[Federal Register Volume 61, Number 168 (Wednesday, August 28, 1996)]
[Notices]
[Pages 44353-44368]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: X96-30828]
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NUCLEAR REGULATORY COMMISSION
Biweekly Notice
Applications and Amendments to Facility Operating Licenses
Involving No Significant Hazards Considerations
I. Background
Pursuant to Public Law 97-415, the U.S. Nuclear Regulatory
Commission (the Commission or NRC staff) is publishing this regular
biweekly notice. Public Law 97-415 revised section 189 of the Atomic
Energy Act of 1954, as amended (the Act), to require the Commission to
publish notice of any amendments issued, or proposed to be issued,
under a new provision of section 189 of the Act. This provision grants
the Commission the authority to issue and make immediately effective
any amendment to an operating license upon a determination by the
Commission that such amendment involves no significant hazards
consideration, notwithstanding the pendency before the Commission of a
request for a hearing from any person.
This biweekly notice includes all notices of amendments issued, or
proposed to be issued from August 3, 1996, through August 16, 1996. The
last biweekly notice was published on August 14, 1996 (61 FR 42274).
Notice Of Consideration Of Issuance Of Amendments To Facility
Operating Licenses, Proposed No Significant Hazards Consideration
Determination, And Opportunity For A Hearing
The Commission has made a proposed determination that the following
amendment requests involve no significant hazards consideration. Under
the Commission's regulations in 10 CFR 50.92, this means that operation
of the facility in accordance with the proposed amendment would not (1)
involve a significant increase in the probability or consequences of an
accident previously evaluated; or (2) create the possibility of a new
or different kind of accident from any accident previously evaluated;
or (3) involve a significant reduction in a margin of safety. The basis
for this proposed determination for each amendment request is shown
below.
The Commission is seeking public comments on this proposed
determination. Any comments received within 30 days after the date of
publication of this notice will be considered in making any final
determination.
Normally, the Commission will not issue the amendment until the
expiration of the 30-day notice period. However, should circumstances
change during the notice period such that failure to act in a timely
way would result, for example, in derating or shutdown of the facility,
the Commission may issue the license amendment before the expiration of
the 30-day notice period, provided that its final determination is that
the amendment involves no significant hazards consideration. The final
determination will consider all public and State comments received
before action is taken. Should the Commission take this action, it will
publish in the Federal Register a notice of issuance and provide for
opportunity for a hearing after issuance. The Commission expects that
the need to take this action will occur very infrequently.
Written comments may be submitted by mail to the Chief, Rules
Review and Directives Branch, Division of Freedom of Information and
Publications Services, Office of Administration, U.S. Nuclear
Regulatory Commission, Washington, DC 20555-0001, and should cite the
publication date and page number of this Federal Register notice.
Written comments may also be delivered to Room 6D22, Two White Flint
North, 11545 Rockville Pike, Rockville, Maryland from 7:30 a.m. to 4:15
p.m. Federal workdays. Copies of written comments received may be
examined at the NRC Public Document Room, the Gelman Building, 2120 L
Street, NW., Washington, DC. The filing of requests for a hearing and
petitions for leave to intervene is discussed below.
By September 27, 1996, the licensee may file a request for a
hearing with respect to issuance of the amendment to the subject
facility operating license and any person whose interest may be
affected by this proceeding and who wishes to participate as a party in
the proceeding must file a written request for a hearing and a petition
for leave to intervene. Requests for a hearing and a petition for leave
to intervene shall be filed in accordance with the Commission's ``Rules
of Practice for
[[Page 44354]]
Domestic Licensing Proceedings'' in 10 CFR Part 2. Interested persons
should consult a current copy of 10 CFR 2.714 which is available at the
Commission's Public Document Room, the Gelman Building, 2120 L Street,
NW., Washington, DC and at the local public document room for the
particular facility involved. If a request for a hearing or petition
for leave to intervene is filed by the above date, the Commission or an
Atomic Safety and Licensing Board, designated by the Commission or by
the Chairman of the Atomic Safety and Licensing Board Panel, will rule
on the request and/or petition; and the Secretary or the designated
Atomic Safety and Licensing Board will issue a notice of a hearing or
an appropriate order.
As required by 10 CFR 2.714, a petition for leave to intervene
shall set forth with particularity the interest of the petitioner in
the proceeding, and how that interest may be affected by the results of
the proceeding. The petition should specifically explain the reasons
why intervention should be permitted with particular reference to the
following factors: (1) the nature of the petitioner's right under the
Act to be made a party to the proceeding; (2) the nature and extent of
the petitioner's property, financial, or other interest in the
proceeding; and (3) the possible effect of any order which may be
entered in the proceeding on the petitioner's interest. The petition
should also identify the specific aspect(s) of the subject matter of
the proceeding as to which petitioner wishes to intervene. Any person
who has filed a petition for leave to intervene or who has been
admitted as a party may amend the petition without requesting leave of
the Board up to 15 days prior to the first prehearing conference
scheduled in the proceeding, but such an amended petition must satisfy
the specificity requirements described above.
Not later than 15 days prior to the first prehearing conference
scheduled in the proceeding, a petitioner shall file a supplement to
the petition to intervene which must include a list of the contentions
which are sought to be litigated in the matter. Each contention must
consist of a specific statement of the issue of law or fact to be
raised or controverted. In addition, the petitioner shall provide a
brief explanation of the bases of the contention and a concise
statement of the alleged facts or expert opinion which support the
contention and on which the petitioner intends to rely in proving the
contention at the hearing. The petitioner must also provide references
to those specific sources and documents of which the petitioner is
aware and on which the petitioner intends to rely to establish those
facts or expert opinion. Petitioner must provide sufficient information
to show that a genuine dispute exists with the applicant on a material
issue of law or fact. Contentions shall be limited to matters within
the scope of the amendment under consideration. The contention must be
one which, if proven, would entitle the petitioner to relief. A
petitioner who fails to file such a supplement which satisfies these
requirements with respect to at least one contention will not be
permitted to participate as a party.
Those permitted to intervene become parties to the proceeding,
subject to any limitations in the order granting leave to intervene,
and have the opportunity to participate fully in the conduct of the
hearing, including the opportunity to present evidence and cross-
examine witnesses.
If a hearing is requested, the Commission will make a final
determination on the issue of no significant hazards consideration. The
final determination will serve to decide when the hearing is held.
If the final determination is that the amendment request involves
no significant hazards consideration, the Commission may issue the
amendment and make it immediately effective, notwithstanding the
request for a hearing. Any hearing held would take place after issuance
of the amendment.
If the final determination is that the amendment request involves a
significant hazards consideration, any hearing held would take place
before the issuance of any amendment.
A request for a hearing or a petition for leave to intervene must
be filed with the Secretary of the Commission, U.S. Nuclear Regulatory
Commission, Washington, DC 20555-0001, Attention: 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 20555-0001,
and to the attorney for the licensee.
Nontimely filings of petitions for leave to intervene, amended
petitions, supplemental petitions and/or requests for a hearing will
not be entertained absent a determination by the Commission, the
presiding officer or the Atomic Safety and Licensing Board that the
petition and/or request should be granted based upon a balancing of
factors specified in 10 CFR 2.714(a)(1)(i)-(v) and 2.714(d).
For further details with respect to this action, see the
application for amendment which is available for public inspection at
the Commission's Public Document Room, the Gelman Building, 2120 L
Street, NW., Washington, DC, and at the local public document room for
the particular facility involved.
Carolina Power & Light Company, et al., Docket No. 50-400, Shearon
Harris Nuclear Power Plant, Unit 1, Wake and Chatham Counties,
North Carolina
Date of amendment request: July 19, 1996
Description of amendment request: The proposed amendment would
revise Technical Specification (TS) Section 3/4.6.2, Containment Spray
System, to extend the surveillance interval for performance of an air
or smoke flow test through containment spray nozzles from once per 5
years to once per 10 years. This change is consistent with the guidance
in NRC Generic Letter 93-05, ``Line Item Technical Specifications
Improvements to Reduce Surveillance Requirements for Testing During
Power Operations,'' and NUREG-1366, ``Improvements To Technical
Specifications Surveillance Requirements.''
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. The proposed amendment does not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
The proposed extended testing frequency of containment spray
nozzles will not affect any initiators of any previously evaluated
accidents or change the manner of operation for any system or
component. The containment spray system serves a mitigating function
by removing heat and fission products from a post accident
containment atmosphere. Increasing the surveillance test interval
will not affect the system's ability to provide this function.
Therefore, there would be no increase in the probability or
consequences of an accident previously evaluated.
[[Page 44355]]
2. The proposed amendment does not create the possibility of a
new or different kind of accident from any accident previously
evaluated.
Since the proposed change affects only a surveillance frequency,
it will not involve any physical alterations to plant equipment or
alter the manner in which any safety-related system performs its
function. Therefore, the proposed change does not create the
possibility of a new or different kind of accident from any accident
previously evaluated.
3. The proposed amendment does not involve a significant
reduction in the margin of safety.
The proposed change does not affect any Final Safety Analysis
Report (FSAR) Chapter 15 accident analyses or impact the margin of
safety for the containment spray system as defined in the Bases to
the Technical Specifications. Therefore, the proposed change does
not involve a significant reduction in the margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: Cameron Village Regional
Library, 1930 Clark Avenue, Raleigh, North Carolina 27605
Attorney for licensee: William D. Johnson, Vice President and
Senior Counsel, Carolina Power & Light Company, Post Office Box 1551,
Raleigh, North Carolina 27602
NRC Project Director: Eugene V. Imbro
Commonwealth Edison Company, Docket Nos. 50-254 and 50-265, Quad
Cities Nuclear Power Station, Units 1 and 2, Rock Island County,
Illinois
Date of amendment request: June 10, 1996
Description of amendment request: To change the technical
specifications to reflect the transition from General Electric Company
(GE) to Siemens Power Corporation (SPC) as the fuel supplier for the
Quad Cities Nuclear Power Station, Units 1 and 2.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1) Involve a significant increase in the probability or
consequences of an accident previously evaluated.
The probability of an evaluated accident is derived from the
probabilities of the individual precursors to that accident. The
consequences of an evaluated accident are determined by the
operability of plant systems designed to mitigate those
consequences. Limits will be established consistent with NRC
approved methods to ensure that fuel performance during normal,
transient, and accident conditions is acceptable. The proposed
Technical Specifications amendment reflects previously approved SPC
methodology used to analyze normal operations, including anticipated
operational occurrences (AOOs), and to determine the potential
consequences of accidents.
Licensing Methods and Models
The proposed amendment is to support operation with NRC approved
fuel and licensing methods supplied from Siemens Power Corporation.
In accordance with FSAR Chapter 15, the same accidents and
transients will be analyzed with the new fuel and methods as were
analyzed by GE for GE fuel. The analysis methods and models are NRC
approved. These approved methods and models are used to determine
the fuel thermal limits (e.g., LHGR, APLHGR, MCPR). The SPC core
monitoring code enables the site to monitor keff as well as rod
density to perform the reactivity anomaly surveillance. This is
consistent with GE methodology. The support systems for minimizing
the consequences of transients and accidents are not affected by the
proposed amendment. Therefore, the change in licensing analysis
methods and models does not significantly increase the probability
of an accident or the consequences of an accident previously
identified.
New Fuel Design
The use of ATRIUM 9B fuel at Quad Cities does not involve a
significant increase in the probability or consequences of any
accident previously evaluated in the FSAR. The ATRIUM-9B fuel is
generically approved for use as a reload BWR fuel type (Reference:
ANF-89-014(P)(A) Rev. 1 Supplement 1, General Mechanical Design for
Advanced Nuclear Fuels 9X9-IX and 9X9-9X BWR Reload Fuel). Limiting
postulated occurrences and normal operation have been analyzed using
NRC-approved methods for the ATRIUM 9B fuel design to ensure that
safety limits are protected and that acceptable transient and
accident performance is maintained.
The reload fuel has no adverse impact on the performance of in-
core neutron flux instrumentation or CRD response. The ATRIUM-9B
fuel design will not adversely affect performance of neutron
instrumentation nor will it adversely affect the movement of control
blades relative to the GE fuel. The exterior dimensions of the
ATRIUM-9B fuel have been evaluated by ComEd; the SPC fuel provides
adequate clearances relative to the GE10 fuel installed at Quad
Cities. Thus, no increased interactions with the adjacent control
blade and nuclear instrumentation are created. Additionally, given
the above mentioned overall envelope similarities, no problems are
anticipated with other station equipment such as the fuel storage
racks, the new fuel inspection stand and the spent fuel pool fuel
preparation machine. Therefore, the probability of adverse
interactions between the Siemens fuel and components in the core and
fuel handling equipment is not significantly increased.
The ATRIUM 9B design is neutronically compatible with the
existing fuel types and core components in the Quad Cities core. SPC
tests have demonstrated that the ATRIUM-9B fuel design is
hydraulically compatible with the GE9/GE10 fuel. The bundle pressure
drop characteristics of the ATRIUM 9B bundle are similar to those of
the GE9/GE10 fuel design, hence core thermal-hydraulic stability
characteristics are not adversely affected by the ATRIUM 9B design.
Cycle stability calculations are performed by SPC. Therefore, the
probability of thermal hydraulic instability is not significantly
increased.
An evaluation of the Emergency Procedures is being performed to
ensure that the use of the ATRIUM-9B fuel at Quad Cities does not
alter any assumptions previously made in evaluating the radiological
consequences of an accident at Quad Cities Station. Therefore, the
radiological consequences of accidents are not significantly
increased.
Methods approved by the NRC are being used in the evaluation of
fuel performance during normal and abnormal operating conditions.
The ComEd and SPC methods to be used for the cycle specific
transient analyses have been previously NRC approved. The proposed
methodologies are administrative in nature and do not significantly
affect any accident precursors or accident results; as such, the
proposed incorporation of the SPC methodologies for Quad Cities does
not significantly increase the probability or consequences of any
previously evaluated accidents. The description of the fuel is
modified to include the water box design of the NRC approved ATRIUM-
9B fuel. This change is administrative.
Review of the above concludes that the probability of occurrence
and the consequences of an accident previously evaluated in the
safety analysis report have not been significantly increased.
* * * * *
2) Create the possibility of a new or different kind of accident
from any accident previously evaluated:
Creation of the possibility of a new or different kind of
accident would require the creation of one or more new precursors of
that accident. New accident precursors may be created by
modifications of the plant configuration, including changes in
allowable modes of operation.
Licensing Methods and Models
The proposed Technical Specification amendment reflects
previously approved SPC methodology used to analyze normal
operations, including AOOs, and to determine the potential
consequences of accidents. In accordance with FSAR Chapter 15, the
same accidents and transients will be analyzed with the new fuel and
methods as were analyzed by GE for GE fuel. As stated above, the
proposed changes do not permit modes of operation which differ from
those currently permitted; therefore, the possibility of a new or
different kind of accident is not created. Plant support equipment
is not affected by the proposed changes; therefore, no new failure
modes are created.
New Fuel Design
The basic design concept of a 9x9 fuel pin array with an
internal water box has been used in various lead assembly programs
and in reload quantities in Europe since 1986.
[[Page 44356]]
WNP-2 has loaded reload quantities since 1991. Approximately 650
water box assemblies have been irradiated in the United States
through 1995, with a substantially higher number being irradiated
overseas. The NRC has reviewed and approved the ATRIUM-9B fuel
design (Reference: ANF-89-014(P)(A) Rev. 1 Supplement 1, Generic
Mechanical Design for Advanced Nuclear Fuels 9X9-IX and 9X9-9X BWR
Reload Fuel). The similarities in fuel design and operation between
GE and SPC, and the previous Boiling Water Reactor experience with
both vendors' fuel indicate there would be no new or different types
of accidents for Quad Cities than have been considered for the
existing fuel. Therefore, the use of ATRIUM-9B fuel at Quad Cities
does not create the possibility of a new or different kind of
accident from any accident previously evaluated.
* * * * *
3) Involve a significant reduction in the margin of safety for
the following reasons:
The existing margin to safety is provided by the existing
acceptance criteria (e.g., 10CFR50.46 limits). The proposed
Technical Specification amendment reflects previously approved SPC
methodology used to demonstrate that the existing acceptance
criteria are satisfied. The revised methodology has been previously
reviewed and approved by the USNRC for application to reload cores
of GE BWRs. References for the Licensing Topical Reports which
document this methodology, and include the Safety Evaluation Reports
prepared by the USNRC, are added to the Reference section of the
Technical Specifications as part of this amendment.
Licensing Methods and Models
The proposed amendment does not involve changes to the existing
operability criteria. NRC approved methods and established limits
(implemented in the COLR) ensure acceptable margin is maintained.
The ComEd and SPC reload methodologies for the ATRIUM-9B reload
design are consistent with the Technical Specification Bases. The
Limiting Conditions for Operation are taken into consideration while
performing the cycle specific and generic reload safety analyses.
NRC approved methods are listed in Section 6 of the Technical
Specifications.
Analyses performed with NRC-approved methodology have
demonstrated that fuel design and licensing criteria will be met
during normal and abnormal operating conditions. The same margins of
safety are utilized by SPC as GE (e.g., limits on peak cladding
temperature, cladding oxidation, plastic strain). Therefore, there
is not a significant reduction in the margin of safety.
New Fuel Design
The exterior dimensions of the ATRIUM-9B fuel assembly result in
equivalent clearances relative to the GE10B. Thus, no increased
interactions with the adjacent control blade and nuclear
instrumentation are created. The change does not adversely impact
equipment important to safety; therefore,the margin of safety is not
significantly reduced.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
requested amendments involve no significant hazards consideration.
Local Public Document Room location: Dixon Public Library, 221
Hennepin Avenue, Dixon, Illinois 61021
Attorney for licensee: Michael I. Miller, Esquire; Sidley and
Austin, One First National Plaza, Chicago, Illinois 60603
NRC Project Director: Robert A. Capra
Duke Power Company, Docket Nos. 50-269, 270 and 50-287, Oconee
Nuclear Station, Units 1, 2 and 3, Oconee County, South Carolina
Date of amendment request: August 12, 1996
Description of amendment request: The proposed change would
implement the performance-based containment leak rate testing
provisions of Option B to 10 CFR Part 50 Appendix J for the Type A
(containment) testing 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:
The following analysis is presented, pursuant to 10 CFR 50.91,
to demonstrate that the proposed change will not create a
Significant Hazard Consideration.
1. The proposed change will not involve a significant increase
in the probability or consequences of an accident previously
evaluated.
Containment leak rate testing is not an initiator of any
accident; the proposed change does not affect reactor operations or
accident analysis, and has no significant radiological consequences.
Therefore, this proposed change will not involve an increase in the
probability or consequences of any previously-evaluated accident.
2. The proposed change will not create the possibility of any
new accident not previously evaluated.
The proposed change does not affect normal plant operations or
configuration, or change any design basis. The proposed changes will
not affect the response of [the] containment during a design basis
accident.
3. There is no significant reduction in a margin of safety.
The proposed changes are based on NRC-accepted provisions, and
maintain necessary levels of reliability of containment integrity.
The performance-based approach to leakage rate testing recognizes
that historically good results of containment testing provide
appropriate assurance of future containment integrity; this supports
the conclusion that the impact on the health and safety of the
public as a result of extended test intervals is negligible.
Based on the above, no significant hazards consideration is
created by the proposed change.
The NRC has reviewed the licensee's analysis and, based on
thisreview, 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., Docket No. 50-313, Arkansas Nuclear One,
Unit No. 1, Pope County, Arkansas
Date of amendment request: May 31, 1996
Description of amendment request: The proposed amendment revises
the surveillance test interval for the reactor protection system
reactor trip breakers, reactor trip modules, and electronic trip relays
from 1 month to 6 months. In addition to requesting a change to the
Arkansas Nuclear One, Unit 1 Technical Specifications, the request also
proposes the same changes to NUREG-1430, Standard Technical
Specifications - Babcock and Wilcox Plants.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
Criterion 1 - Does Not Involve a Significant Increase in the
Probability or Consequences of an Accident Previously Evaluated.
The accident mitigation features of the plant are not affected
by the proposed test interval extension. The results of the B&W
Owners Group Topical Report BAW-10167, Supplement 3, ``Justification
for increasing The Reactor Trip System On-Line Test Intervals,''
show that the test interval extension of the reactor protection
system trip devices is not a significant contributor to trip system
unavailability or the risk of core damage.
Therefore, this change does not involve a significant increase
in the probability or consequences of any accident previously
evaluated.
Criterion 2. Does Not Create the Possibility of a New or
Different Kind of Accident from any Previously Evaluated.
The reactor trip device surveillance test interval is not, in
and of itself, considered to be an accident initiator. Failure of a
trip device to function is an analyzed condition and does not
constitute a new or different kind of accident.
Therefore, this change does not create the possibility of a new
or different kind of accident from any previously evaluated.
Criterion 3. Does Not Involve a Significant Reduction in the
Margin of Safety.
The results of the B&W Owners Group Topical Report BAW-10167,
Supplement 3, ``Justification for Increasing The Reactor Trip
[[Page 44357]]
System On-Line Test Intervals,'' show that the test interval
extension of the reactor protection system trip devices is not a
significant contributor to trip system unavailability or the risk of
core damage. In addition, the uncertainty analysis contained in BAW-
10167 confirms the robustness of the results by demonstrating that
even with an order of magnitude change in the failure data, the
incremental increase due to an increased test interval is
insignificant. Entergy Operations has reviewed BAW-10167 and found
it applicable to ANO-1.
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
Entergy Operations, Inc., Docket Nos. 50-313 and 50-368, Arkansas
Nuclear One, Unit Nos. 1 and 2 (ANO-1&2), Pope County, Arkansas
Date of amendment request: May 9, 1996
Description of amendment request: The proposed amendment changes
the name of Arkansas Power and Light Company (AP&L) to Entergy
Arkansas, Inc. in both the Operating License and the Technical
Specifications. AP&L is licensed to own and possess Arkansas Nuclear
One (ANO). The company licensed to operate ANO, Entergy Operations,
Inc. is unaffected by this change.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. Does Not Involve a Significant Increase in the Probability or
Consequences of an Accident Previously Evaluated.
The proposed change documents changing the legal name of the
company. The proposed change will not affect any other obligations.
The company will continue to own all of the same assets, will
continue to serve the same customers, and will continue to honor all
existing obligations and commitments. 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.
The administrative changes in the operating license requirements
do not involve any change in the design of the plant. Therefore,
this change 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 is administrative in nature and does not
reduce the margin of safety imposed by any current requirements.
Therefore, this change does not involve a significant reduction in
the margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: Tomlinson Library, Arkansas
Tech University, Russellville, AR 72801
Attorney for licensee: Nicholas S. Reynolds, Esquire, Winston and
Strawn, 1400 L Street, N.W., Washington, DC 20005-3502
NRC Project Director: William D. Beckner
Florida Power and Light Company, Docket Nos. 50-250 and 50-251,
Turkey Point Plant Units 3 and 4, Dade County, Florida
Dates of amendment request: July 17, 1996
Description of amendment request: The licensee proposed to change
the Turkey Point Units 3 and 4 Technical Specifications (TS) to
implement 10 CFR 50, Appendix J, Option B, for containment leakage
testing. Changes include relocating the details for containment testing
to the ``containment leakage rate testing program'' and adding the
requirements of the containment leakage rate testing program to TS
6.8.4, which describes facility programs.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below.
(1) Operation of the facility in accordance with the proposed
amendments would not involve a significant increase in the
probability or consequences of an accident previously evaluated.
The proposed amendments do not involve a significant increase in
the probability or consequences of an accident previously evaluated
because:
a) These proposed changes are all consistent with NRC
requirements and guidance for implementation of 10 CFR 50, Appendix
J, Option B.
b) Based on industry and NRC evaluations performed in support of
developing Option B, these changes potentially result in a minor
increase in the consequences of an accident previously evaluated due
to the expanded testing intervals. However, the proposed changes do
not result in an increase in the core damage frequency since the
containment system is used for mitigation purposes only.
c) These changes are expected to result in increased attention
to components with poor leakage test history as part of the
performance-based nature of Option B, such that the marginally
increased consequences from the expanded testing intervals may be
further reduced or negated.
Therefore, these changes do 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 use of the modified specifications can not create the
possibility of a new or different kind of accident from any
previously evaluated since the proposed amendments will not change
the physical plant or the modes of plant operation defined in the
facility operating license. No new failure mode is introduced due to
the implementation of a performance-based program for containment
leakage rate testing, since the proposed changes do not involve the
addition or modification of equipment, nor do they alter the design
or operation of affected plant systems, structures, or components.
(3) Operation of the facility in accordance with the proposed
amendments would not involve a significant reduction in a margin of
safety.
The operating limits and functional capabilities of the affected
systems, structures, and components are basically unchanged by the
proposed amendments due to the following reasons:
a) The acceptance criteria for total integrated containment
leakage of 1.0 La is consistent with the current technical
specifications and is within the design basis accident assumptions,
and therefore does not reduce the margin of safety.
b) The increase in intervals between leak-test surveillances
will not significantly reduce the margin of safety as shown by
findings in NUREG 1493, ``Performance-Based Containment Leak-Test
Program'', which was based on implementation of the performance-
based testing of Option B.
Therefore these changes do not involve a significant reduction
in the margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 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: Frederick J. Hebdon
[[Page 44358]]
Georgia Power Company, Oglethorpe Power Corporation, Municipal
Electric Authority of Georgia, City of Dalton, Georgia, Docket Nos.
50-321 and 50-366, Edwin I. Hatch Nuclear Plant, Units 1 and 2,
Appling County, Georgia
Date of amendment request: May 21, 1996
Description of amendment request: The proposed change to the
condensate storage tank (CST) level indication would ensure that the
water level is sufficient to provide 50,000 gallons of water for core
spray makeup to the reactor pressure vessel.
Technical Specification (TS) Surveillance Requirement (SR)
3.5.2.2.b for ECCS - Shutdown states: ``Condensate storage tank (CST)
water level is [greater than or equal to] 12 feet.'' The corresponding
Bases state: ''... the CST contains [greater than or equal to] 150,000
gallons of water, equivalent to 12 feet, ensures that the CS System can
supply at least 50,000 gallons of makeup water to the RPV.''
Subsequent licensee analyses confirmed that Plant Hatch Units 1 and
2 CST configurations are different; that is, for both CSTs, a water
level of 12 feet is not equivalent to the required capacity of 150,000
gallons of water. Based on these calculations, the correct level for
the Unit 1 CST is 13 feet, and the correct level for the Unit 2 CST is
15 feet.
The proposed change would revise Unit 1 and Unit 2 SR 3.5.2.2.b to
require a CST water level of greater than or equal to 13 feet and
greater than or equal to 15 feet, respectively, to ensure at least
50,000 gallons of water are available for core spray (CS) makeup to the
reactor pressure vessel (RPV).
The associated Bases for each unit will be revised accordingly.
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 change does not involve a significant
increase in the probability or consequences of an accident
previously evaluated, because this administrative change to the CST
water level does not alter the operation of any plant system or
component. The proposed change does not involve a physical
modification to any structure, system, or component. The minimum CST
water level for each unit is being increased to account for the
height of the CS suction standpipe within each CST and the
differences in the Unit 1 and
Unit 2 CST diameters (gallons/ft of water) as follows:
a. Unit 1 - The proposed minimum water level is calculated as:
CS suction standpipe height of 9 ft + (50,000 gallons divided by
12,704 gallons/ft) = 12.93 ft or 13 ft.
b. Unit 2 - The proposed minimum water level is calculated as:
CS suction standpipe height of 10 ft + (50,000 gallons divided by
11,343 gallons/ft) = 14.4 ft or 15 ft.
The revised minimum levels ensure at least 50,000 gallons of
water are provided above the top of the standpipe in each unit's CST
and are available for CS makeup to the RPV, as stated in the
applicable Bases. The TS Limiting Conditions for Operation (LCO)
remain unaffected by the proposed change.
2. The proposed TS change does not create the possibility of a
new or different kind of accident from any accident previously
evaluated. Revising Surveillance Requirement acceptance criteria
does not result in any physical modification to the plant or
operation of any existing equipment.
3. The proposed TS change does not involve a significant
reduction in a margin of safety, since this administrative change
only ensures the existing TS Bases are satisfied by increasing the
minimum CST water level requirement to ensure at least 50,000
gallons of water are available for CS injection to the RPV. The
proposed change does not involve a physical modification to any
structure, system or component, and does not modify the operation of
any existing equipment.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: Appling County Public Library,
301 City Hall Drive, Baxley, Georgia 31513
Attorney for licensee: Ernest L. Blake, Jr., Esquire, Shaw,
Pittman, Potts and Trowbridge, 2300 N Street, NW., Washington, DC 20037
NRC Project Director: Herbert N. Berkow
Houston Lighting & Power Company, City Public Service Board of San
Antonio, Central Power and Light Company, City of Austin, Texas,
Docket Nos. 50-498 and 50-499, South Texas Project, Units 1 and 2,
Matagorda County, Texas
Date of amendment request: July 8, 1996
Description of amendment request: The proposed amendment would
clarify that the component cooling water system surge tank level
instrumentation can be demonstrated operable by performing a channel
calibration test during any plant mode of operation.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. The proposed change does not involve a significant increase
in the probability or consequences of an accident previously
evaluated.
The proposed change to Technical Specification Surveillance
Requirement 4.7.3.b.3 will not effect any accident initiators or
precursors and will not alter the design assumptions for the systems
or components used to mitigate the consequences of an accident.
Calibration is performed on level instrumentation of Component
Cooling Water System trains that are out of service for scheduled
maintenance. Isolation redundancy is provided by instrumentation
associated with the trains that are in service during the
calibration. Since the surveillance will continue to be performed at
the specified interval, this proposed change will not increase the
probability of occurrence of an accident previously evaluated. The
surveillance does not differ from those previously performed;
therefore, there is no impact on the 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.
Clarifying the surveillance interval for surge tank level
instrumentation does not involve installation or operation of new or
different kinds of equipment. There is no change in the procedures
as described in the Technical Specifications. The change only
clarifies the interval at which the subject calibration will be
performed. Therefore, the proposed change does not create the
possibility of a new or different kind of accident from any accident
previously evaluated.
3. The proposed change does not involve a significant reduction
in a margin of safety.
The specified surveillance will remain as stated in the
Technical Specifications. Consequently, there is no reduction in the
effectiveness of the surveillance in ensuring equipment operability.
Calibration is performed on level instrumentation of Component
Cooling Water System trains that are out of service for scheduled
maintenance. Isolation redundancy is provided by instrumentation
associated with the trains that are in service during the
calibration. Consequently, clarifying the interval at which the
calibration is performed will have no significant impact on the
margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
request for amendments involves no significant hazards consideration.
Local Public Document Room location: Wharton County Junior College,
J. M. Hodges Learning Center, 911 Boling Highway, Wharton, TX 77488
Attorney for licensee: Jack R. Newman, Esq., Morgan, Lewis &
Bockius, 1800 M Street, N.W., Washington, DC 20036-5869
NRC Project Director: William D. Beckner
[[Page 44359]]
Houston Lighting & Power Company, City Public Service Board of San
Antonio, Central Power and Light Company, City of Austin, Texas,
Docket Nos. 50-498 and 50-499, South Texas Project, Units 1 and 2,
Matagorda County, Texas
Date of amendment request: August 8, 1996
Description of amendment request: The proposed amendment would
allow the transition from Mode 4 to Mode 3 with the turbine-driven
auxiliary feedwater pump inoperable and allow a 72-hour period after
the entry into Mode 3 to complete all necessary operability testing.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. The proposed change does not involve a significant increase
in the probability or consequences of an accident previously
evaluated.
The proposed change will allow entry into Mode 3 with an
inoperable Turbine Driven Auxiliary Feedwater pump. Since the
operability test on the Turbine Driven Auxiliary Feedwater pump can
only be performed once steam pressure is greater than or equal to
1000 psig, this change will allow the plant to reach the Mode where
steam pressure greater than or equal to 1000 psig is available to
perform the operability testing on the Turbine Driven Auxiliary
Feedwater pump. The allowance of 72 hours to complete the
surveillance testing will make the surveillance requirements
consistent with the allowed outage time already established in the
Action Statements. The proposed change does not affect the
probability of an accident. The Turbine Driven Auxiliary Feedwater
pump is not assumed to be an initiator of any analyzed event. The
consequences of an accident previously evaluated remain unchanged by
allowing the pump to be inoperable until suitable conditions exist
to perform the operability testing. The operability testing will
continue to demonstrate that the Turbine Driven Auxiliary Feedwater
pump will perform as required prior to entry into Mode 2. This
change will not alter assumptions relative to the mitigation of an
accident or transient event. Therefore, this change will 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 previously evaluated.
This change will not physically alter the plant (no new or
different type of equipment will be installed). The changes in
methods governing normal plant operation are consistent with current
safety analysis assumptions. The proposed change will allow entry
into Mode 3 with the Turbine Driven Auxiliary Feedwater pump
inoperable in order to perform the pump Operability Test on the
turbine driven AFW [Auxiliary Feedwater] pump once steam pressure is
greater than or equal to 1000 psig. Therefore, this change does not
create the possibility of a new or different kind of accident from
any accident previously evaluated.
3. The proposed change does not involve a significant reduction
in a margin of safety.
The proposed change will allow entry into Mode 3 with the
Turbine Driven AFW pump inoperable in order to perform the pump
Operability Test on the turbine driven AFW pump once steam pressure
is greater than or equal to 1000 psig. This will allow time for the
plant to obtain suitable test conditions with steam pressure greater
than or equal to 1000 psig. The margin of safety is not affected by
this change. The operability testing will continue to maintain
assurance that the AFW Pumps will perform as required prior to entry
into Mode 2. The safety analysis assumptions will still be
maintained, thus, no question of safety exists. Therefore, this
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 standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
request for amendments involves no significant hazards consideration.
Local Public Document Room location: Wharton County Junior College,
J. M. Hodges Learning Center, 911 Boling Highway, Wharton, TX 77488
Attorney for licensee: Jack R. Newman, Esq., Morgan, Lewis &
Bockius, 1800 M Street, N.W., Washington, DC 20036-5869
NRC Project Director: William D. Beckner
North Atlantic Energy Service Corporation, Docket No. 50-443,
Seabrook Station, Unit No. 1, Rockingham County, New Hampshire
Date of amendment request: June 4, 1996
Description of amendment request: The proposed amendment would
modify the Seabrook Station, Unit No. 1 Technical Specifications to
implement Option B to 10 CFR Part 50, Appendix J by referring to
Regulatory Guide 1.163, ``Performance-Based Containment Leak-Test
Program. The following Technical Specifications would be affected by
the proposed amendment:
1. Definitions: Definition 1.7, Containment Integrity (Item d.)
would be revised to reflect that leakage rates would be in accordance
with the Containment Leakage Rate Testing Program.
2. Limiting Conditions for Operation and Surveillance Requirements:
a. Containment Integrity: Surveillance Requirement 4.6.1.1.c would
be deleted because the specific guidance would be contained in the
Containment Leakage Rate Testing Program.
b. Containment Leakage: Limiting Condition for Operation 3.6.1.2.a
through 3.6.1.2.c and Surveillance Requirements 4.6.1.2.a through
4.6.1.2.h would be revised to replace specific guidance with a
reference to the Containment Leakage Rate Testing Program.
c. Containment Leakage: The Action for Limiting Condition for
Operation 3.6.1.2 would be revised to include the equivalent Action as
required for Limiting Condition for Operation 3.6.1.1 when the overall
integrated containment leak rate exceeds 1.0 La.
d. Containment Air Locks: Limiting Conditions for Operation
3.6.1.3.a and 3.6.1.3.b would be deleted and Surveillance Requirements
4.6.1.3.a and 4.6.1.3.b would be revised to replace specific guidance
with a reference to the Containment Leakage Rate Testing Program. The
footnote addressing the exemption to Appendix J regarding testing the
air locks prior to establishing containment integrity would be
maintained in the Containment Leakage Rate Testing Program.
e. Containment Vessel Structural Integrity: Surveillance
Requirement 4.6.1.6 would be revised to replace specific guidance with
a reference to the Containment Leakage Rate Testing Program.
f. Containment Ventilation System: Limiting Condition for Operation
3.6.1.7, Action b. would be revised to replace specific guidance with a
reference to the Containment Leakage Rate Testing Program. Surveillance
Requirement 4.6.1.7.1 would be revised to replace specific guidance
with a reference to the Containment Leakage Rate Testing Program.
g. Containment Enclosure Building: Limiting Condition for Operation
3.6.5.3 and Surveillance Requirement 4.6.5.3 would be revised to
include a reference to the requirements in the Containment Leakage Rate
Testing Program.
3. Bases: Sections 3/4.6.1.2, Containment Leakage; 3/4.6.1.7,
Containment Ventilation System; and 3/4.6.5.3, Containment Enclosure
Building Structural Integrity, would be revised to reflect the above
changes including a reference to the Containment Leakage Rate Testing
Program. In addition, a statement would be added to Section 3/4.6.1.2
to clarify the operability of containment regarding allowable leakage
rates.
4. Administrative Controls: Section 6.15 would be added to
establish a Containment Leakage Rate Testing Program, as specified in
Regulatory Guide 1.163, dated September 1995.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the
[[Page 44360]]
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 NRC staff's
review is presented below.
A. The changes do not involve a significant increase in the
probability or consequences of an accident previously evaluated (10
CFR 50.92(c)(1)) because the proposed changes merely revise the
testing criteria for containment penetrations. The revised criteria
will be based on the guidance in Regulatory Guide 1.163,
``Performance-Based Containment Leak-Test Program.''
This guidance allows for the use of relaxed testing frequencies
for containment penetrations that have performed satisfactorily on a
historical basis.
To support consideration of Option B to Appendix J, the NRC
staff reviewed the potential impact of performance-based testing
frequencies for containment penetrations. The NRC staff review is
documented in NUREG-1493 ``Performance-Based Containment Leak-Test
Program.'' One of the staff's conclusions was that reducing the
frequency of Type A tests (Integrated Leak Rate Tests) from three
per 10 years to one per 10 years leads to a marginal increase in
risk. For Type B and C testing (Local Leak Rate Tests), the change
in testing frequency will not have significant impact since, under
existing requirements, leakage contributes less than 0.1 percent of
the overall accident risk. The use of a performance-based testing
program will continue to provide assurance that the accident
analysis assumptions remain bounding.
B. The changes do not create the possibility of a new or
different kind of accident from any accident previously evaluated
(10 CFR 50.92(c)(2)) because they do not affect the manner by which
the facility is operated or involve changes to structures, systems,
or components that affect the operational characteristics of the
facility. The changes merely revise the testing criteria for the
containment penetrations, and establish a Containment Leakage Rate
Testing Program to ensure that the performance history of each
penetration is satisfactory prior to changing any test frequency.
Since there is no change to the facility or the way in which the
facility is operated, there is no possibility of creating a new or
different kind of accident than previously analyzed.
C. The changes do not involve a significant reduction in a
margin of safety (10 CFR 50.92(c)(3)). During the development of 10
CFR Part 50, Appendix J, Option B, the NRC staff determined the
reduction in safety associated with the implementation of the
performance-based testing program. The staff concluded that reducing
the frequency of Type A tests (Integrated Leak Rate Tests) from
three per 10 years to one per 10 years would have an imperceptible
impact upon risk. For Type B and C testing (Local Leak Rate Tests),
the change in testing frequency will not have significant impact
since this leakage contributes less than 0.1 percent of the overall
risk based on the existing regulations. The use of Option B will
have minimal impact on the radiological release rates since most
penetration leakage is well below the specified limits. The staff
noted that the accident risk is relatively insensitive to
containment leakage rate because accident risk is dominated by
accident sequences that result in failure of or bypass of the
containment. The use of a performance-based testing program will
continue to provide assurance that the accident analysis assumptions
remain bounding.
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: Exeter Public Library,
Founders Park, Exeter, NH 03833
Attorney for licensee: Lillian M. Cuoco, Esquire, Northeast
Utilities Service Company, Post Office Box 270, Hartford CT 06141-0270
NRC Project Director: Phillip F. McKee
Omaha Public Power District, Docket No. 50-285, Fort Calhoun
Station, Unit No. 1, Washington County, Nebraska
Date of amendment request: May 17, 1996
Description of amendment request: The proposed amendment would
revise the technical specifications (TS) to relocate the operability
requirements for shock suppressors (snubbers) from the TS to the
Updated Safety Analysis Report (USAR) and incorporate snubber
examination and testing requirements into TS 3.3.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. The proposed change does not involve a significant increase
in the probability or consequences of an accident previously
evaluated.
The proposed change will relocate operability requirements for
shock suppressors (snubbers) from the Technical Specifications (TS)
to the Updated Safety Analysis Report (USAR) and/or plant
procedures. On July 16, 1993, the NRC issued a Final Policy
Statement on Technical Specification Improvements for Nuclear Power
Reactors. The Final Policy Statement contains four criteria which
can be used to determine which constraints on the design and
operation of nuclear power plants are appropriate for inclusion in
TS. The NRC has incorporated these criteria into 10 CFR 50.36,
``Technical specifications.'' Snubbers do not meet any of the four
criteria for inclusion as a Limiting Condition for Operations within
the TS, and therefore it is proposed that these requirements be
relocated from the TS. The proposed change would not reduce or
revise any of the current requirements for snubber operability, only
relocate the requirements. Any changes to the requirements contained
in the USAR and/or plant procedures can be made without NRC approval
only when the changes meet the criteria of 10 CFR 50.59. Changes to
the snubber operability requirements that do not meet the criteria
of 10 CFR 50.59 must be approved by the NRC by license amendment.
Therefore, the relocation of the requirements on snubber operability
from the TS to the USAR does not increase the probability or
consequences of any accident previously analyzed.
The proposed change also deletes sections of the TS which are
redundant or in conflict with the American Society of Mechanical
Engineers (ASME) Boiler and Pressure Vessel Code. Snubbers are
required to be examined and tested in accordance with ASME Section
XI by 10 CFR 50.55a. The proposed change will ensure that the TS
implement ASME Section XI examination and testing requirements for
snubbers in accordance with 10 CFR 50.55a. Where differences between
the deleted sections of the TS and ASME Section XI requirements
exist, the Section XI requirements are similar or more conservative
than the TS. For example, although the functional test sample size
differs between the methodologies, both ensure that a very high
percentage of the snubbers in the plant are operable within
acceptance limits. Therefore, the proposed revision does not reduce
the effectiveness of snubber examination and testing.
The proposed change would not reduce the operability
requirements, acceptance criteria, or examination and testing of
snubbers. Therefore, the proposed change would not increase 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.
There will be no physical alterations to the plant
configuration, changes to setpoint values, or changes to the
implementation of setpoints or limits as a result of this proposed
change.
The proposed change deletes duplicate or conflicting
requirements between the TS and the ASME Section XI. In these areas,
the proposed deletions would remove the TS requirements and testing
would be conducted in accordance with ASME Section XI as directed by
10 CFR 50.55a. Although the requirements of ASME Section XI differ
from the TS in some cases, the differences do not decrease the
effectiveness of testing and examination as compared to the TS
requirements. Other areas, such as snubber operability requirements
and service life monitoring, which are presently addressed by TS,
but are not covered under ASME Section XI, will be maintained in the
USAR so that these requirements cannot be deleted without NRC
approval.
Therefore, the proposed change does not create the possibility
of a new or different kind of accident from any previously
evaluated.
3. The proposed change does not involve a significant reduction
in a margin of safety.
The proposed change does not reduce the operability,
examination, or testing requirements for snubbers. Snubbers will
still be required to meet the requirements of ASME Section XI and 10
CFR 50.55a except where specific written relief has been granted by
the NRC. Therefore, the proposed change does not involve a
significant reduction in a margin of safety.
[[Page 44361]]
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: W. Dale Clark Library, 215
South 15th Street, Omaha, Nebraska 68102
Attorney for licensee: Perry D. Robinson, Winston & Strawn, 1400 L
Street, N.W., Washington, DC 20005-3502
NRC Project Director: William H. Bateman
Omaha Public Power District, Docket No. 50-285, Fort Calhoun
Station, Unit No. 1, Washington County, Nebraska
Date of amendment request: May 20, 1996
Description of amendment request: The proposed amendment would
revise the technical specifications (TS) to clarify surveillance test
requirements of TS 3.1, Tables 3-1, 3-2, 3-3, 3-3A, and 3-5.
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 changes to the Table of Contents are administrative in
nature to reflect the removal of incore instrumentation
(Specification 2.10.3) from the TS by Amendment 167 and for
consistency. Amendment 169 inadvertently reinserted incore
instrumentation back into the Table of Contents.
The change to Specification 2.1.7(1)b is necessary because the
requirement to test the signal to alarm meter relay located in
Specification 3.1, Table 3-3, Item 6 is being deleted. The test,
which verifies the high and low pressurizer level alarm settings and
the pressurizer heater cutout function is unnecessary. Operating
experience has shown that a shiftly pressurizer level verification
as proposed for Specification 3.1, Table 3-3, Item 6.a is sufficient
to detect any level deviation and verify that operation is within
safety analyses assumptions. The level alarms serve as early warning
devices but do not provide an accident mitigation function.
Replacing the monthly test with a channel check is in accordance
with NUREG-1432, Combustion Engineering (CE), Standard Technical
Specifications (STS), Surveillance Requirement (SR) 3.3.11.1 (post
accident monitoring instrumentation). The monthly channel check
supplements the shiftly level verification.
The Basis of Specification 3.1 is revised to clarify
expectations regarding a channel check of channels that are normally
off scale when the surveillance is required. In this situation, the
channel check only verifies that they are off scale in the same
direction. Off scale low current loop channels are verified to be
reading at the bottom of the range and not failed downscale. These
statements are taken from the Bases of CE STS SR 3.3.4.1 Engineered
Safety Features Actuation System (ESFAS) Instrumentation (Analog).
In addition, the Basis of Specification 3.1 is revised to
clarify that power operated relief valve (PORV) actuation is not
required during the channel functional test of the PORV low
temperature setpoint (Table 3-3, Item 18.a). PORV actuation is not
required because it could depressurize the reactor coolant system.
This clarification is modeled after a similar statement from the
Bases of SR 3.4.12.6 (Low Temperature Overpressure Protection (LTOP)
System) of the CE STS.
Changing Specification 3.1, Tables 3-1, 3-2, 3-3, and 3-3A by
using defined terms to enable the Surveillance Method to match the
Surveillance Function is an administrative change designed to
simplify the tables. Removal of the extraneous text does not alter
the surveillance because the defined terms are equivalent in meaning
to the deleted text.
The reordering of several items in the tables into a Check-Test-
Calibrate sequence adds consistency to the tables. Text revisions in
the Channel Description or Surveillance Function columns of Tables
3-1 and 3-2 add clarity and/or consistency. Footnote No. 1 in Table
3-1 concerning the bistable trip tester was deleted because it is
unnecessary.
The Surveillance Function of Table 3-1, Item 1.c (Power Range
Safety Channels) is being changed to ``Test'' from ``Calibrate and
Test.'' It is not necessary for Item 1.c to require both because
Item 1.b already requires the power range safety channel adjustment
(calibration) to be performed daily. As stated in the Basis of
Specification 3.1, ``The minimum calibration frequencies of once-
per-day for the power range safety channels, ...are considered
adequate.'' To further clarify the issue, the Basis of Specification
3.1 is being revised to note that the daily calibration is a heat
balance adjustment only.
Changing Table 3-1, Item 4 (Thermal Margin/Low Pressure (TM/LP))
to use the defined term CHANNEL CALIBRATION will allow OPPD to relax
the current TM/LP calibration requirements with a negligible impact
on safety. Calibration of the temperature input and pressure input
will still require calibration to known standards (i.e., resistance
and pressure), but will allow the calibrations to be done separately
instead of coincidently. The channel functional test that follows
the channel calibration verifies proper function of the TM/LP
circuitry.
Removing the word ``Instruments'' from the Channel Description
of Table 3-2, Item 14 makes the Channel Description consistent with
the Surveillance Method. Table 3-2, Item 14 is not intended to
verify safety injection tank (SIT) instrumentation operability but
rather that the parameters level and pressure are within limits.
Generic Letter (GL) 93-05, Item 7.4, states that the operability of
SIT instrumentation is not directly related to the capability of a
SIT to perform its safety function. GL 93-05 concludes that the
surveillance should only confirm that the parameters defining SIT
operability are within their specified limits.
Items 22 & 24 are being added to Table 3-2 to clearly state the
requirement for testing manual actuation of the Engineered Safety
Features (ESF) channels for Off-site Power Low Signal (OPLS) and
Auxiliary Feedwater. Although testing manual actuation of these
channels is done via the existing Specifications, the requirement to
do so is not clearly stated. Reordering Table 3-2, Item 23 into a
Check-Test-Calibrate Surveillance Frequency sequence adds clarity
and consistency.
The addition of Footnote No. 7 to Table 3-2 clarifies that the
refueling frequency ESF channel functional test pertains to the
backup channels such as derived circuits and equipment that cannot
be tested when the plant is at power. Operating certain relays
during power operation could cause plant transients or equipment
damage.
The revisions to Table 3-3, Item 6, clarify that pressurizer
level is the parameter to be verified and not the pressurizer level
instruments. The revision to Item 6.a is consistent with CE STS SR
3.4.9.1 (pressurizer water level). Reordering Item 6 into a Check-
Test-Calibrate Surveillance Function sequence makes Item 6
consistent with the ordering of the other items in Table 3-3. The
requirement to test the signal to alarm meter relay currently
located in Specification 3.1, Table 3-3, Item 6.c is unnecessary.
Operating experience has shown that a shiftly pressurizer level
verification as proposed for Specification 3.1, Table 3-3, Item 6.a
is sufficient to detect any level deviation and verify that
operation is within safety analyses assumptions. Thus, the monthly
``Test'' requirement will be replaced with a ``Check'' to supplement
the less formal but more frequent shiftly level verification of Item
6.a.
Table 3-3, Items 21 (PORV Operation & Acoustic Position
Indication Channel) and 23 (Safety Valve Acoustic Position
Indication Channel) should be revised to a channel functional test
from a channel/circuit check. An oscillator and installed impactors
are used to generate noise signals and therefore, this surveillance
is more accurately described as a channel functional test rather
than a channel check.
Table 3-3, Items 21 and 22 (PORV Block Valve Operation &
Position Indication) should have the requirement to verify operation
on the emergency power supply deleted. Permanent Class 1E power
supplies the PORV and PORV Block Valve. Therefore, verification of
PORV or PORV Block Valve operability while powered from the
emergency power supply system provides no additional benefit.
(Operability of the emergency power supply system is tested in
accordance with Specification 3.7.) The proposed revision is in
accordance with the exception for plants with a permanent Class 1E
power supply to these valves as stated in CE STS, SR 3.4.11.4.
Deletion of the requirement of TS 3.2, Table 3-5, Item 15, to
test spent fuel pool surveillance coupons for a change in hardness
corrects an oversight in the Application for Amendment dated
December 7, 1992.
As stated in the Safety Evaluation Report enclosed with
Amendment 155, ``Each
[[Page 44362]]
coupon, upon its removal from the mounting jacket, will be analyzed
according to the following tests:
visual observation and photography
neutron attenuation
dimensional measurements (length, width, and thickness)
weight and specific gravity.''
The tests listed above are sufficient to detect degradation of
the Boral material and do not require that the surveillance coupons
be tested for hardness.
Based on the above discussion, the proposed changes clarify and
standardize existing surveillance requirements, remove redundant
requirements, correct minor oversights from previous amendment
requests or are in accordance with CE STS. Thus, none of the
requested changes involve a significant increase in the probability
or consequences of an accident previously evaluated.
2. The proposed change does not create the possibility of a new
or different kind of accident from any accident previously
evaluated.
The proposed revisions will not result in any physical
alterations to the plant configuration, changes to setpoint values,
or changes to the application of setpoints or limits. Therefore, the
proposed changes do not create the possibility of a new or different
kind of accident from any accident previously evaluated.
3. The proposed change does not involve a significant reduction
in a margin of safety.
The proposed changes clarify existing surveillance requirements,
remove redundant requirements, correct minor oversights from
previous amendment requests or are in accordance with CE STS. Thus,
none of the requested changes involves a significant reduction in a
margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: W. Dale Clark Library, 215
South 15th Street, Omaha, Nebraska 68102
Attorney for licensee: Perry D. Robinson, Winston & Strawn, 1400 L
Street, N.W., Washington, DC 20005-3502
NRC Project Director: William H. Bateman
Pennsylvania Power and Light Company, Docket No. 50-388 Susquehanna
Steam Electric Station, Unit 2, Luzerne County, Pennsylvania
Date of amendment request: May 20, 1996, as supplemented by letter
dated July 25, 1996
Description of amendment request: This amendment request would
modify the Technical Specifications for the unit by: changing the
Minimum Critical Power Ratio safety limit values, adding a reference to
reflect the use of the ANF-B Critical Power Correlation, and modifying
the associated Bases.
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 change to the ANFB correlation and corresponding MCPR Safety
Limits does not physically change the plant systems, structures, or
components. Thus, the probability of an event evaluated in the SAR
is not increased. The acceptance criterion for the MCPR Safety Limit
(i.e., 99.9% of the fuel rods expected to avoid boiling transition)
is not changed. Only the methodology used to demonstrate compliance
is changed.
Therefore, the consequences of anticipated operational
occurrences (which must show the Safety Limit is not violated) are
not changed. Results of incorporating this change will not
significantly increase 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.
As stated above, this methodology change does not impact the
acceptance criteria for the MCPR Safety Limits and does not
physically change the plant systems, structures, or components.
Since no changes to the physical plant are being made, this 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.
A cycle specific MCPR Safety Limit analysis was performed by SPC
[Siemens Power Corporation]. This analysis used NRC approved methods
described in the SPC report: ANF-524(P)(A), Revision 2 and
Supplement 1, Revision 2. The MCPR Safety Limit value is calculated
such that at least 99.9% of the fuel rods are expected to avoid
boiling transition during normal operation or anticipated operation
occurrences. Both the existing analysis using XN-3 and the new
analysis using ANFB utilize NRC approved methods to accomplish this
same objective. Therefore, the change to an ANFB based Safety Limit
does not involve a significant reduction in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: Osterhout Free Library,
Reference Department, 71 South Franklin Street, Wilkes-Barre, PA 18701
Attorney for licensee: Jay Silberg, Esquire, Shaw, Pittman, Potts
and Trowbridge, 2300 N Street NW., Washington, DC 20037
NRC Project Director: John F. Stolz
TU Electric Company, Docket Nos. 50-445 and 50-446, Comanche
Peak Steam Electric Station (CPSES), Units 1 and 2, Somervell
County, Texas
Date of amendment request: July 31, 1996
Brief description of amendments: Based on analyses of the core
configuration and expected operation for CPSES Unit 1, Cycle 6, the
proposed amendments would revise core safety limit curves and
Overtemperature N-16 reactor trip setpoints. In addition, the TU
Electric Small Break LOCA Topical Report on the Core Operating Limits
Report Technical Specification is incorporated. The topical report
change is applicable to both Units.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1.a. Revision to the Unit 1 Core Safety Limits
Analyses of reactor core safety limits are required as part of
reload calculations for each cycle. TU Electric has performed the
analyses of the Unit 1, Cycle 6 core configuration to determine the
reactor core safety limits. The methodologies and safety analysis
values result in new operating curves which, in general, permit
plant operation over a similar range of acceptable conditions. This
change means that if a transient were to occur with the plant
operating at the limits of the new curve, a different temperature
and power level might be attained than if the plant were operating
within the bounds of the old curves. However, since the new curves
were developed using NRC approved methodologies which are wholly
consistent with and do not represent a change in the Technical
Specification BASES for safety limits, all applicable postulated
transients will continue to be properly mitigated. As a result,
there will be no significant increase in the consequences, as
determined by accident analyses, of any accident previously
evaluated.
1.b. Revision to Unit 1 Overtemperature N-16 Reactor Trip
Setpoints, Parameters and Coefficients
As a result of changes discussed, the Overtemperature N-16
reactor trip setpoint has been recalculated. These trip setpoints
help ensure that the core safety limits are maintained and that all
applicable limits of the safety analysis are met.
Based on the calculations performed, the safety analysis value
for Overtemperature N-16 reactor trip setpoint has changed. This
essentially means if a transient were to occur, the actual
temperature and power level achievable prior to initiating a reactor
trip could be slightly higher. However, the analyses performed show
that, using the TU Electric methodologies, all applicable limits of
the safety analysis are met. This setpoint
[[Page 44363]]
provides a trip function which allows the mitigation of postulated
accidents and has no impact on accident initiation. Therefore, the
changes in safety analysis values do not involve an increase in the
probability of an accident and, based on satisfying all applicable
safety analysis limits, there is no significant increase in the
consequences of any accident previously evaluated.
In addition, sufficient operating margin has been maintained in
the overtemperature setpoint such that the risk of turbine runbacks
or reactor trips due to upper plenum flow anomalies or other
operational transients will be minimized, thus reducing potential
challenges to the plant safety systems.
1.c. Incorporation of TU Electric Small Break LOCA Topical
Report, RXE-95-0001-P.
TU Electric has submitted the topical report ``Small Break Loss
of Coolant Accident Analysis Methodology,'' RXE-95-001-P and plans
to use the report to support Unit 1 Cycle 6. In order to accomplish
this activity, it is necessary to include the topical report in the
list of NRC-approved methodologies in Technical Specification
6.9.1.6b. Use of this topical report is contingent upon NRC
approval; therefore, inclusion of this report in Section 6 of the
Technical Specifications is administrative in nature and does not
change the probability or consequences of an accident.
2. The proposed changes involve the use of revised safety
analysis values and the calculation of new reactor core safety
limits and reactor trip setpoints. As such, the changes play an
important role in the analysis of postulated accidents but none of
the changes effect plant hardware or the operation of plant systems
in a way that could initiate an accident. Therefore, the proposed
changes do not create the possibility of a new or different kind of
accident from any accident previously evaluated.
3. In reviewing and approving the methods used for safety
analyses and calculations, the NRC has approved the safety analysis
limits which establish the margin of safety to be maintained. While
the actual impact on safety is discussed in response to question 1,
the impact on margin of safety is discussed below:
3.a.
Revision to the Unit 1 Reactor Core Safety Limits
The TU Electric reload analysis methods have been used to
determine new reactor core safety limits. All applicable safety
analysis limits have been met. The methods used are wholly
consistent with Technical Specification BASES 2.1 which is the bases
for the safety limits. In particular, the curves assure that for
Unit 1, Cycle 6, the calculated DNBR is no less than the safety
analysis limit and the average enthalpy at the vessel exit is less
than the enthalpy of saturated liquid. The acceptance criteria
remains valid and continues to be satisfied; therefore, no change in
a margin of safety occurs.
3.b. Revision to Unit 1 Overtemperature N-16 Reactor Trip
Setpoints, Parameters and Coefficients
Because the reactor core safety limits for CPSES Unit 1, Cycle 6
are recalculated, the Reactor Trip System instrumentation setpoint
values for the Overtemperature N-16 reactor trip setpoint which
protect the reactor core safety limits must also be recalculated.
The Overtemperature N-16 reactor trip setpoint helps prevent the
core and Reactor Coolant System from exceeding their safety limits
during normal operation and design basis anticipated operational
occurrences. The most relevant design basis analysis in Chapter 15
of the CPSES Final Safety Analysis Report (FSAR) which is affected
by the change in the safety analysis value for the CPSES Unit 1
Overtemperature N-16 reactor trip setpoint is the Uncontrolled Rod
Cluster Control Assembly Bank Withdrawal at Power (FSAR Section
15.4.2). This event has been re-analyzed with the revised safety
analysis value for the Overtemperature N-16 reactor trip setpoint to
demonstrate compliance with event specific acceptance criteria.
Because all event acceptance criteria are satisfied, there is no
degradation in a margin of safety.
The nominal Reactor Trip System instrumentation setpoints values
for the Overtemperature N-16 reactor trip setpoint (Technical
Specification Table 2.2-1) are determined based on a statistical
combination of all of the uncertainties in the channels to arrive at
a total uncertainty. The total uncertainty plus additional margin is
applied in a conservative direction to the safety analysis trip
setpoint value to arrive at the nominal and allowable values
presented in Technical Specification Table 2.2-1. Meeting the
requirements of Technical Specification Table 2.2-1 assures that the
Overtemperature N-16 reactor trip setpoint assumed in the safety
analyses remains valid. The CPSES Unit 1, Cycle 6 Overtemperature N-
16 reactor trip setpoint is different from previous cycles which
provides more operational flexibility to withstand mild transients
without initiating automatic protective actions. Although the
setpoint is different, the Reactor Trip System instrumentation
setpoint values for the Overtemperature N-16 reactor trip setpoint
are consistent with the safety analysis assumption which has been
analytically demonstrated to be adequate to meet the applicable
event acceptance criteria. Thus, there is no reduction in a margin
of safety.
3.c. Revise 6.9.1.6b to include Topical Report RXE-95-001-P,
``Small Break Loss of Coolant Accident Methodology''
TU Electric has submitted the topical report ``Small Break Loss
of Coolant Accident Analysis Methodology,'' RXE-95-001-P and plans
to use the report to support Unit 1 Cycle 6. In order to accomplish
this activity, it is necessary to include the topical report in the
list of NRC-approved methodologies in Technical Specification
6.9.1.6b. Use of this topical report is contingent upon NRC
approval; therefore, inclusion of this report in Section 6 of the
Technical Specifications is administrative in nature and does not
reduce the margin of safety.
Using the NRC approved TU Electric methods, the reactor core
safety limits are determined such that all applicable limits of the
safety analyses are met. Because the applicable event acceptance
criteria continue to be met, there is no significant reduction in
the margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: 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
TU Electric Company, Docket Nos. 50-445 and 50-446, Comanche Peak
Steam Electric Station (CPSES), Units 1 and 2, Somervell County,
Texas
Date of amendment request: July 31, 1996
Brief description of amendments: The proposed amendments would
revise the Technical Specifications by (1) changing the battery charger
ratings; (2) by clarifying the meaning of the term ``associated
inverter''; and by (3) deleting the protection channel and the vital
bus ratings for the instrument busses identified for Mode 1 through 4.
These changes are associated with a plant modification in which the
inverters and battery chargers are being replaced and an installed
spare inverter is being added for each safety train. These changes are
equally applicable to CPSES Units 1 and 2.
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?
CHANGE TO IDENTIFY BATTERY CHARGER RATINGS
The first proposed change replaces the test amperes with the
design value for the replacement battery charger and allows a
voltage range (greater than or equal to 130 volts) instead of a
single value. The intent of the surveillance requirement or the
surveillance frequency is not changed. The replacement inverters and
battery chargers will continue to provide the capacity needed to
perform the required safety functions. The revised surveillance will
continue to assure that the battery chargers are capable of
performing as designed. Therefore this change does not impact the
probability or the consequences of an accident previously evaluated.
CLARIFICATION TO DEFINE ASSOCIATED INVERTER
The second proposed change adds a foot note to clarify the term
``associated inverter'' by describing it as, ''... the dedicated
inverter or installed spare inverter.'' Also the Bases
[[Page 44364]]
for this specification is revised to reflect the basis for this
change. This change allows use of an installed spare inverter (for
each train) having the capability to energize the Instrument Bus for
the protection channel or the vital bus. Procedural controls and
interlocks ensure that the spare is available to feed only one of
the protection channel or vital bus Instrument Bus at a time, in the
event the dedicated inverter is not available. Procedural controls
and interlocks also ensure that the installed spare inverter is fed
from the same power source as that of the dedicated inverter not in
service and whose loads are being fed by the spare inverter. This
proposed design only allows the spare inverter for a safety train to
be manually aligned to replace only one of the four inverters in
that train at a time.
The installation of a spare inverter for each train and the
associated design configuration increases the availability of
energized Instrument Bus for the protection channel and vital bus.
These changes do not involve an increase in the probability or
consequences of an accident previously evaluated.
DELETION OF THE PROTECTION CHANNEL AND VITAL BUS RATINGS FOR
INSTRUMENT BUS
The third proposed change deletes specifying of the protection
channel and vital bus KVA ratings for the Instrument Bus. The
ratings of inverter that feeds these instrument buses are being
described in other Licensing Bases Documents or Design Basis
Documents. There is no change proposed to the intent of the action
statements.
This is considered an administrative change and does not impact
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?
CHANGE TO IDENTIFY BATTERY CHARGER RATINGS
Replacing the inverters and battery chargers and changing the
parameters of the battery charger surveillance test to match the
replacement chargers does not alter the functional modes of this
portion of the design and does not result in any new failure modes.
As such, it does not create the possibility of a new or different
accident from any previously evaluated.
CLARIFICATION TO DEFINE ASSOCIATED INVERTER
The second proposed change allows use of an installed spare
inverter for each train to energize the one of the Instrument Bus
for the protection channel and vital bus at a time for the
respective safety train while its dedicated inverter is not
available. The spare inverter is such that it has the capability to
support the maximum load for the protection channel or vital bus.
Manually aligning the installed inverter to replace on[e] of the
dedicated inverters is essentially equivalent to a repair activity
which replaces a faulted inverter with a new inverter. In addition,
procedural controls and interlocks are provided to ensure the proper
alignment of the installed spare when it is used. The proposed
changes do not create the possibility of a new or different accident
from any previously evaluated.
DELETION OF THE PROTECTION CHANNEL AND VITAL BUS RATINGS FOR
INSTRUMENT BUS
The third proposed change as discussed earlier does not change
intent of the Technical Specifications action statements. This is an
administrative change which does not introduce new failure modes and
has no new or different accidents from any previously evaluated are
created.
3. DO THE PROPOSED CHANGES INVOLVE A SIGNIFICANT REDUCTION IN
MARGIN OF SAFETY?
The relevant Technical Specification sections proposed for
changes: (1) ensure that the battery charger is capable of charging
the battery by performing the surveillance at 18 month frequency;
(2) establish operability requirements of the Instrument Bus for the
protection channel and vital bus in MODES 1 through 6; and (3)
identify the actions required for not meeting item 2.
These proposed changes do not alter the intent of the above
requirements; however replacement of the currently installed
inverters with inverters which are expected to be more reliable and
available and the addition of a spare inverter per safety train to
energize Instrument Bus for protection channel and vital bus does
increase the reliability of the instrument busses for the train.
Allowing credit for this spare inverter in meeting the operability
requirements of Instrument Bus for the protection channel and vital
bus, minimize potential plant shutdowns due to non-energized
instrument from its dedicated inverter. These changes do not involve
a significant reduction in margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 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
Vermont Yankee Nuclear Power Corporation, Docket No. 50-271,
Vermont Yankee Nuclear Power Station, Vernon, Vermont
Date of amendment request: August 9, 1996
Description of amendment request: The proposed amendment would
revise the Safety Limits for Minimum Critical Power Ratio (MCPR) based
upon a Vermont Yankee plant and cycle specific analysis, performed by
General Electric. The revised MCPR Safety Limits are needed to
accommodate Vermont Yankee's core design for upcoming refueling cycle
number 19. Specifically, the MCPR Safety Limits of 1.07 and 1.08 in the
Vermont Yankee Technical Specifications (TS) section 1.1.A are proposed
to be increased to 1.10 and 1.12 for two loop and single loop
operation, respectively.
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 Safety Limit Minimum Critical Power Ratio (MCPR) is
defined to ensure that during normal operation and Anticipated
Operational Transients (AOTs), at least 99.9% of the fuel rods in
the core do not experience transition boiling. Core MCPR operating
limits are developed to ensure these Safety Limits are maintained in
the event of the worst case transient. Since the Safety Limit MCPR
will be maintained at all times, operation under the proposed
changes will ensure at least 99.9% of the fuel rods in the core do
not experience transition boiling and no significant radiological
release will result. Therefore, this Safety Limit MCPR change does
not affect the probability or consequences of a previously evaluated
accident.
(2) The proposed changes do not involve any new modes of
operation or any plant modifications. Establishment and monitoring
of the operating limits will continue as per established procedure.
The proposed changes to these limits do not result in the creation
of any new precursors to an accident. Therefore, the proposed change
does not create the possibility of a new or a different kind of
accident from any previously analyzed.
(3) The Safety Limit MCPR values were evaluated by General
Electric based upon a cycle specific Vermont Yankee analysis, using
NRC approved methods. The resulting limits are more conservative
than the previous generic limits and will continue to assure that at
least 99.9% of the fuel rods in the core do not experience
transition boiling during analyzed transients. This acceptance
criteria ensures the safety design limit of ``no damage to a nuclear
system process barrier shall result from forces associated with
AOTs.'' Therefore, the implementation of the proposed change does
not involve a significant reduction in [a] margin of safety.
The NRC staff has reviewed the licensee's analysis. The staff notes
that, although the proposed change does not involve a plant
modification, the reason for the proposed higher safety limit MCPRs is
the cycle-specific core design and the local power distribution in the
slightly higher enriched fresh GE-9B fuel bundles. This new fuel will
be loaded during the September/October 1996 refueling outage. In
conjunction with the proposed safety limit MCPRs and the core operating
limits determined in accordance with Vermont Yankee TS 6.7.A.4, the new
fuel load will not involve a significant increase in the probability or
consequences of an
[[Page 44365]]
accident previously evaluated nor a significant reduction in a margin
of safety. In addition, the new fuel load does not create the
possibility of a new or different kind of accident from any accident
previously evaluated. 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: Brooks Memorial Library, 224
Main Street, Brattleboro, VT 05301
Attorney for licensee: R. K. Gad, III, Ropes and Gray, One
International Place, Boston, MA 02110-2624
NRC Project Director: Jocelyn A. Mitchell, Acting Director
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.
Boston Edison Company, Docket No. 50-293, Pilgrim Nuclear Power
Station, Plymouth County, Massachusetts
Date of application for amendment: May 1, 1996
Brief description of amendment: The proposed amendment will modify
the definition of ``Core Alteration,'' and the limiting condition for
operation, Surveillance conditions and Bases section associated with
Technical Specification 3.7.C, ``Secondary Containment.''
Date of issuance: August 12, 1996
Effective date: August 12, 1996
Amendment No.: 166
Facility Operating License No. DPR-35: Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: June 5, 1996 (61 FR
28606) The Commission's related evaluation of the amendment is
contained in a Safety Evaluation dated August 12, 1996. No significant
hazards consideration comments received: No
Local Public Document Room location: Plymouth Public Library, 11
North Street, Plymouth, Massachusetts 02360.
Carolina Power & Light Company, Docket No. 50-261, H. B. Robinson
Steam Electric Plant, Unit No. 2, Darlington County, South Carolina
Date of application for amendment: July 17, 1995, as supplemented
May 2, 1996, and July 1, 1996.
Brief description of amendment: The change revises technical
specification (TS) section 3.8 to specify that the spent fuel building
refueling filter fan and at least one containment purge fan shall be
shown to operate within plus or minus 10 percent of the design flow.
Date of issuance: August 6, 1996
Effective date: August 6, 1996
Amendment No. 172
Facility Operating License No. DPR-23. Amendment revises the
Technical Specifications.
Date of initial notice in Federal Register: September 13, 1995 (60
FR 47615). The May 2, and July 1, 1996, letters provided clarifying
information that did not affect the proposed no significant hazards
consideration. The Commission's related evaluation of the amendment is
contained in a Safety Evaluation dated August 6, 1996. No significant
hazards consideration comments received: No
Local Public Document Room location: Hartsville Memorial Library,
147 West College Avenue, Hartsville, South Carolina 29550
Carolina Power & Light Company, Docket No. 50-261, H. B. Robinson
Steam Electric Plant, Unit No. 2, Darlington County, South Carolina
Date of application for amendment: June 6, 1996
Brief description of amendment: The amendment revises technical
specifications (TS) Section 4.2.3 to allow the licensee to defer the
ultrasonic inspection of the reactor coolant pump flywheel for one
operating cycle.
Date of issuance: August 9, 1996
Effective date: August 9, 1996
Amendment No. 173
Facility Operating License No. DPR-23. Amendment revises the
Technical Specifications.
Date of initial notice in Federal Register: July 3, 1996 (61 FR
34888) The Commission's related evaluation of the amendment is
contained in a Safety Evaluation dated August 9, 1996. No significant
hazards consideration comments received: No
Local Public Document Room location: Hartsville Memorial Library,
147 West College Avenue, Hartsville, South Carolina 29550
Carolina Power & Light Company, et al., Docket No. 50-400, Shearon
Harris Nuclear Power Plant, Unit 1, Wake and Chatham Counties,
North Carolina
Date of application for amendment: May 31, 1996
Brief description of amendment: The amendment revises Technical
Specifications (TS) Table 3.3-7, Seismic Monitoring Instrumentation,
and TS Table 4.3-4, Seismic Monitoring Instrumentation Surveillance
Requirements, to correct the location described for one of the three
Triaxial Peak Accelerograph recorders.
Date of issuance: August 7, 1996
Effective date: August 7, 1996
Amendment No. 66
Facility Operating License No. NPF-63. Amendment revises the
Technical Specifications
Date of initial notice in Federal Register: July 3, 1996 (61 FR
34888) The Commission's related evaluation of the amendment is
contained in a Safety Evaluation dated August 7, 1996. No significant
hazards consideration comments received: No
Local Public Document Room location: Cameron Village Regional
Library, 1930 Clark Avenue, Raleigh, North Carolina 27605.
Commonwealth Edison Company, Docket Nos. 50-373 and 50-374, LaSalle
County Station, Units 1 and 2, LaSalle County, Illinois
Date of application for amendments: April 16, 1996
Brief description of amendments: The amendments revise the
Technical
[[Page 44366]]
Specifications (TSs) to eliminate selected response time testing
requirements based on analyses performed by the Boiling Water Reactor
Owners' Group as documented in NEDO-32291. The affected TS sections are
3/4.3.1, ``Reactor Protection System Instrumentation;'' 3/4.3.2,
``Isolation Actuation Instrumentation;'' and 3/4.3.3, ``Emergency Core
Cooling System Actuation Instrumentation.''
Date of issuance: August 14, 1996
Effective date: Immediately, to be implemented within 60 days.
Amendment Nos.: 114 and 99
Facility Operating License Nos. NPF-11 and NPF-18: The amendments
revised the Technical Specifications.
Date of initial notice in Federal Register: May 22, 1996 (61 FR
25702) The Commission's related evaluation of the amendments is
contained in a Safety Evaluation dated August 14, 1996. No significant
hazards consideration comments received: No
Local Public Document Room location: Jacobs Memorial Library,
Illinois Valley Community College, Oglesby, Illinois 61348.
Detroit Edison Company, Docket No. 50-341, Fermi-2, Monroe County,
Michigan
Date of application for amendment: December 21, 1995
Brief description of amendment: The amendment revises the Technical
Specifications (TS) to implement 10 CFR Part 50, Appendix J - Option B,
by referring to Regulatory Guide 1.163, ``Performance-Based Containment
Leak-Test Program.'' Specifically, changes have been made to TS Section
3/4.6.1.2, ``Primary Containment Leakage,'' TS 3/4.6.1.3, ``Primary
Containment Air Locks,'' TS 3/4.6.1.5, ``Primary Containment Structural
Integrity,'' TS 6.0, ``Administrative Controls,'' and their associated
Bases.
Date of issuance: August 8, 1996
Effective date: August 8, 1996, with full implementation within 45
days.
Amendment No.: 108
Facility Operating License No. NPF-43. Amendment revises the
Technical Specifications.
Date of initial notice in Federal Register: February 28, 1996 (61
FR 7551) The Commission's related evaluation of the amendment is
contained in a Safety Evaluation dated August 8, 1996. No significant
hazards consideration comments received: No.
Local Public Document Room location: Monroe County Library System,
3700 South Custer Road, Monroe, Michigan 48161
Florida Power and Light Company, Docket Nos. 50-250 and 50-251,
Turkey Point Plant Units 3 and 4, Dade County, Florida
Date of application for amendments: April 19, 1996, and supplements
dated May 10 and May 28, 1996.
Brief description of amendments: The amendment changes the
Technical Specifications to address frequency extension on a periodic
basis, deletes separate notification requirements for an inoperable
startup transformer, and allows the operating residual heat removal
loop to be removed from operation, under certain conditions, during
refueling.
Date of Issuance: August 6, 1996
Effective Date: August 6, 1996
Amendment Nos.: 189 and 183Facility Operating Licenses Nos. DPR-31
and DPR-41: Amendments revised the Technical Specifications.
Date of initial notice in Federal Register: July 3, 1996 (61 FR
34892) The Commission's related evaluation of the amendments is
contained in a Safety Evaluation dated August 6, 1996. No significant
hazards consideration comments received: No.
Local Public Document Room location: Florida International
University, University Park, Miami, Florida 33199.
Florida Power and Light Company, Docket Nos. 50-250 and 50-251,
Turkey Point Plant Units 3 and 4, Dade County, Florida
Date of application for amendments: July 26, 1995, and supplemented
March 13, May 3, and May 9, 1996.
Brief description of amendments: Change TS 6.9.1.7, Core Operating
Limits Report, resulting from a reanalysis of the small break loss-of-
coolant accident for the Turkey Point Units using the NOTRUMP code
including the COSI safety injection (SI) condensation model.
Date of issuance: August 13, 1996
Effective date: August 13, 1996
Amendment Nos. 190 and 184Facility Operating Licenses Nos. DPR-31
and DPR-41: Amendments revised the Technical Specifications.
Date of initial notice in Federal Register: September 13, 1995 (60
FR 47618). The supplements dated March 13, May 3, and May 9, 1996
provided clarifying information that did not change the initial
proposed no significant hazards consideration determination. The
Commission's related evaluation of the amendments is contained in a
Safety Evaluation dated August 13, 1996. No significant hazards
consideration comments received: No
Local Public Document Room location: Florida International
University, University Park, Miami, Florida 33199.
Houston Lighting & Power Company, City Public Service Board of San
Antonio, Central Power and Light Company, City of Austin, Texas,
Docket Nos. 50-498 and 50-499, South Texas Project, Units 1 and 2,
Matagorda County, Texas
Date of amendment request: May 1, 1996
Brief description of amendments: The amendments changed the
technical specifications to implement 10 CFR Part 50, Appendix J,
Option B, by referring to Regulatory Guide 1.163, ``Performance-Based
Containment Leak-Test Program.'' Part of the requested change, that
regarding the frequency of leakage rate testing the normal containment
purge valves and the supplementary containment purge valves, was
denied.
Date of issuance: August 13, 1996
Effective date: August 13, 1996
Amendment Nos.: 84 and 71
Facility Operating License Nos. NPF-76 and NPF-80. The amendments
revised the Technical Specifications.
Date of initial notice in Federal Register: June 5, 1996 (61 FR
28616) The Commission's related evaluation of the amendments is
contained in a Safety Evaluation dated August 13, 1996. No significant
hazards consideration comments received: No
Local Public Document Room location: Wharton County Junior College,
J. M. Hodges Learning Center, 911 Boling Highway, Wharton, TX 77488
IES Utilities Inc., Docket No. 50-331, Duane Arnold Energy, Center,
Linn County, Iowa
Date of application for amendment: November 30, 1995
Brief description of amendment: The amendment implements the Option
I-D long-term stability solution and removes the existing SIL-380 Rev.
1-based specifications. In addition, the amendment requires a plant
scram be initiated should the plant enter natural circulation
conditions and prohibits restarting a recirculation pump while in
natural circulation. Finally, this amendment deletes Technical
Specification (TS) actions and surveillance requirements related to
core plate differential pressure noise while in single recirculation
pump operation (SLO).
Date of issuance: August 7, 1996
Effective date: August 7, 1996
Amendment No.: 215
Facility Operating License No. DPR-49: Amendment revised the
Technical Specifications.
[[Page 44367]]
Date of initial notice in Federal Register: March 13, 1996 (61 FR
10394) The Commission's related evaluation of the amendment is
contained in a Safety Evaluation dated August 7, 1996. No significant
hazards consideration comments received: No.
Local Public Document Room location: Cedar Rapids Public Library,
500 First Street, S. E., Cedar Rapids, Iowa 52401
IES Utilities Inc., Docket No. 50-331, Duane Arnold Energy, Center,
Linn County, Iowa
Date of application for amendment: November 15, 1995, as
supplemented April 9, 1996
Brief description of amendment: The amendment revises the
requirements for the End of Cycle Recirculation Pump Trip logic to
match more closely the assumptions applicable to the turbine trip
events for which it was installed. The surveillance requirements are
also revised, based on those same assumptions.
Date of issuance: August 8, 1996
Effective date: August 8, 1996
Amendment No.: 216
Facility Operating License No. DPR-49: Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: January 22, 1996 (61 FR
1629) The April 9, 1996, submittal was clarifying in nature and did not
affect the no significant hazards determination. The Commission's
related evaluation of the amendment is contained in a Safety Evaluation
dated August 8, 1996. No significant hazards consideration comments
received: No.
Local Public Document Room location: Cedar Rapids Public Library,
500 First Street, S. E., Cedar Rapids, Iowa 52401
IES Utilities Inc., Docket No. 50-331, Duane Arnold Energy, Center,
Linn County, Iowa
Date of application for amendment: January 18, 1996
Brief description of amendment: The amendment revises the setpoint
at which the Reactor Water Cleanup (RWCU) system isolates, based on
reactor vessel water level. In particular, the amendment changes the
Group 5 isolation from isolating on ``reactor water level low'' to
``reactor water level low-low.''
Date of issuance: August 8, 1996
Effective date: August 8, 1996, and shall be implemented prior to
startup from RFO 14.
Amendment No.: 217
Facility Operating License No. DPR-49: Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: February 14, 1996 (61
FR 5814) The Commission's related evaluation of the amendment is
contained in a Safety Evaluation dated August 8, 1996. No significant
hazards consideration comments received: No.
Local Public Document Room location: Cedar Rapids Public Library,
500 First Street, S. E., Cedar Rapids, Iowa 52401
Indiana Michigan Power Company, Docket Nos. 50-315 and 50-316,
Donald C. Cook Nuclear Plant, Unit Nos. 1 and 2, Berrien County,
Michigan
Date of application for amendments: January 12, 1996 (AEP:NRC:1233)
Brief description of amendments: The amendments modify the
Technical Specifications to delete the surveillance requirement
demonstrating operability of the emergency power supply for the
pressurizer power operated relief valves and block valves.
Date of issuance: August 15, 1996
Effective date: August 15, 1996, with full implementation within 45
days
Amendment Nos.: 211 and 196
Facility Operating License Nos. DPR-58 and DPR-74. Amendments
revised the Technical Specifications.
Date of initial notice in Federal Register: February 28, 1996 (61
FR 7554) The Commission's related evaluation of the amendments is
contained in a Safety Evaluation dated August 15, 1996. No significant
hazards consideration comments received: No.
Local Public Document Room location: Maud Preston Palenske
Memorial Library, 500 Market Street, St. Joseph, Michigan 49085
Niagara Mohawk Power Corporation, Docket No. 50-410, Nine Mile
Point Nuclear Station, Unit 2, Oswego County, New York
Date of application for amendment: February 7, 1996, as
supplemented July 26, 1996.
Brief description of amendment: The amendment revises the operating
license, TSs and associated Bases to implement Option B ``Performance-
Based Requirements'' of Appendix J to 10 CFR Part 50 for Type A, B, and
C leakage rate testing.
Date of issuance: August 13, 1996
Effective date: As of the date of issuance to be implemented within
30 days.
Amendment No.: 74
Facility Operating License No. NPF-69: Amendment revises the
Technical Specifications and operating license.
Date of initial notice in Federal Register: May 8, 1996 (61 FR
20849) The Commission's related evaluation of the amendment is
contained in a Safety Evaluation dated August 13, 1996. No significant
hazards consideration comments received: No
Local Public Document Room location: Reference and Documents
Department, Penfield Library, State University of New York, Oswego, New
York 13126.
Northeast Nuclear Energy Company, et al., Docket No. 50-336,
Millstone Nuclear Power Station, Unit No. 2, New London County,
Connecticut
Date of application for amendment: July 3, 1996
Brief description of amendment: The amendment removes, on a one-
time basis during the cycle 13 mid-cycle offload/reload activities, the
Technical Specification (TS) requirement that the boron concentration
in all filled portions of the reactor coolant system be ``uniform.''
The requested change also adds a footnote indicating that it is
acceptable for the boron concentration of the water volumes in the
steam generators and the connecting piping to be as low as 1300 parts
per million. The TS Bases are also updated to reflect the one-time TS
change.
Date of issuance: August 12, 1996
Effective date: As of the date of issuance, to be implemented
within 30 days.
Amendment No.: 201
Facility Operating License No. DPR-65. Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: July 11, 1996 (61 FR
36583) The Commission's related evaluation of the amendment is
contained in a Safety Evaluation dated August 12, 1996. No significant
hazards consideration comments received: No.
Local Public Document Room location: Learning Resources Center,
Three Rivers Community-Technical College, 574 New London Turnpike,
Norwich, CT 06360, and Waterford Library, ATTN: Vince Juliano, 49 Rope
Ferry Road, Waterford, CT 06385
Pennsylvania Power and Light Company, Docket Nos. 50-387 and 50-388
Susquehanna Steam Electric Station, Units 1 and 2, Luzerne County,
Pennsylvania
Date of application for amendments: February 29, 1996
Brief description of amendments: These amendments relocate
Specification 3/4.9.6, ``Refueling Platform,'' to the Susquehanna Steam
Electric Station Technical Requirements Manual, a document which is
controlled under the requirements of 10 CFR 50.59.
Date of issuance: August 13, 1996
Effective date: August 13, 1996
Amendment Nos.: 159 and 130
[[Page 44368]]
Facility Operating License Nos. NPF-14 and NPF-22. The amendments
revised the Technical Specifications.
Date of initial notice in Federal Register: April 10, 1996 (61 FR
15992) The Commission's related evaluation of the amendments is
contained in a Safety Evaluation dated August 13, 1996. No significant
hazards consideration comments received: No
Local Public Document Room location: Osterhout Free Library,
Reference Department, 71 South Franklin Street, Wilkes-Barre,
Pennsylvania 18701.
Tennessee Valley Authority, Docket Nos. 50-259, 50-260, and 50-296,
Browns Ferry Nuclear Plant, Units 1, 2, and 3, Limestone County,
Alabama
Date of application for amendments: May 20, 1996 (TS 373)
Brief description of amendment: The amendments incorpore the
guidance of Generic Letter 87-09 in the technical specifications,
allowing a 24-hour delay in implementing action requirements due to a
missed surveillance requirement.
Date of issuance: August 5, 1996
Effective Date: August 5, 1996
Amendment Nos.: 230, 245 and 205
Facility Operating License Nos. DPR-33, DPR-52 and DPR-68:
Amendments revised the Technical Specifications.
Date of initial notice in Federal Register: June 19, 1996 (61 FR
31185) The Commission's related evaluation of the amendment is
contained in a Safety Evaluation dated August 5, 1996. No significant
hazards consideration comments received: None
Local Public Document Room location: Athens Public library, South
Street, Athens, Alabama 35611
Union Electric Company, Docket No. 50-483, Callaway Plant, Unit 1,
Callaway County, Missouri
Date of application for amendment: May 29, 1996
Brief description of amendment: The amendment authorizes revision
of the Final Safety Analysis Report (FSAR) to incorporate a
modification to the facility that will reduce the single failure trip
potential for the main feedwater and bypass valves.
Date of issuance: August 13, 1996
Effective date: August 13, 1996
Amendment No.: 115
Facility Operating License No. NPF-30: The amendment revised the
Final Safety Analysis Report.
Date of initial notice in Federal Register: July 3, 1996 (61 FR
34900) The Commission's related evaluation of the amendment is
contained in a Safety Evaluation dated August 13, 1996. No significant
hazards consideration comments received: No.
Local Public Document Room location: Callaway County Public
Library, 710 Court Street, Fulton, Missouri 65251.
Wisconsin Public Service Corporation, Docket No. 50-305, Kewaunee
Nuclear Power Plant, Kewaunee County, Wisconsin
Date of application for amendment: June 4, 1996
Brief description of amendment: The amendment revises the Technical
Specifications by reducing the surveillance test frequencies for the
radiation monitoring system (Table TS 4.1-1) and the control rods
(Table TS 4.1-3) in accordance with the guidance of Generic Letter 93-
05, ``Line-Item Technical Specifications Improvements to Reduce
Surveillance Requirements for Testing During Power Operation,'' dated
September 27, 1993.
Date of issuance: August 7, 1996
Effective date: August 7, 1996
Amendment No.: 125
Facility Operating License No. DPR-43: Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: July 3, 1996 (61 FR
34901) The Commission's related evaluation of the amendment is
contained in a Safety Evaluation dated August 7, 1996. No significant
hazards consideration comments received: No.
Local Public Document Room location: University of Wisconsin,
Cofrin Library, 2420 Nicolet Drive, Green Bay, Wisconsin 54311-7001
Wolf Creek Nuclear Operating Corporation, Docket No. 50-482, Wolf
Creek Generating Station, Coffey County, Kansas
Date of amendment request: July 29, 1994, as superseded by letter
dated September 15, 1995, and subsequently supplemented by letters
dated March 8, 1996, April 18, 1996, June 14, 1996, and July 12, 1996.
Brief description of amendment: The amendment revises TS 3/4.8.1,
``Electric Power Systems - A.C. Sources,'' and its associated Bases to
achieve an overall improvement in emergency diesel generator
reliability and availability.
Date of issuance: August 9, 1996
Effective date: August 9, 1996, to be implemented within 90 days of
the date of issuance.
Amendment No.: 101
Facility Operating License No. NPF-42. The amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: May 22, 1996 (61 FR
25716) The June 14, 1996, and July 12, 1996, supplemental letters
provided Bases page changes 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 August 9, 1996. No significant hazards consideration comments
received: No.
Local Public Document Room locations: Emporia State University,
William Allen White Library, 1200 Commercial Street, Emporia, Kansas
66801 and Washburn University School of Law Library, Topeka, Kansas
66621
Dated at Rockville, Maryland, this 21st day of August 1996.
For the Nuclear Regulatory Commission
Steven A. Varga,
Director, Division of Reactor Projects - I/II Office of Nuclear Reactor
Regulation
[Doc. 96-21813 Filed 8-27-96; 8:45 am]
BILLING CODE 7590-01-F
