[Federal Register Volume 61, Number 80 (Wednesday, April 24, 1996)]
[Notices]
[Pages 18162-18183]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: 96-9925]
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NUCLEAR REGULATORY COMMISSION
Biweekly Notice Applications and Amendments to Facility Operating
Licenses Involving No Significant Hazards Considerations
I. Background
Pursuant to Public Law 97-415, the U.S. Nuclear Regulatory
Commission (the Commission or NRC staff) is publishing this regular
biweekly notice. Public Law 97-415 revised section 189 of the Atomic
Energy Act of 1954, as amended (the Act), to require the Commission to
publish notice of any amendments issued, or proposed to be issued,
under a new provision of section 189 of the Act. This provision grants
the Commission the authority to issue and make immediately effective
any amendment to an operating license upon a determination by the
Commission that such amendment involves no significant hazards
consideration, notwithstanding the pendency before the Commission of a
request for a hearing from any person.
This biweekly notice includes all notices of amendments issued, or
proposed to be issued from March 30, 1996, through April 12, 1996. The
last biweekly notice was published on April 10, 1996 (61 FR 15985).
[[Page 18163]]
Notice of Consideration of Issuance of Amendments to Facility Operating
Licenses, Proposed No Significant Hazards Consideration Determination,
and Opportunity for a Hearing
The Commission has made a proposed determination that the following
amendment requests involve no significant hazards consideration. Under
the Commission's regulations in 10 CFR 50.92, this means that operation
of the facility in accordance with the proposed amendment would not (1)
Involve a significant increase in the probability or consequences of an
accident previously evaluated; or (2) create the possibility of a new
or different kind of accident from any accident previously evaluated;
or (3) involve a significant reduction in a margin of safety. The basis
for this proposed determination for each amendment request is shown
below.
The Commission is seeking public comments on this proposed
determination. Any comments received within 30 days after the date of
publication of this notice will be considered in making any final
determination.
Normally, the Commission will not issue the amendment until the
expiration of the 30-day notice period. However, should circumstances
change during the notice period such that failure to act in a timely
way would result, for example, in derating or shutdown of the facility,
the Commission may issue the license amendment before the expiration of
the 30-day notice period, provided that its final determination is that
the amendment involves no significant hazards consideration. The final
determination will consider all public and State comments received
before action is taken. Should the Commission take this action, it will
publish in the Federal Register a notice of issuance and provide for
opportunity for a hearing after issuance. The Commission expects that
the need to take this action will occur very infrequently.
Written comments may be submitted by mail to the Rules Review and
Directives Branch, Division of Freedom of Information and Publications
Services, Office of Administration, U.S. Nuclear Regulatory Commission,
Washington, DC 20555, and should cite the publication date and page
number of this Federal Register notice. Written comments may also be
delivered to Room 6D22, Two White Flint North, 11545 Rockville Pike,
Rockville, Maryland from 7:30 a.m. to 4:15 p.m. Federal workdays.
Copies of written comments received may be examined at the NRC Public
Document Room, the Gelman Building, 2120 L Street NW., Washington, DC.
The filing of requests for a hearing and petitions for leave to
intervene is discussed below.
By May 24, 1996, the licensee may file a request for a hearing with
respect to issuance of the amendment to the subject facility operating
license and any person whose interest may be affected by this
proceeding and who wishes to participate as a party in the proceeding
must file a written request for a hearing and a petition for leave to
intervene. Requests for a hearing and a petition for leave to intervene
shall be filed in accordance with the Commission's ``Rules of Practice
for Domestic Licensing Proceedings'' in 10 CFR Part 2. Interested
persons should consult a current copy of 10 CFR 2.714 which is
available at the Commission's Public Document Room, the Gelman
Building, 2120 L Street NW., Washington, DC and at the local public
document room for the particular facility involved. If a request for a
hearing or petition for leave to intervene is filed by the above date,
the Commission or an Atomic Safety and Licensing Board, designated by
the Commission or by the Chairman of the Atomic Safety and Licensing
Board Panel, will rule on the request and/or petition; and the
Secretary or the designated Atomic Safety and Licensing Board will
issue a notice of a hearing or an appropriate order.
As required by 10 CFR 2.714, a petition for leave to intervene
shall set forth with particularity the interest of the petitioner in
the proceeding, and how that interest may be affected by the results of
the proceeding. The petition should specifically explain the reasons
why intervention should be permitted with particular reference to the
following factors: (1) The nature of the petitioner's right under the
Act to be made a party to the proceeding; (2) the nature and extent of
the petitioner's property, financial, or other interest in the
proceeding; and (3) the possible effect of any order which may be
entered in the proceeding on the petitioner's interest. The petition
should also identify the specific aspect(s) of the subject matter of
the proceeding as to which petitioner wishes to intervene. Any person
who has filed a petition for leave to intervene or who has been
admitted as a party may amend the petition without requesting leave of
the Board up to 15 days prior to the first prehearing conference
scheduled in the proceeding, but such an amended petition must satisfy
the specificity requirements described above.
Not later than 15 days prior to the first prehearing conference
scheduled in the proceeding, a petitioner shall file a supplement to
the petition to intervene which must include a list of the contentions
which are sought to be litigated in the matter. Each contention must
consist of a specific statement of the issue of law or fact to be
raised or controverted. In addition, the petitioner shall provide a
brief explanation of the bases of the contention and a concise
statement of the alleged facts or expert opinion which support the
contention and on which the petitioner intends to rely in proving the
contention at the hearing. The petitioner must also provide references
to those specific sources and documents of which the petitioner is
aware and on which the petitioner intends to rely to establish those
facts or expert opinion. Petitioner must provide sufficient information
to show that a genuine dispute exists with the applicant on a material
issue of law or fact. Contentions shall be limited to matters within
the scope of the amendment under consideration. The contention must be
one which, if proven, would entitle the petitioner to relief. A
petitioner who fails to file such a supplement which satisfies these
requirements with respect to at least one contention will not be
permitted to participate as a party.
Those permitted to intervene become parties to the proceeding,
subject to any limitations in the order granting leave to intervene,
and have the opportunity to participate fully in the conduct of the
hearing, including the opportunity to present evidence and cross-
examine witnesses.
If a hearing is requested, the Commission will make a final
determination on the issue of no significant hazards consideration. The
final determination will serve to decide when the hearing is held.
If the final determination is that the amendment request involves
no significant hazards consideration, the Commission may issue the
amendment and make it immediately effective, notwithstanding the
request for a hearing. Any hearing held would take place after issuance
of the amendment.
If the final determination is that the amendment request involves a
significant hazards consideration, any hearing held would take place
before the issuance of any amendment.
A request for a hearing or a petition for leave to intervene must
be filed with the Secretary of the Commission, U.S. Nuclear Regulatory
Commission, Washington, DC 20555, Attention: Docketing and Services
Branch, or may be delivered to the Commission's Public
[[Page 18164]]
Document Room, the Gelman Building, 2120 L Street NW., Washington DC,
by the above date. Where petitions are filed during the last 10 days of
the notice period, it is requested that the petitioner promptly so
inform the Commission by a toll-free telephone call to Western Union at
1-(800) 248-5100 (in Missouri 1-(800) 342-6700). The Western Union
operator should be given Datagram Identification Number N1023 and the
following message addressed to (Project Director): petitioner's name
and telephone number, date petition was mailed, plant name, and
publication date and page number of this Federal Register notice. A
copy of the petition should also be sent to the Office of the General
Counsel, U.S. Nuclear Regulatory Commission, Washington, DC 20555, and
to the attorney for the licensee.
Nontimely filings of petitions for leave to intervene, amended
petitions, supplemental petitions and/or requests for a hearing will
not be entertained absent a determination by the Commission, the
presiding officer or the Atomic Safety and Licensing Board that the
petition and/or request should be granted based upon a balancing of
factors specified in 10 CFR 2.714(a)(1) (i)-(v) and 2.714(d).
For further details with respect to this action, see the
application for amendment which is available for public inspection at
the Commission's Public Document Room, the Gelman Building, 2120 L
Street NW., Washington, DC, and at the local public document room for
the particular facility involved.
Carolina Power & Light Company, et al., Docket No. 50-400, Shearon
Harris Nuclear Power Plant, Unit 1, Wake and Chatham Counties, North
Carolina
Date of amendment request: March 20, 1996.
Description of amendment request: The licensee proposes to relocate
Technical Specification (TS) 3.3.3.2, Movable Incore Detectors, to the
Harris Nuclear Plant Core Operating Limits Report (COLR). Future
changes to the relocated provisions will be evaluated in accordance
with 10 CFR 50.59.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. The proposed amendment does not involve a significant increase
in the probability or consequences of an accident previously evaluated.
The proposed change will simplify the Technical Specifications,
while implementing the recommendations of the Commission's Final Policy
Statement on TS Improvements. The changes are administrative in nature
and do not involve any modifications to plant equipment or affect plant
operation. Since the TS provisions are being relocated to a licensee-
controlled document, any future changes will be controlled under 10 CFR
50.59. Therefore, there would be no increase in the probability or
consequences of an accident previously evaluated.
2. The proposed amendment does not create the possibility of a new
or different kind of accident from any accident previously evaluated.
The proposed change is a relocation of existing Technical
Specification provisions. It does not involve any physical alterations
to plant equipment or alter the method by which any safety-related
system performs its function. Therefore, the proposed changes do not
create the possibility of a new or different kind of accident from any
accident previously evaluated.
3. The proposed amendment does not involve a significant reduction
in the margin of safety.
The proposed change does not affect any Final Safety Analysis
Report (FSAR) Chapter 15 accident analyses or have any impact on margin
as defined in the Bases to the Technical Specifications. Therefore, the
proposed changes do not involve a significant reduction in a margin of
safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: Cameron Village Regional
Library, 1930 Clark Avenue, Raleigh, North Carolina 27605
Attorney for licensee: W. D. Johnson, Vice President & Senior
Counsel, Carolina Power & Light Company, Post Office Box 1551, Raleigh,
North Carolina 27602
NRC Project Director: Eugene V. Imbro
Connecticut Yankee Atomic Power Company, Docket No. 50-213, Haddam Neck
Plant, Middlesex County; Northeast Nuclear Energy Company, et al.,
Docket Nos. 50-245, 50-336, 50-423, Millstone Nuclear Power Station,
Units 1, 2, and 3, New London County, Connecticut; and North Atlantic
Energy Service Company, Docket No. 50-443, Seabrook Station, Unit No.
1, Rockingham County, New Hampshire
Date of amendment request: February 1, 1996
Description of amendment request: The amendment request would
revise Section 6 ``Administrative Controls,'' of the Haddam Neck Plant,
Millstone Unit Nos. 1, 2, and 3, and Seabrook Station, Unit 1 Technical
Specifications to reflect several changes in organizational titles. The
proposed changes are administrative title and editorial changes only.
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 (SHC), which is presented below:
* * * The proposed changes do not involve an SHC because the change
would not:
1. Involve a significant increase in the probability or
consequences of an accident previously evaluated.
No design basis accidents are affected by these proposed changes.
The proposed changes are administrative and editorial in nature and are
being proposed to reflect the recently announced organizational changes
which will become effective on February 1, 1996. These changes include:
insertion of the function Chief Nuclear Officer, in lieu of Executive
Vice President--Nuclear; and establishment of a single point of
operational direction for all five units in the position of the Vice
President--Nuclear Operations. This individual is in lieu of the
positions of Vice President--Haddam Neck, Senior Vice President--
Millstone Station, and Executive Director--Nuclear Production. These
latter positions have been eliminated; other changes are: the
appointment of the Haddam Neck Plant Nuclear Unit Director as chairman
of the Haddam Neck PORC [Plant Operations Review Committee]; promotion
of the Shift Supervisor/Shift Superintendent to the position of Shift
Manager; revising the titles of ``additional operator'' and ``auxiliary
operator'' to ``nuclear systems operator''; modifying the phrase
``crewman'' to a gender neutral term ``crewperson'';
[[Page 18165]]
reassignment of the delivery of ISEG [Independent Safety Engineering
Group] reports to the Senior Vice President--Nuclear Safety and
Oversight; and a change to the title of the Seabrook Station Manager to
Station Director. No safety systems are adversely affected by the
proposed changes, and no failure modes are associated with the changes.
Therefore, there is no impact on the probability of occurrence or the
consequences of any accidents previously evaluated.
2. Create the possibility of a new or different kind of accident
from any accident previously evaluated.
Because there are no changes in the way the plants are operated due
to this administrative change, the potential for an unanalyzed accident
is not created. There is no impact on plant response, and no new
failure modes are introduced. These proposed administrative and
editorial changes have no impact on safety limits or design basis
accidents, and they have no potential to create a new or unanalyzed
event.
3. Involve a significant reduction in a margin of safety.
The changes do not directly affect any protective boundaries nor do
they impact the safety limits for the protective boundaries. These
proposed changes are administrative and editorial in nature. Therefore,
there can be no 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 locations: For the Haddam Neck Plant,
Russell Library, 123 Broad Street, Middletown, CT 06457; for Millstone
Nuclear Power Station, Unit Nos. 1, 2, and 3, Learning Resources
Center, Three Rivers Community-Technical College, 574 New London
Turnpike, Norwich, CT 06360; for Seabrook Station, Unit No. 1, Exeter
Public Library, Founders Park, Exeter, NH 03833.
Attorney for Licensees: Lillian M. Cuoco, Esq., Senior Nuclear
Counsel, Northeast Utilities Service Company, P.O. Box 270, Hartford,
CT 06141-0270.
NRC Project Director: Phillip F. McKee
Duke Power Company, et al., Docket No. 50-413, Catawba Nuclear
Station, Unit 1, York County, South Carolina
Date of amendment request: January 26, 1996.
Description of amendment request: The amendment would allow a one-
time change to the Technical Specifications (TS) to allow operation of
the containment purge ventilation system during Modes 3 and 4 during
startup following the forthcoming Unit 1 steam generator replacement
outage. This would alleviate respiratory hazards to personnel who would
enter the containment to perform surveillances during Modes 4 and 3 of
startup operations. Those hazards are expected to result from the
thermal decomposition product gases evolving from the heatup of newly
installed thermal insulation. Operation of the containment purge system
to exhaust these gases would ensure that the air quality meets
applicable standards for personnel safety.
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 activity does not involve a significant increase in the
probability or consequences of an accident previously evaluated.
The VP [Containment Purge] System has no interfaces with any
primary system, secondary system, or power transmission system. It has
no interfaces with any reservoir of radioactive gases or liquids. None
of the systems listed above are modified by the activity. In summary,
no ``accident initiator'' is affected with the proposed operation of
the VP System in Mode[s] 3 and 4. For this reason, the activity does
not involve an increase in the probability of an accident previously
evaluated.
Analyses have been performed to determine upper bounds to the
source term, the offsite doses, and the Control Room dose. The results
of that analyses are reported above. Both the source term and the doses
were found to be significantly lower than the results of the
corresponding design basis analyses. No credit was taken for operation
of the annulus ventilation system (VE) in the dose analysis. In
addition, it has been determined that with no credit taken for any heat
transfer from the fuel and cladding to the moderator channels, that
sufficient time would exist for the operators to initiate recovery of
flow from the ECCS [Emergency Core Cooling System] to the reactor core.
The flow required from the ECCS to maintain the core in a coolable
geometry was found to be well within the capacity of any one ECCS pump.
Furthermore, it was determined that convective heat transfer to steam
would be sufficient to prevent release of significant source term or a
significant degree of fuel damage.
For the above reasons, it is determined that operation of the VP
System in Mode 3 or 4 immediately following the steam generator
replacement outage does not involve a significant increase in either
the probability or the consequences of an accident previously
evaluated.
(2) The activity does not create the possibility of a new or
different type of accident from any accident previously evaluated.
As discussed above, no ``accident initiators'' are affected by the
proposed activity. Operation of the VP System proposed for Modes 3 and
4 will be the same as that routinely carried in other modes of
operation. For these reasons, the activity will not create the
possibility of a new or different type of accident from any previously
evaluated.
(3) The activity does not involve a significant reduction in the
margin of safety.
Margin of safety is associated with confidence in the ability of
the fission product barriers (the fuel and fuel cladding, the Reactor
Coolant System pressure boundary, and the containment) to limit the
level of radiation doses to the public. The proposed operation of the
VP System will occur at the end of an extended outage. The level of
decay heat and activity in the reactor is very low compared to the
level of decay heat and activity associated with full power operations.
For this reason, the likelihood of damage to the fuel following a
DBLOCA [design basis loss-of-coolant analysis] occurring during the
proposed purging is reduced, as determined above. Both offsite doses
and doses to the Control Room were found to be small compared to the
limits of 10 CFR [Part] 100 and GDC [General Design Criterion] 19. For
these reasons, the activity 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: York County Library, 138 East
Black Street, Rock Hill, South Carolina 29730.
Attorney for licensee: Mr. Albert Carr, Duke Power Company, 422
South Church Street, Charlotte, North Carolina 28242.
NRC Project Director: Herbert N. Berkow.
[[Page 18166]]
Duke Power Company, Docket Nos. 50-369 and 50-370, McGuire Nuclear
Station, Units 1 and 2, Mecklenburg County, North Carolina
Date of amendment request: December 12, 1995.
Description of amendment request: The proposed amendments would
correct an error in the Axial Flux Difference (AFD) Equations to more
accurately reflect the proper AFD limit reduction, which is more
conservative than the literal interpretation of the current Technical
Specifications.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
A. The change would not involve a significant increase in the
probability or consequences of an accident previously evaluated.
The monitoring of core power distribution and peaking factors is to
ensure accident analysis assumptions such as maximum local pin power at
the initiation of an accident are satisfied, and are not involved in
the initiation or mitigation of any previously evaluated accident.
The proposed change is actually more conservative than the existing
Technical Specification currently being used at McGuire.
B. The change will not create the possibility of a new or different
kind of accident from any accident previously evaluated.
No plant modifications (hardware or control methods) are involved
with this proposed change. The change is simply to correct an error in
the Specification introduced in Amendments 130 (Unit 1) and 112 (Unit
2). The proposed change is more restrictive than the current
specification. No changes are proposed which could create any new
accident scenarios.
C. The proposed change will not involve a significant reduction in
any margin of safety.
The proposed change ensures the margin of safety is properly
maintained by properly reducing (instead of increasing) the Positive
AFD [Axial Flux Difference] limit if a peaking factor exceeds its
surveillance limit. The change is more conservative than the existing
Specification and will ensure the margins of safety are properly
maintained.
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: Atkins Library, University of
North Carolina, Charlotte (UNCC Station), North Carolina 28223.
Attorney for licensee: Mr. Albert Carr, Duke Power Company, 422
South Church Street, Charlotte, North Carolina 28242.
NRC Project Director: Herbert N. Berkow.
Duke Power Company, Docket Nos. 50-369 and 50-370, McGuire Nuclear
Station, Units 1 and 2, Mecklenburg County, North Carolina
Date of amendment request: March 4, 1996.
Description of amendment request: The proposed amendments would
delete the Flow Monitoring System from Technical Specification (TS)
3.4.6.1 and associated surveillance requirements. The TS requires that
either the Containment Floor and Equipment Sump Level System or the
Flow Monitoring System be used to ensure that Reactor Coolant leakage
is maintained within the specified limits. Duke Power does not use the
Flow Monitoring System as a result of documented instrumentation
inaccuracies due to the as-built piping configuration. The existing
piping configuration does not ensure a water solid line which is
necessary for the correct operation of any type of flow
instrumentation. Modification to add a loop seal downstream of the flow
element would be necessary for operability, which would create access
difficulties as well as increase the potential for a radiological
hazard in the form of a CRUD trap.
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. This amendment will not significantly increase the probability
or consequence of any accident previously evaluated.
This change will not increase the probability or consequences of an
accident since this Reactor Coolant Leakage Detection instrumentation
is not an accident initiator or mitigator.
This proposed Technical Specification change does not decrease the
number of methods for Reactor Coolant leakage detection. This change
will ensure there are still three distinctly separate methods of
detecting NC [reactor coolant] leakage within the Containment Building.
The first method will be detecting liquid leakage inside Containment
via CFAE [Containment Floor and Equipment] level monitoring. The second
method is detecting an increase in Radiation levels inside Containment
and the third method is detecting steam leakage inside Containment. All
three methods satisfy the diversity requirements listed in Regulatory
Guide 1.45 for detecting a Reactor Coolant leak inside Containment.
The sensitivity requirement listed in Regulatory Guide 1.45 is to
detect a Reactor Coolant leak of one (1) gpm in one (1) hour. The first
method meets this by use of the Sump level monitoring and rate of
increase alarm from this level monitoring device. There are two sumps
inside containment and the levels for both sumps are combined for
detecting a one (1) gpm leak. McGuire uses the Sump Level monitoring to
adequately address liquid leakage detection inside Containment;
therefore, a flow monitoring system on the Sump Discharge line is not
necessary and can be deleted.
The Radiation Monitors are also set up to the required Regulatory
Guide 1.45 sensitivity for detecting Reactor Coolant leakage and are
not designed for SSE [safe-shutdown earthquake] events per the McGuire
FSAR [Final Safety Analysis Report] (see McGuire's Request for
Amendment: Reactor Coolant Leakage Detection Systems, dated March 4,
1996).
The third method for detecting Reactor Coolant leakage is to
monitor Containment Ventilation Condensate Drain Tank (VUCDT) flow, for
which McGuire is also using a level monitor. As in the case of the CFAE
Unit Sump Level monitor, level monitoring for leakage detection is more
reliable than flow monitoring.
2. This amendment will not create the possibility of any new or
different kind of accident not previously evaluated.
The CFAE Flow Monitoring System has no control function, ([i.e.,]
it is only a process monitor). Therefore, its deletion cannot create
the pos[s]ibility of a new or different kind of accident.
3. This amendment will not involve a significant reduction in a
margin of safety.
This proposed Tech Spec change does not decrease the number of
methods for Reactor Coolant leakage detection. This change will ensure
there are still three distinctly separate methods of detecting Reactor
Coolant leakage within the Containment Building.
Tech Spec 3.4.6.1 specifies two Radiation Monitors as two separate
[[Page 18167]]
required methods for Reactor Coolant Leakage Detection with the
Containment Ventilation condensate level monitoring as a backup. The
third method is the Containment Sump level monitoring with the flow
monitoring as a backup.
The new standardized Tech Spec 3.4.15, lists method one as
Containment Sump (Level OR Discharge Flow) Monitoring Device. McGuire
proposes to use a Sump Level monitoring device only. The second method
listed is one Containment Radiation Monitor (either the gaseous or
particulate monitor). McGuire will still have both available. The third
method listed is one Containment air cooler condensate flow rate
monitor for which McGuire plans to also use a level monitor. Liquid,
Radiation, and Steam monitoring will still be accounted for in the Tech
Spec, with the additional requirement of running a Reactor Coolant leak
calculation if any of the methods are inoperable.
Since McGuire is retaining three distinct methods of Reactor
Coolant leakage detection per current TS [technical specification]
requirements (and in agreement with current ISTS [improved standard
technical specification] requirements), the proposed Technical
Specification amendment does not cause any reduction in safety margin.
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: Atkins Library, University of
North Carolina, Charlotte (UNCC Station), North Carolina 28223.
Attorney for licensee: Mr. Albert Carr, Duke Power Company, 422
South Church Street, Charlotte, North Carolina 28242.
NRC Project Director: Herbert N. Berkow.
Georgia Power Company, Oglethorpe Power Corporation, Municipal Electric
Authority of Georgia, City of Dalton, Georgia, Docket Nos. 50-321 and
50-366, Edwin I. Hatch Nuclear Plant, Units 1 and 2, Appling County,
Georgia
Date of amendment request: February 21, 1996.
Description of amendment request: The licensee proposes a change to
the Plant Hatch Unit 1 and Unit 2 Technical Specifications. The
proposed revision would change the Drywell Air Temperature Limiting
Condition for Operation (LCO) from less than or equal to 135 deg.F to
less than or equal to 150 deg.F. The proposed change would provide a
margin for the primary containment Drywell Air Temperature LCO when
prolonged summer and high river temperatures are experienced. Also, a
correction to a Final Safety Analysis Report (FSAR) reference would be
made. This typographical error is strictly editorial.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration which is presented below:
1. The proposed change does not involve a significant increase in
the probability or consequences of an accident previously evaluated.
The probability (frequency of occurrence) of previously evaluated
accidents is not a function of the ambient drywell air temperature.
Instrumentation setpoint calculations were assessed, and the increased
ambient drywell air temperature does not affect any instrumentation
setpoints or allowable values.
The design basis accidents were reevaluated utilizing the increased
drywell air temperature as an initial assumption. The results indicated
that no regulatory limits or equipment design requirements will be
exceeded as the result of the proposed change. Therefore, the change in
drywell air temperature does not result in a significant increase in
the probability or consequences of any previously evaluated accidents.
2. The proposed change does not create the possibility of a new or
different kind of accident from any accident previously analyzed.
Revising the Drywell Air Temperature LCO does not physically modify the
plant nor does it modify the operation of any existing equipment.
3. The proposed change does not involve a significant reduction in
a margin of safety. Design bases analyses performed utilizing 150 deg.F
as the initial drywell temperature demonstrate that design and
regulatory limits are not exceeded. Equipment in the drywell required
to mitigate the effects of a DBA [design basis accident] is qualified
to operate under environmental conditions expected for an accident.
Analysis results do not affect instrumentation setpoints or
calibration, or accident equipment qualification.
Equipment qualified life is evaluated by an existing program which
uses elevation-dependent drywell temperature rather than bulk average
temperature. Therefore, the margin of safety associated with safety and
other limits identified in the Technical Specifications are not
significantly reduced.
The correction to an FSAR reference is strictly editorial.
Therefore, it meets the three criteria stated above.
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: Appling County Public Library,
301 City Hall Drive, Baxley, Georgia 31513.
Attorney for licensee: Ernest L. Blake, Jr., Esquire, Shaw,
Pittman, Potts and Trowbridge, 2300 N Street NW., Washington, DC 20037.
NRC Project Director: Herbert N. Berkow.
GPU Nuclear Corporation, et al., Docket No. 50-219, Oyster Creek
Nuclear Generating Station, Ocean County, New Jersey
Date of amendment request: March 28, 1996 (TSCR 234).
Description of amendment request: The proposed amendment modifies
statements in the Technical Specifications and bases to correctly
reflect the reference parameter for anticipatory scram signal bypass.
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. State the basis for the determination that the proposed activity
will or will not increase the probability of occurrence or consequences
of an accident.
This change modifies the terminology in a footnote to a Technical
Specification Table and the bases. The change properly aligns the
footnote and the bases with the FSAR [final safety analysis report] and
the newly revised conservative setpoint which now correctly correlates
the high pressure turbine third stage extraction steam line pressure to
rated reactor thermal power. The change does not modify the function or
operation of the bypass logic. Therefore, the proposed change will not
increase the probability of occurrence or consequences of an accident.
2. State the basis for the determination that the activity does or
does not create the possibility of an accident or malfunction of
equipment of
[[Page 18168]]
a different type than any previously identified in the SAR.
The change does not involve any hardware and does not alter the
functional intent of the pressure switches. The change of the footnote
wording and the bases are primarily administrative and the existing
Technical Specification Limiting Condition for Operation are preserved.
Thus the proposed activity does not create the possibility of an
accident or malfunction of a different type than any previously
identified in the SAR.
3. State the basis for the determination that the margin of safety
as defined in the bases of any Technical Specification is not reduced.
The revised setpoint assures that the anticipatory scram signal
bypass is removed before reaching the Technical Specification limit of
40 percent rated reactor thermal power (during power ascension). Thus,
the margin of safety as stated in the bases of Technical Specification
3.1 is preserved.
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: Ocean County Library,
Reference Department, 101 Washington Street, Toms River, NJ 08753.
Attorney for licensee: Ernest L. Blake, Jr., Esquire. Shaw,
Pittman, Potts & Trowbridge, 2300 N Street NW., Washington, DC 20037.
NRC Project Director: John F. Stolz.
Illinois Power Company and Soyland Power Cooperative, Inc., Docket No.
50-461, Clinton Power Station, Unit No. 1, DeWitt County, Illinois
Date of amendment request: February 22, 1996 (U-602554)
Description of amendment request: The proposed amendment would
modify Technical Specifications 3.3.8.1, ``Loss of Power
Instrumentation,'' and 3.8.1, ``AC Sources-Operating.'' The proposed
changes would delete the Surveillance Requirement (SR) 3.3.8.1.1 which
requires a channel check for Loss of Power instrumentation and change
Technical Specification Table 3.3.8.1-1 to change the allowable value
for the Degraded Voltage Function (items 1.c and 2.c) from ``[greater
than or equal to] 3762V and [less than or equal to] 3832V'' to
``[greater than or equal to] 3876V.'' The amendment would also change
Technical Specification Table 3.3.8-1 to modify the Division 3 degraded
voltage logic to be the same as Divisions 1 and 2 (i.e., two-out-of-two
rather than three-out-of-three), and increase the steady state voltage
from [greater than or equal to] 3740V to [greater than or equal to]
3870V for SRs 3.8.1.2, 3.8.1.7, 3.8.1.11, 3.8.1.12, 3.8.1.15, 3.8.1.19
and 3.8.1.20.
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) None of the proposed changes involve a significant increase in
the probability or consequences of any accident previously evaluated.
Each of the proposed changes is evaluated against this criteria as
discussed below.
The deletion of the channel check surveillance will result in
discontinuing the recording of information that is not effective in
assessing the capability of the degraded voltage relays to perform
their intended function. Deletion of the channel check does not change
the design or the expected performance of the Loss of Power (LOP)
degraded voltage instrumentation, and therefore, the proposed change
does not impact the intended function of this instrumentation to ensure
adequate voltage for the ECCS equipment during DBA and other non-
accident scenarios. This surveillance provides little added assurance
of relay operability since the relay is normally in a ``non-tripped''
state.
The revision of the Allowable Values for the LOP degraded voltage
and increase in the minimum required voltage for testing diesel
generators will not result in any increase in the probability or
consequences of any accident. The revised Allowable Values will
continue to provide assurance that adequate voltage is available to run
ECCS equipment during DBAs or any other non accident scenarios. With
the emergency bus(es) voltage at or greater than the revised Allowable
Values, the operability of required ECCS equipment is assured. The
revised setpoints for the degraded voltage instrumentation, as
controlled under 10CFR50.59 in the Clinton Power Station Operational
Requirements Manual (ORM), are sufficiently low to assure that the
possibility of spurious trips is minimized.
The planned modification for Division 3 LOP degraded voltage
sensor/relay logic will make Division 3 logic identical to the present
designs for Division 1 and 2. The proposed design for Division 3 will
not result in an increase in the probability of any accident because
the proposed LOP Degraded Voltage logic for Division 3 will be
identical to the proven design of Division 1 and 2. There will not be
an increase in the consequences of an accident because the design of
the LOP Degraded Voltage instrumentation will continue to ensure
adequate voltage for ECCS equipment during any DBA and during non-
accident scenarios.
The proposed footnotes merely assure that the proposed changes
become effective upon installation of the corresponding plant
modifications. Thus, these changes are purely administrative.
Chapter 15 of the Clinton Updated Safety Analysis Report (USAR)
discusses the effects of anticipated process disturbances to determine
their consequences and the capability of the plant to control or
accommodate such events. Subsection 15.2.6 discusses loss of AC power,
including loss of grid voltage. This discussion demonstrates that fuel
design limits and reactor coolant pressure boundary design conditions
are not exceeded. The proposed changes do not affect the discussion nor
the conclusion of this evaluation.
(2) None of the proposed changes create the possibility of a new or
different kind of accident from any accident previously evaluated. Each
of the proposed changes is evaluated against this criterion as
discussed below.
The proposed changes (deletion of the channel check, the revised
Allowable Value for the LOP degraded voltage instrumentation, revision
of the minimum required voltage for the diesel generator (DG)
surveillance, and change of the number of required channels for
Division 3) do not alter the intent or purpose of the degraded voltage
instrumentation. The instrumentation will continue to function to
protect the loads on the emergency bus by switching automatically to
the on site power source when the voltage has been at a degraded
condition for greater than the Allowable Value of the time delay. The
LOP instrumentation provides a responsive actuation (trip) to an
accident or scenario where the protection provided by this function
prevents damage to ECCS equipment during undervoltage (degraded
voltage) conditions on the emergency bus(es). Because the
instrumentation will continue to function to ensure that the emergency
bus voltage for all three divisions is sufficient for the proper
operation of all class 1E equipment down to the 120 volt level, the
proposed change does not create the possibility of a new or different
kind of accident from any accident previously evaluated. The change in
the lower voltage for the DG
[[Page 18169]]
surveillances will not impact the way the surveillances are conducted
because the DGs are run as close to the nominal voltage as possible.
The lower voltage is a criterion for evaluating the surveillance and
the revised lower voltage is adequate for its intended purpose.
(3) None of the proposed changes involve a significant reduction in
a margin of safety. Each of the proposed changes is evaluated against
this criterion as discussed below.
The proposed deletion of the channel check SR 3.3.8.1.1 will not
result in any reduction of the margin of safety because the channel
check is ineffective and the status of the channel will continue to be
apparent to plant personnel because of information provided by other TS
required surveillances. The margin of safety is provided by LOP
instrumentation ensuring the emergency bus(es) have adequate voltage to
support ECCS operability. The proposed revision of the Allowable Value
for the LOP degraded voltage will provide assurance that emergency
bus(es) voltage will be adequate for ECCS loads during DBA and other
non-accident scenarios. These setpoints were determined based on
revised voltage calculations and using an NRC-approved setpoint
methodology. Thus, these changes will not involve any reduction of the
margin of safety. The proposed revision of the number of required
channels for Division 3 will not result in a reduction in a margin of
safety because the proposed Division 3 LOP Degraded Voltage
instrumentation logic will be the same as the proven design of Division
1 and 2. This modification will improve plant maintenance and training
by making Divisions 1, 2 and 3 similar thereby enhancing plant
performance and safety.
Similarly, the proposed revision of the lower voltage limit for
voltage for the DG surveillances (SR 3.8.1.2, SR 3.8.1.7, SR 3.8.1.11,
SR 3.8.1.12, SR 3.8.1.15, SR 3.8.1.19, and SR 3.8.1.20) will assure
that the DGs will be capable of controlling voltage to a range that
will be adequate for the loads on the bus. This value was determined
using revised voltage calculations and is consistent with the proposed
degraded voltage setpoints. None of the proposed changes will involve a
significant reduction in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: Vespasian Warner Public
Library, 120 West Johnson Street, Clinton, Illinois 61727.
Attorney for licensee: Leah Manning Stetener, Vice President,
General Counsel, and Corporate Secretary, 500 South 27th Street,
Decatur, Illinois 62525.
NRC Project Director: Gail H. Marcus.
Illinois Power Company and Soyland Power Cooperative, Inc., Docket No.
50-461, Clinton Power Station, Unit No. 1, DeWitt County, Illinois
Date of amendment request: February 22, 1996 (U-602551).
Description of amendment request: The proposed amendment would
change Technical Specification 3.4.11, ``Reactor Coolant System (RCS)
Pressure and Temperature (P/T) Limits,'' to incorporate specific P/T
limits for the bottom head region of the reactor vessel, separate and
apart from the core beltline region of the reactor vessel.
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 results in a specific pressure and
temperature (P/T) limit curve for the bottom head during vessel
pressure testing evolutions, while the P/T limits for the remaining
balance of reactor pressure vessel regions are unchanged. The limits
for the bottom head region, which are only applicable during vessel
system pressure or leak testing, were developed consistent with
Regulatory Guide 1.99, Revision 2; 10CFR50, Appendix G; ASME Section
III, Appendix G; and Welding Research Council (WRC) Bulletin 175.
Additionally, the proposed change does not result in a change to the
way in which the hydrostatic pressure tests are performed. That is,
conformance to the P/T limits specified in Technical Specification
Figure 3.4.11-1 with the proposed bottom head P/T limits incorporated,
will continue to provide protection against brittle fracture of the
vessel system during required testing so that vessel integrity is
maintained. Therefore, this proposed change does not result in an
increase in the probability or consequences of any accident previously
evaluated.
(2) The proposed change does not result in any change to the plant
or the way in which the hydrostatic pressure tests are performed. As a
result, no new failure modes are introduced. Therefore, the proposed
change cannot create the possibility of a new or different kind of
accident from any accident previously evaluated.
(3) The new P/T limit curve for the bottom head has been developed
consistent with Regulatory Guide 1.99, Revision 2; 10CFR50, Appendix G;
ASME Section III, Appendix G; and Welding Research Council (WRC)
Bulletin 175. All other regions of the reactor pressure vessel retain
their applicability to appropriate and previously approved P/T limit
curves which are based on the same methodology. Conformance to the P/T
limit curves, with the proposed changes incorporated, will continue to
provide adequate margins of safety against brittle fracture of the
reactor vessel. Therefore, this proposed change does 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 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: Vespasian Warner Public
Library, 120 West Johnson Street, Clinton, Illinois 61727.
Attorney for licensee: Leah Manning Stetener, Vice President,
General Counsel, and Corporate Secretary, 500 South 27th Street,
Decatur, Illinois 62525.
NRC Project Director: Gail H. Marcus.
Illinois Power Company and Soyland Power Cooperative, Inc., Docket No.
50-461, Clinton Power Station, Unit No. 1, DeWitt County, Illinois
Date of amendment request: February 22, 1996 (U-602522)
Description of amendment request: The proposed amendment would
change Technical Specification 3.3.4.1, ``End of Cycle Recirculation
Pump Trip (EOC-RPT) Instrumentation,'' by deleting Surveillance
Requirement (SR) 3.3.4.1.6. The SR requires the reactor recirculation
pump trip breaker interruption time to be determined at least once per
60 months.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration which is presented below:
(1) End of cycle recirculation pump trip (EOC-RPT) actuation in
response to main generator load rejection and main turbine trip events
has previously been evaluated in Chapter 15 of Clinton Power Station
(CPS) Updated Final
[[Page 18170]]
Safety Analysis. The proposed change does not affect the initiators of
any of these events. In addition, the possibility of failure of the
EOC-RPT breaker to mitigate these events has not been increased because
there has been no change in design and no change to the plant. Deleting
the requirement to periodically measure the breaker arc suppression
time will not impact the EOC-RPT breakers' capability of performing
their intended function because CPS will continue to perform
inspections, testing and maintenance that supports breaker operation as
intended and provides assurance that breaker interruption time will be
within limits. Thus, the EOC-RPT breaker trip may be expected to
operate as before to mitigate pressurization transient effects.
The EOC-RPT breaker trip is also assumed to occur in the analyses
for the loss of feedwater heating, feedwater controller failure,
pressure regulator failure, recirculation flow control failure, and
recirculation pump seizure events. However, the EOC-RPT breaker trip is
not an initiator or mitigating feature for these events. The proposed
change cannot therefore impact the probability or consequences for
these events. Nonetheless, the EOC-RPT breaker trip may be assumed to
function as before for these scenarios.
For scenarios where the EOC-RPT breaker trip could initiate an
event (i.e., inadvertent recirculation pump trip events), the
probability of occurrence is not increased. The design and operation of
the EOC-RPT system has not been changed, and therefore, the
consequences resulting from the EOC-RPT breaker trip are unchanged.
Based on the above, neither the probability nor the consequences of
any accident previously evaluated have been increased.
(2) As noted above, the EOC-RPT breakers will continue to function
as before. The proposed change involves no design change or physical
change in the plant. Therefore, previous accident analyses are
unchanged. Further, no new operations or testing is involved. On this
basis, no new failure modes are introduced. Therefore, this proposed
change does not create the possibility of a new or different kind of
accident from any accident previously evaluated.
(3) This proposed change does not involve a significant reduction
in a margin of safety. The capability of the EOC-RPT breaker trip to
provide additional insertion of negative reactivity for mitigating
design-basis events remains unchanged. That is, the EOC-RPT will
continue to be capable of reducing the peak reactor pressure and power
resulting from turbine trip or generator load rejection transients,
thus providing additional margin to core thermal MCPR Safety Limits.
The margin of safety is assured by the EOC-RPT breaker trip
occurring within established limits such that the overall system
performs its intended safety function within the time analyzed for the
system safety response. No system time limit change is proposed. The
robust design of the breakers, combined with continued performance of
vendor-recommended testing and maintenance that ensures proper
mechanical and electrical performance of the breakers, will continue to
provide assurance that breaker interruption time is within the
acceptable limit. Therefore, there is no significant reduction in the
margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: Vespasian Warner Public
Library, 120 West Johnson Street, Clinton, Illinois 61727.
Attorney for licensee: Leah Manning Stetener, Vice President,
General Counsel, and Corporate Secretary, 500 South 27th Street,
Decatur, Illinois 62525.
NRC Project Director: Gail H. Marcus.
Illinois Power Company and Soyland Power Cooperative, Inc., Docket No.
50-461, Clinton Power Station, Unit No. 1, DeWitt County, Illinois
Date of amendment request: February 22, 1996 (U-602549).
Description of amendment request: The proposed amendment would
revise Technical Specification (TS) 3.6.5.1, ``Drywell,'' to allow
drywell bypass leakage tests to be performed at intervals of up to ten
years based, in part, on the demonstrated performance of the drywell
barrier with respect to leak tightness. The proposed amendment would
also revise TS 3.6.5.2, ``Drywell Air Lock,'' to extend the testing
intervals for the surveillances on drywell air lock overall leakage and
interlock operability, relocate the specific leakage limits on the air
lock barrel and door seals to the TS Bases, relocate the requirement to
pressurize the drywell air lock to 19.7 psid prior to performance of
the overall drywell air lock leakage test to the TS Bases, and other
administrative changes.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration which is presented below:
(1) The proposed changes do not involve a change to the plant
design or operation. As a result, the proposed changes do not affect
any of the parameters or conditions that contribute to initiation of
any accidents previously evaluated. Therefore, the proposed changes
cannot increase the probability of any accidents previously evaluated.
The proposed changes do potentially affect the leaktight integrity
of the drywell, a structure used to mitigate the consequences of a loss
of coolant accident (LOCA). The function of the drywell is to force the
steam released from a LOCA through the suppression pool, limiting the
amount of steam released to the primary containment atmosphere. This
serves to limit the containment pressurization due to the LOCA. The
leakage of the drywell is limited to ensure that the primary
containment does not exceed its design limits of 185 deg.F and 15 psig.
Because the proposed change to replace the current 18-month frequency
for performing drywell bypass leakage tests (DBLRTs) with a
performance-based frequency does not alter the plant design, the
proposed change does not directly result in an increase in the drywell
leakage. However, decreasing the test frequency can increase the
probability that a large increase in drywell bypass leakage could go
undetected for an extended period of time. This potential has been
evaluated, and Illinois Power has determined that the proposed change
to the DBLRT frequency will not result in the potential for undetected,
large increases in leakage, as further discussed below.
There are several potential drywell bypass leakage paths. These
include potential cracks in drywell concrete structure, the drywell
vacuum breakers, and various penetrations through the drywell
structure. Based on the results of the structural integrity test
conducted at the design pressure of 30 psig as part of the
preoperational test program, additional cracking of the drywell is not
expected during the remaining life of the plant. Ventilation and piping
penetrations (including the drywell vacuum breaker penetrations) are
designed to ASME Code Class 2 and Seismic Category 1 requirements.
These penetrations are typically designed with two isolation valves in
series with one valve in the drywell and another either outside primary
containment or in the wetwell. Technical Specification (TS)
Surveillance Requirements (SRs) require, as applicable, periodic
verification of drywell isolation valve
[[Page 18171]]
position, stroke time, and automatic isolation capability. High energy
lines that extend into the wetwell, such as the main steam lines and
feedwater lines, are encapsulated by guard pipes to direct energy back
into the drywell in case of a piping rupture. Electrical penetrations
are sealed with a high strength/density material that will prevent
leakage, as well as provide radiation shielding.
The proposed changes for the drywell air lock involve relocation of
the separate limits on the drywell air lock barrel and seal leakage
rates to the TS Bases, relocation of the requirement to pressurize the
air lock to 19.7 psid prior to performance of the air lock overall
(barrel) leakage test, and changing the frequency for these tests from
18 months to 24 months. While the proposed changes will eliminate
separate TS limits on leakage of the drywell air lock, the overall
drywell bypass leakage TS limit (which includes leakage through the air
lock) is not affected by this proposed change. The limiting scenario
for drywell bypass leakage is a small break LOCA which results in
drywell pressures of approximately 3 psid. Only a large break LOCA can
create drywell pressures of 19.7 psid. For this event, the allowable
drywell bypass leakage rate is over eight times larger than for a small
break LOCA. Thus, relocation of these requirements to the TS Bases will
continue to provide adequate control of these requirements. The
proposed air lock overall leakage rate testing frequency is consistent
with the guidance for testing primary containment air locks in Nuclear
Energy Institute (NEI) 94-01, ``Industry Guideline for Implementing
Performance-Based Option of 10CFR50, Appendix J.'' The drywell air lock
is tested in a manner similar to the primary containment air locks,
even though the drywell air lock is not a direct leakage path from
primary containment and, therefore, 10CFR50, Appendix J test
requirements do not apply. The drywell air lock's use is limited during
plant operation due to radiation and temperature in the drywell. Since
sufficient confidence in the door's sealing capability is assured
considering past performance and the air lock door usage is very low
throughout an operating cycle, it is justified to allow performance of
these tests at refueling-outage intervals, whether the unit is on a 18-
month or a 24-month refueling cycle.
Operational experience has shown that the leak tightness of the
drywell has been maintained well below the allowable leakage limits at
Clinton Power Station. The TS limit of 10% of the design [maximum
allowable leakage path area] provides a large margin for degradation.
Drywell performance to date suggests that drywell degradation, even
with a ten-year interval between tests, will not exceed this margin.
The most recent DBLRT performed during the fourth refueling outage (RF-
4) measured a drywell bypass leakage rate of 0.07% of the design limit.
An analysis was also conducted to determine the potential risk to
the public from unacceptable drywell bypass leakage going undetected as
a result of the proposed change. Based on this probabilistic risk
analysis, for several different accident scenarios, the risk of
radioactivity release from containment was found to be insignificant.
Based on the above, Illinois Power has concluded that the proposed
changes will not result in a significant increase in the consequences
of any accident previously evaluated.
(2) The proposed change does not involve a change to the plant
design or operation. As a result, the proposed change does not affect
any of the parameters or conditions that could contribute to initiation
of any accidents. Drywell bypass leakage cannot, of itself, create an
accident. Thus, it has been concluded that the proposed change cannot
create the possibility of an accident not previously evaluated.
(3) The NRC has provided standards for determining whether a no
significant hazards consideration exists as stated in 10CFR50.92(c).
These proposed changes involve the withdrawal of operating restrictions
previously imposed because acceptable operation of the Mark III primary
containment design had not been demonstrated at the time of initial
licensing. As published in the Federal Register (FR) regarding no
significant hazards consideration criteria, granting of a relief based
upon demonstration of acceptable operation from an operating
restriction that was imposed because acceptable operation had not yet
been demonstrated does not involve a significant hazards consideration
(reference 48 FR 14870).
The proposed change only affects the frequency of measuring the
drywell bypass leakage rate and does not change the bypass leakage rate
limit. The proposed change could potentially increase the probability
that a large increase in drywell bypass leakage could go undetected for
an extended period of time. However, operational experience has shown
that the leaktightness of the drywell has been maintained well below
the allowable leakage limits. In addition, there are TS surveillances
which require, as applicable, periodic verification of drywell
isolation valve position, stroke time, and automatic isolation
capability. Further, qualitative methods (such as periodic verification
that the drywell pressurizes, which ensures that the drywell leak rate
is less than the instrument air leak and usage rates) are available to
provide assurance that the drywell leakage rate is being maintained
within limits. The Clinton Power Station TS require the drywell leakage
rate measured during DBLRTs to be less than or equal to 10% of the
design limit. This request does not affect this required margin. Nor
does it affect the existing margin between the primary containment
design pressure and the actual pressure at which primary containment
would fail.
With respect to proposed changes to the drywell air lock overall
leakage testing and interlock testing requirements, the proposed leak
test frequencies are consistent with the guidance for testing primary
containment air locks in NEI 94-01. Due to the limited use of the
drywell air locks during plant operation, it is justified to allow
performance of interlock operability testing on a refueling outage
basis, whether the unit is on an 18-month or a 24-month refueling
cycle. The separate limits on the drywell air lock and barrel are being
relocated from the TS, these limits are being controlled under
10CFR50.59 and the TS Bases Control program of TS 5.5.11. Leakage
through these pathways will continue to be a part of the overall
drywell bypass leakage limited by LCO 3.6.5.1.
An analysis was also conducted to determine the potential risk to
the public from the proposed change. Based on this probabilistic risk
analysis, for several different accident scenarios, the risk of
radioactivity release from containment was found to be insignificant.
As a result, Illinois Power has concluded that the proposed changes
will continue to assure that the drywell bypass leakage will be within
design limits if challenged and therefore, will 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 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: Vespasian Warner Public
Library, 120 West Johnson Street, Clinton, Illinois 61727.
[[Page 18172]]
Attorney for licensee: Leah Manning Stetener, Vice President,
General Counsel, and Corporate Secretary, 500 South 27th Street,
Decatur, Illinois 62525.
NRC Project Director: Gail H. Marcus.
Indiana Michigan Power Company, Docket Nos. 50-315 and 50-316, Donald
C. Cook Nuclear Plant, Unit Nos. 1 and 2, Berrien County, Michigan
Date of amendment requests: February 26, 1996 (AEP:NRC:1071U).
Description of amendment requests: The proposed amendments would
modify the technical specifications (TS) to increase the current limit
on nominal fuel assembly enrichment for new, Westinghouse-fabricated,
fuel stored in the new fuel storage racks from 4.55 weight percent
uranium-235 isotope to 4.95 weight percent uranium-235 isotope with
certain provisions. Also, TS 5.6.2 would be reformatted similar to that
used in the standard TS (NUREG-1431, Rev. 1).
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
Per 10 CFR 50.92, a proposed amendment will not involve a
significant hazards consideration if the proposed amendment does not:
(1) involve a significant increase in the probability or
consequences of an accident previously evaluated,
(2) create the possibility of a new or different kind of accident
from any accident previously evaluated, or
(3) involve a significant reduction in a margin of safety.
Criterion 1
The proposed changes will not involve a significant increase in the
probability of an accident previously evaluated because similar
administrative controls to those presently used to identify new fuel
storage rack inventory and compliance with T/S limits will be used.
There are no physical changes to the plant associated with this T/S
change. The consequences of an accident previously evaluated will not
be increased because the reactivity of the fuel stored in the new fuel
storage racks under the proposed T/S limits will be no greater than the
reactivity of fuel stored in the new fuel storage racks presently
allowable under the current T/S limits.
Criterion 2
The proposed changes will not create the possibility of a new or
different kind of accident from any accident previously evaluated
because the changes will involve no physical changes to the plant nor
any changes in plant operations. Furthermore, the reactivity of the
fuel stored in the new fuel storage racks under the proposed T/S limits
will be no greater than the reactivity of fuel stored in the new fuel
storage rack presently allowable under the current T/S limits.
Criterion 3
The proposed amendment will not involve a significant reduction in
a margin of safety because the reactivity of the fuel stored in the new
fuel storage racks under the proposed T/S limits will be no greater
than the reactivity of fuel stored in the new fuel storage racks
presently allowable under the current T/S limits.
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. In addition, the reformatting of TS 5.6.2 is a purely
administratiave change having no effect on the physical plant or its
operation. Therefore, the NRC staff proposes to determine that the
amendment requests involve no significant hazards consideration.
Local Public Document Room location: Maud Preston Palenske Memorial
Library, 500 Market Street, St. Joseph, Michigan 49085.
Attorney for licensee: Gerald Charnoff, Esq., Shaw, Pittman, Potts
and Trowbridge, 2300 N Street, NW, Washington, DC 20037.
NRC Project Director: Mark Reinhart, Acting.
Indiana Michigan Power Company, Docket Nos. 50-315 and 50-316, Donald
C. Cook Nuclear Plant, Unit Nos. 1 and 2, Berrien County, Michigan
Date of amendment requests: February 29, 1996 (AEP:NRC:1232).
Description of amendment requests: The proposed amendments would
revise the technical specifications to reduce the boric acid
concentration in the boric acid storage system from approximately 12
percent to approximately 4 percent by weight. Related changes are also
proposed to increase the minimum required flow rate in action
statements for certain affected TS and add an additional surveillance
requirement for this flow rate, and decrease the minimum temperature
requirement in certain affected TS to 63 deg.F. The bases section is
also updated to reflect these proposed changes.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
Per 10 CFR 50.92, a proposed change does not involve a significant
hazards consideration if the change does not:
1. involve a significant increase in the probability or
consequences of an accident previously evaluated,
2. create the possibility of a new or different kind of accident
from any accident previously evaluated, and
3. involve a significant reduction in a margin of safety.
Criterion 1
Does the change involve a significant increase in the probability
or consequences of an accident previously evaluated?
NO. The BAST [boric acid storage tank] water volume and boron
concentration were not credited in any Chapter 14 safety analysis.
Therefore, no change in the probabilities of the accident analysis will
result from the BAST water volume and boron concentration change. In
addition, since the BAST water volume and boron concentration are not
taken into consideration in any safety analysis, the consequences of an
accident previously evaluated in the FSAR [final safety analysis
report] are not increased. The heat tracing system is currently only
necessary to prevent precipitation of existing high boric acid
concentration in the plant systems. The reduction in boron
concentration in this proposal eliminates the need for the heat tracing
system. The existence of the heat tracing system was not part of any
safety analysis and disabling of the heat tracing system will not
result in a significant increase in the probability or consequences of
an accident previously evaluated.
Criterion 2
Does the change create the possibility of a new or different kind
of accident from any accident previously evaluated?
NO. Since the minimum required water flow from the boric acid
storage system to the reactor coolant system was increased to
counteract any possible operational transients, as shown in Attachment
4 [of the application], the change in BAST water volume and boron
concentration and disabling of the heat tracing system do not create
the possibility of an accident which is different from any already
evaluated in the FSAR. No new or different failure modes have been
defined for any system or component nor has any new limiting single
failure been identified.
[[Page 18173]]
Criterion 3
Does the change involve a significant reduction in a margin of
safety?
NO. The margin of safety requirements are not affected by the
removal of the heat tracing system and the reduction of the boric acid
concentration in the boric acid storage system. The required flow paths
and borated water sources are unaffected by this proposal. The required
quantity of borated water is still available based upon the performed
evaluation, and appropriate surveillance requirements ensure the
ability to deliver this borated water. The reduction of the boric acid
concentration in the BASTs will ensure that the boric acid remains in
solution at the normal room temperature in the auxiliary building. With
the above changes, there will be a net improvement in system
reliability and accordingly the proposed changes do not affect 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 requests involve no significant hazards consideration.
Local Public Document Room location: Maud Preston Palenske Memorial
Library, 500 Market Street, St. Joseph, Michigan 49085.
Attorney for licensee: Gerald Charnoff, Esq., Shaw, Pittman, Potts
and Trowbridge, 2300 N Street, NW, Washington, DC 20037.
NRC Project Director: Mark Reinhart, Acting.
Pacific Gas and Electric Company, Docket Nos. 50-275 and 50-323, Diablo
Canyon Nuclear Power Plant, Unit Nos. 1 and 2, San Luis Obispo County,
California
Date of amendment requests: March 13, 1996.
Description of amendment requests: The proposed amendments would
revise the combined Technical Specifications (TS) for the Diablo Canyon
Power Plant, Unit Nos. 1 and 2 to revise TS 4.0.5, ``Surveillance
Requirements,'' to delete reference to prior NRC approval for written
relief from the Inservice Inspection (ISI) and Inservice Testing
Program (IST) requirements and to add ASME Section XI definition of
``Biennially or every 2 years--At least once per 731 days'' in TS
4.0.5b.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. The proposed changes do not involve a significant increase in
the probability or consequences of an accident previously evaluated.
The proposed changes implement the NRC's recommendation contained
in NUREG-1482, ``Guidelines for Inservice Testing Programs at Nuclear
Power Plants,'' endorsed by Generic Letter
89-04, Supplement 1, ``Guidance on Developing Acceptable Inservice
Testing Programs.'' The changes are consistent with 10 CFR 50.55a,
``Codes and Standards,'' which does not prohibit the implementation of
relief from ASME Section XI requirements prior to specific written
approval when those changes are found acceptable by change process
specified in 10 CFR 50.59, ``Changes, Tests and Experiments.'' The
proposed changes are administrative in nature and do not involve any
modifications to any plant equipment or affect plant operation.
Therefore, the proposed changes do not involve a significant
increase in the probability or consequences of an accident previously
evaluated.
2. The proposed changes do not create the possibility of a new or
different kind of accident from any accident previously evaluated.
The proposed changes are administrative in nature, do not involve
any physical alterations to any plant equipment, and cause no change in
the method by which any safety-related system performs its function.
Therefore, the proposed changes do not create the possibility of a
new or different kind of accident from any accident previously
evaluated.
3. The proposed changes do not involve a significant reduction in a
margin of safety.
The proposed changes do not alter the basic regulatory requirements
and do not affect any safety analyses.
Therefore, the proposed changes do not involve a significant
reduction in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment requests involve no significant hazards consideration.
Local Public Document Room location: California Polytechnic State
University, Robert E. Kennedy Library, Government Documents and Maps
Department, San Luis Obispo, California 93407.
Attorney for licensee: Christopher J. Warner, Esq., Pacific Gas and
Electric Company, P.O. Box 7442, San Francisco, California 94120.
NRC Project Director: William H. Bateman.
Pacific Gas and Electric Company, Docket Nos. 50-275 and 50-323, Diablo
Canyon Nuclear Power Plant, Unit Nos. 1 and 2, San Luis Obispo County,
California
Date of amendment request: April 3, 1996.
Description of amendment request: The proposed amendments would
revise the combined Technical Specifications (TS) for the Diablo Canyon
Power Plant, Unit Nos. 1 and 2 to revise Technical Specifications 3/
4.7.5, ``Control Room Ventilation System,'' 3/4.7.6, ``Auxiliary
Building Safeguards Air Filtration System,'' and 3/4.9.12, ``Fuel
Handling Building Ventilation System,'' to clarify the testing
methodology utilized by PG&E to determine the operability of the
charcoal and high-efficiency particulate air (HEPA) filters in the
engineering safeguards features (ESF) air handling units.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. The proposed change does not involve a significant increase in
the probability or consequences of an accident previously evaluated.
The charcoal testing protocol changes will not affect system
operation or performance, nor do they affect the probability of any
event initiators. These changes do not affect any engineered safety
features actuation setpoints or accident mitigation capabilities. The
new charcoal adsorber sample laboratory testing protocol more
accurately demonstrates the required performance of the adsorbers in
the control room ventilation system and auxiliary building safeguards
air filtration system following a design basis loss of coolant accident
or in the fuel handling building ventilation system following a fuel
handling accident outside containment. The decontamination efficiencies
used in the offsite and control room dose analyses are not affected by
these changes. Therefore, offsite and control room dose analyses are
not affected by this change, and all offsite and control room doses
will remain within the limits of 10 CFR 100 and 10 CFR 50, Appendix A,
General Design Criterion (GDC) 19.
[[Page 18174]]
The requirements of ANSI N510-1980 encompass the requirements of
ANSI N510-1975, which is referenced in Regulatory Guide (RG) 1.52, as
it applies to testing at Diablo Canyon Power Plant (DCPP).
Consequently, revising the Technical Specifications (TS) to reference
ANSI N510-1980 will have no effect on filter testing.
The proposed changes are consistent with the new Standard Technical
Specifications (NUREG-1431, Rev. 1).
Therefore, the proposed changes do not involve a significant
increase in the probability or consequences of an accident previously
evaluated.
2. The proposed change does not create the possibility of a new or
different kind of accident from any accident previously evaluated.
The changes to the charcoal sample testing protocol will not affect
the method of operation of the system. The proposed changes only affect
the testing parameters for the charcoal samples. No new or different
accident scenarios, transient precursors, failure mechanisms, or
limiting single failures will be introduced as a result of these
changes.
The requirements of ANSI N510-1980 encompass the requirements of
ANSI N510-1975, which is referenced in RG 1.52, as it applies to
testing at DCPP. Consequently, revising the TSs to reference ANSI N510-
1980 will have no effect on filter testing.
Therefore, the proposed changes do not create the possibility of a
new or different kind of accident from any accident previously
evaluated.
3. The proposed change does not involve a significant reduction in
a margin of safety.
The changes in charcoal sample testing protocol will not affect
system performance or operation. The decontamination efficiencies used
in the offsite and control room dose analyses are not affected by these
changes. Therefore, offsite and control room dose analyses are not
affected by this change, and all offsite and control room doses will
remain within the limits of 10 CFR 100 and 10 CFR 50, Appendix A, GDC
19.
The requirements of ANSI N510-1980 encompass the requirements of
ANSI N510-1975, which is referenced in RG 1.52, as it applies to
testing at DCPP. Consequently, revising the TSs to reference ANSI N510-
1980 will have no effect on filter testing.
Therefore, the proposed changes do not involve a significant
reduction in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment requests involve no significant hazards consideration.
Local Public Document Room location: California Polytechnic State
University, Robert E. Kennedy Library, Government Documents and Maps
Department, San Luis Obispo, California 93407.
Attorney for licensee: Christopher J. Warner, Esq., Pacific Gas and
Electric Company, P.O. Box 7442, San Francisco, California 94120.
NRC Project Director: William H. Bateman.
Pacific Gas and Electric Company, Docket No. 50-133, Humboldt Bay Power
Plant, Unit 3, Humboldt County, California
Date of amendment request: March 13, 1996.
Description of amendment request: The proposed amendment would
revise the Humboldt Bay Power Plant (HBPP), Unit 3, Technical
Specifications (TS) by incorporating position changes to reflect a
proposed plant staff reorganization. The TS changes proposed are as
follows:
(1) TS Section VII.C.2.c and VII.D.1.b--change the position title
from ``Power Plant Engineer'' to ``Senior Power Production Engineer.''
(2) TS Section VII.C.2.d--change the position title from ``Senior
Chemical and Radiological Engineer'' to ``Senior Radiation Protection
Engineer.''
(3) TS Section VII.C.2.e and VII.D.1.b--change the position title
from ``Maintenance Planner'' to ``Supervisor of Maintenance.''
(4) TS Section VII.C.2.g and VII.D.1.b--add the position of
``Assistant Plant Manager/Power Plant Engineer.''
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. Does the change involve a significant increase in the
probability or consequences of an accident previously evaluated?
The proposed administrative and organizational changes provide
editorial corrections and reflect the proposed HBPP and current NRC
organizations. These changes do not affect the operating methodology of
HBPP, and they are not related to the probability or consequences of an
accident previously evaluated.
Therefore, the proposed changes do not involve a significant
increase in the probability or consequences of an accident previously
evaluated.
2. Does the change create the possibility of a new or different
kind of accident from any accident previously evaluated?
The proposed revisions to the HBPP TS are organizational and
administrative in nature, and do not change the method by which any
safety-related system performs its function.
Therefore, the proposed changes do not create the possibility of a
new of different kind of accident from any accident previously
evaluated.
3. Does the change involve a significant reduction in a margin of
safety?
The proposed changes have no effect on the current operating
methodologies or actions that govern plant performance. In addition,
the proposed changes do not affect the margin of safety associated with
parameters for any accident analysis.
Therefore, the proposed changes do not involve a significant
reduction in a margin of safety.
The NRC staff has reviewed the analysis of the licensee 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: Humboldt County Library, 1313
3rd Street, Eureka, California 95501.
Attorney for licensee: Christopher J. Warner, Esquire, Pacific Gas
& Electric Company, P.O. Box 7442, San Francisco, California 94120.
NRC Project Director: Seymour H. Weiss.
Union Electric Company, Docket No. 50-483, Callaway Plant, Unit 1,
Callaway County, Missouri
Date of application request: February 9, 1996, as superseded by
letter dated March 22, 1996.
Description of amendment request: The amendment would revise
Technical Specification (TS) Definition 1.7, TS 3/4.6, TS 6.8, and
their associated bases to directly reference Regulatory Guide 1.163 as
required by 10 CFR 50, Appendix J, Option B, for the Type A containment
integrated leak rate tests (ILRTs) and the Type B and C local leak rate
tests (LLRTs).
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
[[Page 18175]]
consideration, which is presented below:
1. The proposed change does not involve a significant increase in
the probability or consequences of an accident previously evaluated.
The proposed changes to TS 1.7e, 4.6.1.1, 3/4.6.1.3, Bases 3/
4.6.1.1 and the program addition to TS 6.8.4g have no effect on plant
operation. The proposed changes only provide mechanisms within TS for
implementing a performance-based methodology for determining the
frequency of leak rate testing, as allowed by the NRC. The test type,
method, and acceptance criteria will not be changed. Containment
leakage will continue to be maintained within the required limits.
Based on industry and NRC evaluations performed in support of
developing Option B, these changes potentially result in a minor
increase in the consequences of an accident previously evaluated due to
the increased testing intervals. However, the proposed changes do not
result in an increase in the core damage frequency since the
containment system is used for mitigation purposes only.
Directly referencing the Containment Leakage Rate Testing Program
for Containment ILRT and LLRT requirements does not involve any
modification to plant equipment or affect the operation or design basis
of the containment. Leakage rate testing is not a precursor to or an
initiating event for any accident.
Therefore, these changes do not involve a significant increase in
the probability or consequences of an accident previously evaluated.
2. The proposed change does not create the possibility of a new or
different kind of accident from any accident previously evaluated.
The proposed changes only allow for implementation of 10 CFR 50,
Appendix J, Option B and do not involve any modifications to any plant
equipment or affect the operation or design basis of the containment.
The proposed changes do not affect the response of the containment
during a design basis accident.
3. The proposed change does not involve a significant reduction in
a margin of safety.
The proposed changes do not affect or change a safety limit, any
limiting condition for operation or affect plant operations. The
changes only implement the Appendix J, Option B test frequencies that
have been determined by NRC not to involve a safety concern. The
testing methods, acceptance criteria and bases are not changed and
still provide assurance that the containment will provide its intended
function.
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: Callaway County Public
Library, 710 Court Street, Fulton, Missouri 65251.
Attorney for licensee: Gerald Charnoff, Esq., Shaw, Pittman, Potts
& Trowbridge, 2300 N Street, N.W., Washington, D.C. 20037.
NRC Project Director: William H. Bateman.
Virginia Electric and Power Company, Docket Nos. 50-338 and 50-339,
North Anna Power Station, Units No. 1 and No. 2, Louisa County,
Virginia
Date of amendment request: March 21, 1996.
Description of amendment request: The proposed changes to the
Technical Specifications (TS) for the North Anna Power Station, Units
1&2 (NA-1&2) would clarify the requirements for testing charcoal
adsorbent in the Waste Gas Charcoal Filter System, the Control Room
Emergency Habitability System, and the Safeguards Area Ventilation
System. No change in the testing is being proposed, only clarification
of the description of the required testing in TS 3/4.6.4.3, 3/4.7.7.1,
and 3/4.7.8.1.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
The proposed Technical Specifications changes will revise
Surveillance Requirements for the charcoal adsorbent in the Waste Gas
Charcoal Filter System (TS 3/4.6.3.), Control Room Emergency
Habitability System (TS 3/4.7.7.2), and the Safeguards Area Ventilation
System (TS 3/4.7.8.1) to reflect the current testing methodology for
new and used carbon adsorbent. These proposed changes specify ASTM D
3803-1979 as the laboratory testing standard for both new and used
charcoal adsorbent for the ventilation system identified above.
Virginia Electric and Power has evaluated the proposed Technical
Specification changes to the North Anna Units 1 and 2 Technical
Specifications against the Significant Hazards Criteria of 10 CFR 50.92
and determined that the changes do not involve any significant hazard
for the following reasons:
1. The probability or consequences of an accident previously
evaluated is not significantly increased.
The proposed changes are administrative in nature in that the
changes only explicitly specify the current testing methodology for
charcoal adsorbent. The proposed changes will not affect system
operation or performance, nor do they affect the probability of any
event initiators. These changes do not affect any Engineered Safety
Features actuation setpoints or accident mitigation capabilities.
Therefore, the proposed changes will not significantly increase the
consequences of an accident or malfunction of equipment important to
safety previously evaluated in the UFSAR.
2. The possibility of an accident or a malfunction of a different
type than any previously evaluated is not created.
The proposed changes only clarify the requirements for charcoal
testing and will not affect the method of operation of the ventilation
systems. Furthermore, the proposed changes are only intended to clarify
the existing requirements to explicitly specify the current test
methodology. No new or different accident scenarios, transient
precursors, failure mechanisms, or limiting single failures will be
introduced as a result of these changes. Therefore, the possibility of
a new or different kind of accident other than those already evaluated
will not be created by this change.
3. The margin of safety has not been significantly reduced.
The proposed changes which represent the current laboratory testing
methodology for charcoal adsorber samples, demonstrates the required
performance of the adsorbent following a design basis LOCA or Fuel
Handling Accident. Changing the Technical Specification to clarify the
methodology for charcoal sample testing will not affect system
performance or operation.
Therefore, these changes will not result in a significant reduction
in any margin of safety.
Based on the above discussions, it has been determined that the
requested Technical Specification changes do not involve a significant
increase in the probability or consequences of an accident or other
adverse condition over previous evaluations; or create the possibility
of a new or different kind of accident or condition over previous
evaluation; or involve a significant reduction in a margin of safety.
Therefore, the requested license amendment does not involve a
significant hazards consideration.
[[Page 18176]]
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: Michael W. Maupin, Esq., Hunton and
Williams, Riverfront Plaza, East Tower, 951 E. Byrd Street, Richmond,
Virginia 23219
NRC Project Director: Eugene V. Imbro.
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.
Florida Power Corporation, et al., Docket No. 50-302, Crystal River
Nuclear Generating Plant, Unit No. 3, Citrus County, Florida
Date of amendment request: March 21, 1996.
Brief description of amendments: The amendments provide changes to
Technical Specifications (TS) for CR3 relating to the Once Through
Steam Generator's (OTSG's) tube inspection acceptance criteria, and
repair limit for removing steam generator tubes from service. The
proposed TS change would be applicable for one cycle duration, and only
to Inter-Granular-Attack (IGA) degradation mechanism in a limited
region of the OTSG.
Date of publication of individual notice in Federal Register: March
28, 1996 (61 FR 13888)
Expiration date of individual notice: April 29, 1996.
Local Public Document Room location: Coastal Region Library, 8619
W. Crystal Street, Crystal River, Florida 32629.
Houston Lighting & Power Company, City Public Service Board of San
Antonio, Central Power and Light Company, City of Austin, Texas, Docket
Nos. 50-498 and 50-499, South Texas Project, Units 1 and 2, Matagorda
County, Texas
Date of amendment request: May 30, 1995, as supplemented by letter
dated February 8, 1996.
Description of amendment request: The proposed amendment would
increase the spent fuel pool heat load licensing basis to provide
greater flexibility for normal refueling practices.
Date of individual notice in the Federal Register: April 3, 1996
(61 FR 14832)
Expiration date of individual notice: May 3, 1996.
Local Public Document Room location: Wharton County Junior College,
J. M. Hodges Learning Center, 911 Boling Highway, Wharton, TX 77488.
Notice of Issuance of Amendments to Facility Operating Licenses
During the period since publication of the last biweekly notice,
the Commission has issued the following amendments. The Commission has
determined for each of these amendments that the application complies
with the standards and requirements of the Atomic Energy Act of 1954,
as amended (the Act), and the Commission's rules and regulations. The
Commission has made appropriate findings as required by the Act and the
Commission's rules and regulations in 10 CFR Chapter I, which are set
forth in the license amendment.
Notice of Consideration of Issuance of Amendment to Facility
Operating License, Proposed No Significant Hazards Consideration
Determination, and Opportunity for A Hearing in connection with these
actions was published in the Federal Register as indicated.
Unless otherwise indicated, the Commission has determined that
these amendments satisfy the criteria for categorical exclusion in
accordance with 10 CFR 51.22. Therefore, pursuant to 10 CFR 51.22(b),
no environmental impact statement or environmental assessment need be
prepared for these amendments. If the Commission has prepared an
environmental assessment under the special circumstances provision in
10 CFR 51.12(b) and has made a determination based on that assessment,
it is so indicated.
For further details with respect to the action see (1) the
applications for amendment, (2) the amendment, and (3) the Commission's
related letter, Safety Evaluation and/or Environmental Assessment as
indicated. All of these items are available for public inspection at
the Commission's Public Document Room, the Gelman Building, 2120 L
Street, NW., Washington, DC, and at the local public document rooms for
the particular facilities involved.
Arizona Public Service Company, et al., Docket Nos. STN 50-528, STN 50-
529, and STN 50-530, Palo Verde Nuclear Generating Station, Units 1, 2,
and 3, Maricopa County, Arizona
Date of application for amendments: December 20, 1995.
Brief description of amendments: These amendments change the
instrument setpoint for the reactor trip and main steam isolation
signal actuation on low steam generator pressure from greater than or
equal to 919 psia with an allowable value of 911 psia to 895 psia with
an allowable value of greater than or equal to 890 psia.
Date of issuance: April 5, 1996.
Effective date: April 5, 1996, to be implemented within 45 days of
issuance.
Amendment Nos.: Unit 1-105; Unit 2-97; Unit 3-77.
Facility Operating License Nos. NPF-41, NPF-51, and NPF-74: The
amendments revised the Technical Specifications.
Date of initial notice in Federal Register: February 28, 1996 (61
FR 7544) The Commission's related evaluation of the amendments is
contained in a Safety Evaluation dated April 5, 1996.
No significant hazards consideration comments received: No.
Local Public Document Room location: Phoenix Public Library, 1221
N. Central Avenue, Phoenix, Arizona 85004.
Baltimore Gas and Electric Company, Docket Nos. 50-317 and 50-318,
Calvert Cliffs Nuclear Power Plant, Unit Nos. 1 and 2, Calvert County,
Maryland
Date of application for amendments: November 1, 1995 as
supplemented on December 1, 1995.
Brief description of amendments: The amendments reflect the new
plant electrical distribution configuration, surveillance and limiting
condition for operation of the new safety-related (SR) emergency diesel
generator (EDG), the increased electrical capacities for the two of the
three existing SR EDGs, the increased EDG fuel oil storage capacity,
and the fire protection system for the
[[Page 18177]]
new EDG building. The remaining existing SR EDG will be upgraded during
the Unit No. 2 refueling outage scheduled for the spring of 1997.
Date of issuance: April 2, 1996.
Effective date: As of the date of issuance to be implemented within
30 days.
Amendment Nos.: 214 and 191.
Facility Operating License Nos. DPR-53 and DPR-69: Amendments
revised the Technical Specifications.
Date of initial notice in Federal Register: January 3, 1996 (61 FR
175) The Commission's related evaluation of these amendments is
contained in a Safety Evaluation dated April 2, 1996.
No significant hazards consideration comments received: No.
Local Public Document Room location: Calvert County Library, Prince
Frederick, Maryland 20678.
Commonwealth Edison Company, Docket Nos. STN 50-454 and STN 50-455,
Byron Station, Unit Nos. 1 and 2, Ogle County, Illinois
Docket Nos. STN 50-456 and STN 50-457, Braidwood Station, Unit Nos. 1
and 2, Will County, Illinois
Date of application for amendments: December 6, 1995, as
supplemented February 27, 1996, and March 28, 1996.
Brief description of amendments: The amendments modify the
technical specifications to replace the existing scheduling
requirements for overall integrated and local containment leakage rate
testing with a requirement to perform the testing in accordance with 10
CFR Part 50, Appendix J, Option B. Option B allows test scheduling to
be adjusted based on past performance.
Date of issuance: April 4, 1996.
Effective date: April 4, 1996.
Amendment Nos.: 81, 81, 73, and 73.
Facility Operating License Nos. NPF-37, NPF-66, NPF-72 and NPF-77:
The amendments revised the Technical Specifications.
Date of initial notice in Federal Register: February 28, 1996 (61
FR 7547) The February 27, 1996, and March 28, 1996, supplements
modified the Technical Specification pages to be more consistent with
the published guidance, were within this scope of the initial notice,
and did not affect the initial proposed no significant hazards
consideration determination. The Commission's related evaluation of the
amendments is contained in a Safety Evaluation dated April 4, 1996.
No significant hazards consideration comments received: No.
Local Public Document Room location: For Byron, the Byron Public
Library District, 109 N. Franklin, P.O. Box 434, Byron, Illinois 61010;
for Braidwood, the Wilmington Public Library, 201 S. Kankakee Street,
Wilmington, Illinois 60481.
Commonwealth Edison Company, Docket Nos. STN 50-454 and STN 50-455,
Byron Station, Unit Nos. 1 and 2, Ogle County, Illinois
Docket Nos. STN 50-456 and STN 50-457, Braidwood Station, Unit Nos. 1
and 2, Will County, Illinois
Date of application for amendments: October 3, 1995, as
supplemented on February 21, 1996, and April 2, 1996.
Brief description of amendments: The amendments revise the
Technical Specifications (TS) to implement ten of the line-item TS
improvements recommended in Generic Letter (GL) 93-05, ``Line-Item
Technical Specifications Improvements to Reduce Surveillance
Requirements for Testing During Power Operation,'' dated September 27,
1993. The amendments also include editorial changes on the affected TS
pages.
Date of issuance: April 10, 1996.
Effective date: April 10, 1996.
Amendment Nos.: 82, 82 and 74, 74.
Facility Operating License Nos. NPF-37, NPF-66, NPF-72 and NPF-77:
The amendments revised the Technical Specifications.
Date of initial notice in Federal Register: November 27, 1995 (60
FR 58397). The February 21, 1996, and April 2, 1996, submittals 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 April 10, 1996.
No significant hazards consideration comments received: No.
Local Public Document Room location: For Byron, the Byron Public
Library District, 109 N. Franklin, P.O. Box 434, Byron, Illinois 61010;
for Braidwood, the Wilmington Public Library, 201 S. Kankakee Street,
Wilmington, Illinois 60481.
Commonwealth Edison Company, Docket Nos. STN 50-454 and STN 50-455,
Byron Station, Unit Nos. 1 and 2, Ogle County, Illinois
Docket Nos. STN 50-456 and STN 50-457, Braidwood Station, Unit Nos. 1
and 2, Will County, Illinois
Date of application for amendments: May 17, 1995, as supplemented
by letters dated January 17, March 8, March 18, April 4 and April 9,
1996.
Brief description of amendments: The amendments revised the
Facility Operating Licenses and the technical specifications to permit
the steam generator tubes to be repaired using the tungsten inert gas
welded sleeve process developed by ABB-Combustion Engineering and
remove references to the kinetically welded sleeving process.
Date of issuance: April 12, 1996.
Effective date: April 12, 1996.
Amendment Nos.: 83, 83, 75, and 75.
Facility Operating License Nos. NPF-37, NPF-66, NPF-72 and NPF-77:
The amendments revised licenses and the Technical Specifications.
Date of initial notice in Federal Register: July 5, 1995 (60 FR
35064) The additional submittals provided 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 April 12, 1996.
No significant hazards consideration comments received: No.
Local Public Document Room location: For Byron, the Byron Public
Library District, 109 N. Franklin, P.O. Box 434, Byron, Illinois 61010;
for Braidwood, the Wilmington Public Library, 201 S. Kankakee Street,
Wilmington, Illinois 60481.
Commonwealth Edison Company, Docket Nos. 50-237 and 50-249, Dresden
Nuclear Power Station, Units 2 and 3, Grundy County, Illinois
Docket Nos. 50-254 and 50-265, Quad Cities Nuclear Power Station, Units
1 and 2, Rock Island County, Illinois
Date of application for amendments: September 1, 1995, for Dresden
and September 20, 1995, for Quad Cities.
Brief description of amendments: This application upgrades the
current custom Technical Specifications (TS) for Dresden and Quad
Cities to the Standard Technical Specifications contained in NUREG-
0123, ``Standard Technical Specification General Electric Plants BWR/
4.'' This application upgrades only Section 6.0, ``Administrative
Controls.''
Date of issuance: April 2, 1996.
Effective date: Immediately, to be implemented no later than June
30, 1996.
Amendment Nos.: 149, 143, 170, and 166.
Facility Operating License Nos. DPR-19, DPR-25, DPR-29 and DPR-30:
The amendments revised the Technical Specifications.
Date of initial notice in Federal Register: September 20, 1995 (60
FR 48728) for Dresden and October 5, 1995 (60 FR 52226) for Quad
Cities. The Commission's related evaluation of the amendments is
contained in a Safety Evaluation dated April 2, 1996.
[[Page 18178]]
No significant hazards consideration comments received: No.
Local Public Document Room location: for Dresden, Morris Area
Public Library District, 604 Liberty Street, Morris, Illinois 60450;
for Quad Cities, Dixon Public Library, 221 Hennepin Avenue, Dixon,
Illinois 61021.
Commonwealth Edison Company, Docket Nos. 50-373 and 50-374, LaSalle
County Station, Units 1 and 2, LaSalle County, Illinois
Date of application for amendments: January 18, 1996, as
supplemented on March 1, March 22, March 26, and April 3, 1996.
Brief description of amendments: The amendments change the
setpoints for the automatic primary containment isolation signal upon
detection of a high main steamline tunnel differential temperature and
delete the automatic isolation function upon detection of a high main
steamline tunnel temperature. Additionally, the amendments provide a 12
hour allowed outage time for the Main Steam Line Tunnel Differential
Temperature--High isolation signal upon loss of the Reactor Building
Ventilation System.
Date of issuance: April 4, 1996.
Effective date: Immediately, to be implemented prior to restart
from refueling outage L1R07 (Unit 1) and L2R07 (Unit 2).
Amendment Nos.: 111 and 96.
Facility Operating License Nos. NPF-11 and NPF-18: The amendments
revised the Technical Specifications.
Date of initial notice in Federal Register: February 27, 1996 (61
FR 7281). The March 1, March 22, March 26 and April 3, 1996, submittals
provided additional 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 April 4, 1996.
No significant hazards consideration comments received: No.
Local Public Document Room location: Jacobs Memorial Library,
Illinois Valley Community College, Oglesby, Illinois 61348.
Commonwealth Edison Company, Docket Nos. 50-373 and 50-374, LaSalle
County Station, Units 1 and 2, LaSalle County, Illinois
Date of application for amendments: August 25, 1995 as supplemented
on December 15, 1995, February 5, February 9, February 28, March 4,
March 28 and April 3, 1996.
Brief description of amendments: These amendments revise the
LaSalle Facility Operating Licenses and Technical Specifications (TSs)
to reflect the deletion of the leakage control system (LCS) presently
installed to control and contain the leakage past the main steamline
isolation valves (MSIVs) on each of the four main steamlines. The TSs
are also revised to raise the allowable leakage rates from 25 standard
cubic feet per hour (scfh) for each set of MSIVs and a total of 100
scfh from all four main steamlines to values of 100 scfh per steamline
and 400 scfh for all four steamlines.
Date of issuance: April 5, 1996.
Effective date: Immediately, to be implemented by startup from
refueling outage L1R07 (Unit 1) and L2R07 (Unit 2).
Amendment Nos.: 112 and 97.
Facility Operating License Nos. NPF-11 and NPF-18: The amendments
revised the licenses and technical specifications.
Date of initial notice in Federal Register: October 25, 1995 (60 FR
54717). The December 15, 1995, February 5, February 9, February 28,
March 4, March 28 and April 3, 1996, submittals provided additional
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
April 5, 1996.
No significant hazards consideration comments received: No.
Local Public Document Room location: Jacobs Memorial Library,
Illinois Valley Community College, Oglesby, Illinois 61348.
Consolidated Edison Company of New York, Docket No. 50-247, Indian
Point Nuclear Generating Unit No. 2, Westchester County, New York
Date of application for amendment: June 16, 1994, as supplemented
February 6, 1995.
Brief description of amendment: The amendment revises License
Condition 2.K and relocates the Indian Point Nuclear Generating Unit
No. 2 (IP2) fire protection requirements from the IP2 Technical
Specifications to the IP2 fire protection program plan in accordance
with the guidance provided in Generic Letter (GL) 86-10,
``Implementation of Fire Protection Requirements,'' April 24, 1986, and
GL 88-12, ``Removal of Fire Protection Requirements from Technical
Specifications,'' August 2, 1988.
Date of issuance: March 26, 1996.
Effective date: As of the date of issuance to be implemented within
9 months.
Amendment No.: 186.
Facility Operating License No. DPR-26: Amendment revised the
Technical Specifications and the Facility Operating License.
Date of initial notice in Federal Register: August 17, 1994 (59 FR
42335) The February 6, 1995, submittal provided clarifying information
and did not expand the scope of the original application, and did not
change the initial proposed no significant hazards consideration
determination. The Commission's related evaluation of the amendment is
contained in a Safety Evaluation dated March 26, 1996.
No significant hazards consideration comments received: No.
Local Public Document Room location: White Plains Public Library,
100 Martine Avenue, White Plains, New York 10610.
Consumers Power Company, Docket No. 50-255, Palisades Plant, Van Buren
County, Michigan
Date of application for amendment: October 17, 1995.
Brief description of amendment: This amendment revises the
Palisades Facility Operating License to reference 10 CFR Part 40, allow
the use of source materials as reactor fuel, delete references to
specific amendments and specific revisions in the listed titles of the
Physical Security Plan, Suitability Training and Qualification Plan,
and the Safeguards Contingency Plan and make minor editorial changes to
the license. In addition, the Technical Specifications (TS) are
modified as follows: (1) TS 3.1.2 is modified to change the pressurizer
cooldown limit from 100 deg.F to 200 deg.F/hour; (2) the shield cooling
system requirements are relocated to the Final Safety Analysis Report;
(3) several minor editorial changes and corrections are made, including
corrections requested in the licensee's letter of March 24, 1995; and
(4) several TS bases pages have been revised. The portion of the
amendment request deleting license paragraph 2.F on reporting
requirements was denied.
Date of issuance: April 5, 1996.
Effective date: April 5, 1996.
Amendment No.: 171.
Facility Operating License No. DPR-20: Amendment revised the
Facility Operating License and the Technical Specifications.
Date of initial notice in Federal Register: November 27, 1995 (60
FR 58399).
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated April 5, 1996, and an Environmental
Assessment dated March 11, 1996 (61 FR 10811).
[[Page 18179]]
No significant hazards consideration comments received: No.
Local Public Document Room location: Van Wylen Library, Hope
College, Holland, Michigan 49423.
Duquesne Light Company, et al., Docket No. 50-334, Beaver Valley Power
Station, Unit No. 1, Shippingport, Pennsylvania
Date of application for amendment: December 7, 1995, as
supplemented January 4, March 1, March 5, March 7, March 11, March 27,
and March 29, 1996.
Brief description of amendment: The amendment revises Technical
Specifications 3/4.4.5 and 3/4.4.6.2 and their Bases to maintain
voltage-based steam generator tube repair criteria for the tube support
plate elevations for future cycles of operation. The amendment replaces
a 1.0 volt repair limit which had been approved on an interim basis by
License Amendment No. 184 (issued February 3, 1995) with a 2.0 volt
repair limit. The amendment also includes additional changes to reflect
the guidance provided in NRC Generic Letter 95-05, ``Voltage-Based
Repair Criteria for Westinghouse Steam Generator Tubes Affected by
Outside Diameter Stress Corrosion Cracking.''
Date of issuance: April 1, 1996.
Effective date: As of the date of issuance, to be implemented
within 60 days.
Amendment No: 198.
Facility Operating License No. DPR-66: Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: January 3, 1996 (61 FR
178) The January 4, March 1, March 5, March 7, March 11, March 27, and
March 29, 1996, 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
January 3, 1996 notice.
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated April 1, 1996.
No significant hazards consideration comments received: No.
Local Public Document Room location: B. F. Jones Memorial Library,
663 Franklin Avenue, Aliquippa, PA 15001.
Illinois Power Company and Soyland Power Cooperative, Inc., Docket No.
50-461, Clinton Power Station, Unit No. 1, DeWitt County, Illinois
Date of application for amendment: December 14, 1995.
Brief description of amendment: The amendment consists of several
changes to the instrumentation sections of the Clinton Power Station
Technical Specifications. These changes were required due to
engineering reanalyses or plant modifications. The affected
instrumentation includes: (1) steam line flow high channels for the
reactor core isolation cooling (RCIC) system, (2) ambient temperature
channels in the residual heat removal (RHR) system heat exchanger
rooms, (3) reactor vessel pressure channels that provide a permissive
for operation of the shutdown cooling mode of the RHR system, and (4)
RCIC storage tank water level instrument channels.
Date of issuance: April 10, 1996.
Effective date: April 10, 1996.
Amendment No.: 104.
Facility Operating License No. NPF-62: The amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: January 22, 1996 (61 FR
1631) The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated April 10, 1996.
No significant hazards consideration comments received: No.
Local Public Document Room location: The Vespasian Warner Public
Library, 120 West Johnson Street, Clinton, Illinois 61727.
No significant hazards consideration comments received: No.
Northern States Power Company, Docket No. 50-263, Monticello Nuclear
Generating Plant, Wright County, Minnesota
Date of application for amendment: August 15, 1995, as supplemented
November 14, and December 20, 1995.
Brief description of amendment: The amendment modifies the
Monticello Technical Specifications (TS) to: (1) revise the main steam
line isolation valve leak rate test acceptance criterion to be based
upon the combined maximum flow path leakage for all four main steam
lines of 46 standard cubic feet per hour (scfh) in lieu of the current
limit of 11.5 scfh per valve; (2) revise the operability test interval
for the drywell spray header and nozzles from 5 years to 10 years; and
(3) revise TS 3/4.7.a.2, Primary Containment Integrity, to remove
information specific to the primary containment leakage rate testing
program and adopt the requirements of 10 CFR Part 50, Appendix J,
Option B, for Type A testing, while remaining under Appendix J, Option
A, for Type B and C testing.
Date of issuance: April 3, 1996.
Effective date: April 3, 1996.
Amendment No.: 95.
Facility Operating License No. DPR-22: Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: January 22, 1996 (61 FR
1632) The December 20, 1995, letter provided clarifying information
that was within the scope of the initial notice and did not change the
staff's initial proposed no significant hazards considerations
determination. The Commission's related evaluation of the amendment is
contained in a Safety Evaluation dated April 3, 1996.
No significant hazards consideration comments received: No.
Local Public Document Room location: Minneapolis Public Library,
Technology and Science Department, 300 Nicollet Mall, Minneapolis,
Minnesota 55401.
Northern States Power Company, Docket No. 50-263, Monticello Nuclear
Generating Plant, Wright County, Minnesota
Date of application for amendment: March 1, 1996 (supersedes
December 11, 1995, application).
Brief description of amendment: The amendment modifies Technical
Specification Section 4.7, Surveillance Requirements for Primary
Containment Automatic Isolation Valves, by revising Surveillance
Requirement 4.7.D.4 to require that the seat seals of the drywell and
suppression chamber purge and vent valves be replaced every six
operating cycles.
Date of issuance: April 9, 1996.
Effective date: April 9, 1996.
Amendment No.: 96.
Facility Operating License No. DPR-22: Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: March 8, 1996 (61 FR
9504). The Commission's related evaluation of the amendment is
contained in a Safety Evaluation dated April 9, 1996.
No significant hazards consideration comments received: No.
Local Public Document Room location: Minneapolis Public Library,
Technology and Science Department, 300 Nicollet Mall, Minneapolis,
Minnesota 55401.
Philadelphia Electric Company, Docket Nos. 50-352 and 50-353, Limerick
Generating Station, Units 1 and 2, Montgomery County, Pennsylvania
Date of application for amendments: December 22, 1995.
Brief description of amendments: The amendments change Technical
Specification 3.6.1.8, ``Drywell and Suppression Chamber Purge
System,'' increasing the drywell and suppression
[[Page 18180]]
chamber purge system operating time limit from 90 hours each 365 days
to 180 hours each 365 days.
Date of issuance: March 29, 1996.
Effective date: As of date of issuance, to be implemented within 30
days.
Amendment Nos.: 115 and 77.
Facility Operating License Nos. NPF-39 and NPF-85. The amendments
revised the Technical Specifications.
Date of initial notice in Federal Register: February 28, 1996 (61
FR 7555).
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated March 29, 1996.
No significant hazards consideration comments received: No.
Local Public Document Room location: Pottstown Public Library, 500
High Street, Pottstown, Pennsylvania 19464.
Rochester Gas and Electric Corporation, Docket No. 50-244, R. E. Ginna
Nuclear Power Plant, Wayne County, New York.
Date of application for amendment: February 9, 1996, as
supplemented March 20, 1996.
Brief description of amendment: The proposed amendment would revise
the Technical Specifications (TSs) to use an installed retractable
overhead door assembly and change TS 3.9.3 to satisfy closure
requirements for the containment equipment hatch during core
alterations or fuel movement in the containment building. The
retractable door is to be used as a functionally equivalent closure
plate currently required by TS 3.9.3.
Date of issuance: April 1, 1996.
Effective date: April 1, 1996.
Amendment No.: 62.
Facility Operating License No. DPR-18: Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: February 28, 1996 (61
FR 7557). The Commission's related evaluation of the amendment is
contained in a Safety Evaluation dated April 1, 1996.
No significant hazards consideration comments received: No.
Local Public Document Room location: Rochester Public Library, 115
South Avenue, Rochester, New York 14610.
South Carolina Electric & Gas Company, South Carolina Public Service
Authority, Docket No. 50-395, Virgil C. Summer Nuclear Station, Unit
No. 1, Fairfield County, South Carolina
Date of application for amendment: August 18, 1995, as supplemented
on November 1, 1995, February 14, March 14 (there are two supplemental
letters with this date), and March 25, 1996.
Brief description of amendment: The amendment revises the Operating
License (OL) to increase the authorized core power level from 2775
Megawatts thermal (MWt) to 2900 MWt. The amendment also approves
changes to the technical specifications (TS) to implement uprated power
operation.
Date of issuance: April 12, 1996.
Effective date: April 12, 1996.
Amendment No.: 133.
Facility Operating License No. NPF-12: Amendment revises the OL and
TS.
Date of initial notice in Federal Register: December 6, 1995 (60 FR
62495). The original Federal Register notice included information from
the licensee's November 1, 1995 supplemental letter. The February 14,
March 14, and March 25, 1996 supplemental letters provided
clarification and amplification of the analysis in the November 1, 1995
letter and were not outside the scope of the initial Federal Register
notice. The Commission's related evaluation of the amendment is
contained in an Environmental Assessment dated April 12, 1996 and in a
Safety Evaluation dated April 12, 1996.
No significant hazards consideration comments received: No.
Local Public Document Room location: Fairfield County Library, 300
Washington Street, Winnsboro, SC 29180.
Southern California Edison Company, et al., Docket Nos. 50-361 and 50-
362, San Onofre Nuclear Generating Station, Unit Nos. 2 and 3, San
Diego County, California
Date of application for amendments: December 30, 1992, as
supplemented by letters dated September 7, 1993, August 17, 1994, and
March 7, 1996.
Brief description of amendments: These amendments add a new
technical specification (TS) 3/4.7.3.1, ``Component Cooling Water (CCW)
Safety Related Makeup System,'' and its associated Bases. The new TS
will ensure that sufficient CCW capacity is available for continued
operation of safety-related equipment during normal conditions and
design-basis events.
Date of issuance: April 11, 1996.
Effective date: April 11, 1996.
Amendment Nos.: Unit 2-129; Unit 3-118.
Facility Operating License Nos. NPF-10 and NPF-15: The amendments
revised the Technical Specifications.
Date of initial notice in Federal Register: March 3, 1993 (58 FR
12268). The September 7, 1993, August 17, 1994, and March 7, 1996,
letters provided additional clarifying information and did not change
the initial no significant hazards consideration determination.
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated April 11, 1996.
No significant hazards consideration comments received: No.
Local Public Document Room location: Main Library, University of
California, P. O. Box 19557, Irvine, California 92713.
Tennessee Valley Authority, Docket No. 50-328, Sequoyah Nuclear Plant,
Unit 2, Hamilton County, Tennessee
Date of application for amendment: December 12, 1995, and
supplemented March 4, 1996 (TS 95-23).
Brief description of amendment: The amendment revises the TS
surveillance requirements and bases to incorporate alternate S/G tube
plugging criteria at tube support plate (TSP) intersections. The
approach taken is based on guidance given in Generic Letter (GL) 95-05,
``Voltage-Based Repair Criteria for Westinghouse Steam Generator Tubes
Affected by Outside Diameter Stress Corrosion Cracking.'' The amendment
is applicable for Cycle 8 operation only.
Date of issuance: April 3, 1996.
Effective date: April 3, 1996.
Amendment No.: 211.
Facility Operating License Nos. DPR-77: Amendment revises the
technical specifications.
Date of initial notice in Federal Register: January 3, 1996 (61 FR
183) The March 6, 1996 supplemental letter provided clarifying
information which did not change the proposed no significant hazards
consideration.
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated April 3, 1996.
No significant hazards consideration comments received: None
Local Public Document Room location: Chattanooga-Hamilton County
Library, 1101 Broad Street, Chattanooga, Tennessee 37402
The Cleveland Electric Illuminating Company, Centerior Service Company,
Duquesne Light Company, Ohio Edison Company, Pennsylvania Power
Company, Toledo Edison Company, Docket No. 50-440, Perry Nuclear Power
Plant, Unit No. 1, Lake County, Ohio
Date of application for amendment: February 27, 1996, as
supplemented by letter dated March 1, 1996.
Brief description of amendment: The amendment allows the drywell
personnel air lock shield doors to be open during Operational
Conditions 1, 2, and 3 until the end of Operating Cycle 6.
[[Page 18181]]
Date of issuance: March 22, 1996.
Effective date: March 22, 1996.
Amendment No.: 84.
Facility Operating License No. NPF-58: This amendment approved a
change to the design basis as described in the Updated Safety Analysis
Report. Public comments requested as to proposed no significant hazards
consideration: Yes (61 FR 8982 dated March 8, 1996). That notice
provided an opportunity to submit comments on the Commission's proposed
no significant hazards consideration determination. No comments have
been received. The notice also provided for an opportunity to request a
hearing BiWeekly Notice by March 18, 1996, corrected to April 5, 1996
(61 FR 10600 dated March 14, 1996), but indicated that if the
Commission makes a final no significant hazards consideration
determination any such hearing would take place after issuance of the
amendment. The March 1, 1996, supplemental letter provided additional
clarifying information and did not change the staff's original no
significant hazards consideration determination.
The Commission's related evaluation of the amendment and final no
significant hazards consideration determination is contained in a
Safety Evaluation dated March 22, 1996.
Local Public Document Room location: Perry Public Library, 3753
Main Street, Perry, Ohio 44081.
TU Electric Company, Docket Nos. 50-445 and 50-446, Comanche Peak Steam
Electric Station, Unit Nos. 1 and 2, Somervell County, Texas
Date of amendment requests: November 21, 1995 (TXX-95288) as
supplemented by letters dated December 15, 1995 (TXX-95306), and
February 2, 1996 (TXX-96040).
Brief description of amendments: The amendments revised the core
safety limit curves and revised N-16 Overtemperature reactor trip
setpoints as a result of the reload analyses for CPSES Unit 2, Cycle 3.
In addition, the minimum required Reactor Coolant System (RCS) flow was
increased and an administrative enhancement was included in the
footnotes of the RCS flow-low reactor trip function setpoint for both
Units 1 and 2.
Date of issuance: April 1, 1996.
Effective date: April 1, 1996.
Amendment Nos.: Unit 1-49; Unit 2-35.
Facility Operating License Nos. NPF-87 and NPF-89. The amendments
revised the Technical Specifications.
Date of initial notice in Federal Register: January 3, 1996 (61 FR
185) The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated April 1, 1996.
No significant hazards consideration comments received: No.
Local Public Document Room location: University of Texas at
Arlington Library, Government Publications/Maps, 702 College, P.O. Box
19497, Arlington, Texas 76019
Virginia Electric and Power Company, et al., Docket Nos. 50-338 and 50-
339, North Anna Power Station, Units No. 1 and No. 2, Louisa County,
Virginia
Date of application for amendments: July 26, 1995.
Brief description of amendments: The amendments revise the
Technical Specifications to increase the pressurizer safety valve lift
setpoint tolerance and reduce the pressurizer high pressure reactor
trip setpoint and allowable value.
Date of issuance: April 1, 1996.
Effective date: April 1, 1996.
Amendment Nos.: 200 and 181.
Facility Operating License Nos. NPF-4 and NPF-7: Amendments revised
the Technical Specifications.
Date of initial notice in Federal Register: August 30, 1995 (60 FR
45189) The Commission's related evaluation of the amendments is
contained in a Safety Evaluation dated April 1, 1996.
No significant hazards consideration comments received: No.
Local Public Document Room location: The Alderman Library, Special
Collections Department, University of Virginia, Charlottesville,
Virginia 22903-2498.
Wolf Creek Nuclear Operating Corporation, Docket No. 50-482, Wolf Creek
Generating Station, Coffey County, Kansas
Date of amendment request: March 8, 1996, as supplemented by letter
dated March 26, 1996.
Brief description of amendment: This amendment reduces the
calculated thermal design flow of the reactor coolant system and
increases the trip setpoint of the low pressurizer pressure.
Date of issuance: April 4, 1996.
Effective date: April 4, 1996.
Amendment No.: 99.
Facility Operating License No. NPF-42: The amendment revised the
Technical Specifications.
Public comments requested as to proposed no significant hazards
consideration: Yes (61 FR 10389 dated March 13, 1996). The notice
provided an opportunity to submit comments on the Commission's proposed
no significant hazards consideration determination. No comments have
been received. The notice also provided for an opportunity to request a
hearing by April 12, 1996, but indicated that if the Commission makes a
final no significant hazards consideration determination any such
hearing would take place after issuance of the amendment.
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated April 4, 1996.
Local Public Document Room locations: Emporia State University,
William Allen White Library, 1200 Commercial Street, Emporia, Kansas
66801 and Washburn University School of Law Library, Topeka, Kansas
66621.
Notice of Issuance of Amendments to Facility Operating Licenses and
Final Determination of No Significant Hazards Consideration and
Opportunity for a Hearing (Exigent Public Announcement or Emergency
Circumstances)
During the period since publication of the last biweekly notice,
the Commission has issued the following amendments. The Commission has
determined for each of these amendments that the application for the
amendment complies with the standards and requirements of the Atomic
Energy Act of 1954, as amended (the Act), and the Commission's rules
and regulations. The Commission has made appropriate findings as
required by the Act and the Commission's rules and regulations in 10
CFR Chapter I, which are set forth in the license amendment.
Because of exigent or emergency circumstances associated with the
date the amendment was needed, there was not time for the Commission to
publish, for public comment before issuance, its usual 30-day Notice of
Consideration of Issuance of Amendment, Proposed No Significant Hazards
Consideration Determination, and Opportunity for a Hearing.
For exigent circumstances, the Commission has either issued a
Federal Register notice providing opportunity for public comment or has
used local media to provide notice to the public in the area
surrounding a licensee's facility of the licensee's application and of
the Commission's proposed determination of no significant hazards
consideration. The Commission has provided a reasonable opportunity for
the public to comment, using its best efforts to make available to the
public means of communication for the public to respond quickly, and in
the case of telephone comments, the comments have been recorded or
transcribed as appropriate and the licensee has been informed of the
public comments.
[[Page 18182]]
In circumstances where failure to act in a timely way would have
resulted, for example, in derating or shutdown of a nuclear power plant
or in prevention of either resumption of operation or of increase in
power output up to the plant's licensed power level, the Commission may
not have had an opportunity to provide for public comment on its no
significant hazards consideration determination. In such case, the
license amendment has been issued without opportunity for comment. If
there has been some time for public comment but less than 30 days, the
Commission may provide an opportunity for public comment. If comments
have been requested, it is so stated. In either event, the State has
been consulted by telephone whenever possible.
Under its regulations, the Commission may issue and make an
amendment immediately effective, notwithstanding the pendency before it
of a request for a hearing from any person, in advance of the holding
and completion of any required hearing, where it has determined that no
significant hazards consideration is involved.
The Commission has applied the standards of 10 CFR 50.92 and has
made a final determination that the amendment involves no significant
hazards consideration. The basis for this determination is contained in
the documents related to this action. Accordingly, the amendments have
been issued and made effective as indicated.
Unless otherwise indicated, the Commission has determined that
these amendments satisfy the criteria for categorical exclusion in
accordance with 10 CFR 51.22. Therefore, pursuant to 10 CFR 51.22(b),
no environmental impact statement or environmental assessment need be
prepared for these amendments. If the Commission has prepared an
environmental assessment under the special circumstances provision in
10 CFR 51.12(b) and has made a determination based on that assessment,
it is so indicated.
For further details with respect to the action see (1) The
application for amendment, (2) the amendment to Facility Operating
License, and (3) the Commission's related letter, Safety Evaluation
and/or Environmental Assessment, as indicated. All of these items are
available for public inspection at the Commission's Public Document
Room, the Gelman Building, 2120 L Street NW., Washington, DC, and at
the local public document room for the particular facility involved.
The Commission is also offering an opportunity for a hearing with
respect to the issuance of the amendment. By May 24, 1996, the licensee
may file a request for a hearing with respect to issuance of the
amendment to the subject facility operating license and any person
whose interest may be affected by this proceeding and who wishes to
participate as a party in the proceeding must file a written request
for a hearing and a petition for leave to intervene. Requests for a
hearing and a petition for leave to intervene shall be filed in
accordance with the Commission's ``Rules of Practice for Domestic
Licensing Proceedings'' in 10 CFR Part 2. Interested persons should
consult a current copy of 10 CFR 2.714 which is available at the
Commission's Public Document Room, the Gelman Building, 2120 L Street
NW., Washington, DC and at the local public document room for the
particular facility involved. If a request for a hearing or petition
for leave to intervene is filed by the above date, the Commission or an
Atomic Safety and Licensing Board, designated by the Commission or by
the Chairman of the Atomic Safety and Licensing Board Panel, will rule
on the request and/or petition; and the Secretary or the designated
Atomic Safety and Licensing Board will issue a notice of a hearing or
an appropriate order.
As required by 10 CFR 2.714, a petition for leave to intervene
shall set forth with particularity the interest of the petitioner in
the proceeding, and how that interest may be affected by the results of
the proceeding. The petition should specifically explain the reasons
why intervention should be permitted with particular reference to the
following factors: (1) The nature of the petitioner's right under the
Act to be made a party to the proceeding; (2) the nature and extent of
the petitioner's property, financial, or other interest in the
proceeding; and (3) the possible effect of any order which may be
entered in the proceeding on the petitioner's interest. The petition
should also identify the specific aspect(s) of the subject matter of
the proceeding as to which petitioner wishes to intervene. Any person
who has filed a petition for leave to intervene or who has been
admitted as a party may amend the petition without requesting leave of
the Board up to 15 days prior to the first prehearing conference
scheduled in the proceeding, but such an amended petition must satisfy
the specificity requirements described above.
Not later than 15 days prior to the first prehearing conference
scheduled in the proceeding, a petitioner shall file a supplement to
the petition to intervene which must include a list of the contentions
which are sought to be litigated in the matter. Each contention must
consist of a specific statement of the issue of law or fact to be
raised or controverted. In addition, the petitioner shall provide a
brief explanation of the bases of the contention and a concise
statement of the alleged facts or expert opinion which support the
contention and on which the petitioner intends to rely in proving the
contention at the hearing. The petitioner must also provide references
to those specific sources and documents of which the petitioner is
aware and on which the petitioner intends to rely to establish those
facts or expert opinion. Petitioner must provide sufficient information
to show that a genuine dispute exists with the applicant on a material
issue of law or fact. Contentions shall be limited to matters within
the scope of the amendment under consideration. The contention must be
one which, if proven, would entitle the petitioner to relief. A
petitioner who fails to file such a supplement which satisfies these
requirements with respect to at least one contention will not be
permitted to participate as a party.
Those permitted to intervene become parties to the proceeding,
subject to any limitations in the order granting leave to intervene,
and have the opportunity to participate fully in the conduct of the
hearing, including the opportunity to present evidence and cross-
examine witnesses. Since the Commission has made a final determination
that the amendment involves no significant hazards consideration, if a
hearing is requested, it will not stay the effectiveness of the
amendment. Any hearing held would take place while the amendment is in
effect.
A request for a hearing or a petition for leave to intervene must
be filed with the Secretary of the Commission, U.S. Nuclear Regulatory
Commission, Washington, DC 20555, Attention: Docketing and Services
Branch, or may be delivered to the Commission's Public Document Room,
the Gelman Building, 2120 L Street NW., Washington, DC, by the above
date. Where petitions are filed during the last 10 days of the notice
period, it is requested that the petitioner promptly so inform the
Commission by a toll-free telephone call to Western Union at 1-(800)
248-5100 (in Missouri 1-(800) 342-6700). The Western Union operator
should be given Datagram Identification Number N1023 and the following
message addressed to (Project Director): petitioner's name and
telephone number, date petition was mailed, plant name, and publication
date and page number of this Federal Register notice. A copy of the
petition should also be sent to the Office of the
[[Page 18183]]
General Counsel, U.S. Nuclear Regulatory Commission, Washington, DC
20555, and to the attorney for the licensee.
Nontimely filings of petitions for leave to intervene, amended
petitions, supplemental petitions and/or requests for a hearing will
not be entertained absent a determination by the Commission, the
presiding officer or the Atomic Safety and Licensing Board that the
petition and/or request should be granted based upon a balancing of the
factors specified in 10 CFR 2.714(a)(1) (i)-(v) and 2.714(d).
Arizona Public Service Company, et al., Docket No. STN 50-529, Palo
Verde Nuclear Generating Station, Unit 2, Maricopa County, Arizona
Date of application for amendment: April 1, 1996, as supplemented
by letter dated April 3, 1996.
Brief description of amendment: The amendment modifies Technical
Specification (TS) 3/4.9.6 to temporarily allow the use of a hoist
instead of the refueling machine for the movement of the fuel assembly
at core location A-07.
Date of issuance: April 3, 1996.
Effective date: April 3, 1996.
Amendment No.: Unit 2--96.
Facility Operating License No. NPF-51: The amendment revised the
Technical Specifications.
Public comments requested as to proposed no significant hazards
consideration: No.
The Commission's related evaluation of the amendment, finding of
emergency circumstances, and final determination of no significant
hazards consideration are contained in a Safety Evaluation dated April
3, 1996.
Local Public Document Room location: Phoenix Public Library, 1221
N. Central Avenue, Phoenix, Arizona 85004.
Attorney for licensee: Nancy C. Loftin, Esq., Corporate Secretary
and Counsel, Arizona Public Service Company, P.O. Box 53999, Mail
Station 9068, Phoenix, Arizona 85072-3999.
NRC Project Director: William H. Bateman.
Duke Power Company, Docket Nos. 50-269, 50-270, and 50-287, Oconee
Nuclear Station, Units 1, 2, and 3, Oconee County, South Carolina
Date of application of amendments: April 2, 1996.
Brief description of amendments: The amendments revise Technical
Specification (TS) Section 4.5.4, ``Penetration Room Ventilation
System'' and TS Section 4.14, ``Reactor Building Purge Filters and
Spent Fuel Pool Ventilation System.'' The change updates the industry
guidance reference for testing charcoal absorber units for the system
covered by those TS.
Date of Issuance: April 2, 1996.
Effective date: April 2, 1996, to be implemented within 30 days.
Amendment Nos.: 215, 215, and 212.
Facility Operating License Nos. DPR-38, DPR-47, and DPR-55: The
amendments revised the Technical Specifications.
Public comments requested as to proposed no significant hazards
consideration: No.
The Commission's related evaluation of the amendments, finding of
emergency circumstances, and final determination of no significant
hazards consideration are contained in a Safety Evaluation dated April
2, 1996.
Local Public Document Room location: Oconee County Library, 501
West South Broad Street, Walhalla, South Carolina 29691.
Attorney for licensee: J. Michael McGarry, III, Winston and Strawn,
1200 17th Street, NW., Washington, DC 20036.
NRC Project Director: Herbert N. Berkow.
Dated at Rockville, Maryland, this 17th day of April 1996.
For the Nuclear Regulatory Commission.
Steven A. Varga,
Director, Division of Reactor Projects--I/II, Office of Nuclear Reactor
Regulation.
[FR Doc. 96-9925 Filed 4-23-96; 8:45 am]
BILLING CODE 7590-01-P