[Federal Register Volume 60, Number 206 (Wednesday, October 25, 1995)]
[Notices]
[Pages 54714-54733]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: 95-26275]
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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 September 29, 1995, through October 13,
1995. The last biweekly notice was published on October 11, 1995 (60 FR
52927).
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 November 24, 1995, 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
[[Page 54715]]
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 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.
Arizona Public Service Company, et al., Docket Nos. STN 50-528, STN 50-
529, and STN 50-530, Palo Verde Nuclear Generating Station, Units Nos.
1, 2, and 3, Maricopa County, Arizona
Date of amendments request: June 13, 1995, as supplemented by
letter dated August 16, 1995.
Description of amendments request: The proposed amendments would
extend allowed outage times (AOTs) for a safety injection tank (SIT), a
low- pressure safety injection (LPSI) subtrain, and an emergency diesel
generator (EDG) and add the bases for the extended AOTs.
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 proposed amendment does not involve a significant increase
in the probability or consequences of an accident previously
evaluated.
The Safety Injection Tanks (SITs) are passive components in the
Emergency Core Cooling System. The SITs are not an accident
initiator in any accident previously evaluated. Therefore, this
change does not involve a significant increase in the probability of
an accident previously evaluated.
SITs were designed to mitigate the consequences of Loss of
Coolant Accidents (LOCA). These proposed changes do not affect any
of the assumptions used in deterministic LOCA analysis. Hence the
consequences of accidents previously evaluated do not significantly
increase.
The allowed outage time (AOT) extension for boron concentration
outside the prescribed limits does not involve a significant
increase in the consequences of an accident as evaluated and
approved by the NRC in NUREG-1432, ``Standard Technical
Specifications for Combustion Engineering Plants.'' These changes
are applicable to PVNGS.
The changes pertaining to SIT inoperability based solely on
instrumentation malfunction do not involve a significant increase in
the consequences of an accident as evaluated and endorsed by the NRC
in NUREG-1366, ``Improvements to Technical Specifications
Surveillance Requirements,'' and Generic Letter 93-05, ``Line-Item
Technical Specifications Improvements to Reduce Surveillance
Requirements for Testing During Power Operations.'' These changes
are applicable to PVNGS.
The AOT extension from one hour to 24 hours for a SIT that is
inoperable due to reasons other than boron concentration not within
limits or the inability to verify level or pressure does not involve
a significant increase in the consequences of an accident. In order
to fully evaluate the affect of the SIT AOT extension, probabilistic
safety analysis (PSA) methods were utilized. The results of these
analyses show no significant increase in the core damage frequencies
(CDF). As a result, there would be no significant increase in the
consequences of an accident previously evaluated. These analyses are
detailed in CE NPSD-994, Combustion Engineering Owners Group ``Joint
Applications Report for Safety Injection Tank AOT/STI Extension,''
May 1995.
2. The proposed change does not create the possibility of a new
or different kind of accident from any accident previously
evaluated.
This proposed change does not change the design, configuration,
or method of operation of the plant. Therefore, this change does not
create the possibility of a new or different kind of accident from
any previously evaluated.
3. The proposed change does not involve a significant reduction
in a margin of safety.
The proposed changes do not involve a significant reduction in a
margin of safety.
The proposed changes do not affect the limiting conditions for
operation or their bases that are used in the deterministic analyses
to establish the margin of safety. PSA evaluations were used to
evaluate these changes. These evaluations demonstrated that the
changes are either risk neutral or risk beneficial. These
evaluations are detailed in CE NPSD-994.
The NRC staff has reviewed the licensee's analysis and, based on
that review, it appears that the three standards of 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendments request involve no significant hazards consideration.
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.
[[Page 54716]]
Carolina Power & Light Company, Docket No. 50-261, H. B. Robinson Steam
Electric Plant, Unit No. 2, Darlington County, South Carolina
Date of amendment request: September 11, 1995.
Description of amendment request: The proposed change is to (1)
modify a limiting condition for operation (LCO), TS Section 3.10.1.3,
to provide for temporary conditions in which the full length control
rod insertion limits (RILs) are exceeded due to automatic plant
responses or conservative operator actions and (2) add an allowance for
RILs to be exceeded for a time no greater than the time criteria
established by the axial power distribution methodology or 1 hour,
whichever is sooner. An action is added for the reactor to be placed in
the hot shutdown condition within 6 hours if compliance with the RILs
cannot be restored within the specified time period.
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:
This proposed change does not involve a significant hazards
consideration for the following reasons.
1. The proposed change does not involve a significant increase
in the probability or consequences of an accident previously
evaluated. The proposed change does not involve the addition or
modification of plant equipment, nor does it alter the design,
material, or operation of plant systems. No analyzed accidents are
initiated by an entire control rod bank exceeding the RILs, due to
automatic plant responses or conservative operator actions. The
overall performance of the Reactor Control System, Power
Distribution Control procedures, and Control Rod Drive System is not
degraded. There is no increase in fatigue or number of operational
cycles of equipment, and there is no change in system interfaces.
The consequences of previously evaluated accidents are not increased
since exceeding the RILs for a limited period is acceptable as the
probability of a simultaneous occurrence of an independent accident
is low. Therefore, an allowance for RILs to be exceeded for a
maximum of one (1) hour does not affect the probability of
occurrence or consequences of an analyzed accident.
2. The proposed change does not create the possibility of a new
or different kind of accident from any accident previously
evaluated. The proposed change adds an allowance for RILs to be
exceeded for a maximum of one (1) hour. The proposed change does not
involve the addition or modification of plant equipment, nor does it
alter the design or operation of plant systems. The only procedural
changes required will be those associated with recovery from the
infrequent condition of exceeding the RILs. No new accident
scenarios are introduced when the RILs are exceeded for a short
period of time due to automatic plant responses or conservative
operator actions because the probability of a simultaneous
occurrence of an independent accident is low. Therefore, an
allowance for RILs to be exceeded for a maximum of one (1) hour does
not create the possibility of a new or different kind of accident
from any accident previously evaluated.
3. The proposed change does not involve a significant reduction
in the margin of safety. The proposed change adds an allowance for
RILs to be exceeded for a maximum of one (1) hour. The proposed
change does not involve the addition or modification of plant
equipment, nor does it alter the design or operation of plant
systems. The overall performance of the Reactor Control System,
Power Distribution Control, and Control Rod Drive System is not
degraded. There is no increase in fatigue or number of operational
cycles of equipment, and there is no change in system interfaces.
When the RILs are exceeded for a limited time period, due to
automatic plant responses or conservative operator actions, the
margin of safety is not reduced because the probability of a
simultaneous occurrence of an independent accident is acceptably
low. Therefore, an allowance for RILs to be exceeded for a maximum
of one (1) hour does not involve a significant reduction in a margin
of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: Hartsville Memorial Library,
147 West College Avenue, Hartsville, South Carolina 29550.
Attorney for licensee: R.E. Jones, General Counsel, Carolina Power
& Light Company, Post Office Box 1551, Raleigh, North Carolina 27602.
NRC Project Director: David B. Matthews.
Commonwealth Edison Company, Docket Nos. STN 50-454 and STN 50-455,
Byron Station, Unit Nos. 1 and 2, Ogle County, Illinois
Docket Nos. STN 50-456 and STN 50-457, Braidwood Station, Unit Nos. 1
and 2, Will County, Illinois
Date of amendment request: September 14, 1995.
Description of amendment request: The proposed amendment would
allow the use of an alternate zirconium based fuel cladding, ZIRLO, and
permit limited substitution of ZIRLO filler rods for fuel rods. The
proposed amendment also includes a clarification and an editorial
change.
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 methodologies used in the accident analyses remain
unchanged. The proposed changes do not change or alter the design
assumptions for the systems or components used to mitigate the
consequences of an accident. Use of ZIRLO fuel cladding does not
adversely affect fuel performance or impact nuclear design
methodology. Therefore, accident analysis results are not impacted.
The operating limits will not be changed and the analysis
methods to demonstrate operation within the limits will remain in
accordance with NRC-approved methodologies. Other than the changes
to the fuel assemblies, there are no physical changes to the plant
associated with this Technical Specification change. A safety
analysis will continue to be performed for each cycle to demonstrate
compliance with all fuel safety design bases.
VANTAGE 5 fuel assemblies with ZIRLO clad fuel rods meet the
same fuel assembly and fuel rod design bases as other VANTAGE 5 fuel
assemblies. In addition, the 10 CFR 50.46 criteria are applied to
the ZIRLO clad fuel rods. The use of these fuel assemblies will not
result in a change to the reload design and safety analysis limits.
Since the original design criteria are met, the ZIRLO clad fuel rods
will not be an initiator for any new accident. The clad material is
similar in chemical composition and has similar physical and
mechanical properties as Zircaloy-4. Thus, the cladding integrity is
maintained and the structural integrity of the fuel assembly is not
affected. ZIRLO cladding improves corrosion performance and
dimensional stability. No concerns have been identified with respect
to the use of an assembly containing a combination of Zircaloy-4 and
ZIRLO clad fuel rods. Since the dose predictions in the safety
analyses are not sensitive to the fuel rod cladding material used,
the radiological consequences of accidents previously evaluated in
the safety analysis remain valid.
Replacing the reference to the Final Safety Analysis Report
(FSAR) with a reference to the Updated Final Safety Analysis Report
(UFSAR) is an editorial change to reflect the current document.
Adding that reload fuel shall be similar in physical design to the
initial core loading or previous cycle loading is a clarification. A
reload analysis is completed for each cycle, in accordance with
USNRC-approved methodologies.
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.
[[Page 54717]]
VANTAGE 5 fuel assemblies with ZIRLO clad fuel rods satisfy the
same design bases as those used for other VANTAGE 5 fuel assemblies.
All design and performance criteria continue to be met and no new
failure mechanisms have been identified. The ZIRLO cladding material
offers improved corrosion resistance and structural integrity.
The proposed changes do not affect the design or operation of
any system or component in the plant. The safety functions of the
related structures, systems, or components are not changed in any
manner, nor is the reliability of any structure, system, or
component reduced. The changes do not affect the manner by which the
facility is operated and do not change any facility design feature,
structure, or system. No new or different type of equipment will be
installed. Since there is no change to the facility or operating
procedures, and the safety functions and reliability of structures,
systems, or components are not affected, 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 use of Zircaloy-4, ZIRLO, or stainless steel filler rods in
fuel assemblies will not involve a significant reduction in the
margin of safety because analyses using NRC-approved methodology
will be performed for each configuration to demonstrate continued
operation within the limits that assure acceptable plant response to
accidents and transients. These analyses will be performed using
NRC-approved methods that have been approved for application to the
fuel configuration.
Use of ZIRLO cladding material does not change the VANTAGE 5
reload design and safety analysis limits. The use of these fuel
assemblies will take into consideration the normal core operating
conditions allowed in the Technical Specifications. For each cycle
reload core, the fuel assemblies will be evaluated using NRC-
approved reload design methods, including consideration of the core
physics analysis peaking factors and core average linear heat rate
effects.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
requested amendments involve no significant hazards consideration.
Local Public Document Room location: For Byron, the Byron Public
Library District, 109 N. Franklin, P.O. Box 434, Byron, Illinois 61010;
for Braidwood, the Wilmington Public Library, 201 S. Kankakee Street,
Wilmington, Illinois 60481
Attorney for licensee: Michael I. Miller, Esquire; Sidley and
Austin, One First National Plaza, Chicago, Illinois 60603.
NRC Project Director: Robert A. Capra.
Commonwealth Edison Company, Docket Nos. 50-373 and 50-374, LaSalle
County Station, Units 1 and 2, LaSalle County, Illinois
Date of amendment request: August 28, 1995.
Description of amendment request: The proposed amendments would
support elimination of the Main Steam Isolation Valve Leakage Control
System (MSIV LCS) and instead use the main steamline drains and
condenser to process MSIV leakage. The proposed changes would also
increase the allowable MSIV leakage from 100 standard cubic feet per
hour (scfh) for all four main steam lines to 100 scfh per steam line
(400 scfh for all four main steam lines).
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
The proposed changes do not involve a significant increase in
the probability or consequences of an accident previously evaluated
because:
The proposed changes involve eliminating the requirement for the
Main Steam Isolation Valve Leakage Control System (MSIV LCS). This
system is manually initiated following a design basis Loss of
Coolant Accident (LOCA). Since operation of the LCS is initiated
after the accident has already begun, elimination of that system
will not affect the probability of a LOCA. The LCS only interfaces
with the main steamlines, with the exception of one MSIV LCS power
supply which supplies power to the Reactor Protection System Scram
Discharge Volume high level scram. This power supply will remain in
place after the MSIV LCS is isolated from the main steamlines.
Therefore, since the only significant system interface is with the
main steamlines, and the system does not impact the reliability of
any plant equipment, elimination of that system will not cause an
increase in the likelihood that any accident might occur.
The proposed change to increase the allowable MSIV leakage limit
from 100 scfh through all four main steam lines to 100 scfh per main
steam line (400 scfh total) will not increase the probability of an
accident. MSIV operability will not be degraded with the allowed
increased leakage.
The consequences of a LOCA are not significantly increased and
do not exceed the previously accepted licensing criteria for this
accident. General Electric has calculated the revised LOCA doses,
which have been added to the previous LOCA doses. These resulting
values are well below the acceptance criteria of 10CFR100 and
10CFR50, Appendix A.
The proposed changes do not create the possibility of a new or
different kind of accident from any accident previously evaluated
because:
The proposed changes require the use of the main steam piping
and condenser to process MSIV leakage. The analyses presented
provide assurance that this additional function does not compromise
the reliability of those systems. They will therefore continue to
function as intended and not be subject to an increased failure rate
or a failure of a different kind than previously considered.
In addition, MSIV functionality will not be adversely impacted
as a result of the increased leakage limit. The MSIVs are not being
modified in any way and will continue to provide their intended
isolation function.
The MSIV LCS will be cut and capped, which will completely
isolate it from other plant systems. Future degradation of its
associated piping would not impact any other system or create a
failure not previously analyzed. However, piping seismic Class II
over I criteria must be maintained for the abandoned MSIV LCS piping
until it is removed from the plant.
The proposed changes do not involve a significant reduction in a
margin of safety because:
The proposed change has been evaluated with respect to dose
limits contained in 10CFR100 and 10CFR50, Appendix A. The revised
dose calculations verify that the use of the main steam lines and
the condenser for leakage control, in place of the MSIV LCS, and
with an allowable total leakage of 400 scfh, maintains adequate
margins to the criteria listed above.
Even though there is a reduction in the margin to safety, the
new doses remain well within the criteria of 10 CFR 100 and 10 CFR
50, Appendix A. This reduction in margin is not significant when
compared to the increased reliability and capability of the main
steam lines and condenser as a method of treating MSIV leakage. The
new leakage pathway is consistent with the philosophy of protection
by multiple barriers for limiting fission product release to the
environment. In addition, the new method is passive and does not
require any new logic control or interlocks. The new pathway is also
capable of handling a larger amount of leakage than the MSIV LCS,
which was previously subject to concerns that it would not function
at leakage rates higher than its design capacity, or at reactor
pressures greater than 35 psig.
The revised calculated LOCA doses remain well within the
regulatory limits for MSIV leakage rates of 400 scfh for all four
main steam lines (100 scfh per steam line), and the margin to safety
is not significantly reduced as a result of the proposed changes.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
requested amendments involve no significant hazards consideration.
Local Public Document Room location: Jacobs Memorial Library,
Illinois Valley Community College, Oglesby, Illinois 61348.
Attorney for licensee: Michael I. Miller, Esquire; Sidley and
Austin, One First National Plaza, Chicago, Illinois 60603.
[[Page 54718]]
NRC Project Director: Robert A. Capra.
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: September 1, 1995.
Description of amendment request: Generic Letter 88-16 provided
guidance on removing cycle-specific parameters which are calculated
using NRC-approved methodologies from the Technical Specifications
(TS). The parameters are replaced in the TS with a reference to a named
report which contains the parameters, and a requirement that the
parameters remain within the limits specified in the report. The
proposed changes incorporate NRC-approved methodologies, approved
revisions to previously approved methodologies, or republished versions
of previously approved methodologies into section 6.9.2 of the Oconee
TS. The limits to which these methodologies are applied are (1) Axial
Power Imbalance Protective Limits and Variable Low RCS Pressure
Protective Limits, (2) Reactor Protective System Trip Setting Limits
for the Flux/Flow/Imbalance and Variable Low Reactor Coolant System
Pressure Trip Functions, and (3) Power Imbalance Limits. Since the
proposed changes only incorporate NRC-approved methodologies into the
TS, the licensee proposed that the changes are administrative in nature
and can be assumed to have no impact, or potential impact, on the
health and safety of the public.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
The proposed changes will not create a significant hazards
consideration, as defined by 10 CRF 50.92, because:
(1) The proposed changes will not involve a significant increase
in the probability or consequences of an accident previously
evaluated.
The proposed changes are administrative in nature, and do not
affect any system, procedure, or manipulation of any equipment which
could affect the probability or consequences of any accident.
(2) The proposed changes will not create the possibility of any
new or different kind of accident from any accident previously
evaluated.
The proposed changes are administrative in nature, and cannot
introduce any new failure mode or transient which could create any
accident.
(3) The proposed changes will not involve a significant
reduction in a margin of safety.
The proposed changes are administrative in nature, and will not
affect any operating parameters or limits which could result in a
reduction in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: 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.
Entergy Operations, Inc., Docket No. 50-368, Arkansas Nuclear One, Unit
No. 2, Pope County, Arkansas
Date of amendment request: September 25, 1995.
Description of amendment request: The proposed amendment adds a
repair limit for circumferential cracks in steam generator tubes. It
deletes the requirement to repair cracks that are within the repair
limit. The proposed amendment also reduces the primary-to-secondary
leak rate limit.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
Criterion 1--Does Not Involve a Significant Increase in the
Probability or Consequences of an Accident Previously Evaluated.
Consistent with draft Regulatory Guide (RG) 1.121, ``Basis for
Plugging Degraded PWR Steam Generator Tubes,'' the traditional
maximum depth based criteria for steam generator tube repair
implicitly ensures that tubes accepted for continued service will
retain adequate structural and leakage integrity during normal
operating, transient, and postulated accident conditions. It is
recognized that defects in tubes permitted to remain in service
occasionally grow through-wall and develop small leaks. Limits on
allowable primary-to-secondary leakage established in the technical
specifications ensure timely plant shutdown before the structural
and leakage integrity of the affected tube is challenged.
The proposed change to implement a circumferential crack repair
limit in the expansion transition region for ANO-2 meets the
criteria of RG 1.121. The 40% degraded area repair limit was
determined by performing a structural analysis per the
recommendations of the RG and applying the following uncertainties:
95% lower bound material properties, 95% lower bound burst curve,
95% lower bound eddy current measurement uncertainties, and 95%
upper bound crack growth rate. The analysis demonstrates that tube
leakage and conditional probability of burst are acceptably low
during either normal operation or the most limiting accident
condition, a postulated main steam line break (MSLB) event.
As part of the implementation of the circumferential crack
repair limit, the distribution of End-of-Cycle (EOC) circumferential
indications in the expansion transition region will be used to
calculate the primary-to-secondary leakage. The allowable leakage is
bounded by the maximum leakage which results in doses within the
applicable dose limits (10CFR100 and General Design Criteria 19).
The limit is calculated using the technical specification reactor
coolant system (RCS) iodine activity. Application of the
circumferential crack repair limit requires the projection of the
postulated MSLB leakage based on the projected EOC distribution for
the next cycle. The projected EOC distribution is developed using
the most recent EOC eddy current results based on crack arc length.
The reduction in the leak rate limit reduces the possibility
that a defect in a leaking tube will grow to a size that is not
structurally acceptable.
Therefore, this change does not involve a significant increase
in the probability or consequences of any accident previously
evaluated.
Criterion 2--Does not Create the Possibility of a New or
Different Kind of Accident from any Previously Evaluated.
Implementation of the proposed circumferential crack repair
limit does not introduce any significant changes to the plant design
basis. The only accident possible from implementation of this limit
is a tube rupture, which has already been evaluated in the ANO-2
Safety Analysis Report.
The maximum primary-to-secondary leakage rate has been reduced
to 150 gallons per day through any one steam generator to help
preclude the potential for excessive leakage during all plant
conditions. The RG 1.121 criterion for establishing the operational
leak rate limit considers: (1) the detection of a crack before
potential tube rupture as a result of faulted plant conditions; (2)
the maintenance of a margin to tube rupture of not less than three
for normal operating conditions; and (3) that any leakage rate
increase will be gradual to provide time for corrective action. The
150 gallon per day limit is intended to provide for leakage
detection and plant shutdown in the event of an unexpected crack
propagation resulting in excessive leakage.
Steam generator tube integrity is maintained through inservice
inspection and primary-to-secondary leakage monitoring. Any tubes
exceeding the circumferential crack repair limit are removed from
service.
Therefore, this change does not create the possibility of a new
or different kind of accident from any previously evaluated.
Criterion 3--Does not Involve a Significant Reduction in the
Margin of Safety.
The use of the circumferential crack repair limit will maintain
steam generator tube
[[Page 54719]]
integrity commensurate with the criteria of RG 1.121. Upon
implementation of the limit, even under worst case conditions, the
occurrence of circumferential cracking in the expansion transition
region is not expected to lead to a steam generator tube rupture
event during normal or faulted plant conditions. The distribution of
crack indications left in service will result in acceptable primary-
to-secondary leakage and conditional tube burst probability during
all plant conditions.
The installation of steam generator tube plugs and sleeves
reduces RCS flow margin. Implementation of the circumferential crack
repair limit will decrease the number of tubes which must be
repaired by plugging or sleeving, thereby retaining additional flow
margin that would otherwise be reduced.
Therefore, this change does not involve a significant reduction
in the margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: Tomlinson Library, Arkansas
Tech University, Russellville, AR 72801.
Attorney for licensee: Nicholas S. Reynolds, Esquire, Winston and
Strawn, 1400 L Street, N.W., Washington, DC 20005-3502.
NRC Project Director: William D. Beckner.
Florida Power and Light Company, Docket Nos. 50-250 and 50-251, Turkey
Point Plant Units 3 and 4, Dade County, Florida
Date of amendment request: May 5, 1995, as supplemented September
28, 1995.
Description of amendment request: The licensee proposes to change
Turkey Point Units 3 and 4 Technical Specifications (TS) by revising TS
2.1.1, Safety Limit--Reactor Core; TS 2.2, Limiting Safety System
Settings--Reactor Trip System Instrumentation Setpoints; TS 3/4.2.5
Power Distribution Limits-- Departure from Nucleate Boiling (DNB)
Parameters; TS 3/4.3.2 Engineered Safety Features Actuation System
Instrumentation and the associated BASES. The proposed revision to the
TS includes (a) the implementation of Westinghouse's NRC approved
Revised Thermal Design Procedure (RTDP), and (b) a revision to the
Steam Generator Water Level Low-Low trip setpoint.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below. The licensee's analysis was
presented separately for the following areas: core thermal limits,
overtemperature [delta] T and overpower [delta] T reactor trip
setpoint; steam generator process measurement accuracy; and DNB
parameter surveillance requirements.
Core Thermal Limits, overtemperature [delta] T and overpower [delta] T
Reactor Trip Setpoint
(1) Operation of the facility in accordance with the proposed
amendments would not involve a significant increase in the
probability or consequences of an accident previously evaluated.
The revised Overtemperature and Overpower [delta] T reactor trip
functions do not involve an increase in the probability or
consequences of an accident previously evaluated because operation
with these revised values will not cause any design or analysis
acceptance criteria to be exceeded. The structural and functional
integrity of all plant systems is unaffected. The Overtemperature
and Overpower [delta] T reactor trip functions are part of the
accident mitigation response and are not initiators for any
transient. Therefore, the probability of occurrence previously
evaluated are not affected.
The changes to the Overtemperature and Overpower [delta] T
reactor trip functions do not affect the integrity of the fission
product barriers utilized for mitigation of radiological dose
consequences as a result of an accident. In addition, the off-site
mass releases used as input to the dose calculations are unchanged
from those previously assumed. Therefore, the off-site dose
predictions remain within the acceptance criteria of 10 CFR Part 100
limits for each of the transients affected. Since it has been
concluded that the transient analyses results are unaffected by the
parameter modifications, it is concluded that the probability or
consequences of an accident previously evaluated are not increased.
(2) The proposed license amendments do not create the
possibility of a new or different kind of accident from any accident
previously evaluated.
The revised Overtemperature and Overpower [delta] T reactor trip
functions do not create the possibility of a new or different kind
of accident from any accident previously evaluated because the
setpoint adjustments do not affect accident initiation sequences. No
new operating configuration is being imposed by the setpoint
adjustments that would create a new failure scenario. In addition,
no new failure modes or limiting single failures have been
identified. Therefore, the types of accidents defined in the UFSAR
continue to represent the credible spectrum of events to be analyzed
which determine safe plant operation. Therefore, it is concluded
that no new or different kind of accidents from those previously
evaluated have been created as a result of these revisions.
(3) The proposed license amendments do not involve a significant
reduction in a margin of safety.
The changes to the Overtemperature and Overpower [delta] T
reactor trip functions do not involve a reduction in the margin of
safety because the margin of safety associated with the
Overtemperature and Overpower [delta] T reactor trip functions, as
verified by the results of the accident analyses, are within
acceptable limits. All transients impacted by implementation of the
RTDP methodology have been analyzed and have met the applicable
accident analyses acceptance criteria. The margin of safety required
for each affected safety analysis is maintained. This conclusion is
not changed by the Overtemperature and Overpower [delta] T setpoint
modifications. The adequacy of the revised Technical Specifications
values to maintain the plant in a safe operating condition has been
confirmed. Therefore, the changes to the Overtemperature and
Overpower [delta] T reactor trip functions do not involve a
significant reduction in the margin of safety.
Steam Generator Process Measurement Accuracy
(1) Operation of the facility in accordance with the proposed
amendments would not involve a significant increase in the
probability or consequences of an accident previously evaluated.
The revised reactor trip setpoints on Steam Generator water
level do not involve a significant increase in the probability or
consequences of an accident previously evaluated. Operation with
these revised values will not cause any design or analysis
acceptance criteria to be exceeded. The structural and functional
integrity of any plant system is unaffected. The Steam Generator
Water Level trip functions are part of the accident mitigation
response and are not themselves initiators for any transient.
Therefore, the probability of occurrence previously evaluated is not
affected.
The changes to the reactor trip setpoints do not affect the
integrity of the fission product barriers utilized for mitigation of
radiological dose consequences as a result of an accident. The Steam
Generator Water Level Low-Low trip setpoint assumed in the safety
analyses has been revised and acceptable results were obtained. The
Steam Generator Water Level-Low setpoint is not credited in the
safety analysis. Consequently, the required margin of safety for
each affected safety analysis has been maintained. In addition, the
offsite mass releases used as input to the dose calculations are
unchanged from those previously assumed. Therefore, the offsite dose
predictions remain within the acceptance criteria of 10 CFR Part 100
limits for each of the transient analyses affected. Since it has
been determined that the transient analysis results are unaffected
by these parameter modifications, FPL concludes that the
consequences of an accident previously evaluated are not increased.
(2) The proposed license amendments do not create the
possibility of a new or different kind of accident from any accident
previously evaluated.
The setpoint values do not affect the assumed accident
initiation sequences. In addition, no new failure modes or limiting
single failures have been identified for any
[[Page 54720]]
plant equipment. Therefore, the types of accidents defined in the UFSAR
continue to represent the credible spectrum of events to be analyzed
which determine safe plant operation. Therefore, the possibility of
a new or different kind of accident from any accident evaluated is
not increased.
(3) The proposed license amendments do not involve a significant
reduction in the margin to safety.
The current Technical Specification trip setpoints and allowable
values were changed to maintain the current safety analysis limits.
The Steam Generator Water Level Low-Low trip setpoint assumed in the
safety analyses has been revised and acceptable results were
obtained. The Steam Generator Water Level-Low setpoint is not
credited in the safety analysis. Consequently, the required margin
of safety for each affected safety analysis has been maintained.
Thereby, the adequacy of the revised Technical Specification values
to maintain the plant in a safe operating condition is also
confirmed.
DNB Parameter Surveillance Requirements
(1) Operation of the facility in accordance with the proposed
amendments would not involve a significant increase in the
probability or consequences of an accident previously evaluated.
With the retention of the previous Safety Analyses Limits for
Departure from Nucleate Boiling (DNB) (T.S. 3/4.2.5) and the
existing Reactor Coolant System (RCS) low flow trip Nominal Trip
Setpoint (NTS), there is no increase in the probability or
consequences of an accident previously evaluated because there is no
change to any design or analysis acceptance criteria. The structural
and functional integrity of any plant system is unaffected. The
proposed license amendments revise the surveillance requirements for
DNB parameters and incorporate the RTDP uncertainty analysis into
the Westinghouse methodology for the RCS Loss of Flow determination
of the Allowable Value.
The changes to the reactor trip functions do not affect the
integrity of the fission product barriers utilized for mitigation of
radiological dose consequences as a result of an accident. The
margin to safety for the RCS Loss of Flow trip remains protected as
the trip setpoints assumed in the safety analyses are not revised.
In addition, the offsite mass releases used as input to the dose
calculations are unchanged from those previously assumed. Therefore,
the offsite dose predictions remain within the acceptance criteria
of 10 CFR Part 100 limits for each of the transients affected. Since
it has been determined that the transient results are unaffected by
these parameter modifications, it is concluded that the consequences
of an accident previously evaluated are not increased.
(2) Operation of the facility in accordance with the proposed
amendment would not create the possibility of a new or different
kind of accident from any accident previously evaluated.
The revised Allowable Value does not create the possibility of a
new or different kind of accident from any accident previously
evaluated. Revision of the surveillance requirements merely provides
clarification to more accurately reflect the surveillance activity.
The Allowable Value does not affect the assumed accident
initiation sequences. In addition, no new failure modes or single
failures have been identified for any plant equipment. Therefore,
the types of accidents defined in the UFSAR continue to represent
the credible spectrum of events to be analyzed which determine safe
plant operation. Therefore, it is concluded that no new or different
kind of accidents from those previously evaluated have been created
as a result of these revisions.
(3) The proposed license amendments do not involve a significant
reduction in the margin to safety.
The RCS Loss of Flow setpoint assumed in the safety analysis
remains unchanged. Since the safety analysis limit setpoint value is
unchanged and no safety analysis is affected, the required margin of
safety for each affected safety analysis is maintained. Thereby, the
adequacy of the revised Technical Specification values to maintain
the plant in a safe operating condition is also confirmed.
Therefore, the change to the RCS Loss of Flow Allowable Value does
not involve a significant reduction in the margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: Florida International
University, University Park, Miami, Florida 33199.
Attorney for licensee: J. R. Newman, Esquire, Morgan, Lewis &
Bockius, 1800 M Street, NW., Washington, DC 20036.
NRC Project Director: David B. Matthews.
Florida Power and Light Company, Docket Nos. 50-250 and 50-251, Turkey
Point Plant Units 3 and 4, Dade County, Florida
Date of amendment request: July 26, 1995, as supplemented by letter
dated October 4, 1995.
Description of amendment request: The licensee proposes to revise
the technical specifications surveillance intervals and allowed outage
times for the channel operational tests performed on the analog
``bistable'' comparator modules for the reactor trip, reactor trip
permissive functions, engineered safety features actuation and
permissive functions identified below.
TS Table 3.3-1--Revise ACTION Statements 2a, 6, 12 and 13; increase
the time allowed for a channel to be inoperable or out of service in an
untripped condition from 1 hour to 6 hours. Revise ACTION Statement 2b;
increase the time a Nuclear Instrumentation System (NIS) channel in a
functional group may be bypassed to perform testing from 2 to 4 hours.
TS Table 3.3-2--Revise ACTION Statement 14; increase the time to be
in HOT STANDBY with the number of OPERABLE channels one less that the
Minimum Channels OPERABLE requirement from 6 to 12 hours. Revise ACTION
Statements 14, 20 and 22; increase the allowed outage time for test of
the logic trains from 2 hours to 8 hours. Revise ACTION Statements 15,
18 and 25; increase the time allowed for a channel to be inoperable and
out of service in an untripped condition from 1 hour to 6 hours.
TS Table 4.3-1--Revise the surveillance interval for Items 2.a, 4,
7, 8, 10, 11, 12 and Note (9) from monthly to quarterly. Revise the
surveillance interval for Item 2.b from monthly to startup, and Item 3
from monthly/startup to startup only. Revise the surveillance interval
for Items 17.a, 17.b, 17.c and 17.d from monthly to refueling. Revise
Note (1) from ``7 days'' to ``31 days'' and delete Note (8).
TS Table 4.3-2--Revise the surveillance interval for Items 1.d,
1.e, 1.f, 4.d, 5.c, 6.b, and 8.a from monthly to quarterly.
TS BASES 3/4.3.1 and 3/4.3.2--Revise the BASES section for
Technical Specification Sections 3/4.3.1 and 3/4.3.2 to reference the
Westinghouse WCAPs 10271 and 10271, Supplement 2, and associated
Nuclear Regulatory Commission (NRC) safety evaluation reports (SERs).
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
(1) Operation of the facility in accordance with the proposed
amendments would not involve a significant increase in the
probability or consequences of an accident previously evaluated.
The proposed changes in Technical Specification surveillance
intervals and allowed outage times for the subject Reactor
Protection System (RPS)/Nuclear Instrumentation System (NIS)/
Engineered Safety Features Actuation System (ESFAS) analog
instrumentation have been revised in accordance with the
recommendations and criteria of Westinghouse WCAP-10271, WCAP 10271,
Supplement 2, and the NRC's SERs on the same subject dated February
21, 1985 and dated February 22, 1989.
The proposed changes do not involve any hardware or setpoint
changes. Similarly, the proposed changes do not alter the manner in
which safety limits, limiting safety system setpoints or limiting
conditions for operation
[[Page 54721]]
are determined. Implementation of the proposed changes does affect the
probability of failure of the RPS, including NIS, and ESFAS, but
does not alter the manner in which protection is afforded nor the
manner in which limiting setpoint criteria are established for the
RPS/ESFAS instrumentation systems. Consequently, the proposed
changes do not result in an increase in the severity or consequences
of any accident previously evaluated.
Implementation of the proposed changes is expected to result in
an acceptably small increase in total RPS unavailability. This
increase is primarily due to less frequent surveillances and was
generically quantified to be less than 3% within WCAP-10271. WCAP-
10271 also documents that the implementation of the proposed changes
is also expected to result in a significant reduction in the
probability of core melt from inadvertent reactor trips (WCAP-
10271). This is the result of a reduction in the number of
inadvertent reactor trips (0.5 fewer inadvertent reactor trips per
unit per year) occurring during testing of the RPS instrumentation.
This reduction is primarily attributable to testing in bypass for
applicable channels and to less frequent surveillances. WCAP-10271
documents that the reduction in inadvertent core melt probability is
sufficiently large to counter the increased core melt probability,
resulting in an overall reduction in total core melt probability of
approximately 1%.
A corresponding probabilistic risk assessment (WCAP-10271,
Supplement 2) was documented by Westinghouse for the generic
implementation of the proposed changes for ESFAS instrumentation.
This Westinghouse evaluation along with the independent assessments
performed by an NRC contractor demonstrated that a 6% core damage
frequency increase represented an upper bound for Westinghouse
plants. For more realistic testing strategies, the core damage
frequency increase would be substantially less than this.
Consequently, the changes in Technical Specifications associated
with an extension of the surveillance intervals and out of service
times for the RPS/ESFAS instrumentation systems will have only a
small impact on plant risk. On this basis, FPL concludes that the
proposed changes will not have a significant effect on the
probability or consequences of licensing basis events; and the
probability or consequences of an accident previously evaluated for
Turkey Point does not significantly increase.
(2) Operation of the facility in accordance with the proposed
amendments would not create the possibility of a new or different
kind of accident from any accident previously evaluated.
The proposed changes in Technical Specification surveillance
intervals and allowed outage times for the subject RPS/ESFAS analog
instrumentation have been revised in accordance with the
recommendations and criteria of Westinghouse WCAP-10271, WCAP 10271,
Supplement 2, and the NRC's SERs on the same subject dated February
21, 1985 and dated February 22, 1989.
The proposed changes do not involve any hardware or setpoint
changes. Some existing instrumentation is designed to be tested in
bypass and current Technical Specifications allow testing in bypass.
Testing in bypass is also recognized by IEEE Standards.
Therefore, testing in bypass has been previously approved and
implementation of the proposed changes for testing in bypass does
not create the possibility of a new or different kind of accident
from any previously evaluated. Furthermore, since the proposed
changes do not alter the manner in which protection is afforded nor
the manner in which limiting criteria are established for the RPS
and ESFAS instrumentation systems, the possibility of a new or
different kind of accident from any previously evaluated has not
been created.
The proposed changes do not result in a change in the manner in
which the RPS or ESFAS provides plant protection. No change is being
made which alters the function of the RPS or ESFAS (other than in a
test mode). Rather, the likelihood or probability of the RPS and
ESFAS functioning properly is the only effect.
Therefore, the proposed changes do not create the possibility of
a new or different kind of accident nor involve a reduction in a
margin of safety as defined in the Safety Analysis Report.
Consequently, the changes in Technical Specifications associated
with an extension of the surveillance intervals and out of service
times for the RPS/ESFAS instrumentation systems will not create the
possibility of a new or different kind of accident from any
previously evaluated by the NRC, and does not create the possibility
of a new or different kind of accident from any accident previously
evaluated.
(3) Operation of the facility in accordance with the proposed
amendments would not involve a significant reduction in a margin of
safety.
The proposed changes in Technical Specification surveillance
intervals and allowed outage times for the subject RPS/ESFAS analog
instrumentation have been revised in accordance with the
recommendations and criteria of Westinghouse WCAP-10271, WCAP 10271,
Supplement 2, and the NRC's SERs on the same subject dated February
21, 1985 and dated February 22, 1989.
These changes in Technical Specifications only affect the
frequency of the channel operational tests and the allowed outage
times; they do not alter the manner in which protection is afforded
nor the manner in which limiting setpoint criteria are established.
In addition, the fundamental process to implement these channel
operational tests remains the same.
The proposed changes do not alter the manner in which safety
limits, limiting safety system setpoints or limiting conditions for
operation are determined. The impact of reduced testing is to allow
a longer time interval over which instrument uncertainties (e.g.,
drift) may act. The site specific review of historical drift data
and the conservative application of drift in the Westinghouse
methodology are sufficient to demonstrate that the basis of the
Technical Specification setpoint determinations are not adversely
affected by extending the surveillance interval from monthly to
quarterly, that is, quarterly surveillance test intervals would not
exceed the allowable instrument drift of these analog devices.
Implementation of the proposed changes is expected to result in
an overall improvement in safety by:
(a) Fewer inadvertent reactor trips per unit per year. This is
due to less frequent testing which minimizes the time spent in a
partial trip condition.
(b) Higher quality repairs leading to improved equipment
reliability due to longer allowed repair times.
(c) Improvements in the effectiveness of the operating staff in
monitoring and controlling plant operation. This is due to less
frequent distractions of the operator and shift supervisor from
attending to instrumentation testing.
The Westinghouse analysis demonstrates that any expected
increases in probability of core melt or core damage frequency are
small and are therefore acceptable. Consequently, the changes in
Technical Specifications associated with an extension of the
surveillance intervals and out of service times for the RPS/ESFAS
instrumentation systems will not significantly reduce the margin of
plant safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: Florida International
University, University Park, Miami, Florida 33199.
Attorney for licensee: J.R. Newman, Esquire, Morgan, Lewis &
Bockius, 1800 M Street NW., Washington, DC 20036.
NRC Project Director: David B. Matthews.
Maine Yankee Atomic Power Company, Docket No. 50-309, Maine Yankee
Atomic Power Station, Lincoln County, Maine
Date of amendment request: July 24, 1995
Description of amendment request: The proposed amendment would
modify Technical Specification (TS) 3.12.B by adding an Exception to
permit a once-per-operating cycle 10 day restoration time for Remedial
Action statement 3.12.B.2. The extended restoration time would allow
maintenance to be completed on the emergency diesel generators. In
addition, the Basis of TS 3.12 is supplemented in support of the
proposed amendment.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the
[[Page 54722]]
issue of no significant hazards consideration. The NRC staff has
reviewed the licensee's analysis against the standards of 10 CFR
50.92(c). The NRC staff's review is presented below:
1. The proposed amendment does not involve a significant increase
in the probability or consequences of an accident previously evaluated.
The emergency diesel generators (EDG) are not accident initiators
for any accident previously evaluated, nor does the proposed change
affect any of the assumptions used in the deterministic safety
analyses. To evaluate the effect of the proposed extended restoration
time of the EDGs fully, probabilistic safety analysis (PSA) methods
were used. The results of these analyses show no significant increase
in core damage frequency. Thus, the proposed change does not involve a
significant increase in the probability or consequences of an accident
previously evaluated.
2. The proposed amendment does not create the possibility of a new
or different kind of accident from any accident previously evaluated.
The proposed change does not alter the design, configuration, or
method of operation of the plant. Therefore, the proposed change does
not create the possibility of a new or different kind of accident from
any accident previously evaluated.
3. The proposed amendment does not involve a significant reduction
in a margin of safety.
The proposed change does not affect system or component limiting
conditions for operation, or the bases used in the deterministic
analyses to establish the margin of safety. The PSA evaluations used to
evaluate the proposed change demonstrated that the changes are either
risk neutral or risk beneficial. Thus the proposed change does not
involve a significant reduction in a margin of safety.
Based on this review, it appears that the three standards of 10 CFR
50.92(c) are satisfied. Therefore, the NRC staff proposes to determine
that the amendment request involves no significant hazards
consideration.
Local Public Document Room location: Wiscasset Public Library, High
Street, P.O. Box 367, Wiscasset, ME 04578.
Attorney for licensee: Mary Ann Lynch, Esquire, Maine Yankee Atomic
Power Company, 329 Bath Road, Brunswick, ME 04011.
NRC Project Director: Phillip F. McKee.
Maine Yankee Atomic Power Company, Docket No. 50-309, Maine Yankee
Atomic Power Station, Lincoln County, Maine
Date of amendment request: August 15, 1995.
Description of amendment request: The proposed amendment would
allow reduced power operation as a function of total reactor coolant
flow, for flow reductions as much as 5 percent below the currently
specified minimum flow. Specifically, operation would be allowed with
total flow rates below 360,000 gpm, if rated thermal power is reduced
by 1.5 percent for each 1.0 percent that total reactor coolant flow is
reduced.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration. The NRC staff has reviewed the licensee's analysis
against the standards of 10 CFR 50.92(c). The staff's review 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 amendment does not involve any changes in the
configuration of the reactor coolant system. Thus, precursors to
accidents previously evaluated are unchanged. The 5.0 percent reduction
in reactor coolant flow introduces a relatively minor change to the
overall plant heat balance, which is conservatively offset by the
proposed requirement to reduce rated thermal power by 1.5 percent for
each 1.0 percent reduction in reactor coolant system flow. Analysis by
the licensee shows that a 1.0 percent reduction in rated thermal power
for every 1.0 percent reduction in reactor coolant system flow is
sufficient to ensure that the current departure from nuclear boiling
ratio is maintained. The licensee asserts that achieving the reduced
power and other, related limits, within 24-hours of a subject flow
reduction will not significantly increase the probability or
consequences of an accident previously evaluated. Thus, the proposed
amendment does not involve a significant increase in the probability or
consequences of an accident previously evaluated.
2. The proposed amendment does not create the possibility of a new
or different kind of accident from any accident previously evaluated.
The proposed amendment does not involve any modifications or
additions to plant equipment, and the design and operation of the plant
are not affected. The reduction in rated thermal power, reactor
protection system trip points, and operating limits conservatively
offset the reduction in reactor coolant system flow. Plant operating
conditions remain bounded by Final Safety Analysis Report (FSAR)
Chapter 14, Safety Analysis. Thus, the proposed amendment does not
create the possibility of a new or different kind of accident from any
accident previously evaluated.
3. The proposed amendment does not involve a significant reduction
in a margin of safety.
Plant rated power is conservatively reduced, consistent with the
reactor coolant flow reduction. The power reduction is specifically
designed to maintain the margin to the specified acceptable fuel design
limit on the departure from nuclear boiling ratio (DNBR), as defined in
MY TS 2.2. The licensee has evaluated this margin using the
methodologies identified in Maine Yankee Technical Specification 5.14.
The reduction in power level, operating limits, and reactor protection
system setpoints ensures that the DNBR margin is maintained for those
FSAR Chapter 14 events that rely on automatic reactor trip protection.
Power level reductions ensure that the total sensible heat in the
reactor coolant system is conservative for those events dependent on
initial system energy. Thus, the proposed change does not involve a
significant reduction in a margin of safety.
Based on this review, it appears that the three standards of 10 CFR
50.92(c) are satisfied. Therefore, the NRC staff proposes to determine
that this amendment request involves no significant hazards
consideration.
Local Public Document Room location: Wiscasset Public Library, High
Street, P.O. Box 367, Wiscasset, ME 04578.
Attorney for licensee: Mary Ann Lynch, Esquire, Maine Yankee Atomic
Power Company, 329 Bath Road, Brunswick, ME 04011.
NRC Project Director: Phillip F. McKee.
Northeast Nuclear Energy Company, et al., Docket No. 50-336, Millstone
Nuclear Power Station, Unit Nos. 2, New London, Connecticut
Date of amendment request: September 19, 1995.
Description of amendment request: The proposed amendment would
reduce the frequency of the surveillance interval of the Safety
Injection Tanks (SITs) boron concentration from once per 31 days to
once per 6 months.
Basis for proposed no significant hazards consideration
determination:
[[Page 54723]]
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:
Pursuant to 10CFR50.92, Northeast Nuclear Energy Company (NNECO)
has reviewed the proposed change. NNECO concludes that the change
does not involve a significant hazards consideration since the
proposed change satisfies the criteria in 10CFR50.92(c). That is,
the proposed change does not:
1. Involve a significant increase in the probability or
consequences of an accident previously analyzed.
The revised Safety Injection Tank (SIT) surveillance
requirements meet all design and performance criteria. The change
has no [e]ffect on the ability of the SIT to perform its designed
function of providing borated water to the core following a
depressurization as a result of a Loss of Coolant Accident (LOCA).
Therefore, the changes to SIT surveillance requirements will not
increase the probability or consequences of an accident previously
evaluated.
2. Create the possibility of a new or different kind of accident
from any previously analyzed.
The revised SIT surveillance requirements meet all design and
performance criteria. The change has no [e]ffect on the ability of
the SIT to perform its design function of providing borated water to
the core following a depressurization as a result of a LOCA. The
change to the SIT surveillance requirement will not create the
possibility of a new or different kind of accident from any
previously analyzed.
3. Involve a significant reduction in the margin of safety.
The boron concentration of the SIT will not be affected by the
change to the surveillance requirement. The boron concentration
within the SIT will continue to be monitored on a basis consistent
with the historical performance. These changes will have no impact
on 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: Learning Resources Center,
Three Rivers Community-Technical College, 574 New London Turnpike,
Norwich, CT 06360.
Attorney for licensee: Lillian M. Cuoco, Esq., Senior Nuclear
Counsel, Northeast Utilities Service Company, P.O. Box 270, Hartford,
CT 06141-0270.
NRC Project Director: Phillip F. McKee.
Northeast Nuclear Energy Company, et al., Docket No. 50-336, Millstone
Nuclear Power Station, Unit Nos. 2, New London, Connecticut
Date of amendment request: September 29, 1995.
Description of amendment request: The proposed amendment would
modify the Technical Specifications 3.4.2.1, 3.4.2.2, 3.7.1.1, and
Table 4.7-1.
The proposed license amendment combines three separate changes to
the Millstone Unit No. 2 Technical Specifications which pertain to
safety valves. The first proposed modification would expand the as-
found tolerance of the lift setting pressure for the pressurizer and
the main steam safety valves from the current value of plus or minus 1
percent to plus or minus 3 percent. Clarifications have also been
proposed by specifying that the lift setting pressure shall be
determined at normal operating conditions and shall be set within plus
or minus 1 percent of the required lift setting. The second portion of
the modification would eliminate the need to verify the main steam
safety valve orifice size. The third modification would modify the main
steam safety valve action statement to reflect that if a main steam
safety valve is inoperable and compensating action cannot be taken that
the plant must be brought to hot shutdown (Mode 4) in 12 hours instead
of cold shutdown (Mode 5) in 30 hours.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration (SHC), which is presented below:
* * * The proposed changes do not involve an SHC because the changes
would not:
1. Involve a significant increase in the probability or
consequences of an accident previously evaluated.
The change in the as-found pressurizer safety valve tolerance
will not increase the probability of occurrence of any of the design
basis accidents. Even with the larger tolerance, the setpoint will
provide margin to normal operation, the reactor setpoint, and PORV
[power-operated relief valve setpoint]. This minimizes the
challenges to safety valves and assures that there is no increase in
the probability of an inadvertent opening of a pressurizer safety
valve. Similarly, even with the increase in allowed as-found
tolerance for the main steam safety valves, the setpoints will still
provide margin to normal operation. Thus, there is no impact on the
probability of an inadvertent opening of a steam generator safety
valve.
The loss of load event and the inadvertent closure of one main
steam isolation valve have been reanalyzed to show that even with a
[plus or minus] 3 percent tolerance for the pressurizer safety
valves and the main steam safety valves, that both the peak RCS
[reactor coolant system] pressure and the peak steam generator
pressure remain below 110 percent of design. Thus, even with the
larger as-found tolerances, the margin of safety for RCS and steam
generator overpressurization is maintained.
The steam generator tube rupture has been reanalyzed to take
into account the [plus or minus] 3 percent as-found tolerance and to
extend the margin for operator action to one hour. A comparison of
the calculated doses shows that with the new assumptions, there
would be a very small increase in calculated doses. The increased
calculated doses, however, remain well below the Standard Review
Plan acceptance criteria.
The proposed change in the shutdown mode does not impact the
probability or consequences of an accident previously evaluated. The
proposed change makes the action required for inoperable main steam
safety valves consistent with the modes that the technical
specification is applicable and would not modify the assumptions
made in any accident previously analyzed.
The change to delete the main steam safety valve orifice size
from technical specifications has no impact on any design basis
accident analysis.
Based upon these evaluations, it is concluded that the proposed
changes do not significantly increase the probability or
consequences of any design basis accident.
2. Create the possibility of a new or different kind of accident
from any previously analyzed.
The proposed changes do not create the possibility of a new or
different kind of accident from any previously analyzed.
The proposed changes do not change the as-left setpoints. The
change in as-found tolerances for the safety valves is being made to
reflect the results of past surveillances that indicate that the
setpoints can drift more than the current criteria. However, there
is no change in the plant configuration or in as-left setpoints.
The proposed change which requires the plant to go to Mode 4 in
12 hours instead of Mode 5 in 30 hours if the action statement is
not met, is consistent with the applicable modes of the technical
specification (i.e., the technical specification is not applicable
in Mode 4). No new or different kind of accident from those
previously analyzed can be postulated as a result of this proposed
change.
Thus, the changes do not create the possibility of a new or
different kind of accident from any previously analyzed.
3. Involve a significant reduction in the margin of safety.
As discussed above, the loss of load event and the inadvertent
closure of one main steam isolation valve have been reanalyzed to
show that even with a [plus or minus] 3 percent tolerance for the
pressurizer safety valves and the main steam safety valves, that
both the peak RCS pressure and the peak steam generator pressure
remain below 110 percent of design. Thus, even with the larger as-
found tolerances, the margin of safety for RCS and steam generator
overpressurization
[[Page 54724]]
is maintained. In addition, the steam generator tube rupture has been
reanalyzed with a [plus or minus] 3 percent tolerance on the steam
generator safety valves and the results show an insignificant
increase in the calculated doses.
The proposed change also directs the operator to bring the plant
to hot shutdown instead of cold shutdown to be consistent with the
applicable modes of the technical specification. There is no impact
on the assumptions made or the results of any accident previously
analyzed.
Therefore, it is concluded that the changes do not involve a
significant reduction in the margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: Learning Resources Center,
Three Rivers Community-Technical College, 574 New London Turnpike,
Norwich, CT 06360.
Attorney for licensee: Lillian M. Cuoco, Esq., Senior Nuclear
Counsel, Northeast Utilities Service Company, P.O. Box 270, Hartford,
CT 06141-0270.
NRC Project Director: Phillip F. McKee
Pennsylvania Power and Light Company, Docket Nos. 50-387 and 50-388
Susquehanna Steam Electric Station, Units 1 and 2, Luzerne County,
Pennsylvania
Date of amendment request: September 18, 1995
Description of amendment request: The proposed amendment would
relocate Fire Protection requirements from the Technical Specifications
to the Technical Requirements Manual. In addition, the proposed
amendment would revise Technical Specifications to include the
requirement for a program and procedure to implement the Technical
Requirements Program, and also revises Technical Specifications to add
the requirement for the Plant Operations Review Committee to review all
proposed changes to the Technical Requirements Program and to forward
copies of reviewed changes to the Susquehanna Review Committee.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
The proposed changes do not:
1. Involve a significant increase in the probability or
consequences of an accident previously evaluated.
The proposed change relocates the provisions of the Fire
Protection Program that are contained in the Technical
Specifications and places them in the Technical Requirements Manual.
No requirements are being added or deleted. Review and approval of
those portions of the Fire Protection Program contained in the
Technical Requirements Manual and revisions thereto will be the
responsibility of the Plant Operations Review Committee just as it
was their responsibility to review changes to the fire protection
Limiting Condition for Operation and Surveillance Requirements when
they were part of the Technical Specifications. Requiring review by
the Plant Operations Review Committee reinforces the importance of
the Technical Requirements Manual and the requirements controlled by
it and assures a multidisciplined review. Approved Technical
Requirements or changes thereto are provided to the Susquehanna
Review Committee for information. No design basis accidents are
affected by the change, nor are safety systems adversely affected by
the change. Therefore, there is no impact on the probability of
concurrence [occurrence] or the consequences of any design basis
accidents.
2. Create the possibility of a new or different kind of accident
from any previously evaluated.
The proposed changes relocate the provisions of the Fire
Protection Program that are contained in the Technical
Specifications and places them in the Technical Requirements Manual.
No requirements are being added or deleted by the Technical
Requirements Manual. There are no new failure modes associated with
the proposed changes. Therefore, since the plant will continue to
operate as designed, the proposed changes will not modify the plant
response to an accident.
3. Involve a significant reduction in a margin of safety.
No change is being proposed for the Fire Protection Program
requirements themselves. The relevant Technical Specifications are
being relocated, and the requirements contained therein are being
incorporated into the Technical Requirements Manual. Plant
procedures will continue to provide the specific instructions
necessary for the implementation of the requirements, just as when
the requirements resided in the Technical Specifications. Fire
Protection Program changes will be subject to the provisions of 10
CFR 50.59 and the current fire protection license condition. As
such, the changes do not directly affect any protective boundaries
nor does it [do they] impact the safety limits for the boundary.
Review and approval of those portions of the Fire Protection Program
contained in the Technical Requirements Manual and the revisions
thereto will be the responsibility of the Plant Operations Review
Committee just as it was their responsibility to review changes to
the fire protection Limiting Condition for Operation and
Surveillance Requirements when they were part of the Technical
Specification. Approved Technical Requirements or changes thereto
are provided to the Susquehanna Review Committee for information.
Thus, there are no adverse impacts on the protective boundaries,
safety limits, or margin of safety.
Since operability and surveillance requirements will remain in a
controlled document, the changes do not reduce the effectiveness of
Technical Specification requirements. Any changes to the Fire
Protection Program requirements will be made in accordance with the
provisions of 10 CFR 50.59 and the fire protection license
condition.
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: Osterhout Free Library,
Reference Department, 71 South Franklin Street, Wilkes-Barre,
Pennsylvania 18701.
Attorney for licensee: Jay Silberg, Esquire, Shaw, Pittman, Potts
and Trowbridge, 2300 N Street NW., Washington, DC 20037.
NRC Project Director: John F. Stolz.
Virginia Electric and Power Company, Docket No. 50-338, North Anna
Power Station, Unit No. 1, Louisa County, Virginia
Date of amendment request: September 19, 1995.
Description of amendment request: The proposed change would revise
the Technical Specifications (TS) for the North Anna Power Station,
Units 1 & 2 (NA-1 & 2). Specifically, the proposed changes would revise
TS Limiting Condition for Operation (LCO) 3.7.1.1 Action Statements, TS
Table 3.7-1, dually entitled ``Maximum Allowable Power Range Neutron
Flux High Setpoint With Inoperable Steam Line Safety Valves During 3
Loop Operation'' and ``Maximum Allowable Power Range Neutron Flux High
Setpoint With Inoperable Steam Line Safety Valves During 2 Loop
Operation,'' and the TS Bases 3/4.7.1.1, ``Safety Valves'' for NA-1 &
2. Table 3.7-1 provides the maximum allowable power range neutron flux
high setpoints with one or more main steam safety valves (MSSVs)
inoperable during two loop and three loop operation. The proposed
changes provide more conservative power range neutron flux high
setpoints calculated utilizing the Westinghouse Electric Corporation
(Westinghouse) recommended methodology and delete the information for
setpoints for two loop operation. The proposed changes also revise the
TS Bases to reflect the
[[Page 54725]]
methodology used to establish the new setpoints, and delete the LCO
Action Statement and the TS Bases for two loop operation.
Additionally, the information in Table 3.7-1 and the LCO Action
Statement associated with two loop operation have been deleted since
Virginia Electric and Power Company is prohibited by the license from
operating in this configuration.
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:
Specifically, operation of the North Anna Power Station in
accordance with the proposed Technical Specifications changes will
not:
1. Involve a significant increase in the probability or
consequences of an accident previously evaluated.
This change reduces the power level at which the reactor may be
operated with one or more main steam safety valves (MSSVs)
inoperable to ensure that the secondary system is not
overpressurized during the most severe pressurization transient of
the secondary side. There is no change to the function of the MSSVs
by the proposed change and will not alter any accident analysis
assumptions or results. The proposed changes will provide
conservative power range neutron flux high trip setpoints such that
the maximum power level allowed for operation with inoperable MSSVs
is below the heat removing capability of the operable MSSVs.
Therefore, this change will not increase the probability of an
accident.
This change is consistent with the current accident analysis
assumptions for the MSSVs and does not change the containment
response for any design basis event. Therefore, no change in the
mitigation of an accident will result from this proposed change and
no change will occur in the consequences of any accident currently
analyzed.
2. Create the possibility of a new or different kind of accident
from any accident previous[ly] evaluated.
Since the implementation of the proposed changes to the
setpoints will not require hardware modifications (i.e., alterations
to plant configuration), operation of the facilities with these
proposed Technical Specifications does not create the possibility
for any new or different kind of accident which has not already been
evaluated.
The proposed revision to the Technical Specifications will not
result in any physical alteration to any plant system, nor would
there be a change in the method by which any safety-related system
performs its function. The design and operation of the main steam
system is not being changed.
These changes do not change the design, operation, or failure
modes of the main steam system. Therefore, the proposed change does
not create the possibility of a new or different kind of accident
from any accident previously evaluated.
3. Involve a significant reduction in a margin of safety.
The proposed change reduces the total energy of the reactor
coolant system that will ensure the ability of the MSSVs to perform
their intended function as assumed in the current accident analyses.
Correcting this non conservatism restores the margin of safety to
what was originally envisioned. In addition, the results of the
accident analyses which are documented in the UFSAR bound operation
under the proposed changes, so that there is no safety margin
reduction. Therefore, the proposed change does not involve a
significant reduction in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment 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., Riverfront Plaza,
East Tower, 951 E. Byrd Street, Richmond, Virginia 23219.
NRC Project Director: David B. Matthews.
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: September 19, 1995.
Description of amendment request: The proposed change would revise
the Technical Specifications (TS) for the North Anna Power Station,
Units No. 1 and No. 2 (NA-1&2). Specifically, the proposed change would
increase the surveillance test interval for the turbine reheat stop and
intercept valves to once per 18 months and extend the visual and
surface inspection interval to 60 months. The proposed change would
also remove the requirement to perform additional visual and surface
inspections on the remaining turbine overspeed protection system
control valves of that type when unacceptable flaws or excessive
corrosion are identified which can be directly attributed to a service
condition specific to the inspected valve.
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:
Specifically, operation of the North Anna Power Station in
accordance with the proposed Technical Specifications changes will
not:
1. Involve a significant increase in the probability or
consequences of an accident previously evaluated.
No new or unique accident precursors are introduced by these
changes in surveillance requirements. The probability of turbine
missile ejection with an extended 18-month test interval for the
reheat stop and intercept valves has been determined to be within
the applicable acceptance criteria.
The heavy hub design of the turbine rotors provides further
assurance that the probability of the ejection of destructive
missiles remains minimal.
Based upon the results of the probabilistic evaluation, the
probability of a turbine generated missile is less than 10-5
per year which the Commission has endorsed as the acceptable level
for turbine operation.
The reheat stop and intercept valve inspection interval
extension and the elimination of the additional visual/surface
inspections do not change the design, operation, or failure modes of
the valves and other components in the turbine overspeed protection
system.
Therefore, these changes do not involve a significant increase
in the probability or consequences of an accident previously
evaluated.
The demonstrated high reliability of the turbine reheat stop and
intercept valves and the verification of the operability of the
other turbine control valves provide adequate assurance that the
turbine overspeed protection system will operate as designed, if
needed. Turbine reheat stop and intercept valve testing performed to
date has demonstrated the reliability of these valves. In addition,
the operability of the other turbine valves (i.e., turbine throttle
valves and governor valves) will continue to be verified every 31
days or as required by the Technical Specifications.
2. Create the possibility of a new or different kind of accident
from any accident previous[ly] evaluated.
Since the implementation of the proposed change to the
surveillance requirements will not require hardware modifications
(i.e., alterations to plant configuration), operation of the
facilities with these proposed Technical Specifications does not
create the possibility for any new or different kind of accident
which has not already been evaluated in the Updated Final Safety
Analysis Report (UFSAR). In addition, the results of the
probabilistic evaluation indicate that no additional transients have
been introduced.
The proposed revision to the Technical Specifications will not
result in any physical alteration to any plant system, nor would
there be a change in the method by which any safety-related system
performs its function. The design and operation of the turbine
overspeed protection and turbine control systems are not being
changed.
The proposed Technical Specifications changes do not affect the
design, operation, or failure modes of the valves and other
components of the turbine overspeed protection system.
[[Page 54726]]
Therefore, the proposed change does not create the possibility
of a new or different kind of accident from any accident previously
evaluated.
3. Involve a significant reduction in a margin of safety.
The proposed changes would not reduce the margin of safety as
defined in the basis for any Technical Specifications. The design
and operation of the turbine overspeed protection and turbine
control systems are not being changed and the operability of the
turbine reheat stop and intercept valves will be demonstrated on a
refueling outage basis. In addition, the results of the accident
analyses which are documented in the UFSAR continue to bound
operation under the proposed changes, so that there is no safety
margin reduction. Therefore, the proposed change does not involve a
significant reduction in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment 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: David B. Matthews.
Previously Published Notices of Consideration of Issuance of Amendments
to Facility Operating Licenses, Proposed No Significant Hazards
Consideration Determination, and Opportunity for a Hearing
The following notices were previously published as separate
individual notices. The notice content was the same as above. They were
published as individual notices either because time did not allow the
Commission to wait for this biweekly notice or because the action
involved exigent circumstances. They are repeated here because the
biweekly notice lists all amendments issued or proposed to be issued
involving no significant hazards consideration.
For details, see the individual notice in the Federal Register on
the day and page cited. This notice does not extend the notice period
of the original notice.
Commonwealth Edison Company, Docket Nos. STN 50-456 and STN 50-457,
Braidwood Station, Unit Nos. 1 and 2, Will County, Illinois
Date of amendment request: August 15, 1995.
Description of amendment request: This application to revise the
Braidwood, Unit 1, Technical Specifications (TSs) proposes to continue
to use the voltage-based repair criteria which were added to the
Braidwood, Unit 1, TSs by a license amendment issued on August 18,
1994. This August 15, 1995, request will be considered by the staff
only in the event that the staff can not reach a timely decision on
your pending request for license amendments dated September 1, 1995, to
raise the present lower voltage repair limit from 1.0 volt to 3.0
volts.
Date of publication of individual notice in Federal Register:
October 5, 1995 (60 FR 52222).
Expiration date of individual notice: November 6, 1995.
Local Public Document Room location: 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 amendment request: September 15, 1995.
Description of amendment request: To close out open items
identified by the NRC staff's review of the upgrade of sections 1.0, 3/
4.4, 3/4.10, and 5.0 of the Dresden and Quad Cities Technical
Specifications to the BWR Standard Technical Specifications.
Date of publication of individual notice in Federal Register:
October 5, 1995 (60 FR 52220).
Expiration date of individual notice: November 6, 1995.
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. 50-254 and 50-265, Quad Cities
Nuclear Power Station, Units 1 and 2, Rock Island County, Illinois
Date of amendment request: September 20, 1995.
Description of amendment request: The proposed amendment would
upgrade the Quad Cities TS to the Standard Technical Specifications
(STS) contained in NUREG-0123. The Technical Specification Upgrade
Program (TSUP) is not a complete adaption of the STS. The TS upgrade
focuses on (1) integrating additional information such as equipment
operability requirements during shutdown conditions, (2) clarifying
requirements such as limiting conditions for operation and action
statements utilizing STS terminology, (3) deleting superseded
requirements and modifications to the TS based on the licensee's
responses to Generic Letters (GL), and (4) relocating specific items to
more appropriate TS locations. The September 20, 1995, application
proposed to upgrade only Section 6.0 (Administrative Controls) of the
Quad Cities TS.
Date of publication of individual notice in Federal Register:
October 5, 1995 (60 FR 52226).
Expiration date of individual notice: November 6, 1995.
Local Public Document Room location: Dixon Public Library, 221
Hennepin Avenue, Dixon, Illinois 61021.
North Atlantic Energy Service Corporation, Docket No. 50-443, Seabrook
Station, Unit No. 1, Rockingham County, New Hampshire
Date of amendment request: September 5, 1995.
Description of amendment request: The proposed amendment would
modify the Appendix A Technical Specifications (TSs) for the Turbine
Cycle Safety Valves. Specifically, the proposed amendment would change
Seabrook Station Appendix A Technical Specification Table 3.7-1 to
reduce the maximum allowable Power Range Neutron Flux--High setpoints
with inoperable Main Steam Safety Valves (MSSVs) and Table 3.7-2 to
reduce the opening setpoints of the MSSVs.
Date of publication of individual notice in Federal Register:
October 2, 1995 (60 FR 51505).
Expiration date of individual notice: November 1, 1995.
Local Public Document Room location: Exeter Public Library,
Founders Park, Exeter, NH 03833.
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
[[Page 54727]]
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: August 3, 1995.
Brief description of amendments: These amendments add the
analytical method supplement entitled ``Fuel Rod Maximum Allowable Gas
Pressure,'' CEN-372-P-A, dated May 1990, and its associated NRC Safety
Evaluation, dated April 10, 1990, to the list of analytical methods in
Technical Specification 6.9.1.10 used to determine the Palo Verde
Nuclear Generating Station core operating limits.
Date of issuance: October 4, 1995.
Effective date: October 4, 1995, to be implemented prior to startup
from RF06 for Units 1 and 2, and RF5 for Unit 3.
Amendment Nos.: Unit 1--Amendment No. 101; Unit 2--Amendment No.
89; Unit 3--Amendment No. 72.
Facility Operating License Nos. NPF-41, NPF-51, and NPF-74: The
amendments revised the Technical Specifications.
Date of initial notice in Federal Register: August 30, 1995 (60 FR
45173) The Commission's related evaluation of the amendments is
contained in a Safety Evaluation dated October 4, 1995.
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: June 2, 1995.
Brief description of amendments: The amendments revise the
tolerances for the pressurizer safety valve as-found acceptance
criterion.
Date of issuance: September 26, 1995.
Effective date: As of the date of issuance to be implemented within
30 days.
Amendment Nos.: 206 and 184.
Facility Operating License Nos. DPR-53 and DPR-69: Amendments
revised the Technical Specifications.
Date of initial notice in Federal Register: July 5, 1995 (60 FR
35060) The Commission's related evaluation of these amendments is
contained in a Safety Evaluation dated September 26, 1995.
No significant hazards consideration comments received: No.
Local Public Document Room location: Calvert County Library, Prince
Frederick, Maryland 20678.
Baltimore Gas and Electric Company, Docket Nos. 50-317 and 50-318,
Calvert Cliffs Nuclear Power Plant, Unit No. 1, Calvert County,
Maryland
Date of application for amendments: January 31, 1995.
Brief description of amendments: The amendments revise the
Technical Specifications (TSs) to increase the amount of Trisodium
Phosphate Dodecahydrate located in the containment sump baskets which
is required to be verified by TS surveillance. The test requirements
for verifying that the appropriate pH (acidity/alkalinity) would be
maintained in the containment sump water following a design-basis
accident are moved from the TSs to the TS Bases section; however, the
requirement to perform the test remains in the TSs. The associated TS
Bases sections are updated to reflect the changes.
Date of issuance: October 5, 1995.
Effective date: As of the date of issuance to be implemented within
30 days.
Amendment Nos.: 207 and 185.
Facility Operating License No. DPR-53 and DPR-69: Amendment revised
the Technical Specifications.
Date of initial notice in Federal Register: March 15, 1995 (60 FR
14016) The Commission's related evaluation of these amendments is
contained in a Safety Evaluation dated October 5, 1995.
No significant hazards consideration comments received: No.
Local Public Document Room location: Calvert County Library, Prince
Frederick, Maryland 20678
Commonwealth Edison Company, Docket Nos. 50-295 and 50-304, Zion
Nuclear Power Station Units 1 and 2, Lake County, Illinois
Date of application for amendments: March 8, 1995, as supplemented
on June 1, 1995.
Brief description of amendments: The amendments revise the
secondary undervoltage setpoint.
Date of issuance: October 2, 1995.
Effective date: October 2, 1995
Amendment Nos.: 169 and 156.
Facility Operating License Nos. DPR-39 and DPR-48: The amendments
revised the Technical Specifications.
Date of initial notice in Federal Register: August 30, 1995 (60 FR
45178) The Commission's related evaluation of the amendments is
contained in a Safety Evaluation dated October 2, 1995.
No significant hazards consideration comments received: No.
Local Public Document Room location: Waukegan Public Library, 128
N. County Street, Waukegan, Illinois 60085.
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: February 18, 1994, as
supplemented June 3, November 1, December 2, December 14, and December
16, 1994, and August 25, 1995.
Brief description of amendment: The amendment revises the
surveillance intervals for the Boric Acid Tank Level, the Service Water
Inlet Temperature Monitor Instrument, the Boric Acid Makeup Flow
System, the Plant Noble Gas Activity Monitor, the Condenser Evacuation
System Activity Monitor, the Low Turbine Auto Stop Oil Pressure Trip,
the 6.9 kv Undervoltage Monitor, the Sampler Flow Rate Monitor, and the
Refueling Water Storage Tank.
Date of issuance: October 12, 1995.
Effective date: As of the date of issuance to be implemented within
30 days.
Amendment No.: 184.
[[Page 54728]]
Facility Operating License No. DPR-26: Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: April 28, 1994 (59 FR
22003) The Commission's related evaluation of the amendment is
contained in a Safety Evaluation dated October 12, 1995.
No significant hazards consideration comments received: No.
Local Public Document Room location: White Plains Public Library,
100 Martine Avenue, White Plains, New York 10610.
Entergy Operations, Inc., Docket No. 50-368, Arkansas Nuclear One, Unit
No. 2, Pope County, Arkansas
Date of application for amendment: March 17, 1995.
Brief description of amendment: The amendment revises requirements
associated with channel functional tests of the core protection
calculator following a high temperature alarm.
Date of issuance: October 11, 1995.
Effective date: October 11, 1995, to be implemented within 30 days.
Amendment No.: 168.
Facility Operating License No. NPF-6. Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: August 2, 1995 (60 FR
39437) The Commission's related evaluation of the amendment is
contained in a Safety Evaluation dated October 11, 1995.
No significant hazards consideration comments received: No.
Local Public Document Room location: Tomlinson Library, Arkansas
Tech University, Russellville, AR 72801.
Entergy Operations, Inc., Docket No. 50-382, Waterford Steam Electric
Station, Unit 3, St. Charles Parish, Louisiana
Date of amendment request: July 18, 1991, as supplemented by
letters dated March 16, and December 2, 1994, and March 9, and August
30, 1995.
Brief description of amendment: The amendment changes the Appendix
A TSs by subdividing TS 3/4.7.6, ``Control Room Air Conditioning
System,'' into five separate TSs covering the following three distinct
functions: control room emergency air filtration, control room air
temperature, and control room isolation and pressurization. The
amendment also changes the Bases sections of the TS to reflect the
above changes.
Date of issuance: October 4, 1995.
Effective date: October 4, 1995.
Amendment No.: 115.
Facility Operating License No. NPF-38: Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: September 4, 1991 (56
FR 43808) and July 6, 1995 (60 FR 29875).
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated October 4, 1995.
No significant hazards consideration comments received: No.
Local Public Document Room location: University of New Orleans
Library, Louisiana Collection, Lakefront, New Orleans, LA 70122.
GPU Nuclear Corporation, et al., Docket No. 50-289, Three Mile Island
Nuclear Station, Unit No. 1, Dauphin County, Pennsylvania
Date of application for amendment: June 1, 1995, as supplemented
August 23, 1995.
Brief description of amendment: The amendment changes the Technical
Specifications to relocate the procedural details of the Radiological
Effluent Technical Specifications to the Offsite Dose Calculation
Manual. With these changes, the specifications related to RETS
reporting requirements were simplified and changes to the definition of
the ODCM were made to make the definition consistent with the
amendment.
Date of Issuance: October 2, 1995.
Effective date: As of the date of issuance to be implemented within
120 days.
Amendment No.: 197.
Facility Operating License No. DPR-50. Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: July 5, 1995 (60 FR
35078) The August 23, 1995, letter provided supplemental information
that did not change the initial proposed no significant hazards
consideration determination.
The Commission's related evaluation of this amendment is contained
in a Safety Evaluation dated October 2, 1995.
No significant hazards consideration comments received: No.
Local Public Document Room location: Law/Government Publications
Section, State Library of Pennsylvania, (Regional Depository) Walnut
Street and Commonwealth Avenue, Box 1601, Harrisburg, PA 17105.
GPU Nuclear Corporation, et al., Docket No. 50-289, Three Mile Island
Nuclear Station, Unit No. 1, Dauphin County, Pennsylvania
Date of application for amendment: January 16, 1995, as
supplemented June 22 and September 20, 1995.
Brief description of amendment: The amendment revises the Technical
Specifications for TMI-1 to incorporate seven improvements from the
Revised Standard Technical Specifications for Babcock & Wilcox Nuclear
Power Plants (NUREG-1430). The amendment also changes the Bases
incorporating the results of analyses to support allowance for drift of
the Pressurizer Code Safety Valve setpoint. The remaining portion of
the request relating to revisions to Control Room Emergency Ventilation
system are being reviewed separately.
Date of Issuance: October 10, 1995.
Effective date: October 10, 1995.
Amendment No.: 198.
Facility Operating License No. DPR-50. Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: March 15, 1995 (60 FR
14021). The June 22 and September 20, 1995, letter provided clarifying
information that did not change the initial proposed no significant
hazards consideration determination.
The Commission's related evaluation of this amendment is contained
in a Safety Evaluation dated October 10, 1995.
No significant hazards consideration comments received: No.
Local Public Document Room location: Law/Government Publications
Section, State Library of Pennsylvania, (Regional Depository) Walnut
Street and Commonwealth Avenue, Box 1601, Harrisburg, PA 17105.
Indiana Michigan Power Company, Docket Nos. 50-315 and 50-316, Donald
C. Cook Nuclear Plant, Unit Nos. 1 and 2, Berrien County, Michigan
Date of application for amendments: November 12, 1993, as
supplemented November 18, 1994, May 30, 1995, and August 8, 1995.
Brief description of amendments: The amendments delete from the
Technical Specifications the sections and tables entitled ``Component
Cyclic or Transient Limits'' and relocate the information to the
Updated Final Safety Analysis Report.
Date of issuance: September 28, 1995.
Effective date: September 28, 1995, with full implementation within
45 days.
Amendment Nos.: 201 and 186.
Facility Operating License Nos. DPR-58 and DPR-74. Amendments
revised the Technical Specifications.
Date of initial notice in Federal Register: December 22, 1993 (58
FR 67849). The November 18, 1994, May 30, 1995, and August 8, 1995,
supplements provided clarifying information and corrections to
additional pages which referenced the table to be deleted. This
information was within the scope of the original
[[Page 54729]]
application and did not change the staff's initial proposed no
significant hazards consideration determination. The Commission's
related evaluation of the amendments is contained in a Safety
Evaluation dated September 28, 1995.
No significant hazards consideration comments received: No.
Local Public Document Room location: Maud Preston Palenske Memorial
Library, 500 Market Street, St. Joseph, Michigan 49085.
Indiana Michigan Power Company, Docket Nos. 50-315 and 50-316, Donald
C. Cook Nuclear Plant, Unit Nos. 1 and 2, Berrien County, Michigan
Date of application for amendments: May 26, 1995.
Brief description of amendments: The amendments modify Technical
Specification Sections 3/4.3.1 and 3/4.3.2 and their accompanying
Bases, to relocate the tables of response time limits for the reactor
trip system and engineered safety feature acutation system
instrumentation to the Updated Final Safety Analysis Report.
Date of issuance: October 10, 1995.
Effective date: October 10, 1995.
Amendment Nos.: 202 and 187.
Facility Operating License Nos. DPR-58 and DPR-74. Amendments
revised the Technical Specifications.
Date of initial notice in Federal Register: July 5, 1995 (60 FR
35082) The Commission's related evaluation of the amendments is
contained in a Safety Evaluation dated October 10, 1995.
No significant hazards consideration comments received: No.
Local Public Document Room location: Maud Preston Palenske Memorial
Library, 500 Market Street, St. Joseph, Michigan 49085.
Maine Yankee Atomic Power Company, Docket No. 50-309, Maine Yankee
Atomic Power Station, Lincoln County, Maine
Date of application for amendment: May 5, 1995.
Brief description of amendment: The amendment revises the
surveillance frequency of radiation area, and effluent and process
monitors from monthly to quarterly; and the required frequency for
minimum exercise of control element assemblies also from monthly to
quarterly.
Date of issuance: October 2, 1995.
Effective date: As of the date of issuance, to be implemented
within 30 days.
Amendment No.: 153.
Facility Operating License No. DPR-36: Amendment revised the
Technical Specifications and/or License.
Date of initial notice in Federal Register: August 30, 1995 (60 FR
45179). The Commission's related evaluation of the amendment is
contained in a Safety Evaluation dated October 2, 1995.
No significant hazards consideration comments received: No.
Local Public Document Room location: Wiscasset Public Library, High
Street, P.O. Box 367, Wiscasset, ME 04578.
Northern States Power Company, Docket No. 50-263, Monticello Nuclear
Generating Plant, Wright County, Minnesota
Date of application for amendment: June 8, 1994, as superseded by
letter dated April 20, 1995, and supplemented by letter dated August
18, 1995.
Brief description of amendment: The amendment revises Sections 3.7/
4.7, which pertain to the standby gas treatment system (SGTS) and
secondary containment. The amendment revises the surveillance
requirements for both SGTS and the secondary containment and revises
the performance requirements for the SGTS filters and process stream
electric heaters.
Date of issuance: October 2, 1995.
Effective date: October 2, 1995.
Amendment No.: 94.
Facility Operating License No. DPR-22. Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: July 20, 1994 (59 FR
37075). The April 20 and August 18, 1995, submittals provided
clarifying information within the scope of the original submittal and
did not change the staff's initial proposed no significant hazards
consideration determination.
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated October 2, 1995.
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 Nos. 50-282 and 50-306, Prairie
Island Nuclear Generating Plant, Unit Nos. 1 and 2, Goodhue County,
Minnesota.
Date of application for amendments: July 11, 1994, as supplemented
April 18, 1995 (supersedes the February 10, 1993, application).
Brief description of amendments: The amendments change license
condition 2.C.(4) of each license to conform to the standard fire
protection license condition as stated in Generic Letter 86-10. In
addition, the amendments delete fire protection program elements from
the Technical Specifications and incorporate, by reference, the NRC-
approved Fire Protection Program and major commitments, including the
fire hazards analysis, into the Updated Safety Analysis Report.
Date of issuance: October 6, 1995.
Effective date: October 6, 1995, with full implementation within 30
days.
Amendment Nos.: 120 and 113.
Facility Operating License Nos. DPR-42 and DPR-60. Amendments
revised the Technical Specifications.
Date of initial notice in Federal Register: December 21, 1994 (59
FR 65818). The April 18, 1995, letter provided clarifying information
within the scope of the original submittal and did not change the
staff's initial proposed no significant hazards consideration
determination. The Commission's related evaluation of the amendments is
contained in a Safety Evaluation dated October 6, 1995.
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.
Sacramento Municipal Utility District, Docket No. 50-312, Rancho Seco
Nuclear Generating Station, Sacramento County, California
Date of application for amendment: June 20, 1995.
Brief description of amendment: This amendment modifies the
technical specifications on spent fuel storage building load handling
limits to allow the placement of the top shield plug on a dry shielded
canister containing spent fuel which is being prepared for transfer to
the Rancho Seco Independent Spent Fuel Storage Installation.
Date of issuance: October 5, 1995.
Effective date: October 5, 1995.
Amendment No.: 123.
Facility Operating License No. NPF-1: The amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: August 30, 1995 (60 FR
45184). The Commission's related evaluation of the amendment is
contained in a Safety Evaluation dated October 5, 1995.
No significant hazards consideration comments received: No.
Local Public Document Room location: Central Library, Government
Documents, 828 I Street, Sacramento, California 95814.
[[Page 54730]]
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: June 30, 1995, as supplemented
on August 11, 1995.
Brief description of amendment: The amendment revises the Technical
Specifications (TS) for the pressurizer power operated relief valves to
follow the NRC's guidance of Generic Letter 90-06 (Generic Issue 70),
and the improved Westinghouse Standard TS (NUREG-1431, Rev. 1).
Date of issuance: September 18, 1995.
Effective date: September 18, 1995.
Amendment No.: 129.
Facility Operating License No. NPF-12. Amendment revises the TS.
Date of initial notice in Federal Register: Auust 16, 1995 (60 FR
42608).
The August 11, 1995, supplemental letter corrected an error in the
original submittal and did not change the initial proposed no
significant hazards consideration. The Commission's related evaluation
of the amendment is contained in a Safety Evaluation dated September
18, 1995.
No significant hazards consideration comments received: No.
Local Public Document Room location: Fairfield County Library, 300
Washington Street, Winnsboro, SC 29180.
Southern Nuclear Operating Company, Inc., Docket Nos. 50-348 and 50-
364, Joseph M. Farley Nuclear Plant, Units 1 and 2, Houston County,
Alabama
Date of amendments request: August 17, 1994, as supplemented by
letters dated June 15 and August 11, 1995.
Brief Description of amendments: The amendments eliminate periodic
pressure sensor response time testing surveillance requirements for
specific Reactor Trip System and Engineered Safety Feature Actuation
System instrumentation specified in Technical Specification Sections
4.3.1.3 and 4.3.2.3.
Date of issuance: September 28, 1995.
Effective date: As of the date of issuance to be implemented within
30 days from the date of issuance.
Amendment Nos.: 116 and 108.
Facility Operating License Nos. NPF-2 and NPF-8. Amendments revise
the Technical Specifications.
Date of initial notice in Federal Register: September 28, 1994 (59
FR 49434) The June 15 and August 11, 1995, letters provided clarifying
information that did not change the scope of the August 17, 1994,
application and the initial proposed no significant hazards
consideration determination.
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated September 28, 1995.
No significant hazards consideration comments received: No.
Local Public Document Room location: Houston-Love Memorial Library,
212 W. Burdeshaw Street, Post Office Box 1369, Dothan, Alabama 36302.
Southern Nuclear Operating Company, Inc., Docket No. 50-348, Joseph M.
Farley Nuclear Plant, Unit 1, Houston County, Alabama
Date of amendment request: December 7, 1994, as supplemented by
letter dated May 31, 1995.
Brief Description of amendment: The amendment revised Farley Unit 1
Technical Specifications 4.4.6.2, 4.4.6.4, 4.4.6.5, 3.4.7.2, and 3.4.9
for Cycle 14 operation to permit the use of steam generator tube repair
criteria for defects confined within the thickness of the tube support
plate.
Date of issuance: September 28, 1995.
Effective date: As of the date of issuance to be implemented prior
to the start of Unit 1, Cycle 14 operation.
Amendment No.: 117.
Facility Operating License No. NPF-2: Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: February 15, 1995 (60
FR 8754) The May 31, 1995, letter provided clarifying information that
did not change the scope of the December 7, 1994, application and the
initial proposed no significant hazards consideration determination.
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated September 28, 1995.
No significant hazards consideration comments received: No.
Local Public Document Room location: Houston-Love Memorial Library,
212 W. Burdeshaw Street, Post Office Box 1369, Dothan, Alabama 36302.
Tennessee Valley Authority, Docket Nos. 50-327 and 50-328, Sequoyah
Nuclear Plant, Units 1 and 2, Hamilton County, Tennessee
Date of application for amendments: August 7, 1995 (TS 95-18).
Brief description of amendments: The amendments revise the titles
of various administrative positions found in Section 6.0 of the
Technical Specifications.
Date of issuance: October 2, 1995.
Effective date: October 2, 1995,
Amendment Nos.: 212 and 202.
Facility Operating License Nos. DPR-77 and DPR-79: Amendments
revise the technical specifications.
Date of initial notice in Federal Register: August 30, 1995 (60 FR
45186) The Commission's related evaluation of the amendment is
contained in a Safety Evaluation dated October 2, 1995.
No significant hazards consideration comments received: None.
Local Public Document Room location: Chattanooga-Hamilton County
Library, 1101 Broad Street, Chattanooga, Tennessee 37402.
Tennessee Valley Authority, Docket Nos. 50-327 and 50-328, Sequoyah
Nuclear Plant, Units 1 and 2, Hamilton County, Tennessee
Date of application for amendments: August 7, 1995 (TS 95-12).
Brief description of amendments: The amendments correct various
editorial errors in the text of the technical specifications and remove
provisions that have expired or are no longer applicable.
Date of issuance: October 4, 1995.
Effective date: October 4, 1995.
Amendment Nos.: 213 and 203.
Facility Operating License Nos. DPR-77 and DPR-79: Amendments
revise the technical specifications.
Date of initial notice in Federal Register: August 30, 1995 (60 FR
45185) The Commission's related evaluation of the amendment is
contained in a Safety Evaluation dated October 4, 1995.
No significant hazards consideration comments received: None.
Local Public Document Room location: Chattanooga-Hamilton County
Library, 1101 Broad Street, Chattanooga, Tennessee 37402.
Tennessee Valley Authority, Docket No. 50-327, Sequoyah Nuclear Plant,
Unit 1, Hamilton County, Tennessee
Date of application for amendment: July 19, 1995, superseded
September 7, 1995 and supplemented September 15 and 26, 1995 (TS 95-
15).
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 similar to guidance given in Generic Letter (GL) 95-
05, ``Voltage-Based Repair Criteria for Westinghouse Steam Generator
Tubes Affected by Outside Diameter Stress Corrosion Cracking.''
Date of issuance: October 11, 1995.
Effective date: October 11, 1995.
Amendment No.: 214.
[[Page 54731]]
Facility Operating License Nos. DPR-77: Amendment revises the
technical specifications.
Date of initial notice in Federal Register: August 1, 1995 (60 FR
39189) The letters dated September 7, 15 and 26, 1995 provided
information that did not change the initial proposed no significant
hazards consideration. The Commission's related evaluation of the
amendment is contained in a Safety Evaluation dated October 11, 1995.
No significant hazards consideration comments received: No.
Local Public Document Room location: Chattanooga-Hamilton County
Library, 1101 Broad Street, Chattanooga, Tennessee 37402.
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: March 30, 1995, as supplemented
August 24, 1995
Brief description of amendments: The amendments revise the North
Anna 1 and 2 Technical Specifications to allow one of the two service
water loops to be isolated from the component cooling water head
exchangers during power operations in order to refurbish the isolated
service water headers.
Date of issuance: October 11, 1995.
Effective date: October 11, 1995.
Amendment Nos.: 194 and 175.
Facility Operating License Nos. NPF-4 and NPF-7. Amendments revised
the Technical Specifications.
Date of initial notice in Federal Register: May 10, 1995 (60 FR
24923). The Commission's related evaluation of the amendments is
contained in a Safety Evaluation dated October 11, 1995.
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.
Washington Public Power Supply System, Docket No. 50-397, Nuclear
Project No. 2, Benton County, Washington
Date of application for amendment: December 6, 1993.
Brief description of amendment: The amendment changes the
surveillance requirements in Technical Specification 4.6.6.1.b.3 to
provide more appropriate acceptance criteria for demonstrating
operability of the primary containment hydrogen recombiner systems.
Date of issuance: October 5, 1995.
Effective date: October 5, 1995, to be implemented within 30 days
of issuance.
Amendment No.: 142.
Facility Operating License No. NPF-21: The amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: July 6, 1994 (59 FR
34670). The Commission's related evaluation of the amendment is
contained in a Safety Evaluation dated October 5, 1995.
No significant hazards consideration comments received: No.
Local Public Document Room location: Richland Public Library, 955
Northgate Street, Richland, Washington 99352.
Wolf Creek Nuclear Operating Corporation, Docket No. 50-482, Wolf Creek
Generating Station, Coffey County, Kansas
Date of amendment request: May 24, 1994, as supplemented by letter
dated April 6, 1995.
Brief description of amendment: This amendment revises the
technical specifications (TS) to implement the NRC's revised 10 CFR
50.36 on technical specification improvements for nuclear power
reactors. Specifications that do not meet any of the four criteria or
regulatory requirements related to inclusion in the TS are relocated to
Chapter 16 of the Updated Safety Analysis Report.
Date of issuance: October 2, 1995.
Effective date: October 2, 1995, to be implemented within 120 days
from the date of issuance.
Amendment No.: 89.
Facility Operating License No. NPF-42. The amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: July 6, 1994 (59 FR
34671). The April 6, 1995, supplemental letter provided additional
clarifying information and did not change the initial no significant
hazards consideration determination.
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated October 2, 1995.
No significant hazards consideration comments received: No.
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.
Wolf Creek Nuclear Operating Corporation, Docket No. 50-482, Wolf Creek
Generating Station, Coffey County, Kansas
Date of amendment request: July 25, 1995.
Brief description of amendment: The amendment deletes a clause from
Section 4.0.5a, ``Surveillance Requirements for Inservice Inspection
and Testing Program.'' This clause required prior NRC approval before
implementation of a relief request upon finding an ASME Code
requirement impractical because of prohibitive dose rates or
limitations in the design, construction, or system configuration.
Date of issuance: October 4, 1995.
Effective date: October 4, 1995, to be implemented within 30 days
of issuance.
Amendment No.: 90.
Facility Operating License No. NPF-42. The amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: August 30, 1995 (60 FR
45191). The Commission's related evaluation of the amendment is
contained in a Safety Evaluation dated October 4, 1995.
No significant hazards consideration comments received: No.
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
[[Page 54732]]
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.
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 November 24, 1995, 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
[[Page 54733]]
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 the
factors specified in 10 CFR 2.714(a)(1) (i)-(v) and 2.714(d).
Detroit Edison Company, Docket No. 50-341, Fermi-2, Monroe County,
Michigan
Date of application for amendment: October 2, 1995.
Brief description of amendment: The amendment revises the Technical
Specifications to allow deferral until the next plant outage of certain
portions of logic system functional surveillance testing for the diesel
generator 480-volt load sequencer and output breaker reclosure logic
circuitry.
Date of issuance: October 13, 1995.
Effective date: October 13, 1995, with full implementation within
45 days.
Amendment No.: 105.
Facility Operating License No. NPF-43: Amendment revises 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
October 13, 1995.
Local Public Document Room location: Monroe County Library System,
3700 South Custer Road, Monroe, Michigan 48161.
Attorney for licensee: John Flynn, Esq., Detroit Edison Company,
2000 Second Avenue, Detroit, Michigan 48226.
NRC Project Director: Brian E. Holian, 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 application for amendments: September 30, 1995.
Brief description of amendments: The amendments increase the
setpoint tolerance of the main steam safety valves (MSSVs) from plus or
minus 1 percent to plus or minus 3 percent, with the exception that the
lowest set MSSVs would have a tolerance of -2 percent/+3 percent.
Date of issuance: October 1, 1995.
Effective date: October 1, 1995.
Amendment Nos.: Unit 1--Amendment No. 108; Unit 2--Amendment No.
107.
Facility Operating License Nos. DPR-80 and DPR-82: The amendments
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
October 1, 1995.
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.
Dated at Rockville, Maryland, this 18th day of October 1995.
For the Nuclear Regulatory Commission.
Elinor G. Adensam,
Deputy Director, Division of Reactor Projects--III/IV, Office of
Nuclear Reactor Regulation.
[FR Doc. 95-26275 Filed 10-24-95; 8:45 am]
BILLING CODE 7590-01-P
