Effective date: As of the date of issuance to be implemented within
30 days.
Amendment No.: 166.
Facility Operating License No. DPR-63: Amendment revises the
Technical Specifications.
Date of initial notice in Federal Register: December 30, 1998 (63
FR 71970).
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated April 2, 1999.
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. 50-423, Millstone
Nuclear Power Station, Unit No. 3, New London County, Connecticut
Date of application for amendment: April 1, 1998, as supplemented
May 29, June 26, and August 4, 1998.
Brief description of amendment: The amendment revises the Millstone
Unit 3 final safety analysis report (FSAR) by adding a new sump pump
subsystem to address groundwater inleakage through the containment
basemat.
Date of issuance: March 17, 1999.
Effective date: As of the date of issuance, to be implemented
within 60 days from the date of issuance.
[[Page 19569]]
Amendment No.: 168.
Facility Operating License No. NPF-49: Amendment authorized changes
to the FSAR.
Date of initial notice in Federal Register: April 22, 1998 (63 FR
19974).
The May 29, June 26, and August 4, 1998, letters provided
clarifying information that did not change the scope of the April 1,
1998, application and the initial proposed no significant hazards
consideration determination.
The Commission's related evaluation of the amendment, state
consultation, and final determination of no significant hazards
consideration are contained in a Safety Evaluation dated March 17,
1999.
No significant hazards consideration comments received: No public
comments received. A petition to intervene was received from the
Citizens Regulatory Commission that was dismissed and terminated by the
NRC Atomic Safety Licensing Board (LBP-98-22).
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.
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 application for amendments: February 4, 1998, as revised
September 29, 1998.
Brief description of amendments: The amendments revise the
Technical Specifications surveillance requirements concerning secondary
containment doors.
Date of issuance: April 7, 1999.
Effective date: As of the date of issuance, to be implemented
within 30 days.
Amendments Nos.: 227 and 230.
Facility Operating License Nos. DPR-44 and DPR-56: The amendments
revised the Technical Specifications.
Date of initial notice in Federal Register: August 14, 1998 (63 FR
38202).
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated April 7, 1999.
No significant hazards consideration comments received: No.
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.
Public Service Electric & Gas Company, Docket No. 50-354, Hope Creek
Generating Station, Salem County, New Jersey
Date of application for amendment: October 22, 1998.
Brief description of amendment: This amendment revises Technical
Specification (TS) 4.8.2.1.b.3 to increase the minimum battery
electrolyte temperature limit from 60 deg.F to 72 deg.F. This change
resolves a discrepancy in the electrolyte temperature assumed in the
Class 1E battery sizing calculations versus the limit specified in the
TSs.
Date of issuance: March 25, 1999.
Effective date: As of the date of issuance, to be implemented
within 60 days.
Amendment No.: 118.
Facility Operating License No. NPF-57: This amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: December 2, 1998 (63 FR
66602).
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated March 25, 1999.
No significant hazards consideration comments received: No.
Local Public Document Room location: Pennsville Public Library, 190
S. Broadway, Pennsville, NJ 08070.
Southern Nuclear Operating Company, Inc., 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: December 4, 1998.
Brief description of amendments: The amendments make two changes to
the TS. The first change revises the Unit 1 TS Section 2.1.1.2 to
delete the footnote that specifies that the Safety Limit Minimum
Critical Power Ratios are for Cycle 18 only. The second change revises
the TS for both units by deleting Section 5.6.5.b.2) and incorporating
Section 5.6.5.b.1) into Section 5.6.5.b.
Date of issuance: April 1, 1999.
Effective date: As of the date of issuance to be implemented within
30 days from the date of issuance.
Amendment Nos.: Unit 1-215; Unit 2-156.
Facility Operating License Nos. DPR-57 and NPF-5: Amendments
revised the Technical Specifications.
Date of initial notice in Federal Register: January 27, 1999 (64 FR
4161).
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated April 1, 1999.
No significant hazards consideration comments received: No.
Local Public Document Room location: Appling County Public Library,
301 City Hall Drive, Baxley, Georgia.
Southern Nuclear Operating Company, Inc., et al., Docket Nos. 50-424
and 50-425, Vogtle Electric Generating Plant, Units 1 and 2, Burke
County, Georgia
Date of application for amendments: October 15, 1998, as
supplemented by letter dated November 11, 1998.
Brief description of amendments: The amendments change the Vogtle
Electric Generating Plant Unit 1 and 2 Facility Operating Licenses to
delete or modify certain license conditions that have become obsolete
or inappropriate. In addition, the Technical Specifications and Bases
are reissued to reflect new word processing software.
Date of issuance: March 26, 1999.
Effective date: As of the date of issuance to be implemented within
30 days from the date of issuance.
Amendment Nos.: Unit 1-107; Unit 2-85.
Facility Operating License Nos. NPF-68 and NPF-81: Amendments
revised Facility Operating Licenses and the Technical Specifications.
Date of initial notice in Federal Register: December 2, 1998 (63 FR
66602).
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated March 26, 1999.
No significant hazards consideration comments received: No.
Local Public Document Room location: Burke County Library, 412
Fourth Street, Waynesboro, Georgia
Union Electric Company, Docket No. 50-483, Callaway Plant, Unit 1,
Callaway County, Missouri
Date of applications for amendment: October 31, 1997, as
supplemented by letter dated September 29, 1998, and application dated
July 30, 1998.
Brief description of amendment: The amendment revised Tables 3.3-3,
3.3-4, and 4.3-2 of the technical specifications regarding the
engineered safety feature actuation system (ESFAS) Functional Unit 6.f,
``Loss of Offisite Power--Start Turbine-Driven Pump,'' by establishing
separate requirements for the analog and digital portions of the
associated circuit. The amendment also adds a note to TS Table 4.3-2 to
clarify that the verification of time delays associated
[[Page 19570]]
with ESFAS Functional Units 8.a and 8.b, ``Loss of Power,'' is only
performed as part of the channel calibration.
Date of issuance: April 2, 1999.
Effective date: April 2, 1999, to be implemented within 30 days of
the date of issuance.
Amendment No.: 130.
Facility Operating License No. NPF-30: The amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: December 16, 1998 (63
FR 69348).
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated April 2, 1999.
No significant hazards consideration comments received: No.
Local Public Document Room location: Elmer Ellis Library,
University of Missouri, Columbia Missouri 65201.
Wisconsin Public Service Corporation, Docket No. 50-305, Kewaunee
Nuclear Power Plant, Kewaunee County, Wisconsin
Date of application for amendment: November 18, 1998, as
supplemented with additional information by letters dated March 1,
1999, and March 9, 1999.
Brief description of amendment: The amendment revises the pressure/
temperature limits and the low-temperature overpressure protection
requirements in the facility technical specifications.
Date of issuance: April 1, 1999.
Effective date: April 1, 1999.
Amendment No.: 144.
Facility Operating License No. DPR-43: Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: December 30, 1998.
(63FR71978)
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated April 1, 1999.
No significant hazards consideration comments received: No.
Local Public Document Room location: University of Wisconsin,
Cofrin Library, 2420 Nicolet Drive, Green Bay, WI 54311-7001.
Dated at Rockville, Maryland, this 14th day of April 1999.
For the Nuclear Regulatory Commission.
John A. Zwolinski,
Director, Division of Licensing Project Management, Office of Nuclear
Reactor Regulation.
[FR Doc. 99-9839 Filed 4-20-99; 8:45 am]
BILLING CODE 7590-01-P