[Federal Register Volume 64, Number 105 (Wednesday, June 2, 1999)]
[Notices]
[Pages 29707-29721]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: 99-13765]
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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 May 8, 1999, through May 20, 1999. The last
biweekly notice was published on May 19, 1999.
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
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 July 2, 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.
[[Page 29708]]
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.
Boston Edison Company, Docket No. 50-293, Pilgrim Nuclear Power
Station, Plymouth County, Massachusetts
Date of amendment request: May 5, 1999.
Description of amendment request: The proposed amendment would
change the technical specifications (TS) and licensing basis for the
required amount of diesel fuel to be stored on-site and its sources.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
The proposed amendment does not involve a significant increase
in the probability or consequences of an accident previously
evaluated.
The proposed change affects only the on-site diesel fuel storage
capacity for the operation of emergency diesel generators [EDG]. The
on-site storage capacity is not associated with an accident
precursor/initiator; thus, it has no impact on the probability of
[an] accident occurring. The consequences of an accident would not
be significantly increased because reasonable measures will be
available to ensure the EDGs are supplied with enough fuel from the
on-site sources to operate for seven days at rated capacity.
The proposed amendment does not create the possibility of a new
or different kind of accident from any accident previously
evaluated.
The proposed change does not affect normal plant operation or
the immediate response to an accident. The only change is the
proposed refilling operation to transfer fuel from the Class II
SBODG [Station Blackout Diesel Generators] storage tanks to the
Class I EDG tanks. The refilling operation would occur entirely
outdoors through above ground hoses connecting the EDG and SBODG
tanks. This operation would only be required following a LOCA [loss-
of-coolant accident], an accident already analyzed. Since the
proposed refilling operation is a post-accident evolution, it would
not be in place to cause an accident of a different type during non-
accident conditions. No reasonable malfunction of equipment
associated with the evolution could create a new or different kind
of accident than previously evaluated.
The proposed amendment does not involve a significant reduction
in the margin of safety.
The proposed amendment for licensing basis change and TS change
does not significantly reduce the margin of safety. The proposed
change restores the licensing basis to provide sufficient fuel in
on-site storage tanks for continuous operation of each EDG for
approximately seven days. The revised licensing basis requires
36,800 gallons of fuel per EDG to be stored on-site. A minimum of
19,800 gallons of fuel will be stored in Class I EDG storage tanks
and the remaining will be stored in Class II SBODG on-site storage
tanks. The storage of fuel in Class I tanks does not reduce the
margin of safety. The only potential reduction in the margin of
safety is due to the use of Class II SBODG tanks and associated
transfer equipment for the storage and transfer of additional fuel.
These Class II tanks are rugged, double-wall fiberglass tanks. While
not designed to safety-related requirements, the failure of these
tanks under extreme environmental conditions, such as an earthquake,
has been evaluated to be very unlikely. Thus, on-site storage of
sufficient fuel for operation of both EDGs is assured to mitigate
the consequences of an accident previously evaluated. All stored
fuel is maintained at the same quality standard. The proposed diesel
fuel refilling operation is a post design basis accident activity,
which does not create the possibility of a new accident or impact an
accident previously evaluated. Therefore, there is no significant
reduction in the safety margin.
The NRC 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: Plymouth Public Library, 132
South Street, Plymouth, Massachusetts 02360.
Attorney for licensee: J. Fulton, Boston Edison Company, 800
Boylston Street, 36th Floor, Boston, Massachusetts 02199.
NRC Section Chief: James W. Clifford.
[[Page 29709]]
Commonwealth Edison Company, Docket Nos. 50-295 and 50-304, Zion
Nuclear Power Station, Units 1 and 2, Lake County, Illinois
Date of application of amendment request: October 2, 1998, as
supplemented by letter dated April 19, 1999.
Description of amendment request: By letter dated February 13,
1998, Commonwealth Edison Company (ComEd) certified that they have
permanently ceased operations at Zion Nuclear Power Station (ZNPS),
Units 1 and 2. Since ComEd has permanently ceased operations at ZNPS,
they have requested an amendment to the Facility Operating Licenses to
eliminate license conditions that are no longer applicable and to
replace the existing technical specifications in their entirety with
permanently defueled technical specifications (PDTS). The PDTS reflect
the permanently shutdown and defueled condition of the ZNPS.
Basis for a 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 and has determined that the proposed changes do not:
1. The proposed change does not involve a significant increase
in the probability or consequences of an accident previously
evaluated.
The administrative changes remove requirements that are not
invoked with the reactors permanently defueled. The editorial
changes alter format, word choice, grammar, terminology, etc., but
do not change requirements. The more restrictive changes add new
requirements, remove existing exceptions, or make existing limits
more conservative. The relocation or redundancy changes remove
requirements from the facility operating licenses or technical
specifications because they exist in another document controlled by
other approved methods. None of these types of changes affect the
probability or consequences of a previously evaluated accident since
there is no functional reduction in the limitations imposed on
structures, systems, components or activities with the reactors
permanently defueled.
The less restrictive changes to the license conditions eliminate
requirements for programs and commitments that address hazards or
conditions that are no longer credible with both reactors
permanently defueled. Since these hazards or conditions are not
credible, no increase in the probability or consequences of a
previously evaluated accident will result from the elimination of
these requirements.
The less restrictive changes to the equipment-related technical
specifications eliminate or modify restrictions involving certain
structures systems and components (SSCs). Some of the equipment-
related technical specifications have been eliminated because, with
both reactors permanently defueled, the spectrum of previously
evaluated credible accidents has been significantly reduced and many
of the associated hazards (such as reactor coolant gaseous activity,
hydrogen, and radioactive iodine) will not occur. Since those
previously evaluated accidents and associated hazards are no longer
credible, their probability and consequences are not increased by
the changes eliminating the associated technical specifications.
Other equipment-related technical specifications have been modified
to address previously evaluated accidents that are still relevant in
the permanently defueled condition more logically and consistently,
without increasing their probability or consequences.
The less restrictive changes to the Administrative Control
technical specifications affect a variety of functions. They provide
flexibility in Quality Assurance Program administration, allow a
reduction in shift staffing, eliminate certain training requirements
for personnel who have little or no safety involvement, change
certain procedure processing requirements, provide consistency in
scheduling certain radiological surveillances and reports, eliminate
reports that are no longer needed, eliminate unnecessary flood door
requirements, and allow alternative methods of administering Process
Control Program changes. Since none of these changes directly
involve the previously evaluated accidents that remain credible with
both reactors permanently defueled, the changes will not increase
the probability or consequences of any previously evaluated
accident.
2. The proposed change does not create the possibility of a new
or different kind of accident from any accident previously
evaluated. The administrative changes do not alter any SSCs or
activities involved with the safe storage of nuclear fuel. The
editorial changes do not alter any requirements. The more
restrictive changes make the technical specifications more limiting.
The relocation/redundancy changes only change the location of
requirements. None of these types of changes create the possibility
of a new or different kind of accident from any accident previously
evaluated.
The less restrictive changes to the license conditions eliminate
requirements for programs and commitments involving hazards or
conditions that are no longer credible with both reactors
permanently defueled. Since these changes do not result in any new
programs or activities, they do not create the possibility of a new
or different kind of accident from any accident previously
evaluated.
The less restrictive changes to the equipment-related technical
specifications do not alter any SSC or cause any SSC to be operated
in a manner that could initiate any event or accident. Therefore,
these changes do not create the possibility of a new or different
kind of accident from any accident previously evaluated.
The less restrictive changes to the Administrative Control
technical specifications do not change the design, function, or
operation of any SSC except the flood doors and the change involving
the flood doors does not introduce any new type of event. Therefore,
the less restrictive changes to the Administrative Control technical
specifications do not create the possibility of a new or different
kind of accident from any accident previously evaluated.
3. The proposed change does not involve a significant reduction
in a margin of safety.
The administrative changes do not alter any SSCs or activities
involved with the safe storage of nuclear fuel. The editorial
changes do not alter any requirements. The more restrictive changes
make the technical specifications more limiting. The relocation/
redundancy changes only change the location of requirements. None of
these types of changes reduce any safety margin.
The less restrictive changes to the license conditions eliminate
requirements that apply to hazards or conditions that are no longer
relevant with both reactors permanently defueled. The safety margins
that may have been associated with those license conditions are no
longer relevant.
There are no longer any relevant margins of safety associated
with the less restrictive changes to the equipment-related technical
specifications except for those involving criticality control and
seismic criteria. The proposed technical specifications maintain the
same margin of safety for criticality control in the spent fuel
pool, and the Defueled Safety Analysis Report imposes seismic
criteria that provide an adequate safety margin.
The less restrictive changes to the Administrative Control
technical specifications do not directly involve any limits or
parameters and therefore cannot affect any 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: Waukegan Public Library, 128
N. County Street, Waukegan, Illinois 60085.
Attorney for licensee: Pamela B. Strobel, Senior Vice President and
General Counsel, Commonwealth Edison Company, P.O. Box 767, Chicago,
Illinois 60690-076.
NRC Project Director: Stuart A. Richards.
Entergy Operations, Inc., Docket No. 50-313, Arkansas Nuclear One, Unit
No. 1, Pope County, Arkansas
Date of amendment request: May 14, 1999.
Description of amendment request: The proposed amendment would
revise the Technical Specification requirements affecting the
surveillance criteria for that portion of the once-through steam
generator tubes regarded
[[Page 29710]]
as a primary-to-secondary pressure boundary located within the upper
tube sheet and impacted by a specific degradation mechanism, namely,
outside diameter intergranular attack.
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:
Criterion 1--Does Not Involve a Significant Increase in the Probability
or Consequences of an Accident Previously Evaluated
The once-through steam generators OTSG are used to remove heat
from the reactor coolant system during normal operation and during
accident conditions. The OTSG tubing forms a substantial portion of
the reactor coolant pressure boundary. An OTSG tube failure is a
breach of the reactor coolant pressure boundary and is a specific
accident analyzed in the ANO-1 [Arkansas Nuclear One, Unit 1] Safety
Analysis Report.
The purpose of the periodic surveillance performed on the OTSGs
in accordance with ANO-1 Technical Specification (TS) 4.18 is to
ensure that the structural integrity of this portion of the reactor
coolant system will be maintained. The TS plugging limit of 40% of
the nominal tube wall thickness requires tubes to be repaired or
removed from service because the tube may become unserviceable prior
to the next inspection. Unserviceable is defined in the TS as the
condition of a tube if it leaks or contains a defect large enough to
affect its structural integrity in the event of an operating basis
earthquake, a loss-of-coolant accident, or a steam line or feedwater
line break. The proposed TS change allows OTSG tubes with ODIGA
[outside diameter intergranular attack] indications contained within
a defined area of the UTS [upper tube sheet] to remain in service
with existing degradation exceeding the existing 40% through-wall
(TW) plugging limit.
Extensive testing and plant experience has illustrated that
ODIGA flaws confined to this area within the OTSG will not result in
tube burst or tube leakage. Therefore, allowing ODIGA flaws in this
specific region to remain in service will not alter the conditions
assumed in the current ANO-1 accident analysis for OTSG tube
failures under postulated accident conditions. In addition, the
condition of the OTSG tubes in this region are monitored during
regular inspection intervals to assess for evidence of growth. Any
growth noted will be addressed through testing and the operational
assessment * * *.
Application of the ODIGA alternate repair criteria will allow
leaving tubes with ODIGA indications found in the defined area of
the UTS in service while ensuring safe operation by monitoring and
assessing the present and future conditions of the tubes. ANO-1 has
operated since 1984 with ODIGA affected tubes in service with no
appreciable effect on structural integrity or indications of tube
leakage from ODIGA sources within the UTS. Through the inspection,
testing, monitoring, and assessment program previously mentioned,
and the on-line leak detection capabilities available during plant
operation, continued safe operation of ANO-1 is reasonably assured.
Therefore, the application of the ODIGA alternate repair
criteria...does not involve a significant increase in the
probability or consequences of any accident previously evaluated.
Criterion 2--Does Not Create the Possibility of a New or Different Kind
of Accident from any Previously Evaluated
The implementation of the ODIGA alternate repair criteria will
not result in any failure mode not previously analyzed. The OTSGs
are passive components. The intent of the TS surveillance
requirements are being met by these proposed changes in that
adequate structural and leak integrity will be maintained.
Additionally, the proposed change does not introduce any new modes
of plant operation.
Therefore, this change does not create the possibility of a new
or different kind of accident from any previously evaluated.
Criterion 3--Does Not Involve a Significant Reduction in the Margin of
Safety
The application of an alternate repair criteria for ODIGA
provides adequate assurance with margin that ANO-1 steam generator
tubes will retain their integrity under normal and accident
conditions. The structural requirements of ODIGA affected tubes have
been evaluated satisfactorily and meet or exceed regulatory
requirements. Leakage rates for these tubes within the defined
region of the upper tubesheet are essentially zero and are
reasonably assured to remain within the assumptions of the accident
analysis by proper application of the ODIGA alternate repair
criteria program. Because no appreciable impact is evidenced on the
tubes structural integrity or its resulting leak rate, the margin to
safety remains effectively unaltered.
Therefore, this change 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 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: Tomlinson Library, Arkansas
Tech University, Russellville, Arkansas 72801.
Attorney for licensee: Nicholas S. Reynolds, Esquire, Winston and
Strawn, 1400 L Street, NW., Washington, DC 20005-3502.
NRC Section Chief: Robert A. Gramm.
Florida Power Corporation, et al., Docket No. 50-302, Crystal River
Nuclear Generating Plant, Unit No. 3, Citrus County, Florida
Date of amendment request: May 5, 1999.
Description of amendment request: The proposed amendment would
revise the Crystal River Unit 3 (CR-3) Improved Technical
Specifications to approve an alternate repair criteria (ARC) for axial
tube end crack-like indications in the upper and lower tubesheets of
the CR-3 Once Through Steam Generators (OTSGs). The ARC will allow
leaving OTSG tubes with axially oriented tube end cracks located within
the clad region of the tube-to-tubesheet roll joint in service. Tubes
with crack-like indications within the carbon steel portion of the
tubesheet, or tubes with circumferentially oriented tube end cracks or
volumetric indications within the Inconel clad region of the tubesheet,
would be repaired or removed from service.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below.
1. Involve a significant increase in the probability or
consequences of an accident previously evaluated.
This LAR [License Amendment Request] proposes to implement an
alternate repair criteria (ARC) for Once Through Steam Generator
(OTSG) tubes with axial tube end crack (TEC) indications.
Application of the ARC will allow tubes with axially oriented TEC to
remain in service in accordance with specific conditions. Based on a
combination of structural analyses, mock-up testing and inservice
inspections, as detailed in Topical Report BAW-2346P, allowing tubes
with TEC indications to remain in service is safe and justified.
Potential leakage from tubes with TEC will be bounded by the
main steam line break (MSLB) evaluation presented in the Final
Safety Analysis Report (FSAR). The proposed change requires
inspections during subsequent outages of tubes remaining in-service
with the TEC indications. The addition of this inspection does not
change any accident initiators. The proposed inspection of these
indications during the subsequent OTSG inservice inspections assures
continuous monitoring of these tubes such that degradation of tubes
containing TEC indications will be detected. 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 any accident previously evaluated.
The proposed alternate repair criteria for axial TEC indications
introduces no new failure modes or accident scenarios. Topical
Report BAW-2346P demonstrated structural and leakage integrity for
all normal operating and accident conditions for Crystal River Unit
3 (CR-3). Furthermore, leaving TEC in service does not change the
design or operating characteristics of the OTSGs. In the unlikely
event that a tube with a TEC should
[[Page 29711]]
fail and sever completely, 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 a
possibility of a new or different kind of accident from any
previously evaluated.
3. Involve a significant reduction in a margin of safety.
The mechanical joint is constrained within the tubesheet bore;
thus, there is no additional risk associated with tube rupture. ITS
[Improved Technical Specifications] Bases 3.4.12 contains relevant
information pertaining to limitations on Reactor Coolant System
leakage. 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
proposed alternate repair criteria 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.
Florida Power and Light Company, Docket Nos. 50-250 and 50-251, Turkey
Point Plant Units 3 and 4, Dade County, Florida
Date of amendment request: April 26, 1999.
Description of amendment request: The proposed amendments would
revise the Turkey Point Plant, Units 3 and 4, Facility Operating
Licenses and the Technical Specifications (TS): (1) To remove a part of
license condition 3.L that is obsolete, (2) to update the TS Index to
reflect all changes made to the TS Sections, TS Figures, and TS Tables
by previously approved license amendments, and (3) to remove Table and
Figure numeration inconsistencies found in TS 3/4.1.2.5 and TS 3/4.7.6.
These proposed changes represent an administrative update to the Turkey
Point Plant, Units 3 and 4, Facility Operating Licenses and to the TS.
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) Operation of the facility in accordance with the proposed
amendments would not involve a significant increase in the
probability or consequences of an accident previously evaluated.
The proposed amendments do not involve a significant increase in
the probability or consequences of an accident previously evaluated
because the proposed changes are administrative in nature removing
obsolete references in the license conditions, updating the
Technical Specification (TS) Index to reflect the revisions made to
the TS Sections, Tables, and Figures via previous TS amendments.
These amendments will not involve a significant increase in the
probability or consequences of an accident previously evaluated
because they do not affect assumptions contained in plant safety
analyses, the physical design and/or operation of the plant, nor do
they affect Technical Specifications that preserve safety analysis
assumptions. Therefore, the proposed changes do not affect the
probability or consequences of accidents previously analyzed.
(2) Operation of the facility in accordance with the proposed
amendments would not create the possibility of a new or different
kind of accident from any accident previously evaluated.
The use of the modified specifications can not create the
possibility of a new or different kind of accident from any
previously evaluated since the proposed amendments will not change
the physical plant or the modes of plant operation defined in the
facility operating license. No new failure mode is introduced due to
the administrative changes since the proposed changes do not involve
the addition or modification of equipment nor do they alter the
design or operation of affected plant systems, structures, or
components.
(3) Operation of the facility in accordance with the proposed
amendments would not involve a significant reduction in a margin of
safety.
The operating limits and functional capabilities of the affected
systems, structures, and components are unchanged by the proposed
amendments. The proposed changes to the Facility Operating License
Conditions and to the Technical Specifications are administrative
and do not significantly reduce any of the margins 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: Florida International
University, University Park, Miami, Florida 33199.
Attorney for licensee: M.S. Ross, Attorney, Florida Power & Light,
P.O. Box 14000, Juno Beach, Florida 33408-0420.
NRC Project Director: Herbert N. Berkow.
PECO Energy Company, Public Service Electric and Gas Company, Delmarva
Power and Light Company, and Atlantic City Electric Company, Dockets
Nos. 50-277 and 50-278, Peach Bottom Atomic Power Station, Units Nos. 2
and 3, York County, Pennsylvania
Date of application for amendments: March 1, 1999.
Description of amendment request: Changes are proposed to support a
modification which will install a digital Power Range Neutron
Monitoring (PRNM) system and incorporate long-term thermal-hydraulic
stability solution hardware.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
i. The proposed changes do not involve a significant increase in
the probability or consequences of an accident previously evaluated.
As discussed in the LTR [licensing topical report], the [Nuclear
Measurements Analysis and Control] NUMAC PRNM modification and
associated changes to the TS [technical specifications] involve
equipment that is designed to detect the symptoms of certain events
or accidents and initiate mitigating actions. The worst case failure
of the equipment involved in the modification is a failure to
initiate mitigating action (scram or rod block), but no failure can
cause an accident. The PRNM replacement system is designed to
perform the same operations as the existing Power Range Monitor
system and meets or exceeds all operational requirements. Therefore,
it is concluded that the probability of an accident previously
evaluated is not increased as a result of replacing the existing
equipment with the PRNM equipment.
The PRNM system reduces the need for tedious operator actions
during normal conditions and allows the operator to focus more on
overall plant conditions. The automatic self-test and increased
operator information provided with the replacement system are likely
to reduce the burden during off-normal conditions as well. The
replacement equipment qualifications fully envelope the
environmental conditions, including electromagnetic interference, in
the PBAPS [Peach Bottom Atomic Power Station] control room.
The replacement equipment has been specifically designed to
assure that it fully meets the response time requirements in the
worst case. As a result, due to statistical variations resulting
from the sampling and update cycles, the response time is typically
faster than required in order to assure that the required response
time is always met.
[[Page 29712]]
Setpoints are changed only when justified by the improved equipment
performance specifications and by setpoint calculations which show
that safety margins are maintained. There is no impact to the
Control Rod Drop accident analysis because the PRNM system maintains
all existing system functions with a reliability equal to or better
than the existing Power Range Monitor system.
The replacement equipment includes up to 5 LPRM [Local Power
Range Monitor] inputs on a single module compared to one per module
on the current system. Up to 17 LPRM signals are processed through
one preprocessor. The recirculation flow signals are processed in
the same hardware as the LPRM processing. The net effect of these
architectural aspects is that there are some single failures that
can cause a greater loss of ``sub-functionality'' than in the
current system. Other architectural and functional aspects, however,
have an offsetting effect. Redundant power supplies are used so that
a single failure of AC power has no effect on the overall PRNM
system functions while still resulting in a half scram as does the
current system. Continuous automatic self-test also assures that if
a single failure does occur, it is much more likely to be detected
immediately. The net effect is that from a total system level,
unavailability of the safety-related functions in the replacement
system is equal to or better than the current Power Range Monitor
system.
Based on the extensive and through [sic] [thorough] verification
and validation program used in the PRNM design and field operating
experience, common cause failures in software controlled functions
are judged to not be a significant failure mode. However, in spite
of that conclusion, means are provided within the system to mitigate
the effects of such a failure and alert the operator. Therefore,
such a failure, even if it occurred, will not increase the
consequences of a previously evaluated accident.
To reduce the likelihood of common cause failure of software
controlled functions, thorough and careful verification and
validation (V&V) activities are performed both for the requirements
and the implementing software design. In addition, the software is
designed to limit the loading that external systems or equipment can
place on the system, thus significantly reducing the risk that some
abnormal dynamic condition external to the system can cause system
functional performance problems due to processing ``overload''
(i.e., ``slowing down'' or stopping the processing).
As a conservatism, however, despite these V&V activities, common
cause failures of software controlled functions due to residual
software design faults are assumed to occur. Both the software and
hardware are designed to manage the consequences of such failure
(and also cover potential common cause hardware failures). Safety
outputs are designed to be fail safe by requiring dynamic update of
output modules or data signals, where failure to update the
information is detected by simple receiving hardware, which, in
turn, forces a trip. This aspect covers all but rather complex
failures where the software or hardware executes a portion of the
overall logic but fails to process some portion of new information
(inputs ``freeze'') or some portion of the logic (outputs
``freeze'').
To help reduce the likelihood of complex failures, a watchdog
timer is used which is updated by a very simple software routine
that in turn monitors the operational cycle time of all tasks in the
system. The software design is such that as long as all tasks are
updated at the design rate, it is likely that software controlled
functions are executing as intended. Conversely, if any task fails
to update at the design rate, that is a strong indication of at
least some unanticipated condition. If such a condition occurs, the
watchdog timer will not be updated, the computer will be
automatically restarted, and the system will detect an abnormal
condition and provide an alarm and trip.
The information available to the operator is at least the same
as with the current system and, in many cases, improved. No actions
are required by the operator to obtain information normally used and
equivalent to that available with the current equipment. However,
the replacement system does provide more directly accessible
information regarding the condition of the equipment, including
automatic self-test, which can aid the operator in diagnosing
unusual situations beyond those defined in the licensing basis.
In summary, the reliability of the new PRNM system and its
ability to detect and mitigate abnormal flux transients have either
remained the same or improved over the existing Power Range Monitor
system. Since these postulated reactivity transients are mitigated
by the new system as effectively and reliability [sic] [reliably] as
the existing system, the consequences of these transients have not
changed. Therefore, the proposed changes do not involve a
significant increase in the probability or consequences of an
accident previously evaluated.
ii. The proposed changes do not create the possibility of a new
or different kind of accident from any accident previously
evaluated.
PBAPS Modification P00507 uses digital processing with software
(firmware) control for the main signal processing part of the
modification. The remainder of the equipment in the modification
uses conventional equipment similar to the current system (e.g.,
penetrations, cables, interface panels).
The digital equipment has ``control'' processing points and
software controlled digital processing where the current system has
analog and discrete component processing. The result is that the
specific failures of hardware and potential software common cause
failure are different from the current system. The effects of
software common cause failure are mitigated by hardware design and
system architecture, but are of a ``different type'' of failure than
those evaluated in the PBAPS Updated Final Safety Analysis Report
(UFSAR). In general, the PBAPS UFSAR assumes simplistic failure
modes (relays for example) but does not specifically evaluate such
effects as self-test detection and automatic trip or alarm.
Therefore, the replacement system may have a malfunction of a
different type from those evaluated in the PBAPS UFSAR [* * *].
However, when these PRNM failures are evaluated at the system level,
there are no new effects.
PBAPS Modification P00507 involves equipment that is intended to
detect the symptoms of certain transients and accidents and initiate
mitigating action. The worst case failure of the equipment involved
in the modification is a failure to initiate mitigating action
(scram), but no failure can cause an accident. This is unchanged
from the current system. Software common cause failures could cause
the system to fail to perform its safety function, but this
possibility is addressed in Section (i) above. In that case, it
might fail to initiate action to mitigate the consequences of an
accident, but would not cause one. No new system level failure modes
are created with the PRNM system.
Therefore, PBAPS Modification P00507 does not create the
possibility of a new or different kind of accident from any accident
previously evaluated.
iii. The proposed changes do not involve a significant reduction
in a margin of safety.
The PRNM system response time and operator information is either
maintained or improved over the current Power Range Monitor system.
The PRNM system has improved channel trip accuracy compared to
the current system and meets or exceeds system requirements assumed
in setpoint analysis. The channel response time exceeds the
requirements.
The channel indicated accuracy is improved over the current
system and meets or exceeds all of the system requirements.
The PRNM system was developed to detect the presence of thermal-
hydraulic instabilities and automatically initiate the necessary
actions to suppress the oscillations prior to violating the MCPR
Safety Limit. The NRC has reviewed and approved the LTR concluding
that the PRNM system will provide the intended protection.
Therefore, PBAPS Modification P00507 does not result in 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: Government Publications
Section, State Library of Pennsylvania, (REGIONAL DEPOSITORY) Education
Building, Walnut Street and Commonwealth Avenue, Box 1601, Harrisburg,
PA 17105.
Attorney for Licensee: J.W. Durham, Sr., Esquire, Sr. V.P. and
General Counsel, PECO Energy Company, 2301 Market Street, Philadelphia,
PA 19101.
NRC Section Chief: James W. Clifford.
[[Page 29713]]
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, as supplemented April
29, 1999, and May 17, 1999. This notice supersedes a previous notice
(64 FR 19563) published April 21, 1999, which was based upon the
licensee's application for amendment dated 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 probability or consequences of an accident
previously analyzed?
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 safety 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 pit] 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 along with necessary portions of the
AC, DC and 120 VAC vital instrument bus electrical power
distribution subsystems ensures the availability of sufficient
electrical sources to operate the unit in a safe manner and to
mitigate the consequences of postulated accidents during shutdown
(e.g., Fuel Handling Accidents), as well as other postulated events.
Action statements provide prompt, specific guidance to ensure
sufficiently conservative plant response should the expected EDG
power supply or required offsite power supply feeders or necessary
portions of AC, DC and 120 VAC vital instrument bus electrical power
distribution subsystems 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 necessary
energized 480VAC bus(es) 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 in the time frame required to
prevent reactor core/cavity/SFP heatup concerns, with sufficient
``kW loading''. Action statements provide prompt, specific guidance
to ensure sufficiently conservative plant response should the
expected EDG or offsite supply feeder or the necessary portions of
the AC, DC and 120 VAC vital instrument bus electrical power
distribution subsystems 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 3.7.F.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 license amendment involve a significant
reduction in the 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, along with the necessary portions of the AC, DC and 120 VAC
vital instrument electrical power distribution subsystems available,
can supply the necessary electrical power requirements during these
plant operating conditions, and in the time frame required to
prevent reactor core/cavity/SFP heatup concerns, with sufficient
``kW loading''. 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
operator actions to ensure sufficiently conservative plant response
should the expected EDG power supply or the required offsite power
supply feeders or AC, DC and 120 VAC vital instrument bus electrical
power distribution subsystems 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: April 6, 1999.
Description of amendment request: This application for amendment to
the Indian Point 3 (IP3) Technical Specifications (TSs) proposes to
change sections 3.7.A.5 and 3.7.F.4 by removing the words ``three
individual underground'' and ``underground'' from the limiting
conditions for operation (LCO) when referring to the emergency diesel
generator (EDG) fuel oil storage tanks (FOSTs).
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?
[[Page 29714]]
No. The proposed change would not change the design
configuration or function of the permanently installed EDG FOSTs.
The revision of TS 3.7.A.5 and 3.7.F.4 to remove the descriptive
words ``three individual underground'' and ``underground'' from the
text of the two LCOs is intended as a line item change, to remove
unnecessarily restrictive wording in the TS. While the Standard
Technical Specifications (STS), NUREG-1431, mentions in the Bases
section that ``all outside tanks, pumps, and piping are located
underground'', the specification itself does not contain this
requirement. The intent of this TS change is to allow for, if
acceptable under 10CFR50.59, the potential installation of an
alternate above ground FOST to an EDG if needed to perform repairs/
testing of the permanently installed FOST. This alternate tank would
need to be qualified and have the required capacity to maintain the
associated EDG operable. This potential modification would include
design of the temporary tank to preclude winds loads from a tornadic
event causing the associated EDG to become inoperable. Installation
of this temporary tank would then permit repair work or replacement
of an installed EDG FOST, or subsequent similar work on either of
the other EDG FOSTs, one at a time. The changes to the Bases for
Specification 3.7 are consistent with the change in the LCO
Specification and do not alter the design or functionality of the
existing EDG FOSTs. The revised LCOs are consistent with the STS in
that the FOSTs will no longer be identified as ``three individual
underground''. Control of future modifications to support EDG FOST
work would ensure proper licensing and design basis compliance in
accordance with the change process of 10CFR50.59. The associated
changes of the TS Bases provide clarification regarding the normal
underground configuration of the EDG FOSTs. The proposed TS change
will not reduce the ability of any system, structure, or component
in preventing or mitigating a design basis accident since no plant
features are being altered in conjunction with this change, and
future changes would be evaluated under 10CFR50.59. The description
of the FOSTs, including the fact that they are underground, remains
part of the current licensing basis because it is described in FSAR
[final safety analysis report] section 8.2.
Therefore, the proposed changes to the TS will not result in an
increase in the probability or consequences of any previously
evaluated accidents. The other changes to the TS pages are editorial
only, moving text to different pages.
2. Does the proposed license amendment create the possibility of
a new or different kind of accident from any accident previously
evaluated?
No. The proposed change would not change the design
configuration or function of the permanently installed EDG FOSTs.
The changes to TS 3.7 and its bases in describing the physical
location of the EDG FOSTs will not alter the required design
criteria of these tanks nor their ability to withstand the effects
of a tornado. These changes will not reduce the ability of the EDG's
in meeting their design requirements of providing emergency power
towards mitigating an accident. The intent of these changes is to
permit the potential use of a temporary above ground FOST(s) to
supply the EDGs and to fulfill the intent and requirements of the
present EDG fuel oil storage system while allowing for maintenance
on an EDG FOST. The 10CFR50.59 change process will be used to
determine this potential modification acceptability. The intent of
the temporary configuration of an above ground FOST would be to
maintain the fuel oil system and EDG operable. The associated
changes to the Bases section of TS 3.7 provide additional
clarification of the ``underground'' nature of the EDG FOSTs.
Neither the changes to the LCO in describing the EDG FOSTs (whether
the normal underground tanks or any temporary above ground FOSTs)
nor any changes to the TS Bases (which do not alter the design or
operation of the EDG fuel oil transfer system) will affect the
ability of the EDGs to provide the necessary power for operation of
equipment required for mitigating previously analyzed accident
scenarios. No plant features, or FSAR description of such, are being
altered in conjunction with this change, and future changes would be
evaluated under 10CFR50.59. Therefore, the proposed changes will not
result in an unanalyzed condition and does not create the
possibility of a new or different type of accident from any accident
previously evaluated.
3. Does the proposed amendment involve a significant reduction
in the margin of safety?
No. The proposed changes will not alter any assumptions, initial
conditions, or the results of any accident analyses. The design and
licensing requirements for the EDG fuel oil storage system are
defined in other parts of the IP3 licensing and design basis,
specifically in FSAR section 8.2. Potential modifications supported
by this change would require a subsequent safety evaluation in
accordance with 10 CFR 50.59 regarding the design requirements
(e.g., fire loads, tornadic wind loads, tornado missile criteria,
security, etc.) for an alternate FOST if repairs to present
``underground'' FOSTs are undertaken. The proper design criteria for
the presently installed EDG FOSTs or for potential, alternate EDG
FOSTs will be maintained via present licensing and design basis
requirements and through the 10 CFR 50.59 change process as
required. No plant features are being altered in conjunction with
this change, and future changes would be evaluated under 10 CFR
50.59. Therefore, this proposed license amendment will 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: April 9, 1999.
Description of amendment request: This application for amendment to
the Indian Point 3 Technical Specifications (TSs) proposes to increase
the allowed outage time (AOT) for any one safety injection pump from 24
hours to 72 hours.
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 analyzed?
Response: The proposed 72-hour allowed outage time for any one
safety injection pump does not involve a significant increase in the
probability or consequences of an accident previously analyzed. The
plant Technical Specifications provides allowed outage times for
systems and components to accommodate preventive or corrective
maintenance. A variation in the allowed outage time is not an
accident initiator and thus does not result in a significant
increase in the probability of an accident previously analyzed. The
proposed change provides for an increase in allowed outage time for
any one safety injection pump. The operability of the remaining two
safety injection pumps is required by the Technical Specifications
during this period. The Indian Point 3 High Head Safety Injection
System consists of three safety injection pumps, each capable of
providing 50 percent of the Emergency Core Cooling System [ECCS]
design flow requirement. Therefore, with only one pump inoperable
the remaining two pumps are capable (assuming that no single failure
occurs during the period of the allowed outage time) of mitigating
the consequences of previously analyzed accidents. In addition, a
72-hour allowed outage time for safety injection pumps was evaluated
by the NRC (Reference 3) [NRC Memorandum, R.L. Baer to V. Stello,
``Recommended Interim Revisions to LCOs for ECCS Components,'' dated
December 1, 1975] and generically approved in the Standard Technical
Specifications (Reference 1) [NUREG-1431 ``Standard Technical
Specifications--Westinghouse Plants,'' Revision 1, dated April
1995].
(2) Does the proposed license amendment create the possibility
of a new or different kind of accident from any accident previously
evaluated?
[[Page 29715]]
Response: The proposed 72-hour allowed outage time for any one
safety injection pump does not create the possibility of a new or
different kind of accident from any accident previously evaluated.
Changing the allowed outage time is accomplished through
administrative changes, such as changes to plant procedures that
implement Technical Specification requirements for allowed outage
time. This change does not require physical changes to plant systems
or components and also does not involve changes to plant setpoints.
This change also does not affect how the safety injection pumps are
operated under design basis accident conditions. Therefore there are
no changes resulting from the proposed new allowed outage time that
alter system operation or that could create the possibility of a new
or different kind of accident. In addition, a 72-hour allowed outage
time safety injection [pump] was generically approved in the
Standard Technical Specifications.
(3) Does the proposed amendment involve a significant reduction
in a margin of safety?
Response: The proposed 72-hour allowed outage time for any one
safety injection pump does not involve a significant reduction in a
margin of safety. With one safety injection pump inoperable, the
remaining two pumps are capable of providing 100% of the fuel
cooling flow assumed for pertinent accident analyses with the
provision that the single-failure assumption is relaxed during the
time period of the allowed outage time. The acceptability of a 72-
hour allowed outage time for ECCS components was established in an
NRC reliability analysis (Reference 3) [NRC Memorandum, R.L. Baer to
V. Stello, ``Recommended Interim Revisions to LCOs for ECCS
Components,'' dated December 1, 1975]. The use of the 72-hour
allowed outage time was generically approved in the Standard
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.
Public Service Electric & Gas Company, Docket Nos. 50-272 and 50-311,
Salem Nuclear Generating Station, Unit Nos. 1 and 2, Salem County, New
Jersey
Date of amendment request: April 14, 1999.
Description of amendment request: The proposed amendments would
revise Technical Specification (TS) 3/4.9.12, ``Fuel Handling Area
Ventilation System (FHAVS),'' to (1) reflect the latest filter testing
standards in the test requirements, (2) add, modify, or delete certain
surveillance test requirements, and (3) clarify the information in the
applicable TS Bases section. The proposed amendments would also make
the TS requirements more consistent with the system design basis.
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.
A Fuel Handling Accident, as described in the Updated Final
Safety Analysis Report (UFSAR) Section 15.4.6, is the design basis
accident considered for establishing system configuration and
performance capability for the FHAVS. This accident is defined as
the dropping of a spent fuel assembly onto the spent fuel rack
resulting in a rupture of the cladding of all the spent fuel rods in
the assembly.
The probability of a fuel handling accident is independent of
the changes proposed in this submittal and it is unaffected by this
submittal. The consequences of a dropped fuel rod are significantly
reduced by pre-aligning the system to its design basis function
prior to moving fuel in the fuel handling building. Pre-aligning the
system eliminates the potential detrimental consequences associated
with a single failure of an active component on the filter train.
The proposed change will not change the way the FHAVS functions to
control the release of radioactive gaseous effluents. Filter testing
is improved by applying more current filter testing requirements to
both Units 1 and 2.
The proposed change will not modify equipment used to store or
move irradiated fuel assemblies, or equipment used to move heavy
loads in the Fuel Handling Building. The proposed new surveillance
will be incorporated into a new or existing procedure.
Therefore, the proposed amendment 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 does not result in any design or physical
configuration changes to the FHAVS, or to the equipment used to
store or move irradiated fuel within the Fuel Handling Building.
Pre-aligning the system to its design basis function prior to moving
fuel in the fuel handling building eliminates the potential
detrimental consequences associated with a single failure of an
active component on the filter train. The system will not be
operated or placed in a configuration that is different from the
configuration that it was designed to operate.
Therefore, the proposed amendment will not create the
possibility of a new or different kind of accident from any
previously evaluated.
3. The proposed change does not involve a significant reduction
in a margin of safety.
The proposed changes will ensure that the FHAVS is operated and
tested in accordance to its design basis requirements as specified
in the Salem UFSAR.
The proposed changes will clarify the requirements of the system
to be considered operable to ensure that the FHAVS will perform its
intended safety function in the event of a Fuel Handling Accident.
These changes ensure that the existing margin is maintained and
improved by pre-aligning the system to its accident configuration.
The proposed change does not involve the addition or
modification of plant equipment. It is consistent with the intent of
the existing TS, the design basis of the FHAVS as described in the
UFSAR, and the [Standard Technical Specifications Westinghouse
Plants, NUREG-1431] ITS and associated Bases.
Therefore, the proposed change does not involve a significant
reduction in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: Salem Free Public Library, 112
West Broadway, Salem, NJ 08079.
Attorney for licensee: Jeffrie J. Keenan, Esquire, Nuclear Business
Unit--N21, P.O. Box 236, Hancocks Bridge, NJ 08038.
NRC Section Chief: James W. Clifford.
TU Electric Company, Docket Nos. 50-445 and 50-446, Comanche Peak Steam
Electric Station (CPSES), Units 1 and 2, Somervell County, Texas
Date of amendment request: May 4, 1999.
Brief description of amendments: The proposed license amendments
would change the CPSES Units 1 and 2 Technical Specifications. The
first change revises Surveillance Requirement (SR) 3.8.4.7 to allow the
unrestricted substitution of the modified battery performance discharge
test in lieu of the service discharge test. The second change revises
SRs 3.8.1.7, 3.8.1.12 , 3.8.1.15, and 3.8.1.20 to separate the voltage
and frequency acceptance criteria for the Diesel Generator (DG) start
surveillances into two sets of criteria; those criteria required to be
met within 10 seconds, and those criteria required to be met following
achievement of steady state conditions. The third change corrects
[[Page 29716]]
miscellaneous editorial errors resulting from issuance of Amendment No.
64.
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. Do the proposed changes involve a significant increase in the
probability or consequence of an accident previously evaluated?
(1) Batteries are used to support mitigation of the consequences
of an accident, and are not considered to be an initiator of any
previously analyzed accident. The proposed change would not effect
the design or performance of the batteries. The allowance to perform
the modified performance discharge test in lieu of the service test
at any time is permissible since the test's discharge rate envelopes
the duty cycle of the service test. Therefore, the allowance for
unrestricted substitution of the modified performance discharge test
in lieu of the service discharge test does not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
(2) The diesel generators are used to support mitigation of the
consequences of an accident, and are not considered to be an
initiator of any previously analyzed accident. The proposed change
does not affect the accident analysis assumption that the DG reaches
minimum conditions to accept load within 10 seconds. The ability of
the DG to maintain steady state operation within 10 seconds is not
an accident analysis assumption and is primarily used to identify
degradation of governor and voltage regulator performance.
Therefore, the proposed change does not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
(3) The editorial changes are non-technical and therefore do not
involve a significant increase in the probability or consequences of
an accident previously evaluated.
2. Do the proposed changes create the possibility of a new or
different kind of accident from any accident previously evaluated?
(1) The allowance for unrestricted substitution of the modified
performance discharge test in lieu of the service discharge test
does not involve any physical alteration to the plant. No new
failure mechanisms will be introduced and the change does not affect
the ability of the batteries to fulfill their safety-related
function. Therefore, this change does not create the possibility of
a new or different kind of accident from any accident previously
evaluated.
(2) The separation of the DG start surveillance criteria into
those criteria required to be met within 10 seconds, and those
criteria required to be met following achievement of steady state
conditions, does not involve any physical alteration to the plant.
No new failure mechanisms will be introduced and the change does not
affect the ability of the DGs to fulfill their safety-related
function. Therefore, this change does not create the possibility of
a new or different kind of accident from any accident previously
evaluated.
(3) The editorial changes are non-technical and therefore do not
create the possibility of a new or different kind of accident from
any accident previously evaluated.
3. Do the proposed changes involve a significant reduction in a
margin of safety?
(1) The allowance for unrestricted substitution of the modified
performance discharge test in lieu of the service discharge test
will not alter any accident analysis assumptions, initial
conditions, or results. Consequently, it does not have any effect on
the margin of safety. Therefore, this change does not involve a
significant reduction in a margin of safety.
(2) The proposed change to delete the requirement to demonstrate
that the DG can achieve and maintain steady state operation within
10 seconds is not an accident analysis assumption. The accident
analysis assumption that the DG reaches minimum conditions to accept
load within 10 seconds is preserved. Consequently, it does not have
any effect on the margin of safety. Therefore, this change does not
involve a significant reduction in a margin of safety.
(3) The editorial changes are non-technical and therefore 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: University of Texas at
Arlington Library, Government Publications/Maps, 702 College, P.O. Box
19497, Arlington, Texas 76019.
Attorney for licensee: George L. Edgar, Esq., Morgan, Lewis and
Bockius, 1800 M Street, NW., Washington, DC 20036.
NRC Section Chief: Robert A. Gramm.
Virginia Electric and Power Company, Docket Nos. 50-338 and 50-339,
North Anna Power Station, Unit No. 1 and Unit No. 2, Louisa County,
Virginia
Date of amendment request: May 3, 1999.
Description of amendment request: The proposed changes will delete
and/or relocate the additional primary-to-secondary leak rate limits
and enhanced leakage monitoring requirements imposed following the 1987
steam generator tube rupture event.
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:
[O]peration of the North Anna Power Station in accordance with
the proposed Technical Specification changes will not:
Involve a significant increase in the probability or
consequences of an accident previously evaluated.
Eliminating the conservative primary-to-secondary leakage limits
associated with the replaced steam generators and the operability
requirements for the leakage monitoring instrumentation does not
change the operation of the plant. The steam generators will be
operated, inspected, and maintained in the same manner. No new
accident initiators are established as a result of the proposed
changes. Therefore, the probability of occurrence is not increased
for any accident previously evaluated.
Removing the conservative primary-to-secondary leakage limits
associated with the replaced steam generators and the operability
requirements for the leakage monitoring instrumentation does not
change the operation of the plant. Although the conservative leakage
limits are being deleted, the remaining leakage limits will maintain
the dose rate, in the event of a tube rupture, within the analyzed
limits. Therefore, there is no increase in the consequences of any
accident previously analyzed[.]
Create the possibility of a new or different kind of accident
from any accident previously evaluated.
The proposed changes do not affect the operation of the plant.
The steam generators will be operated, inspected, and maintained in
the same manner. There are no modifications to the plant or steam
generators as a result of the change. No new accident or event
initiators are created by the removal of the conservative primary-
to-secondary leakage limits associated with the replaced steam
generators and the operability requirements for the leakage
monitoring instrumentation. Therefore, the proposed changes do not
create the possibility of any accident or malfunction of a different
type.
Involve a significant reduction in the margin of safety as
defined in the bases on any Technical Specifications.
The proposed changes have no effect on any safety analyses
assumptions. The remaining limits maintain primary-to-secondary
leakage within the accident analysis assumptions. The proposed
changes only eliminate overly conservative primary-to-secondary
leakage requirements and the operability and surveillance
requirements for the leakage monitoring system associated with the
replaced steam generators. Therefore, the proposed changes do not
result in 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: The Alderman Library, Special
Collections Department, University of Virginia, Charlottesville,
Virginia 22903-2498.
Attorney for licensee: Mr. Donald P. Irwin, Esq., Hunton and
Williams,
[[Page 29717]]
Riverfront Plaza, East Tower, 951 E. Byrd Street, Richmond, Virginia.
NRC Section Chief: Richard L. Emch, Jr.
Yankee Atomic Electric Co., Docket No. 50-029, Yankee Nuclear Power
Station (YNPS) Franklin County, Massachusetts
Date of amendment request: March 24, 1999.
Description of amendment request: The licensee submitted a request
to delete License Condition 2.C.(10), which states: ``The licensee
shall maintain a Fitness for Duty Program in accordance with the
requirements of 10 CFR Part 26.''
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
The proposed change is administrative in nature in that it
removes a reference in the YNPS Part 50 License to a regulatory
requirement no longer applicable to a plant which has permanently
ceased power operations and permanently removed fuel from its
reactor vessel. This will permit more cost beneficial use of
available resources with no diminution in the YNPS staff's ability
to maintain the safe operation of the YNPS SFP [spent fuel pool].
The change will not:
1. Involve a significant increase in the probability or
consequences of an accident previously evaluated. Each potential
accident in the YNPS FSAR [final safety analysis report] projects a
maximum release of activity and no prompt mitigation actions. None
of the analyzed scenarios resulted in a situation which could
significantly [a]ffect the public health and safety. Removal of a
regulatory requirement which does not apply to a plant which has
permanently ceased power operations and permanently removed fuel
from its reactor vessel cannot be deemed to involve a significant
increase in the probability or consequences of an accident
previously evaluated.
2. Create the possibility of a new or different accident from
any previously evaluated. The proposed change will not modify any
plant systems or components and, therefore will not create the
possibility of a new or different accident from any previously
evaluated.
3. Involve a significant reduction in the margin of safety.
Removal of a regulatory requirement which does not apply to a plant
which has permanently ceased power operations and permanently
removed fuel from its reactor vessel cannot be deemed to involve a
significant reduction in the margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: Greenfield Community College,
1 College Drive, Greenfield, Massachusetts 01301.
Attorney for licensee: Thomas Dignan, Esquire, Ropes and Gray, One
International Place, Boston, Massachusetts 02110-2624.
NRC Section Chief: Michael T. Masnik.
Previously Published Notices of Consideration of Issuance of Amendments
to Facility Operating Licenses, Proposed No Significant Hazards
Consideration Determination, and Opportunity for a Hearing
The following notices were previously published as separate
individual notices. The notice content was the same as above. They were
published as individual notices either because time did not allow the
Commission to wait for this biweekly notice or because the action
involved exigent circumstances. They are repeated here because the
biweekly notice lists all amendments issued or proposed to be issued
involving no significant hazards consideration.
For details, see the individual notice in the Federal Register on
the day and page cited. This notice does not extend the notice period
of the original notice.
Texas Utilities Electric Company, et al., Docket Nos. 50-445 and 50-
446, Comanche Peak Steam Electric Station, Unit Nos. 1 and 2, Somervell
County, Texas
Date of amendment request: February 11, 1999.
Description of amendment request: The proposed amendments would
credit soluble boron in the spent fuel pool water, in the maintenance
of a subcritical condition, and allow an increase in spent fuel storage
from 1291 to 2026 fuel assemblies.
Date of publication of individual notice in Federal Register: May
12, 1999 (64 FR 25522).
Expiration date of individual notice: June 11, 1999.
Local Public Document Room location: University of Texas at
Arlington Library, Government Publications/Maps, 702 College, P. O. Box
19497, Arlington, Texas.
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.
Duquesne Light Company, et al., Docket No. 50-334, Beaver Valley Power
Station, Unit No. 1, Shippingport, Pennsylvania
Date of application for amendment: January 17, 1998, as
supplemented by letters dated February 10, 1998, November 9, 1998,
February 8, 1999, and February 26, 1999.
Brief description of amendment: This amendment authorizes changes
to the Beaver Valley Power Station, Unit No. 1 (BVPS-1) Updated Final
Safety Analysis Report (UFSAR). Specifically, the authorized changes to
the UFSAR reflect revisions to the control room radiological dose
calculations for the waste gas system line break accident analysis to
correct a mathematical error discovered in a previous calculation, and
use of more conservative assumptions in the revised analysis.
Date of issuance: May 12, 1999.
[[Page 29718]]
Effective date: As of the date of issuance.
Amendment No: 222.
Facility Operating License No. DPR-66. Amendment approved changes
to the UFSAR.
Date of initial notice in Federal Register: February 25, 1998 (63
FR 9601).
The February 10, 1998, November 9, 1998, February 8, 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
initial notice.
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated May 12, 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 Operations, Inc., Docket No. 50-368, Arkansas Nuclear One, Unit
No. 2, Pope County, Arkansas
Date of application for amendment: August 23, 1996, as supplemented
on April 9, 1999.
Brief description of amendment: The amendment revises Section 5.0,
``Design Features,'' and Section 6.0, ``Administrative Controls,'' of
the Technical Specifications, adopting, for the most part, the format
and content of the NUREG-1432, Revision 1, ``Standard Technical
Specifications [STS] for Combustion Engineering Plants'' for the
changes requested. This amendment also relocates certain portions of
the design features section to other licensee-controlled documents in
accordance with the STS.
Date of issuance: May 19, 1999.
Effective date: As of its date of issuance and shall be implemented
within 30 days from the date of issuance: May 19, 1999.
Amendment No: 205.
Facility Operating License No. NPF-6: Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: October 9, 1996 (61 FR
52965).
The April 9, 1999, letter provided clarifying information that did
not change the scope of the original application and initial proposed
no significant hazards consideration determination.
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated May 19, 1999.
No significant hazards consideration comments received: No.
Local Public Document Room location: Tomlinson Library, Arkansas
Tech University, Russellville, Arkansas 72801.
Entergy Operations, Inc., Docket No. 50-313 and 50-368, Arkansas
Nuclear One, Units 1 and 2, Pope County, Arkansas
Date of amendment request: December 19, 1996, as supplemented by
letters dated August 6, 1998, and December 3, 1998.
Brief description of amendments: The amendments change requirements
for the control room ventilation system for both Units 1 and 2.
Date of issuance: May 19, 1999.
Effective date: As of the date of issuance and shall be implemented
within 30 days of the date of issuance.
Amendment Nos.: 196 and 206.
Facility Operating License Nos. DPR-51 and NPF-6: Amendments
revised the Technical Specifications.
Date of initial notice in Federal Register: January 29, 1997 (62 FR
4348).
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated May 19, 1999.
No significant hazards consideration comments received: No.
Local Public Document Room location: Tomlinson Library, Arkansas
Tech University, Russellville, Arkansas 72801.
Entergy Operations, Inc., System Energy Resources, Inc., South
Mississippi Electric Power Association, and Entergy Mississippi, Inc.,
Docket No. 50-416, Grand Gulf Nuclear Station, Unit 1, Claiborne
County, Mississippi
Date of application for amendment: January 12, 1999, which
superseded 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 Grand Gulf 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: May 7, 1999.
Effective date: As of the date of issuance and shall be implemented
within 30 days of issuance.
Amendment No.: 138.
Facility Operating License No. NPF-29: Amendment revises the
Technical Specifications.
Date of initial notice in Federal Register: February 10, 1999 (64
FR 6695), which superseded original notice of June 16, 1996 (61 FR
31178).
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated May 7, 1999.
No significant hazards consideration comments received: No.
Local Public Document Room location: Judge George W. Armstrong
Library, 220 S. Commerce Street, Natchez, Mississippi 39120.
FirstEnergy Nuclear Operating Company, Docket No. 50-346, Davis-Besse
Nuclear Power Station, Unit 1, Ottawa County, Ohio
Date of application for amendment: October 28, 1998, as modified by
letter dated March 19, 1999.
Brief description of amendment: This amendment revises
administrative requirements relating to: TS 6.5.1.6, Station Review
Board Responsibilities; TS 6.8.4.d, Radioactive Effluent Controls
Program; TS 6.10, Records Retention; TS 6.11, Radiation Protection
Program; TS 6.12, High Radiation Area; and TS 6.15, Offsite Dose
Calculation Manual.
Date of issuance: May 19, 1999.
Effective date: May 19, 1999.
Amendment No.: 231.
Facility Operating License No. NPF-3: Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: December 18, 1998 (63
FR 64126).
The supplemental information contained clarifying information and
did not change the initial proposed no significant hazards
consideration determination and did not expand the scope of the
application as described in the original Federal Register notice.
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated May 19, 1999.
No significant hazards consideration comments received: No.
Local Public Document Room location: University of Toledo, William
Carlson Library, Government Documents Collection, 2801 West Bancroft
Avenue, Toledo, OH 43606.
GPU Nuclear, Inc. et al., Docket No. 50-219, Oyster Creek Nuclear
Generating Station, Ocean County, New Jersey
Date of application for amendment: May 5, 1998, as supplemented
August 3 (2 letters), September 14, and December 22, 1998.
[[Page 29719]]
Brief description of amendment: The amendment approves the use of a
small amount of containment overpressure to ensure sufficient net
positive suction head for the emergency core cooling system pumps.
Date of Issuance: May 13, 1999.
Effective date: As of the date of issuance, to be implemented
within 30 days.
Amendment No.: 206.
Facility Operating License No. DPR-16. Amendment authorizes changes
to the Updated Final Safety Analysis Report.
Date of initial notice in Federal Register: October 21, 1998 (63 FR
56250).
The supplemental letters provided additional information that was
within the scope of the original application and did not change the
staff's proposed no significant hazards consideration determination.
The Commission's related evaluation of this amendment is contained
in a Safety Evaluation dated May 13, 1999.
No significant hazards consideration comments received: No.
Local Public Document Room location: Ocean County Library,
Reference Department, 101 Washington Street, Toms River, NJ 08753.
GPU Nuclear, Inc. et al., Docket No. 50-219, Oyster Creek Nuclear
Generating Station, Ocean County, New Jersey
Date of application for amendment: November 10, 1998.
Brief description of amendment: The proposed Technical
Specification (TS) change would remove the restriction on the sale or
lease of property within the exclusion area and replace the restriction
with a requirement to retain complete authority to determine and
maintain sufficient control of all activities including the authority
to exclude or remove personnel and property within the minimum
exclusion distance. A TS Bases page for the proposed change is
included. Also included are clarifications and administrative changes
which: (1) clarify TS definition 1.38 to become ``Site Boundary''
rather than the current term ``Exclusion Area'' to be consistent with
the 10 CFR 20.1003 definition for Site Boundary and the 10 CFR 100.3
definition of Exclusion Area, (2) revise the TS definition from
Exclusion Area to Site Boundary in TS 6.8.4(a)(9), and (3) revise and
update the TS Table of Contents for Section I Definitions.
Date of Issuance: May 12, 1999.
Effective date: As of the date of issuance, to be implemented
within 30 days.
Amendment No.: 205.
Facility Operating License No. DPR-16. Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: December 12, 1998 (63
FR 66595).
The Commission's related evaluation of this amendment is contained
in a Safety Evaluation dated May 12, 1999.
No significant hazards consideration comments received: No.
Local Public Document Room location: Ocean County Library,
Reference Department, 101 Washington Street, Toms River, NJ 08753.
GPU Nuclear, Inc. et al., Docket No. 50-219, Oyster Creek Nuclear
Generating Station, Ocean County, New Jersey
Date of application for amendment: February 12, 1999.
Brief description of amendment: The amendment deletes the
organizational chart and related references from the Appendix B
Environmental Technical Specifications (ETS). In addition, the
appearance and format of the ETS have been extensively revised.
Date of Issuance: May 18, 1999.
Effective date: As of the date of issuance, to be implemented
within 30 days from the date of issuance.
Amendment No.: 207.
Facility Operating License No. DPR-16. Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: April 7, 1999 (64 FR
17026).
The Commission's related evaluation of this amendment is contained
in a Safety Evaluation dated May 18, 1999.
No significant hazards consideration comments received: No.
Local Public Document Room location: Ocean County Library,
Reference Department, 101 Washington Street, Toms River, NJ 08753.
Northeast Nuclear Energy Company, et al., Docket No. 50-336, Millstone
Nuclear Power Station, Unit No. 2, New London County, Connecticut
Date of application for amendment: January 18, 1999, as
supplemented February 3 and March 17, 1999.
Brief description of amendment: The amendment removes Technical
Specification (TS) 3/4.6.4.3, ``Containment Systems, Hydrogen Purge
System,'' from the TS and allows downgrading the system to a non-
safety-related system.
Date of issuance: April 12, 1999.
Effective date: As of the date of issuance to be implemented within
60 days from the date of issuance.
Amendment No.: 233.
Facility Operating License No. DPR-65: Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: February 10, 1999 (64
FR 6704).
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated April 12, 1999.
No significant hazards consideration comments received: No.
Local Public Document Room location: Learning Resources Center,
Three Rivers Community-Technical College, 574 New London Turnpike,
Norwich, Connecticut, and the Waterford Library, ATTN: Vince Juliano,
49 Rope Ferry Road, Waterford, Connecticut.
Northeast Nuclear Energy Company, et al., Docket No. 50-423, Millstone
Nuclear Power Station, Unit No. 3, New London County, Connecticut
Date of application for amendment: January 18, 1999.
Brief description of amendment: The amendment modifies Technical
Specification 3/4.2.2 to be in accordance with NRC-approved
Westinghouse methodologies for the heat flux hot channel factor--
FQ(Z). In addition, the amendment makes changes to the core
operating limits and the analytical methods used to determine core
operating limits contained in Section 6.9.1.6.a and b, respectively, by
adding, modifying, or deleting references.
Date of issuance: May 10, 1999.
Effective date: As of the date of issuance to be implemented within
30 days from the date of issuance.
Amendment No.: 170.
Facility Operating License No. NPF-49: Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: February 10, 1999 (64
FR 6705).
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated May 10, 1999.
No significant hazards consideration comments received: No.
Local Public Document Room location: Learning Resources Center,
ThreeRivers 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, Docket Nos. 50-352 and 50-353, Limerick Generating
Station, Units 1 and 2, Montgomery County, Pennsylvania.
Date of application for amendments: October 30, 1998, as
supplemented February 22, 1999.
Brief description of amendments: These amendments revised the
overvoltage, undervoltage, and underfrequency allowable values
[[Page 29720]]
associated with the reactor protection system monitoring channels and
add supporting details to the Technical Specifications Bases 3/4.8.4.
Date of issuance: May 13, 1999.
Effective date: Units 1 and 2, as of date of issuance and shall be
implemented within 30 days.
Amendment Nos.: 134 and 96.
Facility Operating License Nos. NPF-39 and NPF-85. The amendments
revised the Technical Specifications.
Date of initial notice in Federal Register: November 18, 1998 (64
FR 64120)
The February 22, 1999, letter provided clarifying information that
did not change the initial proposed no significant hazards
consideration determination.
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated May 13, 1999.
No significant hazards consideration comments received: No.
Local Public Document Room location: Pottstown Public Library, 500
High Street, Pottstown, PA 19464.
PECO Energy Company, Docket No. 50-353, Limerick Generating Station,
Unit 2, Montgomery County, Pennsylvania
Date of application for amendment: January 12, 1999, as
supplemented January 29 and March 10, 1999.
Brief description of amendment: This amendment revised Technical
Specifications (TSs) Section 3/4.4.2, ``Safety/Relief Valves,'' and TS
Bases Sections B 3/4.4.2, B 3/4.5.1 and B 3/4.5.2 to increase the
allowable as-found main steam safety relief valve (SRV) code safety
function lift setpoint tolerance from plus or minus 1% to plus or minus
3%. Also, the required number of operable SRVs in operational
conditions 1, 2 and 3 will be increased from 11 to 12.
Date of issuance: May 17, 1999.
Effective date: May 17, 1999.
Amendment No.: 98.
Facility Operating License No. NPF-85. The amendment revises the
Technical Specifications.
Date of initial notice in Federal Register: February 24, 1999 (64
FR 9194)
The January 29 and March 10, 1999, letters provided clarifying
information that did not change the initial proposed no significant
hazards consideration determination or expand the scope of the original
Federal Register notice.
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated May 17, 1999.
No significant hazards consideration comments received: No.
Local Public Document Room location: Pottstown Public Library, 500
High Street, Pottstown, PA 19464.
PECO Energy Company, Docket No. 50-353, Limerick Generating Station,
Unit 2, Montgomery County, Pennsylvania
Date of application for amendment: March 11, 1999, as supplemented
April 21, 1999.
Brief description of amendment: The amendment revised the minimum
critical power ratio safety limits and revised the associated Technical
Specification Bases.
Date of issuance: May 14, 1999.
Effective date: As of the date of issuance and shall be implemented
prior to restart following completion of the April 1999 refueling
outage.
Amendment No.: 97.
Facility Operating License No. NPF-85. The amendment revised the
Technical Specifications and/or License.
Date of initial notice in Federal Register: April 7, 1999 (64 FR
17028). The April 21, 1999, letter provided clarifying information that
did not change the initial no significant hazards consideration
determination.
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated May 14, 1999.
No significant hazards consideration comments received: No.
Local Public Document Room location: Pottstown Public Library, 500
High Street, Pottstown, PA 19464.
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 application for amendments: May 22, 1997, as supplemented
by letters dated June 12, 1997, August 28, 1997, January 29, 1998, July
9, 1998, and March 12, 1999.
Brief description of amendments: The amendments authorize revisions
to the licensing basis as described in the Final Safety Analysis Report
(FSAR) Update to incorporate a modification to the Diablo Canyon Power
Plant, Unit Nos. 1 and 2 component cooling water system.
Date of issuance: May 13, 1999.
Effective date: May 13, 1999, and shall be implemented in the next
periodic update to the FSAR Update in accordance with 10 CFR 50.71(e).
Amendment Nos.: Unit 1-134; Unit 2-132.
Facility Operating License Nos. DPR-80 and DPR-82: The amendments
revised the Final Safety Analysis Report Update.
Date of initial notice in Federal Register: July 29, 1998 (63 FR
40558).
The supplemental letters dated July 9, 1998, and March 12, 1999
provided additional clarifying information, did not expand the scope of
the application as originally noticed, and did not change the staff's
initial no significant hazards consideration determination.
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated May 13, 1999.
No significant hazards consideration comments received: No.
Local Public Document Room location: California Polytechnic State
University, Robert E. Kennedy Library, Government Documents and Maps
Department, San Luis Obispo, California 93407.
Southern California Edison Company, et al., Docket Nos. 50-361 and 50-
362, San Onofre Nuclear Generating Station, Unit Nos. 2 and 3, San
Diego County, California
Date of application for amendments: May 8, 1996, as supplemented
January 13, 1999.
Brief description of amendments: The amendments modified the
technical specifications to allow refueling operation with 20 feet of
water level in the refueling cavity for many operating conditions and
at 12 feet of water level for certain specified conditions. The
amendments also restored a phrase to a note to Limiting Conditions for
Operation for TSs 3.9.4 and 3.9.5 that was inadvertently deleted by
previous amendments.
Date of issuance: May 13, 1999.
Effective date: May 13, 1999, to be implemented within 30 days from
the date of issuance.
Amendment Nos.: Unit 2-153; Unit 3-144.
Facility Operating License Nos. NPF-10 and NPF-15: The amendments
revised the Technical Specifications.
Date of initial notice in Federal Register: March 24, 1999 (64 FR
14285).
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated May 13, 1999.
No significant hazards consideration comments received: No.
Local Public Document Room location: Main Library, University of
California, P.O. Box 19557, Irvine, California 92713.
Southern Nuclear Operating Company, Inc., Docket Nos. 50-348 and 50-
364, Joseph M. Farley Nuclear Plant, Units 1 and 2, Houston County,
Alabama.
Date of amendments request: April 2, 1999.
Brief Description of amendments: The amendment changes TS 3/4.4.9,
``Specific Activity,'' and the associated bases to increase the limit
associated with dose equivalent iodine-131. The
[[Page 29721]]
steady-state dose equivalent iodine-131 limit would be increased from
0.15Curie/gram to 0.3 Curie/gram and the transient
limit for 80 percent to 100 percent power provided by Technical
Specificaton Figure 3.4-1 will increase 9 Curie/gram to 18
Curie/gram with a corresponding increase in the 0 percent to
80 percent power limits.
Date of issuance: May 10, 1999.
Effective date: As of the date of issuance, and shall be
implemented within 30 days from the date of issuance.
Amendment Nos.: Unit 1-142; Unit 2-134.
Facility Operating License Nos. NPF-2 and NPF-8: Amendments revise
the Technical Specifications.
Date of initial notice in Federal Register: April 8, 1999 (64 FR
17201).
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated May 10, 1999.
No significant hazards consideration comments received: No.
Local Public Document Room location: Houston-Love Memorial Library,
212 W. Burdeshaw Street, Post Office Box 1369, Dothan, Alabama.
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: September 29, 1998.
Brief description of amendments: The amendments authorize the
revision of the South Texas Project updated final safety analysis
report (UFSAR) to incorporate the revised methodology to calculate the
mass and energy release following a postulated large-break loss-of-
coolant accident.
Date of issuance: May 20, 1999.
Effective date: May 20, 1999 Revisions will be incorporated into
the next UFSAR update in accordance with the schedule in 10 CFR
50.71(e).
Amendment Nos.: Unit 1-110; Unit 2-97.
Facility Operating License Nos. NPF-76 and NPF-80: The amendments
authorize the revision of the UFSAR to incorporate the revised
methodology.
Date of initial notice in Federal Register: November 18, 1998 (63
FR 64123).
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated May 20, 1999.
No significant hazards consideration comments received: No.
Local Public Document Room location:
Wharton County Junior College, J. M. Hodges.
Learning Center, 911 Boling Highway, Wharton, Texas 77488.
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: October 29, 1998, supplemented by letter
dated March 15, 1999.
Brief description of amendments: The amendments relocate the
requirements in Technical Specifications 3/4.7.9 and 6.10.3.l for
snubbers to the Technical Requirements Manual.
Date of issuance: May 17, 1999.
Effective date: As of the date of issuance to be implemented within
30 days from the date of issuance.
Amendment Nos.: Unit 1-109; Unit 2-96.
Facility Operating License Nos. NPF-76 and NPF-80: The amendments
revised the Technical Specifications.
Date of initial notice in Federal Register: December 16, 1998 (63
FR 69346); renoticed April 7, 1999 (64 FR 17031).
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated May 17, 1999.
No significant hazards consideration comments received: No.
Local Public Document Room location: Wharton County Junior College,
J. M. Hodges Learning Center, 911 Boling Highway, Wharton, Texas 77488.
Wisconsin Public Service Corporation, Docket No. 50-305, Kewaunee
Nuclear Power Plant, Kewaunee County, Wisconsin
Date of application for amendment: February 15, 1999.
Brief description of amendment: This amendment revised Technical
Specification Section 6, ``Administrative Controls,'' to reflect
organizational changes, to relocate certain review and audit functions
to the Operational Quality Assurance Program Description, and to
eliminate redundant requirements.
Date of issuance: May 11, 1999.
Effective date: May 11, 1999.
Amendment No.: 145.
Facility Operating License No. DPR-43: Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: April 7, 1999 (64 FR
17031).
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated May 11, 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 25th day of May 1999.
For the Nuclear Regulatory Commission.
John A. Zwolinski,
Director, Division of Licensing Project Management, Office of Nuclear
Reactor Regulation.
[FR Doc. 99-13765 Filed 6-1-99; 8:45 am]
BILLING CODE 7590-01-P