Biweekly Notice
Federal RegisterOct 11, 1995
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[[Page 52927]]
NUCLEAR REGULATORY COMMISSION
Biweekly Notice
Applications and Amendments to Facility Operating LicensesInvolving
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 16, through September 28, 1995.
The last biweekly notice was published on Septmeber 27, 1995 (60 FR
49929).
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 10, 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
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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.
Connecticut Yankee Atomic Power Company, Docket No. 50-213, Haddam
Neck Plant, Middlesex County, Connecticut
Date of amendment request: August 10, 1995
Description of amendment request: The proposed amendment will add a
footnote to Technical Specification (TS) Section 3/4.4.3,
``Pressurizer,'' to allow the pressurizer level to be controlled,
outside of the programmed level, between 25 to 50 percent, plus or
minus 5 percent in Mode 3 when the reactor coolant system is borated to
the required Mode 5 concentrations.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
...The proposed change does not involve an SHC because the
change would not:
1. Involve a significant increase in the probability or
consequences of an accident previously evaluated.
The design basis accidents analyzed in Mode 3 are steam line
break, control rod withdrawal from subcritical, boron dilution and
control rod ejection. Of these four analyzed accidents, the relaxing
of the pressurizer level requirement can only impact the steam line
break accident analyses. The initial pressurizer level can impact
the timing of the safety injection signal and the subsequent boron
addition from the HPSI [high pressure safety injection] system. The
proposed change requires that the boron concentration be equal to
the Mode 5 required concentration in order for the pressurizer level
to be higher than the current requirement. The Mode 5 boron
concentration ensures that there is sufficient negative reactivity
in the core due to boron that a steam line break from this condition
would not need the boron addition from the HPSI system and would be
bounded by the design basis analyses. Thus the proposed change
cannot increase the probability or consequences of the design basis
accidents.
2. Create the possibility of a new or different kind of accident
from any previously analyzed.
The proposed change only modifies the Mode 3 pressurizer level
requirement. This change does not impact the lower bound but
provides flexibility to the plant operators in the maximum
pressurizer level. The upper limit still provides margin to
pressurizer overfill. This cannot cause an accident nor introduce a
new type of malfunction. The modified level would allow for a higher
initial pressurizer level in Mode 3. This higher level is already
used in the accident analyses which result in an increase in
pressurizer level. Therefore, the change does not modify the plant's
response to accidents.
3. Involve a significant reduction in the margin of safety.
The proposed change is consistent with or bounded by the design
basis analyses. The higher shutdown margin required in order to
relax the upper bound of the pressurizer level assures that a steam
line break from these conditions is bounded by the design basis
analyses. Therefore, the proposed change cannot impact 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: Russell Library, 123 Broad
Street, Middletown, CT 06457.
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
Duke Power Company, Docket Nos. 50-269, 50-270 and 50-287, Oconee
Nuclear Station, Units 1, 2 and 3, Oconee County, South Carolina
Date of 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
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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: Oconee County Library, 501
West South Broad Street, Walhalla, South Carolina 29691
Attorney for licensee: J. Michael McGarry, III, Winston and Strawn,
1200 17th Street, NW., Washington, DC 20036
NRC Project Director: Herbert N. Berkow
Entergy Operations, Inc., et al., Docket No. 50-416, Grand Gulf
Nuclear Station, Unit 1, Claiborne County, Mississippi
Date of amendment request: November 9, 1994, as supplemented by
letter dated August 4, 1995
Description of amendment request: This supplement revises the
licensee's November 9, 1994, application by updating the request to
reflect implementation of the Improved Standard Technical
Specifications on March 20, 1995, and by deleting the request for a
definition of the term RECENTLY IRRADIATED FUEL. The proposed amendment
revises those specifications associated with various engineered safety
feature systems following a design basis fuel handling accident. The
proposed changes affect conditions where irradiated fuel is handled in
the primary or secondary containment and when fuel is handled over the
reactor vessel with fuel in the vessel. These changes are based on a
recent re-analysis of the fuel handling accident for Grand Gulf Nuclear
Station (GGNS).
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. The proposed changes do not significantly increase the
probability or consequences of an accident previously evaluated.
A new term to describe irradiated fuel is used to establish
operational conditions where specific activities represent
situations where significant radioactive releases can be postulated.
These operational conditions are consistent with the design basis
analysis. Because the equipment affected by the revised operational
conditions is not considered an initiator to any previously analyzed
accident, inoperability of the equipment cannot increase the
probability of any previously evaluated accident. The proposed
requirements in conjunction with existing administrative controls on
light loads, bounds the conditions of the current design basis fuel
handling accident analysis which concludes that the radiological
consequences are within the acceptance criteria of NUREG 0800,
Section 15.7.4 and General Design Criteria 19. Therefore, the
proposed changes do not significantly increase the probability or
consequences of any previously evaluated accident.
Based on the above, the proposed changes do not significantly
increase the probability or consequences of any accident previously
evaluated.
2. The proposed changes would not create the possibility of a
new or different kind of accident from any previous analyzed.
The new term to describe irradiated fuel is used to establish
operational conditions where specific activities represent
situations where significant radioactive releases can be postulated.
These operational conditions are consistent with the design basis
analysis. The proposed changes do not introduce any new modes of
plant operation and do not involve physical modification of the
plant. Therefore, the proposed changes do not create the possibility
of a new or different kind of accident from any previous analyzed.
Therefore, the proposed changes do not create the possibility of
a new or different kind of accident from any accident previously
analyzed.
3. The proposed changes do not involve a significant reduction
in a margin of safety.
The new term to describe irradiated fuel is used to establish
operational conditions where specific activities represent
situations where significant radioactive releases can be postulated.
These operational conditions are consistent with the design basis
analysis and are established such that the radiological consequences
are at or below the current GGNS licensing limit. Safety margins and
analytical conservatisms have been evaluated and are well
understood. Substantial margins are retained to ensure that the
analysis adequately bounds all postulated event scenarios. The
proposed change only eliminates the excess margin from the analysis.
The current margin of safety is retained.
Specifically, the margin of safety for the fuel handling
accident is the difference between the 10 CFR 100 limits and the
licensing limit defined by NUREG 0800, Section 15.7.4. With respect
to the control room personnel doses, the margin of safety is the
difference between the 10 CFR 100 limits and the licensing limit
defined by 10 CFR 50, Appendix A, Criterion 19 (GDC 19). Excess
margin is the difference between the postulated doses and the
corresponding licensing limit.
The proposed applicability continues to ensure that the
whole-body and thyroid dose at the exclusion area and low
population zone boundaries as well as control room, doses are at or
below the corresponding licensing limit. The margin of safety is
unchanged; therefore, the proposed changes do not involve a
significant reduction in a margin of safety.
Therefore, the proposed changes do not result in a significant
reduction in a margin of safety.
Based on the above evaluation, operation in accordance with the
proposed amendment involves no significant hazards considerations.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: Judge George W. Armstrong
Library, 220 S. Commerce Street, Natchez, MS 39120
Attorney for licensee: Nicholas S. Reynolds, Esquire, Winston and
Strawn, 1400 L Street, N.W., 12th Floor, Washington, DC 20005-3502
NRC Project Director: William D. Beckner
Entergy Operations, Inc., Docket No. 50-368, Arkansas Nuclear One,
Unit No. 2, Pope County, Arkansas
Date of amendment request: July 19, 1995
Description of amendment request: The proposed amendment reduces
requirements associated with the exercise frequency of control element
assemblies from once per 31 days to once per 92 days.
Basis for proposed no significant hazards consideration
determination:
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As required by 10 CFR 50.91(a), the licensee has provided its analysis
of the issue of no significant hazards consideration, which is
presented below:
1. Does not Involve a Significant Increase in the Probability or
Consequences of an Accident Previously Evaluated.
Changing the frequency of the control element assemblies (CEA)
exercise test surveillance introduces no new failure mechanism for
the system, so the consequences of a postulated stuck CEA are no
different than those previously evaluated.
As explained in NUREG-1366, ``Improvements to Technical
Specifications Surveillance Requirements,'' the purpose of this test
is to identify immovable CEAs. NUREG-1366 goes on to explain that
the majority of CEA problems are identified during the performance
of startup physics testing and during CEA withdrawal for startup,
not during the exercise test. The incidence of electrical
malfunctions which will still allow CEA insertion is much greater
than the incidence of mechanically bound CEAs. As stated in NUREG-
1366, there has only been one incidence of multiple CEAs failing to
fully insert upon a reactor trip (Point Beach Nuclear Plant, May
1985) and in this case the two affected CEAs partially inserted.
Based on this history, simply reducing the test frequency will not
increase the probability of a stuck CEA.
Therefore, this change does not involve a significant increase
in the probability or consequences of any accident previously
evaluated.
2. Does Not Create the Possibility of a New or Different Kind of
Accident from any Previously Evaluated.
Because the proposed change does not alter the design,
configuration, or method of operation of the plant, it does not
create the possibility of a new or different kind of accident from
any previously evaluated.
3. Does Not Involve a Significant Reduction in the Margin of
Safety.
The proposed change does not alter the acceptance criteria of
any surveillance requirements, alter any assumptions used in
accident analysis, change any actuation setpoints, nor allow
operations in any configuration not previously evaluated. This
change in surveillance frequency is based on a satisfactory
operating history of CEAs. Additionally, the number of problems
created by this test when compared with the number of problems
identified by this test indicate that reducing the test frequency
will have no adverse impact on the continued safe operation of the
unit.
Therefore, this change does not involve a significant reduction
in the margin of safety.
Therefore, based upon the reasoning presented above and the
previous discussion of the amendment request, Entergy Operations had
determined that the requested change does not involve a significant
hazards consideration.
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: September 11, 1995
Description of amendment request: The licensee proposes to change
Turkey Point Units 3 and 4 Technical Specifications (TS) to incorporate
line-item improvements to Specifications 3/4.8.1, ``Electrical Power
Systems-A.C. Sources,'' and the associated BASES. The licensee stated
that the proposed changes are consistent with the guidance provided by
the NRC in GL 93-05, ``Line-Item Technical Specifications Improvements
to Reduce Surveillance Requirements for Testing During Power
Operation,'' and the corresponding recommendations contained in NUREG-
1366, ``Improvements to Technical Specifications Surveillance
Requirements.''
In addition, line-item improvements are proposed following the
guidance in GL 94-01, ``Removal of Accelerated Testing and Special
Reporting Requirements for Emergency Diesel Generators.'' The
implementation of a maintenance program for monitoring and maintaining
Emergency Diesel Generator (EDG) performance for Turkey Point Units 3
and 4, consistent with the provisions of 10 CFR 50.65 ``Requirements
for Monitoring the Effectiveness of Maintenance at Nuclear Power
Plants'' and the associated guidance of Regulatory Guide (RG) 1.160
will be met by FPL within 90 days following issuance of the proposed
amendments.
The licensee also requested to revise the current wording used in
the Turkey Point Units 3 and 4 TS to require testing of remaining
required diesel generators ``[i]f the diesel generator became
inoperable due to any cause other than planned preventative
maintenance...''. The licensee requested that TS 3.8.1.1, ACTION
statements b. and c. be amended such that the word 'preventative' is
deleted. Deleting this wording will reduce unnecessary testing of
diesel generators as a result of planned corrective maintenance.
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 license amendments proposed for Turkey Point Units 3 and 4
will incorporate line-item Technical Specification (TS) improvements
for Emergency Diesel Generators (EDG) pursuant to guidance provided
in Generic Letters (GL) 93-05 and 94-01. The EDGs are not accident
initiators, the proposed TS changes do not involve any assumptions
relative to accident initiators in the plant safety analyses, and
therefore the proposed amendments will not impact the probability of
occurrence for accidents previously analyzed.
The EDG line-item TS improvements associated with GL 93-05 are
based on recommendations designed to remove unwarranted requirements
for testing during power operation and other factors that are
counter-productive to safety in terms of equipment degradation and
availability. These recommendations resulted from a comprehensive
study of industry-wide EDG surveillance requirements and subsequent
findings reported by the NRC in NUREG-1366. The proposed amendments
are consistent with the guidance of GL 93-05 for implementing such
recommendations as well as contemporary licensing actions by the NRC
on other light water reactors.
Similarly, GL 94-01 provides guidance for a line-item TS
improvement that will remove accelerated testing requirements from
the TS provided that the licensee commits to a maintenance program
for monitoring and maintaining EDG performance that includes the
applicable provisions of the maintenance rule (10 CFR 50.65). Such a
program will further assure EDG availability. Since the availability
of EDGs is assumed in certain success paths for mitigating analyzed
accidents, an improvement in EDG availability will enhance accident
mitigation capabilities.
Therefore, 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.
(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 amendments incorporate line-item TS and other
improvements to EDG surveillance testing requirements, and will not
change the physical plant or the modes of plant operation defined in
the Facility License. The changes do not involve the addition or
modification of equipment, nor do they alter the design or methods
of operation of plant systems. Plant configurations that are
prohibited by TS will
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not be created by the amendments. Therefore, 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.
(3) Operation of the facility in accordance with the proposed
amendments would not involve a significant reduction in a margin of
safety.
The proposed amendments are designed to improve EDG availability
by eliminating unwarranted surveillance testing. The currently
specified surveillance intervals are not changed, except to delete
the requirement for accelerated testing under certain circumstances.
The proposed changes do not otherwise alter the basis for any
Technical Specification that is related to the establishment of, or
the maintenance of a nuclear safety margin. Therefore, operation of
the facility in accordance with the proposed amendment would 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: 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 3.6.C to allow up to 7 days to restore
low pressure safety injection (LPSI) pump subsystem operability, and up
to 24 hours to restore safety injection tank (SIT) operability.
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 LPSI system is designed primarily to
mitigate the consequences of a large loss-of-coolant accident
(LOCA). Inoperable LPSI components are not accident initiators in
any accident previously evaluated, and the proposed change does not
affect any of the assumptions relative to accident initiators in the
plant's safety analysis. Probabilistic safety analysis (PSA) methods
were used to fully evaluate the extension of the LPSI system allowed
outage time (AOT). The licensee asserts that the results of these
analyses show no significant increase in the consequences of an
accident previously evaluated. The SITs were designed to mitigate
the consequences of a LOCA. The proposed amendment does not affect
any of the assumptions used in the deterministic LOCA analysis.
Probabilistic safety analysis methods were used to fully evaluate
the effect of the SIT allowable outage time (AOT). The licensee
asserts that the results of these analyses show no significant
increase in the consequences of an accident previously evaluated.
Thus, there is no 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 change the design,
physical configuration, or modes of operation of the plant. Plant
configurations that are prohibited by TS will not be created by this
proposed amendment. Thus, the proposed amendment does not create the
possibility or consequences of an accident previously evaluated.
3. The proposed amendment does not involve a significant
reduction in a margin of safety. The proposed amendment does not
affect the limiting conditions for operation or the bases used in
the deterministic analyses to establish the margin of safety. The
licensee asserts that PSA methods were used to evaluate these
changes and demonstrate that the changes are either risk neutral or
risk beneficial. Thus, the proposed amendment 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
determination.
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 8, 1995
Description of amendment request: The proposed amendment would
modify the definition of Transthermal (Condition 4), Hot Shutdown
(Condition 5), and Hot Standby (Condition 6) reactor operating
conditions. The Transthermal and Hot Shutdown conditions are modified
to establish an applicable range of subcriticality and be consistent
with other Definitions. The wording of Hot Standby is modified to
remove reference to control rod position, consistent with NUREG-1432,
Standard Technical Specifications for Combustion Engineering Plants,
Revision 1 dated April 1995.
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 changes to these Definitions are
administrative in nature. The Transthermal and Hot Shutdown
conditions are changed by adding ``at least'' to establish a range
of subcriticality. The current Definitions for the Transthermal and
Hot Shutdown conditions set one minimum value for subcriticality;
the change to these two Definitions would allow a range of values
for subcriticality. All values of subcriticality that may be
established by this change are below the current Definitions (more
subcritical). The change to the wording of Hot Standby removes
confusion about the Conditions during which control rods may be
withdrawn and is consistent with current NRC guidance. All current
plant analyses, requirements and acceptance criteria on
subcriticality conditions remain in effect. The changes to these
Definitions have no impact on event probabililty. 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 clarifies the subject Definitions.
Limits on subcriticality requirements are unaffected, as are
reactivity transients previously evaluated. Plant procedures
currently require that minimum values for subcriticality be
established. All values of subcriticality that may be established by
this change are below the current Definitions (more subcritical).
Further, the change to the wording of Hot Standbyis consistent with
current NRC guidance. 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. Adding the words ``at least'' to
the Transthermal and Hot Shutdown conditions establishes a range of
subcriticality to the
[[Page 52932]]
Definitions for these terms. All values of subcriticality are below
(more subcritical) than the current value, thus the margin of safety
is increased. All current plant analyses, requirements and
acceptance criteria on subcriticality conditions remain in effect.
The change to the wording of Hot Standby removes confusion about the
Conditions during which control rods may be withdrawn and is
consistent with current NRC guidance. Thus, there is no 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
Maine Yankee Atomic Power Company, Docket No. 50-309, Maine Yankee
Atomic Power Station, Lincoln County, Maine
Date of amendment request: August 30, 1995
Description of amendment request: The proposed amendment would
change Technical Specification (TS) 1.3.A, Reactor Core, to allow the
use of fuel rods clad with zirconium alloy, rather than restrict fuel
rod cladding to Zircaloy-4. In addition, the fuel enrichment limit
described in this specification would be changed to more closely agree
with the wording found in NUREG-1432, ``Standard Technical
Specifications for Combustion Engineering Plants,'' dated April 1995.
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 change does not involve a significant increase
in the probability or consequences of an acident previously
evaluated. Maine Yankee (MY) reload cores containing fuel rods clad
with zirconium alloy and having higher fuel enrichments will be
analyzed using NRC-approved methods and applicable acceptance
criteria. In addition, the impact of fuel assembly design changes on
fuel storage will be analyzed using NRC-approved methods and
acceptance criteria. Compliance with the acceptance criteria for the
applicable analysis for a given core design must be determined for
each core prior to reloading. The material used to clad the fuel and
the fuel enrichment are only two of the factors considered in this
determinination. The application of approved methods ensures that
all appropriate variables are addressed and their acceptance
criteria satisfied. Thus, the proposed change does not involve a
significant increase in the probability or consequences of an
accident previously evaluated.
2. The proposed change does not create the possibility of a new
or different kind of accident from any accident previously
evaluated. The determination of compliance with the acceptance
criteria of the approved safety evaluation for any given core reload
design is performed for each MY reload core prior to loading. In
addition, determination of compliance with the acceptance criteria
of the approved safety evaluation for fuel storage is performed for
each core prior to receipt of the fuel. The use of approved methods
and their acceptance criteria ensures that new or different
accidents will not be encountered by the use of fuel rods clad with
zirconium alloy and having higher fuel enrichments. Further, the
proposed change does not involve any altertions to plant equipment
that would affect any operational modes or accident precursors.
Finally, the proposed change does not involve, or require secondary
involvement of, any equipment important to safety. Thus the proposed
change does not create the possibility of a new or different kind of
accident from any accident previously evaluated.
3. The proposed change does not involve a significant reduction
in a margin of safety. Maine Yankee reload cores containing fuel
rods clad with zirconium alloy and having higher fuel enrichments
will be analyzed using NRC-approved methods and applicable
acceptance criteria. Safety evaluations performed for each core
reload ensure that the core design meets appropriate safety
assessment acceptance criteria. In addition, the impact of fuel
assembly design changes on fuel storage also will be analyzed using
NRC-approved methods and aceptance criteria. Application of the
approved methods ensures that the requirements of MY TS 1.1, Fuel
Storage, are achieved. Because these requirements are not changed,
the margin of safety remains the same. Thus there is no 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
Maine Yankee Atomic Power Company, Docket No. 50-309, Maine
YankeeAtomic Power Station, Lincoln County, Maine
Date of amendment request: August 31, 1995
Description of amendment request: The proposed amendment would
relocate fire protection requirements from the Maine Yankee (MY) Atomic
Power Station Technical Specifications (TS) to other, licensee-
controlled documents. The proposed amendment is consistent with the
guidance of U.S. NRC Generic Letters 86-10, Implementation of Fire
Protection Requirements, and 88-12, Removal of Fire Protection
Requirements from the Technical Specifications.
Basis for proposed no significant hazards consideration
etermination: As required by 10 CFR 50.91(a), the licensee has provided
its analysis if 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 NRC staff's review 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 change is administrative and consistent with
the guidance provided by the U.S. NRC. Removing fire protection
requirements from the TS does not affect any fire protection
equipment, or involve any physical modifications to plant
structures, systems or components. The proposed change is not
associated with accident initiation or mitigation and cannot affect
the probability of occurrence of an accident, or increase the
consequences of an accident. The licensee's fire protection plan
contains the relocated requirements.
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 introduces no new mode of plant
operation, does not involve physical modification of any structure,
system or component, and does not affect the function, operation or
surveillance requirements of any equipment necessary for safe
operation or shutdown. Further, the proposed change does not involve
any change to equipment setpoints or operating parameters. The
proposed change is administrative in nature. Existing plant fire
protection equipment requirements are retained. Thus, the proposed
change does not create the possibility for a new or different kind
of accident from any accident previously evaluated.
3. The proposed change does not involve a significant reduction
in a margin of safety. No margins of safety established by system or
component design, or verified by testing to ensure operability of
fire protection systems or components, are affected. Fire protection
requirements currently found in the TS will be relocated in their
entirety to the Maine Yankee Fire Protection Plan. Any future
[[Page 52933]]
changes to the Plan will be evaluated in accordance with the
requirements of 10 CFR 50.59, Changes, tests and experiments. 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
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 11, 1995
Description of amendment request: The proposed changes affect
Technical Specification Sections 3.4.8 and 3.9.9, Tables 2.2-1, 3.3-3,
3.3-5 and 3.3-8, and Bases Sections 3/4.2.1, 3/4.4.8 and 3/4.11.2.1.
These changes combine several different administrative changes which
will correct typographical errors, provide clarifications, or make
editorial changes.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration (SHC), which is presented below:
Pursuant to 10CFR50.92, NNECO has reviewed the proposed changes.
NNECO concludes that these changes do not involve a significant
hazards consideration since the proposed change satisfies the
criteria in 10CFR50.92(c). That is, the proposed changes do not:
1. Involve a significant increase in the probability or
consequences of an accident previously analyzed.
The proposed changes are administrative in nature and do not
result in changes to plant configuration, operation, accident
mitigation, or analysis assumptions. Thus, it cannot increase the
probability or consequence of an accident.
2. Create the possibility of a new or different kind of accident
from any previously analyzed.
The proposed changes are administrative in nature and do not
result in changes to plant configuration, operation, accident
mitigation, or analysis assumptions. The intent and application of
the proposed specification will not change. Therefore, the proposal
does 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.
Since the proposed change[s] are administrative in nature and do
not result in changes to plant configuration, operation, accident
mitigation, or analysis assumptions, there is no reduction in the
margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room 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
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 amendment requests: July 17, 1995
Description of amendment requests: The proposed amendments would
revise the Prairie Island Radiological Effluent Technical
Specifications and other sections relating to radiological controls to
conform to NUREG-1431, Standard Technical Specifications, Westinghouse
Plants, Revision 1, and Generic Letter 89-01, ``Implementation of
Programmatic Controls for Radiological Effluent Technical
Specifications in the Administrative Controls Section of the Technical
Specifications and the Relocation of Procedural Details of RETS to the
Offsite Dose Calculation Manual or to the Process Control Program.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. The proposed amendment will not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
The proposed changes do not involve a significant increase in
the probability or consequences of an accident previously evaluated.
The proposed changes are administrative in nature and alter only the
format and location of programmatic controls and procedural details
relative to radioactive effluents, radiological environmental
monitoring, radioactive source leakage testing, solid radioactive
wastes, and associated reporting requirements. Existing Technical
Specifications containing procedural details on radioactive
effluents, radiological environmental monitoring, radioactive source
leakage testing, explosive gas monitoring, storage tank radioactive
content limits, solid radioactive wastes and associated reporting
requirements are being relocated to the Offsite Dose Calculation
Manual, Process Control Program or other new programs as
appropriate. Compliance with applicable regulatory requirements will
continue to be maintained. In addition, the proposed changes do not
alter the conditions or the assumptions in any of the previous
accident analyses. Since the previous accident analyses remain
bonding, the radiological consequences previously evaluated are not
adversely affected by the proposed changes.
Therefore, the probability or consequences of an accident
previously evaluated are not affected by any of the proposed
amendments.
2. The proposed amendment will not create the possibility of a
new or different kind of accident from any accident previously
analyzed.
The proposed changes do not create the possibility of a new or
different kind of accident from any accident previously evaluated.
The proposed changes do not involve any change to the configuration
or method of operation of any plant equipment. Accordingly, no new
failure modes have been defined for any plant system or component
important to safety nor has any new limiting single failure been
identified as a result of the proposed changes. Also, there will be
no change in types or increase in the amounts of any effluents
released offsite.
Therefore, the possibility of a new or different kind of
accident from any accident previously evaluated would not be
created.
3. The proposed amendment will not involve a significant
reduction in the margin of safety.
The proposed changes do not involve a significant reduction in a
margin of safety. The proposed changes do not involve any actual
change in the methodology used in the control of radioactive
effluents, radioactive sources, solid radioactive wastes, or
radiological environmental monitoring. These changes are considered
administrative in nature and provide for the relocation of
procedural details outside of the technical specifications but add
appropriate administrative controls to provide continued assurance
of compliance to applicable regulatory requirements. These proposed
changes also comply with the guidance contained in Generic Letter
89-01 and the Standard Technical Specifications.
Therefore, it can be concluded a significant reduction in the
margin of safety would not be involved.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment requests involve no significant hazards consideration.
[[Page 52934]]
Local Public Document Room location: Minneapolis Public Library,
Technology and Science Department, 300 Nicollet Mall, Minneapolis,
Minnesota 55401
Attorney for licensee: Jay Silberg, Esq., Shaw, Pittman, Potts, and
Trowbridge, 2300 N Street, NW, Washington, DC 20037
NRC Project Director: John N. Hannon
Philadelphia Electric Company, Docket No. 50-352, Limerick
Generating Station, Unit 1, Montgomery County, Pennsylvania
Date of amendment request: June 19, 1995
Description of amendment request: The proposed amendment would
revise Technical Specification Section 2.1, ``Safety Limits,'' to
change the Minimum Critical Power Ratio Safety Limit due to the use of
General Electric 13 fuel product line.
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 Technical Specifications (TS) change does not
involve a significant increase in the probability or consequences of
an accident previously evaluated.
The derivation of the revised GE13 [General Electric] Minimum
Critical Power Ratio (MCPR) Safety Limit for incorporation into the
Technical Specifications, and its use to determine cycle-specific
thermal limits have been performed using NRC-approved methods within
the existing design and licensing basis, and cannot increase the
probability or severity of an accident.
The basis of the MCPR Safety Limit calculation is to ensure that
greater than 99.9% of all fuel rods in the core avoid boiling
transition if the limit is not violated. The new MCPR Safety Limit
preserves the existing margin to transition boiling and fuel damage
in the event of a postulated accident.
All design bases of the MCPR Safety Limit calculation apply to
GE13 fuel in the same manner that they have applied to previous fuel
designs. The probability of fuel damage is not increased.
Therefore, the proposed TS change does not involve an increase
in the probability or consequences of an accident previously
evaluated.
2. The proposed TS change does not create the possibility of a
new or different kind of accident from any accident previously
evaluated.
The MCPR Safety Limit for the GE13 fuel design is a Technical
Specification numerical value, designed to ensure that fuel damage
from transition boiling does not occur as a result of the limiting
postulated accident. It cannot create the possibility of any new
type of accident. The new Minimum Critical Power Ratio (MCPR) Safety
Limit is calculated using NRC-approved methods and has the same
calculational basis as the MCPR Safety Limit for other GE fuel
designs currently used at LGS [Limerick Generating Station] Unit 1.
Therefore, the proposed TS change does not create the
possibility of a new or different kind of accident, from any
accident previously evaluated.
3. The proposed TS change does not involve a significant
reduction in a margin of safety.
The following TS Bases were reviewed for potential reduction in
the margin of safety:
2.1 ``Safety Limits''
3/4.2.1 ``Average Planar Linear Heat Generation Rate''
3/4.2.3 ``Minimum Critical Power Ratio''
3/4.2.4 ``Linear Heat Generation Rate''
3/4.4.1 ``Recirculation System''
3/4.9 ``Refueling Operations''
The margin of safety as defined in the TS Bases will remain the
same. The new Minimum Critical Power Ratio (MCPR) Safety Limit is
calculated using NRC approved methods which are in accordance with
the current fuel design and licensing criteria. The MCPR Safety
Limit for GE13 fuel remains high enough to ensure that greater than
99.9% of all fuel rods in the core will avoid boiling transition if
the limit is not violated, thereby preserving the fuel cladding
integrity.
Therefore, the proposed TS change does not involve a reduction
in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 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: Pottstown Public Library, 500
High Street, Pottstown, Pennsylvania 19464.
Attorney for licensee: J. W. Durham, Sr., Esquire, Sr. V. P. and
General Counsel, Philadelphia Electric Company, 2301 Market Street,
Philadelphia, Pennsylvania 19101
NRC Project Director: John F. Stolz
Philadelphia Electric Company, Docket Nos. 50-352 and 50-353,
Limerick Generating Station, Units 1 and 2, Montgomery County,
Pennsylvania
Date of amendment request: September 14, 1995
Description of amendment request: The amendments change the
Technical Specifications (TS) by removing the Reactor Enclosure and
Refueling Area Secondary Containment Isolation Valve Tables 3.6.5.2.1-1
and 3.6.5.2.2-1 from TS in accordance with NRC Generic Letter (GL) 91-
08, ``Removal of Component Lists from Technical Specifications.''
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. The proposed TS changes do not involve a significant increase
in the probability or consequences of an accident previously
evaluated.
The proposed changes will remove component tables from TS. The
component lists will be retained in licensee controlled documents
(UFSAR [Updated Final Safety Analysis Report] and a plant procedure)
which will be maintained under the requirements of TS Administrative
Controls Section 6.0 and the provisions of 10 CFR 50.59. Since any
changes to licensee controlled documents are required to be
evaluated per 10 CFR 50.59, no increase (significant or
insignificant) in the probability or consequences of an accident
previously evaluated will be allowed.
In addition, these proposed changes will not affect any
equipment important to safety, in structure or operation. These
changes will not alter operation of process variables, structures,
systems, or components as described in the safety analysis and
licensing basis. The changes will not increase the probability or
consequences of occurrence of a malfunction of equipment important
to safety previously evaluated in the SAR [Safety Analysis Report].
2. The proposed TS changes do not create the possibility of a
new or different kind of accident from any accident previously
evaluated.
The proposed changes will not alter the plant configuration or
change the methods governing normal plant operation. The changes
will not impose different operating requirements and adequate
control of information will be retained. The changes will not alter
assumptions made in the safety analysis and licensing basis. Since
the proposed changes cannot cause an accident, and the plant
response to the design basis events is unchanged, the changes do not
create the possibility of a new or different kind of accident from
any accident previously evaluated.
3. The proposed TS changes do not involve a significant
reduction in a margin of safety.
The proposed changes to remove the component tables from TS have
been performed under the guidance of NRC GL 91-08. The component
lists will be retained in licensee controlled documents (UFSAR and a
plant procedure) which will be maintained under the requirements of
TS Administrative Controls Section 6.0 and the provisions of 10 CFR
50.59. These changes will not reduce the margin of safety since they
have no impact on any safety analysis assumptions. Since any future
changes to the removed tables will be evaluated under the
requirements of 10 CFR 50.59, no reduction (significant or
insignificant) in a margin of safety will be allowed. Therefore, the
proposed TS changes do not involve a significant reduction in a
margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
[[Page 52935]]
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: Pottstown Public Library, 500
High Street, Pottstown, Pennsylvania 19464.
Attorney for licensee: J. W. Durham, Sr., Esquire, Sr. V. P. and
General Counsel, Philadelphia Electric Company, 2301 Market Street,
Philadelphia, Pennsylvania 19101
NRC Project Director: John F. Stolz
Rochester Gas and Electric Corporation, Docket No. 50-244, R. E.
Ginna Nuclear Power Plant, Wayne County, New York
Date of amendment request: May 26, 1995
Brief description of amendment: The proposed amendment would
represent a full conversion from the current Technical Specifications
(TSs) to a set of TS based on NUREG-1431, ``Standard Technical
Specifications, Westinghouse Plants,'' Revision 0, dated September
1993, together with approved travellers used in the issuance of
Revision 1, dated April 1995. NUREG-1431 was developed through working
groups composed of NRC staff members and industry representatives and
has been endorsed by the staff as part of an industry-wide initiative
to standardize and improve the TSs. As part of this submittal, the
licensee has applied the criteria contained in the Commission's Final
Policy Statement on Technical Specification Improvements for Nuclear
Power Reactors of July 22, 1993, to the current Ginna TSs, and using
NUREG-1431 as a basis, developed a proposed set of improved TSs for
Ginna.Date of publication of individual notice in Federal Register:
September 26, 1995 (60 FR 49636)
Expiration date of individual notice: October 26, 1995
Local Public Document Room location: Rochester Public Library, 115
South Avenue, Rochester, New York 14610
Tennessee Valley Authority, Docket No. 50-296, Browns Ferry Nuclear
Plant, Unit 3, Limestone County, Alabama
Date of amendment request: September 13, 1995 (TS 368)
Description of amendment request: The proposed amendment deletes
requirements for daily checks for certain instruments that do not have
indications, and provides editorial changes.
Basis for proposed no significant hazards consideration
determination:As required by 10 CFR 50.91(a), the licensee has provided
its analysis of the issue of no significant hazards consideration,
which is presented below:
A. The proposed amendment does not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
The proposed changes are administrative in nature and correct
errors that were introduced by previous changes to the TSs. These
changes do not affect any of the design basis accidents nor do they
involve an increase in the probability or consequences of an
accident previously evaluated.
B. The proposed amendment does not create the possibility of a
new or different kind of accident from any accident previously
evaluated.
The proposed changes are administrative in nature. These changes
do not change the operation or function of the affected
instrumentation. The deletion of the RCIC and HPCI instrument checks
reflects the actual installed configuration of this instrumentation
(no indication) and the change to Table 4.2.C corrects the
referenced note for the SRM Upscale function. Therefore, the
possibility for an accident or malfunction of a different type than
any evaluated previously is not created by this change.
C. The proposed amendment does not involve a significant
reduction in a margin of safety.
The proposed changes are administrative in nature. The proposed
changes to TS Tables 4.2.B and 4.2.C do not affect any acceptable
limit of operation, instrument setpoint, or analysis assumption in
the TS or Bases. Therefore, this change does not reduce the margin
of safety as defined in the basis for any TS.
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: Athens Public Library, South
Street, Athens, Alabama 35611
Attorney for licensee: General Counsel, Tennessee Valley Authority,
400 West Summit Hill Drive, ET llH, Knoxville, Tennessee 37902
NRC Project Director: Frederick J. Hebdon
TU Electric Company, Docket Nos. 50-445 and 50-446, Comanche Peak
Steam Electric Station, Units 1 and 2, Somervell County, Texas
Date of amendment request: August 15, 1995
Brief description of amendments: The proposed amendment would
relocate the Shutdown Margin limits from the Technical Specifications
(TSs) to the Core Operating Limits Report. The proposed changes are
consistent with the intent of Generic Letter (GL) 88-16 which provides
guidelines for the removal of cycle-specific parameter limits from the
TSs.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. Do the proposed changes involve a significant increase in the
probability or consequences of an accident previously evaluated?
The proposed changes remove cycle-specific parameter limits from
the Technical Specifications, add them to the list of limits
contained in the Core Operating Limits Report (COLR), and revise the
Administrative Controls section of the Technical Specifications. The
changes do not, by themselves, alter any of the parameter limits.
The changes are administrative in nature and have no adverse effect
on the probability of an accident or on the consequences of an
accident previously evaluated. The removal of parameter limits from
the Technical Specifications does not eliminate the requirement to
comply with the parameter limits.
The parameter limits in the COLR may be revised without prior
NRC approval. However, Specification 6.9.1.6c continues to ensure
that the parameter limits are developed using NRC-approved
methodologies and that applicable limits of the safety analyses are
met. While future changes to the COLR parameter limits could result
in event consequences which are either slightly less or slightly
more severe than the consequences for the same event using the
present parameter limits, the differences would not be significant
and would be bounded by the requirement of specification 6.9.1.6c to
meet the applicable limits in the safety analysis.
Based on the above, removal of the parameter limits from the
Technical Specifications and the addition of these limits the list
of limits in the COLR, thus allowing revision of the parameter
limits without prior NRC approval, has no significant effect on the
probability or consequences of an accident previously evaluated.
2. Do the proposed changes create the possibility of a new or
different kind of accident from any accident previously evaluated?
The proposed changes remove certain parameter limits from the
Technical Specifications and add these limits to the list of limits
in the COLR, removing the requirement for prior NRC approval of
revisions to those parameters. The changes do not add new hardware
or change plant operations and therefore cannot initiate an event
nor cause an analyzed event to progress differently. Thus, the
possibility of a new or different kind of accident is not created.
3. Do the proposed changes involve a significant reduction in a
margin of safety?
The margin of safety, as it relates to a parameter limit, is the
difference between the
[[Page 52936]]
acceptance criterion for that parameter and its failure value. The
proposed changes do not affect the failure values for any system.
Through the accident analyses, all relevant event acceptance
criteria (as described in the NRC-approved analysis methodologies)
are shown to be satisfied; therefore, there is no impact on an event
acceptance criteria. Because neither the failure values nor the
acceptance criteria are affected, the proposed change has no effect
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: University of Texas at
Arlington Library, Government Publications/Maps, 702 College, P.O. Box
19497, Arlington, TX 76019
Attorney for licensee: George L. Edgar, Esq., Morgan, Lewis and
Bockius, 1800 M Street, N.W., Washington, DC 20036
NRC Project Director: William D. Beckner
Wisconsin Public Service Corporation, Docket No. 50-305, Kewaunee
Nuclear Power Plant, Kewaunee County, Wisconsin
Date of amendment request: September 19, 1995
Description of amendment request: The proposed amendment would make
administrative changes to the Kewaunee Nuclear Power Plant (KNPP)
Technical Specifications (TS) to improve their clarity and consistency.
The proposed amendment includes changes to reflect revisions to 10 CFR
Part 20, and changes to correct minor typographical and format
inconsistencies as part of an ongoing effort to convert the TS to the
WordPerfect format.
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 were reviewed in accordance with the
provisions of 10 CFR 50.92 to show no significant hazards exist. The
proposed changes will not:
1. involve a significant increase in the probability or
consequences of an accident previously evaluated.
The likelihood that an accident will occur is neither increased
or decreased by these TS changes. These TS changes will not impact
the function or method of operation of plant equipment. Thus, there
is not a significant increase in the probability of a previously
analyzed accident due to these changes. No systems, equipment, or
components are affected by the proposed changes. Thus, the
consequences of the malfunction of equipment important to safety
previously evaluated in the Updated Safety Analysis Report (USAR)
are not increased by these changes.
The proposed changes are administrative in nature and,
therefore, have no impact on accident initiators or plant equipment,
and thus, do not affect the probabilities or consequences of an
accident.
2. create the possibility of a new or different kind of accident
from any accident previously evaluated.
Operation of the facility in accordance with the proposed TS
changes would not create the possibility of a new or different kind
of accident from any accident previously evaluated.
The proposed changes do not involve changes to the physical
plant or operations. Since these administrative changes do not
contribute to accident initiation, they do not produce a new
accident scenario or produce a new type of equipment malfunction.
Also, these changes do not alter any existing accident scenarios;
they do not affect equipment or its operation, and thus, do not
create the possibility of a new or different kind of accident.
3. involve a significant reduction in the margin of safety.
Operation of the facility in accordance with the proposed TS
would not involve a significant reduction in a margin of safety. The
proposed changes do not affect plant equipment or operation. Safety
limits and limiting safety system settings are not affected by these
proposed changes.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: University of Wisconsin,
Cofrin Library, 2420 Nicolet Drive, Green Bay, Wisconsin 54311-7001.
Attorney for licensee: Bradley D. Jackson, Esq., Foley and Lardner,
P. O. Box 1497, Madison, Wisconsin 53701-1497.
NRC Project Director: Gail H. Marcus
Wolf Creek Nuclear Operating Corporation, Docket No. 50-482, Wolf
Creek Generating Station, Coffey County, Kansas
Date of amendment request: September 14, 1995
Description of amendment request: The proposed amendment would
revise Technical Specification 3/4.5.5 to increase the outage time
allowed for adjusting the boron concentration of the refueling water
storage tank (RWST) from 1 hour to 8 hours.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. The proposed change does not involve a significant increase
in the probability or consequences of an accident previously
evaluated.
The increase in the RWST allowed outage time does not alter the
plant configuration or operation. The potential for the RWST boron
concentration to be outside the technical specification limits is
small because the RWST and its contents are not involved with normal
plant operation and are not subject to process variations associated
with plant operation.
The potential causes of boron concentration deviation have been
evaluated with the conclusion that any deviation in RWST boron
concentration would not be expected to increase significantly during
the proposed 7 hour allowed outage time increase.
Therefore, the proposed change does not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
2. The proposed change does not create the possibility of a new
or different kind of accident from any accident previously
evaluated.
Increasing the RWST allowed outage time from 1 hour to 8 hours
for reasons directly related to boron concentration does not require
physical alteration to any plant system and does not change the
method by which any safety related system performs its functions.
Therefore, the proposed change does not create the possibility of a
new or different kind of accident from any accident previously
evaluated.
3. The proposed change does not involve a significant reduction
in a margin of safety.
Increasing the RWST allowed outage time for reasons directly
related to boron concentration does not affect any accident analysis
assumptions, initial conditions, or results. The margins of safety
reflected in the Wolf Creek Generating Station Technical
Specifications are not compromised by the 7 hour allowed outage time
increase. Therefore, the proposed change does not involve a
significant reduction in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room locations: Emporia State University,
William Allen White Library, 1200 Commercial Street, Emporia, Kansas
66801 and Washburn University School of Law Library, Topeka, Kansas
66621
Attorney for licensee: Jay Silberg, Esq., Shaw, Pittman, Potts and
Trowbridge, 2300 N Street, N.W., Washington, D.C. 20037
NRC Project Director: William H. Bateman
[[Page 52937]]
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 50-454 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 1, 1995
Description of amendment request: The proposed amendments would
revise the present voltage-based repair criteria in the Byron 1 and
Braidwood 1 Technical Specifications (TSs). These proposed revisions
would raise the lower voltage limit from its present value of 1.0 volt
to 3.0 volts; there would no longer be an upper voltage limit.
The Braidwood 1 TSs were revised by License Amendment No. 54,
issued on August 18, 1994, to add voltage-based repair criteria to the
existing steam generator (SG) tube repair criteria. The Byron 1 TSs
were revised in a similar manner by License Amendment No. 66, issued on
October 24, 1994.
The voltage-based repair criteria in the subject TSs are applicable
only to a specific type of SG tube degradation which is predominantly
axially-oriented outer diameter stress corrosion cracking (ODSCC). This
particular form of SG tube degradation occurs entirely within the
intersections of the SG tubes with the tube support plates (TSPs).
The present voltage values for the ODSCC repair criteria are based
on the assumption of a ``free span'' exposure of the SG tube flaw;
i.e., no credit is given for any constraint against burst or leakage,
which may be provided by the presence of the TSPs. This approach is, in
turn, based on the assumption that under postulated accident
conditions, the TSPs may be displaced sufficiently by blowdown
hydrodynamic loads such that a SG tube flaw which was fully confined
within the thickness of the TSP prior to the accident would then be
fully exposed. This approach was first advanced by the NRC staff in a
draft generic letter issued on August 12, 1994, which was subsequently
modified slightly and issued as Generic letter (GL) 95-05, ``Voltage-
Based Repair Criteria For Westinghouse Steam Generator Tubes Affected
by Outside Diameter Stress Corrosion Cracking,'' dated August 3, 1995.
The previous license amendments related to the issue of ODSCC were
based to a large extent on the draft generic letter cited above.
The fundamental difference between the pending proposal to raise
the lower voltage repair limit to 3.0 volts and the methodology
contained in GL 95-05, is that the licensee proposes to install certain
modifications to the SG internal structures, thereby limiting to a
small value, the maximum displacement of the TSPs under accident
conditions. The proposed structural modifications consist of expanding
a limited number of SG tubes only on the hot leg side of the TSP, at
each of the intersections of the tubes with the TSPs. The purpose of
this approach would be to greatly reduce the probability of SG tube
burst under postulated accident conditions by several orders of
magnitude. There would be a negligible impact on the primary-to-
secondary SG tube leakage under accident conditions.
While the voltage-based repair criteria for ODSCC flaws are
applicable only to Byron 1 and Braidwood 1, the pending request for
license amendments involves all four units in that both stations have a
common set of TSs. Date of publication of individual notice in Federal
Register: September 27, 1995 (60 FR 49963)
Expiration date of individual notice: October 27, 1995
Local Public Document Room location: For Byron, the Byron Public
Library District, 109 N. Franklin, P.O. Box 434, Byron, Illinois 61010;
for Braidwood, the Wilmington Public Library, 201 S. Kankakee Street,
Wilmington, Illinois 60481
Commonwealth Edison Company, Docket Nos. 50-237 and 50-249, Dresden
Nuclear Power Station, Grundy County, Illinois
Date of amendment request: September 1, 1995
Description of amendment request: The proposed amendment would
upgrade the Dresden 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 1, 1995, application proposed to upgrade only
Section 6.0 (Administrative Controls) of the Dresden TS.Date of
publication of individual notice in Federal Register: September 20,
1995 (60 FR 48728)
Expiration date of individual notice: October 20, 1995
Local Public Document Room location: Morris Area Public Library
District, 604 Liberty Street, Morris, Illinois 60450
Duquesne Light Company, et al., Docket Nos. 50-334 and 50-412,
Beaver Valley Power Station, Unit Nos. 1 and 2, Shippingport,
Pennsylvania
Date of amendment request: September 13, 1995
Brief description of amendment request: The proposed amendments
would revise the Administrative Controls section and the Bases section
of the Beaver Valley Power Station, Unit Nos. 1 and 2 (BVPS-1 and BVPS-
2), technical specifications to be consistent with the requirements of
the Offsite Dose Calculation Manual (ODCM). The ODCM was recently
updated to reflect the radioactive liquid and gaseous effluent release
limits and the liquid holdup tank activity limit of BVPS-1 License
Amendment No. 188 and BVPS-2 License Amendment No. 70 which were issued
June 12, 1995.Date of publication of individual notice in Federal
Register: September 22, 1995 (60 FR 49292)
Expiration date of individual notice: October 23, 1995
Local Public Document Room location: B. F. Jones Memorial Library,
663 Franklin Avenue, Aliquippa, Pennsylvania 15001
[[Page 52938]]
PECO Energy Company, Public Service Electric and Gas Company,
Delmarva Power and Light Company, and Atlantic City Electric
Company, Docket No. 50-278, Peach Bottom Atomic Power Station, Unit
No. 3, York County, Pennsylvania
Date of amendment request: September 1, 1995
Brief description of amendment request: The proposed amendment
would delete License Condition 2.C.(5) from Facility Operating License
DPR-56 which restricts power levels to no less than seventy percent in
the coastdown condition.
Date of publication of individual notice in Federal Register:
September 19, 1995 (60 FR 48530)
Expiration date of individual notice: October 18, 1995
Local Public Document Room location: Government Publications
Section, State Library of Pennsylvania, (REGIONAL DEPOSITORY) Education
Building, Walnut Street and Commonwealth Avenue, Box 1601, Harrisburg,
Pennsylvania 17105
Notice Of Issuance Of Amendments To Facility Operating Licenses
During the period since publication of the last biweekly notice,
the Commission has issued the following amendments. The Commission has
determined for each of these amendments that the application complies
with the standards and requirements of the Atomic Energy Act of 1954,
as amended (the Act), and the Commission's rules and regulations. The
Commission has made appropriate findings as required by the Act and the
Commission's rules and regulations in 10 CFR Chapter I, which are set
forth in the license amendment.
Notice of Consideration of Issuance of Amendment to Facility
Operating License, Proposed No Significant Hazards Consideration
Determination, and Opportunity for A Hearing in connection with these
actions was published in the Federal Register as indicated.
Unless otherwise indicated, the Commission has determined that
these amendments satisfy the criteria for categorical exclusion in
accordance with 10 CFR 51.22. Therefore, pursuant to 10 CFR 51.22(b),
no environmental impact statement or environmental assessment need be
prepared for these amendments. If the Commission has prepared an
environmental assessment under the special circumstances provision in
10 CFR 51.12(b) and has made a determination based on that assessment,
it is so indicated.
For further details with respect to the action see (1) the
applications for amendment, (2) the amendment, and (3) the Commission's
related letter, Safety Evaluation and/or Environmental Assessment as
indicated. All of these items are available for public inspection at
the Commission's Public Document Room, the Gelman Building, 2120 L
Street, NW., Washington, DC, and at the local public document rooms for
the particular facilities involved.
Arizona Public Service Company, et al., Docket Nos. STN 50-528, STN
50-529, and STN 50-530, Palo Verde Nuclear Generating Station,
Units 1, 2, and 3, Maricopa County, Arizona
Date of application for amendments: December 7, 1994, as
supplemented by letter dated August 1, 1995.
Brief description of amendments: The amendments change Note 5 to
Table 4.3-1 of Technical Specification 3/4.3.1 to allow verification of
the shape-annealing matrix elements used in the core protection
calculators. This provides the option of using generic shape-annealing
matrix elements in the core protection calculators. Presently, cycle-
specific shape-annealing elements are determined during startup testing
after each core reload. Use of a generic shape-annealing matrix
eliminates several hours of critical path work during startup after a
refueling outage.
Date of issuance: September 20, 1995
Effective date: September 20, 1995
Amendment Nos.: Unit 1 - Amendment No. 100; Unit 2 - Amendment No.
88; Unit 3 - Amendment No. 71
Facility Operating License Nos. NPF-41, NPF-51, and NPF-74: The
amendments revised the Technical Specifications.
Date of initial notice in Federal Register: January 4, 1995 (60 FR
495). The August 1, 1995, supplemental letter provided clarifying
information and did not change the original no significant hazards
consideration determination. The Commission's related evaluation of the
amendments is contained in a Safety Evaluation dated September 20,
1995.No significant hazards consideration comments received: No
Local Public Document Room location: Phoenix Public Library, 1221
N. Central Avenue, Phoenix, Arizona 85004
Commonwealth Edison Company, Docket Nos. 50-237 and 50-249, Dresden
Nuclear Power Station, Units 2 and 3, Grundy County, Illinois
Docket Nos. 50-254 and 50-265, Quad Cities Nuclear Power Station,
Units 1 and 2, Rock Island County, Illinois
Date of application for amendments: March 26, 1993, as
supplemented May 15, 1995
Brief description of amendments: These amendments upgrade the
current custom Technical Specifications (TS) for Dresden and Quad
Cities to the Standard Technical Specifications contained in NUREG-
0123, ``Standard Technical Specification General Electric Plants BWR/
4.'' These amendments upgrade only Section 3/4.9 (Electrical Power
Systems). These amendments include the relocation of some TS
requirements to licensee-controlled documents.
Date of issuance: September 18, 1995
Effective date: Immediately, to be implemented no later than
December 31, 1995, for Dresden Nuclear Power Station and June 30, 1996,
for Quad Cities Nuclear Power Station.
Amendment Nos.: 138, 132, 160, 156
Facility Operating License Nos. DPR-19, DPR-25, DPR-29 and DPR-30.
The amendments revised the Technical Specifications.
Date of initial notice in Federal Register: January 19, 1994 (59 FR
2864) The Commission's related evaluation of the amendments is
contained in a Safety Evaluation dated September 18, 1995. No
significant hazards consideration comments received: No
Local Public Document Room location: for Dresden, Morris Area
Public Library District, 604 Liberty Street, Morris, Illinois 60450;
for Quad Cities, Dixon Public Library, 221 Hennepin Avenue, Dixon,
Illinois 61021
Commonwealth Edison Company, Docket Nos. 50-237 and 50-249, Dresden
Nuclear Power Station, Units 2 and 3, Grundy County, Illinois,
Docket Nos. 50-254 and 50-265, Quad Cities Nuclear Power Station,
Units 1 and 2, Rock Island County, Illinois
Date of application for amendments: December 8, 1992, as
supplemented September 10, 1993, and May 17, 1995.
Brief description of amendments: This application upgrades the
current custom Technical Specifications (TS) for Dresden and Quad
Cities to the Standard Technical Specifications (STS) contained in
NUREG-0123, ``Standard Technical Specification General Electric Plants
BWR/4.'' This application upgrades only Section 3/4.1 (Reactor
Protection System). Date of issuance:
[[Page 52939]]
September 20, 1995Effective date: Immediately, to be implemented no
later than December 31, 1995, for Dresden Station and June 30, 1996,
for Quad Cities Station.
Amendment Nos.: 139, 133, 161, and 157
Facility Operating License Nos. DPR-19, DPR-25, DPR-29 and DPR-30.
The amendments revised the Technical Specifications.
Date of initial notice in Federal Register: June 6, 1995 (60 FR
29872) The Commission's related evaluation of the amendments is
contained in a Safety Evaluation dated September 20, 1995. No
significant hazards consideration comments received: No
Local Public Document Room location: for Dresden, Morris Area
Public Library District, 604 Liberty Street, Morris, Illinois 60450;
for Quad Cities, Dixon Public Library, 221 Hennepin Avenue, Dixon,
Illinois 61021
Commonwealth Edison Company, Docket Nos. 50-237 and 50-249, Dresden
Nuclear Power Station, Units 2 and 3, Grundy County, Illinois
Docket Nos. 50-254 and 50-265, Quad Cities Nuclear Power Station,
Units 1 and 2, Rock Island County, Illinois
Date of application for amendments: September 17, 1993, as
supplemented June 30, 1995.
Brief description of amendments: This application upgrades the
current custom Technical Specifications (TS) for Dresden and Quad
Cities to the Standard Technical Specifications (STS) contained in
NUREG-0123, ``Standard Technical Specification General Electric Plants
BWR/4.'' This application upgrades only Section 3/4.6.
Date of issuance: September 21, 1995
Effective date: Immediately, to be implemented no later than
December 31, 1995, for Dresden Station and June 30, 1996, for Quad
Cities Station.
Amendment Nos.: 140, 134, 162, and 158
Facility Operating License Nos. DPR-19, DPR-25, DPR-29 and DPR-30.
The amendments revised the Technical Specifications.
Date of initial notice in Federal Register: July 19, 1995 (60 FR
37087) The Commission's related evaluation of the amendments is
contained in a Safety Evaluation dated September 21, 1995. No
significant hazards consideration comments received: No
Local Public Document Room location: for Dresden, Morris Area
Public Library District, 604 Liberty Street, Morris, Illinois 60450;
for Quad Cities, Dixon Public Library, 221 Hennepin Avenue, Dixon,
Illinois 61021
Commonwealth Edison Company, Docket Nos. 50-373 and 50-374, LaSalle
County Station, Units 1 and 2, LaSalle County, Illinois
Date of application for amendments: April 11, 1995
Brief description of amendments: The amendments allow a one-time
extension of specific LaSalle, Units 1 and 2, 18-month Technical
Specification Surveillance Requirements to allow surveillance testing
to coincide with the LaSalle, Unit 1, seventh refueling outage (L1R07).
The shutdown for L1R07 has been rescheduled from September 1995 until
early 1996. The proposed extensions apply to calibrations and
functional testing of isolation actuation instrumentation, emergency
core cooling system actuation instrumentation, and recirculation pump
trip actuation instrumentation; leakage testing of reactor coolant
system isolation valves; inspection of fire-rated seals; functional
testing of mechanical snubbers; inspections of emergency diesel
generators; and testing of batteries, battery chargers, and other
electrical components.
Date of issuance: September 27, 1995
Effective date: Immediately, to be implemented within 30 days.
Amendment Nos.: 106 and 92
Facility Operating License Nos. NPF-11 and NPF-18: The amendments
revised the Facility Operating Licenses.
Date of initial notice in Federal Register: July 5, 1995 (60 FR
35066) The Commission's related evaluation of the amendments is
contained in a Safety Evaluation dated September 27, 1995. No
significant hazards consideration comments received: No
Local Public Document Room location: Jacobs Memorial Library,
Illinois Valley Community College, Oglesby, Illinois 61348.
Duke Power Company, et al., Docket Nos. 50-413 and 50-414, Catawba
Nuclear Station, Units 1 and 2, York County, South Carolina
Date of application for amendments: June 17, 1993, as supplemented
July 5, 1995
Brief description of amendments: The amendments revise Technical
Specification Section 5.3.1 ``Fuel Assemblies'' in accordance with
Generic Letter 90-02, Supplement 1, ``Alternative Requirements For Fuel
Assemblies in The Design Features Section of Technical
Specifications.''
Date of issuance: September 18, 1995
Effective date: As of the date of issuance to be implemented within
30 days from the date of issuance
Amendment Nos.: 135 and 129
Facility Operating License Nos. NPF-35 and NPF-52: Amendments
revised the Technical Specifications.
Date of initial notice in Federal Register: July 21, 1993 (58 FR
39048) and ReNoticed August 16, 1995 (60 FR 42601) The Commission's
related evaluation of the amendments is contained in a Safety
Evaluation dated September 18, 1995. No significant hazards
consideration comments received: No
Local Public Document Room location: York County Library, 138 East
Black Street, Rock Hill, South Carolina 29730
Duquesne Light Company, et al., Docket No. 50-334, Beaver Valley
Power Station, Unit No. 1, Shippingport, Pennsylvania
Date of application for amendment: July 11, 1995
Brief description of amendment: This amendment revised the required
area of the reactor coolant system overpressure protection system vent
from 3.14 square inches to 2.07 square inches which is equal to the
relief area of a single power-operated relief valve.
Date of issuance: September 26, 1995
Effective date: As of the date of issuance, to be implemented
within 60 days.
Amendment No.: 193
Facility Operating License No. DPR-66. Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: August 16, 1995 (60 FR
42603) The Commission's related evaluation of the amendment is
contained in a Safety Evaluation dated September 26, 1995. No
significant hazards consideration comments received: No
Local Public Document Room location: B. F. Jones Memorial Library,
663 Franklin Avenue, Aliquippa, Pennsylvania 15001
Duquesne Light Company, et al., Docket No. 50-412, Beaver Valley
PowerStation, Unit 2, Shippingport, Pennsylvania
Date of application for amendment: July 24, 1995
Brief description of amendment: This amendment revises TS 3/4.4.11,
``Relief Valves,'' and associated Bases to make Unit 2 TS 3/4.4.11
consistent with Unit 1 TS 3/4.4.11 which was revised by Unit 1 License
Amendment No. 187 issued on May 15, 1995. The amendment generally
reflects the guidance provided in NRC Generic Letter 90-06 and in the
NRC's Improved Standard Technical Specifications (NUREG-1431).
Date of issuance: September 18, 1995
[[Page 52940]]
Effective date: As of the date of issuance, to be implemented
within 60 days.
Amendment No.: 76
Facility Operating License No. NPF-73: Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: August 16, 1995 (60 FR
42604) 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: B. F. Jones Memorial Library,
663 Franklin Avenue, Aliquippa, Pennsylvania 15001
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 the frequency of containment post-entry visual
inspections.
Date of issuance: September 15, 1995
Effective date: September 15, 1995
Amendment No.: 162
Facility Operating License No. NPF-6. Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: July 19, 1995 (60 FR
37089) The Commission's related evaluation of the amendment is
contained in a Safety Evaluation dated September 15, 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. 50-368, Arkansas Nuclear One,
Unit No. 2, Pope County, Arkansas
Date of application for amendment: October 27, 1993
Brief description of amendment: The amendment relocated reactor
incore detector requirements from the TSs to the safety analysis
report.
Date of issuance: September 15, 1995
Effective date: September 15, 1995
Amendment No.: 163
Facility Operating License No. NPF-6. Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: December 8, 1993 (58 FR
64606) The Commission's related evaluation of the amendment is
contained in a Safety Evaluation dated September 15, 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. 50-368, Arkansas Nuclear One,
Unit No. 2, Pope County, Arkansas
Date of application for amendment: March 17, 1995
Brief description of amendment: The amendment transfers
requirements for cycle specific core operating limits from the
Technical Specifications to the Core Operating Limits Report.
Additionally, a reference to a statistical methodology for determining
uncertainties is being changed to reference a methodology that was
recently approved by the NRC.
Date of issuance: September 19, 1995
Effective date: September 19, 1995
Amendment No.: 164
Facility Operating License No. NPF-6. Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: July 19, 1995 (60 FR
37088) The Commission's related evaluation of the amendment is
contained in a Safety Evaluation dated September 19, 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. 50-368, Arkansas Nuclear One,
Unit No. 2, Pope County, Arkansas
Date of application for amendment: April 4, 1995, as supplemented
August 25, 1995
Brief description of amendment: The amendment provides a one-time
extension of the reactor coolant pump flywheel inservice inspection.
Date of issuance: September 22, 1995
Effective date: September 22, 1995
Amendment No.: 165
Facility Operating License No. NPF-6. Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: July 5, 1995 (60 FR
35069) The August 25, 1995, submittal did not change the original no
significant hazards consideration determination.The Commission's
related evaluation of the amendment is contained in a Safety Evaluation
dated September 22, 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. 50-368, Arkansas Nuclear One,
Unit No. 2, Pope County, Arkansas
Date of application for amendment: May 19, 1995 as supplemented
July 21, 1995.
Brief description of amendment: The amendment revises the
specifications to permit the containment personnel airlock doors to
remain open during fuel handling.
Date of issuance: September 28, 1995
Effective date: September 28, 1995
Amendment No.: 166
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 July 22, 1995, supplement provided clarifying information
and did not change the original 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: Tomlinson Library, Arkansas
Tech University, Russellville, AR 72801
Entergy Operations, Inc., Docket No. 50-368, Arkansas Nuclear One,
Unit No. 2, Pope County, Arkansas
Date of application for amendment: April 4, 1995, as
supplementedSeptember 28, 1995
Brief description of amendment: The amendment removes the
requirement to maintain water level 23 feet above irradiated fuel
assemblies in the reactor while latching and unlatching control element
assemblies.
Date of issuance: September 28, 1995
Effective date: September 28, 1995
Amendment No.: 167
Facility Operating License No. NPF-6. Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: August 16, 1995 (60 FR
42604) The September 28, 1995, submittal provided clarifying
information and did not change the original 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: Tomlinson Library, Arkansas
Tech University, Russellville, AR 72801
[[Page 52941]]
Entergy Operations, Inc., Docket No. 50-382, Waterford Steam
Electric Station, Unit 3, St. Charles Parish, Louisiana
Date of amendment request: June 22, 1994, as supplemented by
letters dated June 28, 1995 and August 22, 1995
Brief description of amendment: The amendment changes the Appendix
A TSs by increasing the control room radiation monitor setpoint (CRRMS)
to a fixed value of 5.45E-6 micro curies per cubic centimeters instead
of being set at two times the background.
Date of issuance: September 27, 1995
Effective date: Septembe 27, 1995
Amendment No.: 114
Facility Operating License No. NPF-38. Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: August 3, 1994 (59 FR
39586) The June 28, 1995 and August 22, 1995, letters provided
clarifying information that did not change the originial proposed no
significant hazards consideration determination. The Commission's
related evaluation of the amendment is contained in a Safety Evaluation
dated September 27, 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: August 11, 1995
Brief description of amendment: The amendment removes the Technical
Specifications for the Makeup, Purification, and Chemical Addition
Systems from the Technical Specifications (Section 3.2) and relocates
the pertinent design information, including tank volume and boron
concentrations, to the TMI-1 Updated Final Safety Analysis Report.
Date of issuance: September 19, 1995
Effective date: September 19, 1995
Amendment No.: 196
Facility Operating License No. DPR-50. Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: August 18, 1995 (60 FR
43172) The Commission's related evaluation of this amendment is
contained in a Safety Evaluation dated September 19, 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
Illinois Power Company and Soyland Power Cooperative, Inc., Docket
No. 50-461, Clinton Power Station, Unit No. 1, DeWitt County,
Illinois
Date of application for amendment: June 9, 1995
Brief description of amendment: The amendment modifies Technical
Specification 4.1, ``Site Location,'' to incorporate a description of
the exclusion area boundary. The change is necessary to ensure the
content of the technical specifications conform to Section 182 of the
Atomic Energy Act of 1954.
Date of issuance: September 14, 1995
Effective date: September 14, 1995
Amendment No.: 101
Facility Operating License No. NPF-62: The amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: July 19, 1995 (60 FR
37093) The Commission's related evaluation of the amendment is
contained in a Safety Evaluation dated September 14, 1995. No
significant hazards consideration comments received: No
Local Public Document Room location: The Vespasian Warner Public
Library, 120 West Johnson Street, Clinton, Illinois 61727
Power Authority of The State of New York, Docket No. 50-286, Indian
Point Nuclear Generating Unit No. 3, Westchester County, New York
Date of application for amendment: July 21, 1995
Brief description of amendment: The amendment revised Technical
Specifications Section 6.0 (Administrative Controls) to replace the
title-specific list of members on the Plant Operating Review Committee
(PORC) with a more general statement of membership requirements. The
scope of disciplines represented on the PORC was also expanded to
include nuclear licensing and quality assurance. The amendment also
changed the title ``Resident Manager'' to ``Site Executive Officer.''
This title change was an administrative change that did not affect the
reporting relationship, authority, or responsibility of the position.
Date of issuance: September 20, 1995
Effective date: As of the date of issuance to be implemented within
30 days.
Amendment No.: 163
Facility Operating License No. DPR-64: Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: August 16, 1995 (60 FR
42606) The Commission's related evaluation of the amendment is
contained in a Safety Evaluation dated September 20, 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
Public Service Electric & Gas Company, Docket No. 50-354, Hope
Creek Generating Station, Salem County, New Jersey
Date of application for amendment: April 25, 1994
Brief description of amendment: This amendment revises TS Section
3.8.1.1, ``A.C. Sources - Operating,'' TS Section 3.8.1.2, ``A.C.
Sources - Shutdown,'' and associated Bases, to increase the required
quantity of fuel in the Emergency Diesel Generator Fuel Oil Day Tanks
from 200 to 360 gallons.
Date of issuance: September 15, 1995
Effective date: As of the date of issuance and shall be implemented
within 60 days.
Amendment No.: 79
Facility Operating License No. NPF-57: This amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: June 8, 1994 (59 FR
29632)The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated September 15, 1995. No significant hazards
consideration comments received: No
Local Public Document Room location: Pennsville Public Library,
190 S. Broadway, Pennsville, New Jersey 08070
Public Service Electric & Gas Company, Docket No. 50-354, Hope
Creek Generating Station, Salem County, New Jersey
Date of application for amendment: January 20, 1995
Brief description of amendment: This amendment changes Technical
Specification (TS) 4.1.3.1.2.b, ``Control Rods - Surveillance
Requirement'' to change the required action to be taken when a control
rod becomes immovable due to excessive friction from ``at least once
per'' 24 hours to ``within'' 24 hours.
Date of issuance: September 20, 1995
Effective date: As of its date of issuance, to be implemented
within 60 days.
Amendment No.: 80
Facility Operating License No. NPF-57: This amendment revised the
Technical Specifications.
[[Page 52942]]
Date of initial notice in Federal Register: August 2, 1995 (60 FR
39452) The Commission's related evaluation of the amendment is
contained in a Safety Evaluation dated September 20, 1995. No
significant hazards consideration comments received: No
Local Public Document Room location: Pennsville Public Library, 190
S. Broadway, Pennsville, New Jersey 08070
Public Service Electric & Gas Company, Docket No. 50-354, Hope
Creek Generating Station, Salem County, New Jersey
Date of application for amendment: January 11, 1995
Brief description of amendment: This amendment changes Technical
Specification (TS) 3/4.3.8, ``Turbine Overspeed Protection System,''
removing these requirements from the TS and relocating the Bases to the
Hope Creek Updated Final Safety Analysis Report (UFSAR) and the
Surveillance Requirements to the applicable surveillance procedures.
The Limiting Conditions for Operation (LCOs) are eliminated.
Date of issuance: September 25, 1995
Effective date: As of the date of issuance and shall be implemented
within 60 days.
Amendment No.: 81
Facility Operating License No. NPF-57: This amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: August 2, 1995 (60 FR
39451). The Commission's related evaluation of the amendment is
contained in a Safety Evaluation dated September 25, 1995. No
significant hazards consideration comments received: No
Local Public Document Room location: Pennsville Public Library, 190
S. Broadway, Pennsville, New Jersey 08070
Public Service Electric & Gas Company, Docket No. 50-354, Hope
Creek Generating Station, Salem County, New Jersey
Date of application for amendment: September 29, 1994
Brief description of amendment: This amendment changes Technical
Specification (TS) Sections 3/4.3.7.2, ``Seismic Monitoring
Instrumentation,'' and 3/4.3.7.3, ``Meteorological Instrumentation,''
to remove the requirements from the TS and relocate the appropriate
descriptive information and testing requirements to the Hope Creek
Updated Final Safety Analysis Report.
Date of issuance: September 25, 1995
Effective date: As of the date of issuance, to be implemented
within 60 days.
Amendment No.: 82
Facility Operating License No. NPF-57: This amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: August 2, 1995 (60 FR
39449). The Commission's related evaluation of the amendment is
contained in a Safety Evaluation dated September 25, 1995. No
significant hazards consideration comments received: No
Local Public Document Room location: Pennsville Public Library, 190
S. Broadway, Pennsville, New Jersey 08070
Public Service Electric & Gas Company, Docket Nos. 50-272 and 50-
311, Salem Nuclear Generating Station, Unit Nos. 1 and 2, Salem
County, New Jersey
Date of application for amendments: September 20, 1994
Brief description of amendments: The amendments change the Channel
Functional Test surveillance frequency for the Manual Reactor Trip
Switches and Reactor Trip Breakers (RTB) and relocate the RTB
maintenance requirements from the Technical Specifications to the Salem
Updated Final Safety Analysis Report.
Date of issuance: September 18, 1995
Effective date: Both units, as of the date of issuance, to be
implemented within 60 days.
Amendment Nos.: 176 and 157
Facility Operating License Nos. DPR-70 and DPR-75. The amendments
revised the Technical Specifications.
Date of initial notice in Federal Register: November 9, 1994 (59 FR
55890 The Commission's related evaluation of the amendments is
contained in a Safety Evaluation dated September 18, 1995. No
significant hazards consideration comments received: No
Local Public Document Room location: Salem Free Public Library, 112
West Broadway, Salem, New Jersey 08079
Public Service Electric & Gas Company, Docket Nos. 50-272 and 50-
311, Salem Nuclear Generating Station, Unit Nos. 1 and 2, Salem
County, New Jersey
Date of application for amendments: January 21, 1994, as
supplemented June 28 and September 13, 1994, and April 4, 1995.
Brief description of amendments: Revised Technical Specifications
3.8.2.3, ``125-Volt D.C. DISTRIBUTION - OPERATING.''
Date of issuance: September 19, 1995
Effective date: Both units, as of the day of issuance and shall be
implemented within 60 days.
Amendment Nos.: 177 and 158
Facility Operating License Nos. DPR-70 and DPR-75. The amendments
revised the Technical Specifications.
Date of initial notice in Federal Register: April 28, 1994 (58 FR
22012) The June 28 and September 13, 1994, and April 4, 1995 letters
provided clarifying information that did not change the scope of the
January 21, 1994 application and initial proposed no significant
hazards consideration determination, nor go beyond the scope of the
Federal Register notice. The Commission's related evaluation of the
amendments is contained in a Safety Evaluation dated September 19,
1995. No significant hazards consideration comments received: No
Local Public Document Room location: Salem Free Public Library, 112
West Broadway, Salem, New Jersey 08079
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 19, 1995, as supplemented
on August 21, 1995.
Brief description of amendment: The amendment revises the Technical
Specifications to change the required test frequency for the reactor
building spray nozzle flow test from once per five years to once per
ten years.
Date of issuance: September 18, 1995
Effective date: September 18, 1995
Amendment No.: 127
Facility Operating License No. NPF-12: Amendment revises the
Technical Specifications.
Date of initial notice in Federal Register: July 19, 1995 (60 FR
37100). The August 21, 1995 letter provided supplemental 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 September 18, 1995. No
significant hazards consideration comments received: No
Local Public Document Room location: Fairfield County Library, 300
Washington Street, Winnsboro, SC 29180
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: July 28, 1995
[[Page 52943]]
Brief description of amendment: The amendment revises the Technical
Specifications to exclude the requirement to perform the slave relay
test of the 36-inch containment purge supply and exhaust valves on a
quarterly basis while the plant is in Modes 1, 2, 3, or 4.
Date of issuance: September 18, 1995
Effective date: September 18, 1995
Amendment No.: 128
Facility Operating License No. NPF-12. Amendment revises the
Technical Specifications.
Date of initial notice in Federal Register: August 16, 1995 (60 FR
42608) 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
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 19, 1995, as supplemented
on August 21, 1995.
Brief description of amendment: The amendment revises the Technical
Specifications to change the required test frequency for the reactor
building spray nozzle flow test from once per five years to once per
ten years.
Date of issuance: September 18, 1995
Effective date: September 18, 1995
Amendment No.: 129
Facility Operating License No. NPF-12. Amendment revises the
Technical Specifications.
Date of initial notice in Federal Register: July 19, 1995 (60 FR
37100). The August 21, 1995 letter provided supplemental 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 September 18, 1995. No
significant hazards consideration comments received: No
Local Public Document Room location: Fairfield County Library, 300
Washington Street, Winnsboro, SC 29180
The Cleveland Electric Illuminating Company, Centerior Service
Company, Duquesne Light Company, Ohio Edison Company, Pennsylvania
Power Company, Toledo Edison Company, Docket No. 50-440, Perry
Nuclear Power Plant, Unit No. 1, Lake County, Ohio
Date of application for amendment: April 3, 1995
Brief description of amendment: The amendment revised the Technical
Specifications (TS) to relocate radiological effluent and radiological
environmental monitoring TS to the Offsite Dose Calculation Manual or
to the Process Control Program. Programmatic controls for radioactive
effluent and radiological environmental monitoring were included in TS
6.8.4.
Date of issuance: September 15, 1995
Effective date: September 15, 1995
Amendment No.: 72
Facility Operating License No. NPF-58: This amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: May 10, 1995 (60 FR
24921) The Commission's related evaluation of the amendment is
contained in a Safety Evaluation dated September 15, 1995. No
significant hazards consideration comments received: No
Local Public Document Room location: Perry Public Library, 3753
Main Street, Perry, Ohio 44081
The Cleveland Electric Illuminating Company, Centerior Service
Company, Duquesne Light Company, Ohio Edison Company, Pennsylvania
Power Company, Toledo Edison Company, Docket No. 50-440, Perry
Nuclear Power Plant, Unit No. 1, Lake County, Ohio
Date of application for amendment: June 1, 1995
Brief description of amendment: The amendment revised the Technical
Specifications to make them more restrictive regarding control rod
drive scram time testing. CRD scram time testing would be required
following maintenance prior to considering the CRD operable, and could
be performed at any reactor pressure. Additional testing would be
required when reactor coolant pressure is greater than or equal to 950
psig and prior to 40 percent rated thermal power.
Date of issuance: September 26, 1995
Effective date: September 26, 1995
Amendment No. 73
Facility Operating License No. NPF-58: This amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: August 2, 1995 (60 FR
39452) The Commission's related evaluation of the amendment is
contained in a Safety Evaluation dated September 26, 1995. No
significant hazards consideration comments received: No
Local Public Document Room location: Perry Public Library, 3753
Main Street, Perry, Ohio 44081
Washington Public Power Supply System, Docket No. 50-397, Nuclear
Project No. 2, Benton County, Washington
Date of application for amendment: January 14, 1992, as
supplemented by letters dated February 10, 1995, and August 16, 1995.
Brief description of amendment: The amendment revises technical
specification surveillance requirements regarding demonstration of jet
pump operability and corrects several administrative discrepancies.
Date of issuance: September 18, 1995
Effective date: September 18, 1995, to be implemented within 30
days of issuance
Amendment No.: 141
Facility Operating License No. NPF-21: The amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: May 27, 1992 (57 FR
22272) and March 29, 1995 (60 FR 16204). The August 16, 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 September 18, 1995.No
significant hazards consideration comments received: No
Local Public Document Room location: Richland Public Library, 955
Northgate Street, Richland, Washington 99352
Dated at Rockville, Maryland, this 3rd 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
[Doc. 95-25006 Filed 10-10-95; 8:45 am]
BILLING CODE 7590-01-F
This is a copy of a public record, reproduced as it was published. It is not legal advice, and it may not be the version a court would rely on. Check the official source before you cite it.