[Federal Register Volume 61, Number 100 (Wednesday, May 22, 1996)]
[Notices]
[Pages 25696-25720]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: X96-20522]
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UNITED STATES 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 April 27, 1996, through May 10, 1996. The
last biweekly notice was published on May 8, 1996 (61 FR 20842).
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 June 21, 1996, the licensee may file a request for a hearing
with respect to issuance of the amendment to the subject facility
operating license and any person whose interest may be affected by this
proceeding and who wishes to participate as a party in the proceeding
must file a written request for a hearing and a petition for leave to
intervene. Requests for a hearing and a petition for leave to intervene
shall be filed in accordance with the Commission's ``Rules of Practice
for Domestic Licensing Proceedings'' in 10 CFR Part 2. Interested
persons should consult a current copy of 10 CFR 2.714 which is
available at the Commission's Public Document Room, the Gelman
Building, 2120 L Street, NW., Washington, DC and at the local public
document room for the particular facility involved. If a request for a
hearing or petition for leave to intervene is filed by the above date,
the Commission or an Atomic Safety and Licensing Board, designated by
the Commission or by the Chairman of the Atomic Safety and Licensing
Board Panel, will rule on the request and/or petition; and the
Secretary or the designated Atomic Safety and Licensing Board will
issue a notice of a hearing or an appropriate order.
As required by 10 CFR 2.714, a petition for leave to intervene
shall set forth with particularity the interest of the petitioner in
the proceeding, and how that interest may be affected by the results of
the proceeding. The petition should specifically explain the reasons
why intervention should be permitted with particular reference to the
following factors: (1) the nature of the petitioner's right under the
Act to be made a party to the proceeding; (2) the nature and extent of
the petitioner's property, financial, or other interest in the
proceeding; and (3) the possible effect of any order which may be
entered in the proceeding on the petitioner's interest. The petition
should also identify the specific aspect(s) of the subject matter of
the proceeding as to which petitioner wishes to intervene. Any person
who has filed a petition for leave to intervene or who has been
admitted as a party may amend the petition without requesting leave of
the Board up to 15 days prior to the first prehearing conference
scheduled in the proceeding, but such an amended petition must satisfy
the specificity requirements described above.
Not later than 15 days prior to the first prehearing conference
scheduled in the proceeding, a petitioner shall file a supplement to
the petition to intervene which must include a list of the contentions
which are sought to be litigated in the matter. Each contention must
consist of a specific statement of the issue of law or fact to be
raised or controverted. In addition, the petitioner shall provide a
brief explanation of the bases of the contention and a concise
statement of the alleged facts or expert opinion which support the
contention and on which the petitioner intends to rely in proving the
contention at the hearing. The petitioner must also provide references
to those specific sources and documents of which the petitioner is
aware and on which the petitioner intends to rely to establish those
facts or expert opinion. Petitioner must provide sufficient information
to show that a genuine dispute exists with
[[Page 25697]]
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.
Baltimore Gas and Electric Company, Docket Nos. 50-317 and 50-318,
Calvert Cliffs Nuclear Power Plant, Unit Nos. 1 and 2, Calvert
County, Maryland
Date of amendments request: April 5, 1996
Description of amendments request: Pursuant to 10 CFR 50.80 and
50.90, the Baltimore Gas and Electric Company (BGE) hereby requests the
transfer and amendment of Operating License Nos. DPR-53 and DPR-69 for
Calvert Cliffs Nuclear Power Plant, Unit Nos. 1 and 2.
The proposed license transfers and amendments are requested as part
of the pending merger between BGE and Potomac Electric Power Company
into Constellation Energy Corporation. The proposed license transfers
would transfer authority to possess and operate Calvert Cliffs from BGE
to Constellation Energy Corporation. The proposed amendments would
change the licenses as well as the related Technical Specifications, to
reflect this transfer by submitting Constellation Energy Corporation in
place of BGE as the licensee for Calvert Cliffs.
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. Would not involve a significant increase in the probability
or consequences of an accident previously evaluated.
The proposed amendment will change the name of the licensee
authorized to possess and operate Calvert Cliffs Nuclear Power Plant
from Baltimore Gas and Electric Company (BGE) to Constellation
Energy Corporation. This amendment request is necessary because of a
proposed merger of BGE and Potomac Electric Power Company into
Constellation Energy Corporation. As a result of the savings
achieved through a reduction in operating costs due to the merger,
Constellation Energy Corporation will have the financial resources
to possess and operate Calvert Cliffs.
In addition, Constellation Energy Corporation personnel will be
technically qualified to operate the plant. Baltimore Gas and
Electric Company nuclear personnel have been named to management
positions in Constellation Energy Corporation, and will remain
responsible for Calvert Cliffs operation and maintenance. The
proposed amendment involves no changes in the training program or
operating organization for Calvert Cliffs.
The proposed amendment does not require any physical change to
the facilities or substantive modifications to the Technical
Specifications or to procedures. The proposed change does not
increase the probability of an accident previously evaluated because
it does not affect any initiators in any previously evaluated
accidents. The proposed change does not increase the consequences of
an accident previously evaluated because it does not affect any of
the items on which the consequences depend.
Therefore, the proposed amendment does not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
2. Would not create the possibility of a new or different kind
of accident from any accident previously evaluated.
The proposed amendment does not modify the plant's configuration
or operations. As a result, no new accident initiators are
introduced. Therefore, the proposed amendment does not create the
possibility of a new or different kind of accident from any accident
previously evaluated.
3. Would not involve a significant reduction in a margin of
safety.
This amendment request is necessary because of a proposed merger
of BGE and Potomac Electric Power Company into Constellation Energy
Corporation. As a result of the savings achieved through a reduction
in operating costs due to the merger, Constellation Energy
Corporation will have the financial resources to possess and operate
Calvert Cliffs. Also, Constellation Energy Corporation personnel
will be technically qualified to operate the plant. Baltimore Gas
and Electric Company nuclear personnel have been named to management
positions in Constellation Energy Corporation, and will remain
responsible for Calvert Cliffs' operation and maintenance. The
proposed amendment involves no changes in the training program or
operating organization for Calvert Cliffs. In addition, the proposed
amendment to substitute Constellation Energy Corporation for BGE
does not result in any changes to the physical design or operation
of the plant. Therefore, the proposed amendment does not involve a
significant reduction in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendments request involves no significant hazards consideration.
Local Public Document Room location: Calvert County Library, Prince
Frederick, Maryland 20678.
[[Page 25698]]
Attorney for licensee: Jay E. Silbert, Esquire, Shaw, Pittman,
Potts and Trowbridge, 2300 N Street, NW., Washington, DC 20037.
NRC Project Director: Susan Frant Shankman, Acting
Carolina Power & Light Company, et al., Docket Nos. 50-325 and 50-
324, Brunswick Steam Electric Plant, Units 1 and 2, Brunswick
County, North Carolina
Date of amendments request: April 2, 1996
Description of amendments request: The proposed amendments revise
the Brunswick Steam Electric Plant, Units 1 and 2, Technical
Specifications (TS) to allow uprate of the units to 105 percent of
rated thermal power.
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:
. May the proposed activity involve a significant increase in
the probability or consequences of an accident evaluated previously
in the Safety Analysis Report?
The increase in power level, steam flow, feedwater flow and
associated instrument setpoint changes will not significantly
increase the probability or consequences of an accident previously
evaluated.
The probability (frequency of occurrence) of Design Basis
Accidents occurring is not affected by the increase in power level,
as plant equipment will remain in compliance with the applicable
regulatory criteria (ASME Codes, IEEE Standards, NEMA Standards,
Regulatory Guide criteria, etc.). The physical plant changes
necessary to support power uprate include instrument setpoint
changes, indicating meter scale changes for the RWCU [reactor water
cleanup] System flow and Main Steam Flow indicators, Leak Detection,
Process Computer, ERFIS [emergency response facility information
system], and Feedwater System software changes, and SRV [safety/
relief valve] setpoint changes. The setpoints were calculated in
accordance with the CP&L Setpoint Methodology. Utilizing this
methodology ensures scram setpoints (instrument settings that
initiate automatic plant shutdowns) will be established such that
there is no significant increase in scram frequency due to uprate.
No new challenges to safety related equipment will result from power
uprate.
The changes in consequences of hypothetical accidents which
would occur from 102% of the uprated power (2609 MWt), compared to
those previously evaluated from [greater than or equal to] 102% of
the original power (2485 MWt), are not significant, because the
accident evaluations at uprated power will not result in exceeding
the NRC approved acceptance limits. The spectrum of hypothetical
accidents and transients has been investigated, and those accidents/
transients currently evaluated in the UFSAR [Updated Final Safety
Analysis Report] were shown to meet the plant's current regulatory
criteria at uprated conditions (105%). In the area of core design,
for example, the fuel operating limits will still be met at the
uprated power level, and fuel reload analyses show plant transients
will still meet the criteria accepted by the NRC as specified in
NEDO-24011, ``GESTAR II.'' Challenges to fuel or ECCS [emergency
core cooling system] performance have been evaluated and shown to
meet the criteria of 10CFR50 Appendix K. Challenges to the
containment have been evaluated and still meet 10CFR50 Appendix A
Criterion 38, Long Term Cooling, and Criterion 50, Containment.
Bounding events involving radiological releases have been evaluated
and were shown to be well within the criteria of 10CFR100.
2. May the proposed activity create the possibility of a new or
different kind of accident from any accident previously evaluated in
the Safety Analysis Report?
The change in reactor thermal power will not create the
possibility of a new or different kind of accident from any accident
previously evaluated.
Equipment that could be affected by power uprate has been
evaluated. No new operating mode, safety related equipment lineup,
accident scenario, or equipment failure mode was identified. The
full spectrum of accident considerations defined in the BNP
[Brunswick Nuclear Plant] UFSAR has been evaluated and no new or
different kind of accident has been identified. Uprate uses
developed technology and applies it within the capabilities of
existing plant equipment in accordance with existing regulatory
criteria including NRC approved codes, standards, and methods.
General Electric has designed BWRs [Boiling Water Reactors] of
higher power levels than the uprated power of any of the currently
uprated BWR/4 fleet and has not identified new power dependent
accidents.
The changes to the Technical Specifications required to
implement power uprate make little change to the plant's
configuration. These changes fall into three major categories. The
first includes those changes resulting from power uprate parameter
changes. These parameter changes, such as the increase in vessel
pressure, temperature and piping system flows are minor in nature.
The evaluations have shown the plant is still within its design
capabilities when operating under these conditions. The changes
required as a result of power uprate will not affect the design
function(s) of currently installed equipment; therefore, there is no
possibility of a new or different kind of failure mode. The second
set of changes is a result of applying setpoint methodology to
calculate TS Allowable Values and Normal Trip Setpoints for
instruments that are directly affected by the parameter changes due
to power uprate. By using CP&L's methodology, the TS values were
calculated to ensure adequate margin exists between the analytical
limit and the TS Allowable Value. The third change include [sic]
setpoints that were reconstituted by the power uprate project.
Again, CP&L methodology was applied and the results show the
setpoints have moved to a more conservative value. This will reduce
the likelihood of spurious scrams and unnecessary challenges to
safety systems while ensuring initiation/actuation equipment
continues to function consistent with existing accident analyses.
3. Does the proposed activity involve a significant reduction in
a margin of safety defined in the basis of any Operating License
Technical Specification?
Power Uprate will not involve a significant reduction in a
margin of safety. The bounding events which had been analyzed in the
UFSAR were reevaluated to demonstrate that power uprate can be
implemented without exceeding any analyzed limit. Because the
applicable safety analysis criteria and limits are satisfied for
power uprate, the margin of safety associated with the safety limits
and other limits identified in the Technical Specifications will be
maintained.
As discussed in Section 5 of GE Nuclear Energy's License Topical
Report NEDO-31984P ``Generic Evaluations of General Electric Boiling
Water Reactor Power Uprate,'' the safety margins prescribed by the
Code of Federal Regulations (CFR) have been maintained by meeting
the appropriate regulatory criteria. Similarly, the margins provided
by the application of the ASME design criteria have been maintained.
The Brunswick unique analysis NEDC-32466P ``Power Uprate Safety
Analysis Report for Brunswick Steam Electric Plant Units 1 and 2''
discusses the effects of power uprate on safety margins for (1) fuel
thermal limits, (2) design basis accidents and the challenges for
fuel, containment and radiological releases, (3) transient analysis,
(4) non-LOCA radiological releases, and (5) environmental
consequences. These evaluations conclude that applicable safety
analysis criteria and limits are satisfied, and thus, the margins of
safety will be maintained.
The changes to the Technical Specification instrumentation will
not involve a reduction in the margin of safety. The calculations
performed for power uprate have established an analytical limit and
calculated the TS Allowable Value and Nominal Trip Setpoint using
formal setpoint methodology. This ensures the instrumentation
functional requirements are met.
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 North Carolina
at Wilmington, William Madison Randall Library, 601 S. College Road,
Wilmington, North Carolina 28403-3297.
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
[[Page 25699]]
Carolina Power & Light Company, Docket No. 50-261, H. B. Robinson
Steam Electric Plant, Unit No. 2, Darlington County, South Carolina
Date of amendment request: March 29, 1996
Description of amendment request: The proposed amendment would
revise the technical specifications (TS) to add an allowance to
complete a TS required surveillance within 24 hours of discovery of a
missed surveillance in accordance with the guidance of Generic Letter
(GL) 87-09, ``Sections 3.0 and 4.0 of the Standard Technical
Specifications (STS) on the Applicability of Limiting Conditions for
Operation and Surveillance Requirements.'' The wording specifying
intervals for testing has been changed to reflect wording consistent
the new STS. Typographical errors in the basis are also being
corrected.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. The proposed changes do not involve a significant increase in the
probability or consequences of an accident previously evaluated.
The proposed changes clarify and incorporates [sic] NRC guidance
for application of extending or moving surveillance intervals by
plus or minus 25%, by elimination of restrictive surveillance
interval descriptions that conflict with NRC guidance, by allowing
for an additional 24 hours to perform missed surveillances, and by
providing a defined finite period for the term ``immediate'' for
Technical Specification (TS) and Inservice Inspection (ISI)
surveillances. The basis for extending or moving surveillances, as
stated in GL 89-14, ``Line-Item Improvements in Technical
Specifications - Removal of the 3.25 Limit on Extending Surveillance
Intervals,'' is to provide plants flexibility for scheduling the
performance of surveillances and to permit consideration of plant
operating conditions that may not be suitable for conducting a
surveillance at the specified time interval. Such operating
conditions include transient plant operation or ongoing surveillance
or maintenance activities. Extending surveillance intervals during
plant operation can result in a benefit to safety when a scheduled
surveillances [sic] is due at a time that is not suitable for
conducting the scheduled surveillance. NUREG-1431, ``Standard
Technical Specifications - Westinghouse Plants,'' states ``the 25%
extension does not significantly degrade the reliability that
results from performing the surveillance at its specified
frequency.'' This is based on the recognition that the most probable
result of any particular surveillance being performed is the
verification of conformance with the surveillance requirements. The
basis for the 24 hour delay period, as stated in the basis for
NUREG-1431, includes consideration of unit conditions, adequate
planning, availability of personnel, the time required to perform
the surveillance, the recognition that the most probable result of
any particular surveillance being performed is the verification of
conformance with the requirements.'' The basis for defining the term
``immediate'' is to provide guidance to plant personnel for
conducting operability testing of the Steam Driven Auxiliary
Feedwater pump after extended shutdown periods in order to minimize
plant risks and not pose an unsafe operational transient during an
unstable plant configuration (i.e., during plant startup). Since
these changes do not affect plant design, operation, or the manner
in which testing is performed, the proposed changes do not involve a
significant increase in the probability or consequences of an
accident previously evaluated.
2. The proposed changes do not create the possibility of a new
or different kind of accident from any accident previously
evaluated.
The proposed changes clarify and incorporates [sic] NRC guidance
for application of extending or moving surveillance intervals by
plus or minus 25%, by elimination of restrictive surveillance
interval descriptions that conflict with NRC guidance, by allowing
for an additional 24 hours to perform missed surveillances, and by
providing a defined finite period for the term ``immediate'' for TS
and ISI surveillances. Since these changes do not affect plant
design, operation, or the manner in which testing is performed, the
proposed changes do not create the possibility of a new or different
kind of accident from any accident previously evaluated.
3. The proposed changes do not involve a significant reduction
in the margin of safety.
The changes proposed, with the exception of allowing an
additional 24 hours to complete missed surveillances, are to clarify
existing surveillance intervals and to provide more specific and
detailed criteria without changing current surveillance scheduling
methodologies. The NRC has determined that allowing an additional 24
hours to complete missed surveillance tests minimizes additional
challenges to plant operations such that there is a conservative
balance between the risk associated with performing the surveillance
during stable plant conditions and the risk of imposing a plant
transient due to TS action statements or changing ``modes'' of
operation. These extensions are current industry practices endorsed
by the NRC which provide flexibility for scheduling and performing
surveillances and permit consideration of plant operating conditions
that may not be suitable for conducting a surveillance at either the
specified time interval or inadvertently missing the surveillance
interval. The risk to safety is low in contrast to the alternatives;
therefore, the proposed changes do not involve a significant
reduction in the margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: Hartsville Memorial Library,
147 West College Avenue, Hartsville, South Carolina 29550
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. 50-373 and 50-374, LaSalle
County Station, Units 1 and 2, LaSalle County, Illinois
Date of amendment request: April 8, 1996
Description of amendment request: The proposed amendments would
change various sections of the Technical Specifications (TS) to reflect
the transition of fuel supplier from Generic Electric to Siemens Power
Corporation (SPC). The amendments would revise the definitions and
Limiting Conditions for Operation related to Linear Heat Generation
Rate, Critical Power Ratio, Maximum Critical Power Ratio, and Fraction
of Limiting Power Density to incorporate SPC terms and methodology or
to make the TS vendor neutral. Section 6.0 of the TS would be revised
to include SPC references. The proposed amendment also adds a
requirement to adjust the Average Planar Linear Heat Generation Rate
when the reactor is in single loop operation since SPC methodologies
may require this reduction factor for SPC fuel. The SPC methodologies
to be added to the TS have previously been approved by the NRC. The
proposed amendment would also relocate requirements for the traversing
in-core probe system from the TS to the Core Operating Limits Report
and would upgrade the fuel description in Section 5.0 as a line item
from the Improved Technical Specifications.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. 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
[[Page 25700]]
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 (Note the mixed core treatment of critical power ratio is
being addressed under separate correspondence). These approved
methods and models are used to determine the fuel thermal limits.
Traversing In-core Probe (TIP) uncertainty are assumptions in the
approved Siemens core monitoring methodologies. 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. 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. The support systems for minimizing the
consequences of transients and accidents are not affected by the
proposed amendment.
New Fuel Design
The use of ATRIUM 9B fuel at LaSalle 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. (See Boiling
Water Reactor Licensing Methodology Summary, Siemens Power
Corporation, EMF-94-217(NP)). 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 control rod drive response. The
ATRIUM-9B fuel design will not adversely affect performance of
neutron instrumentation nor will it adversely affect the movement of
control blades. The exterior dimensions of the ATRIUM-9B fuel
assembly are essentially identical to the GE9B; the ATRIUM-9B fuel
assembly for LaSalle uses a standard fuel channel and normal control
cell positioning (i.e., no offset). Thus, no adverse interactions
with the adjacent control blade and nuclear instrumentation are
anticipated. 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.
The ATRIUM 9B design is neutronically compatible with the
existing fuel types and core components in the LaSalle core. SPC
tests have demonstrated that the ATRIUM-9B fuel design is
hydraulically compatible with the GE9 fuel. The bundle pressure drop
characteristics of the ATRIUM 9B bundle are similar to those of the
GE9 fuel design, hence core thermal-hydraulic stability
characteristics are not adversely affected by the ATRIUM 9B design.
An evaluation of the Emergency Procedures is being performed to
ensure that the use of the ATRIUM-9B fuel at LaSalle does not alter
any assumptions previously made in evaluating the radiological
consequences of an accident at LaSalle Station.
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 exception
is the mixed core treatment of critical power ratio, which is being
addressed under separate correspondence.
The description of the fuel is expanded to be consistent with
NUREG-1434. The description of the fuel materials, lead test
assembly use, and stating that designs must have been analyzed with
NRC Staff approved codes does not change existing methods; it only
describes them.
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. As stated above, the proposed changes do
not permit modes of reactor operation which differ from those
currently permitted.
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. 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. (See Boiling Water
Reactor Licensing Methodology Summary, Siemens Power Corporation,
EMF-94-217(NP)). The similarities in fuel design and operation
indicate there would be no expectation of introducing new or
different types of accidents than have been considered for the
existing fuel. Therefore, the use of ATRIUM-9B fuel at LaSalle 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 Core Operating Limits Report) 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
Specification 6.0 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. Therefore, there is
not a significant reduction in the margin of safety.
New Fuel Design
The exterior dimensions of the ATRIUM-9B fuel assembly are
essentially identical to the GE9B; the ATRIUM-9B fuel assembly for
LaSalle uses a standard fuel channel and normal control cell
positioning; i.e., no offset. Thus, no adverse interactions with the
adjacent control blade and nuclear instrumentation are anticipated.
The change does not adversely impact equipment important to safety
and, therefore does not reduce the margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
requested amendments involve no significant hazards consideration.
Local Public Document Room location: Jacobs Memorial Library,
Illinois Valley Community College, Oglesby, Illinois 61348.
Attorney for licensee: Michael I. Miller, Esquire; Sidley and
Austin, One
[[Page 25701]]
First National Plaza, Chicago, Illinois 60603
NRC Project Director: Robert A. Capra
Commonwealth Edison Company, Docket Nos. 50-373 and 50-374, LaSalle
County Station, Units 1 and 2, LaSalle County, Illinois
Date of amendment request: April 9, 1996
Description of amendment request: The proposed amendments would
eliminate the automatic reactor scram function and the group 1 and 3
isolation valve closure functions associated with the Main Steam Line
Radiation Monitoring (MSLRM) system high radiation setpoint.
Elimination of these functions will eliminate potential spurious scrams
and isolations caused by increased main steam line radiation levels
during hydrogen injection. The licensee also proposes to raise the
MSLRM system alarm setpoints which are not part of the Technical
Specifications to include increased background radiation during
hydrogen injection. The proposed amendment would also delete the
surveillance requirements for the associated instruments.
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 because:
Redefining the full power radiation background, thus changing
the MSLRM alarm setpoint, does not change the probability of
occurrence of any accident which has been postulated and analyzed in
the UFSAR, but will reduce the probability of the inadvertent MSIV
closure transient which is an analyzed transient in the UFSAR. It
does not change the probability of malfunction of any equipment
important to safety associated with [loss of coolant accident] LOCA,
fuel handling accident or [control rod drop accident] CRDA. It also
does not change the resultant offsite radiological dose from the
bounding design basis CRDA. This is based upon all radioactivity,
resulting from the design basis CRDA, going to the condenser
instantaneously (or independent of the actual MSLRM setpoint) in the
offsite dose calculation.
The elimination of reactor scram and isolation of MSIVs,
isolation of main steam line drain valves and reactor water sample
line valves, associated with the MSLRM system actuation do not
introduce, mitigate, or reduce the probability of any design basis
accident, or any accident, evaluated in the UFSAR. The topical
report NEDO-31400A has shown that there is essentially no reasonable
radiological consequence benefit in a design basis CRDA of retaining
the MSLRM associated reactor scram and MSIV isolation function. In
addition, the probability of inadvertent scram and isolation is
reduced. The proposed change will not adversely impact the operation
of the [reactor protection system] RPS or [primary containment
isolation system] PCIS with respect to performing its other intended
safety functions. The proposed change will not affect the operation
of other plant systems or equipment important to safety. The
consequences of eliminating the automatic closure of the main steam
line drain isolation valves and reactor recirculation water sample
line isolation valves along with the MSIVs has been evaluated to be
negligible additions to the CRDA doses. A [LaSalle County Station]
LSCS unique analysis has demonstrated that the radiological doses as
a result of design basis CRDA are acceptable.
The MSLRM system high radiation trip was intended to function in
response to a CRDA which has been previously evaluated. No credit
for MSIV closure was taken in the CRDA analysis since it postulates
that all the radioactive material assumed to be released from the
fuel is transported to the main condenser prior to MSIV closure.
Furthermore, the probability of a fuel failure is independent of the
operation of the MSLRM system.
By eliminating the MSLRM induced MSIV closure, the Offgas system
can be utilized to reduce potential offsite doses after a CRDA. The
[mechanical vacuum pump] MVP is tripped no later than 15 minutes of
a Hi-Hi radiation alarm but analytically results in acceptable
offsite doses.
Thus the proposed amendment will not increase the probability of
any accident previously evaluated, and the elimination of the MSLRM
isolation signal for MSIVs and other small containment valves will
not significantly increase the consequences of a CRDA as previously
evaluated.
2) Create the possibility of a new or different kind of accident
from any accident previously evaluated because:
Redefining the full power radiation background, thus changing
the actual MSLRM alarm setpoint, does not alter the configuration of
the plant. It does not revise any logic or function of the MSLRM
trip channels or add, replace, or delete any equipment important to
safety. Therefore it does not introduce any new failure modes or
create any possibility of a new accident which may challenge safety
to the public and has not been previously analyzed. It also does not
involve any equipment which either has not been evaluated
previously, or may have any safety consequences to the public.
The proposed Technical Specification changes involve eliminating
the MSLRM system high radiation trip function for initiating an
automatic reactor scram, and automatic isolations. The proposed
changes will not affect the operation of other plant systems or
equipment important to safety. The MSLRM system will continue to
initiate alarms as before. Plant procedures will be in place to take
appropriate mitigative measures in response to a high alarm.
The isolation and reactor scram functions associated with the
MSLRM system actuation were originally intended to mitigate, not
prevent, a potential accident scenario such as a CRDA or gross fuel
failure event. Adding or removing an electronic signal, such as the
one from the MSLRM system, does not change system or hardware design
within the reactor vessel pressure boundary, and therefore will not
create the possibility of a new or different kind of accident from
those evaluated in the UFSAR like a LOCA or CRDA during power
operation. It also does not create the possibility of a new or
different kind of accident outside the reactor vessel pressure
boundary from those evaluated in the UFSAR, such as a LOCA or Fuel
Handling Accident. Removing the isolation signal also reduces the
probability of inadvertent scram and isolation.
Therefore the proposed amendment will not create the possibility
of a new or different kind of accident from any accident previously
analyzed.
3) Involve a significant reduction in the margin of safety
because:
The current MSLRM trip Hi-Hi alarm setpoint (about 4 R/hour with
full power background at 1.3 R/hour) is at 3 times the full power
radiation background. As indicated in the plant unique analytical
result for LSCS, the radiological reading at the MSLRMs for design
basis CRDA is equivalent to over 1200 times the normal full power
radiation background (1600 R/hour divided by 1.3 R/hour), or 150
times the full power radiation background during peak HWC
environment (since the radiation background is 8 times the normal
background). Thus the safety margin was very large, and would still
be quite large with the HWC background factored into the MSLRM
actuation setpoint (3 x 8 x 1.3 = about 50). The Hi alarm setpoint
of 1.5 times full power background likewise will have a higher
safety margin. Thus there is basically no adverse consequence to the
margin of safety in the basis for the LaSalle technical
specifications.
The proposed Technical Specification changes to eliminate the
MSLRM system high radiation trip function for initiating an
automatic reactor scram, and automatic closure of the MSIVs, main
steam line drain isolation valves, and reactor recirculation water
sample line isolation valves do not cause radiological dose
consequences to exceed the limit established by SRP 15.4.9.
Per NEDO-31400A, the elimination of MSLRM trip/scram signal will
result in the reduction of potential inadvertent scrams, unnecessary
safety-related actuations, undue vessel isolation, and duty
challenges during normal plant operation. These can be interpreted
to be a potential reduction in core damage frequency, which
translates to an improvement in the margin of safety.
Thus the margin of safety as defined in the basis of the
technical specifications is essentially unaffected, and is therefore
acceptable.
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
[[Page 25702]]
requested amendments involve no significant hazards consideration.
Local Public Document Room location: Jacobs Memorial Library,
Illinois Valley Community College, Oglesby, Illinois 61348.
Attorney for licensee: Michael I. Miller, Esquire; Sidley and
Austin, One First National Plaza, Chicago, Illinois 60603
NRC Project Director: Robert A. Capra
Commonwealth Edison Company, Docket Nos. 50-373 and 50-374, LaSalle
County Station, Units 1 and 2, LaSalle County, Illinois
Date of amendment request: April 16, 1996
Description of amendment request: The proposed amendments would
eliminate the Technical Specification requirement to perform response
time testing for selected instruments. The instruments affected are the
sensors for selected reactor protection system instrumentation, main
steam isolation actuation instrumentation, and all sensors for
emergency core cooling system (ECCS) actuation instrumentation. The
proposed changes are supported by analyses performed by the Boiling
Water Reactor Owners' Group as documented in NEDO-32291-A which was
approved by the NRC for use in license amendment applications.
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 because:
The purpose of the proposed Technical Specification (TS) change
is to eliminate response time testing requirements for selected
components in the Reactor Protection System (RPS), Isolation
Actuation instrumentation and Emergency Core Cooling System (ECCS)
actuation instrumentation. The Boiling Water Reactor Owners' Group
(BWROG) has completed an evaluation which demonstrates that response
time testing is redundant to the other TS-required testing. These
other tests, in conjunction with actions taken in response to NRC
Bulletin 90-01, ``Loss of Fill-Oil in Transmitters Manufactured by
Rosemount,'' and Supplement 1, are sufficient to identify failure
modes or degradations in instrument response time and ensure
operation of the associated systems within acceptable limits. There
are no known failure modes that can be detected by response time
testing that cannot also be detected by the other TS-required
testing. This evaluation was documented in NEDO-32291-A, ``System
Analyses for the Elimination of Selected Response Time Testing
Requirements,'' dated October 1995. LaSalle County Station, LaSalle,
has confirmed the applicability of this evaluation to LaSalle. In
addition, LaSalle will complete the actions identified in the NRC
staffs safety evaluation of NEDO-32291-A.
Because of the continued application of other existing TS-
required tests such as channel calibrations, channel checks, channel
functional tests, and logic system functional tests, the response
time of these systems will be maintained within the acceptance
limits assumed in plant safety analyses and required for successful
mitigation of an initiating event. The proposed changes do not
affect the capability of the associated systems to perform their
intended function within their required response time, nor do the
proposed changes themselves affect the operation of any equipment.
As a result, LaSalle has concluded that the proposed changes do not
involve a significant increase in the probability or the
consequences of an accident previously evaluated.
2) Create the possibility of a new or different kind of accident
from any accident previously evaluated because:
The proposed changes only apply to the testing requirements for
the components identified above and do not result in any physical
change to these or other components or their operation. As a result
no new failure modes are introduced. Therefore, the proposed changes
do 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
because:
The current TS-required response times are based on the maximum
allowable values assumed in the plant safety analyses. These
analyses conservatively establish the margin of safety. As described
above, the proposed changes do not affect the capability of the
associated systems to perform their intended function within the
allowed response time used as the basis for the plant safety
analyses. The potential failure modes for the components within the
scope of this request were evaluated for impact on instrument
response time. This evaluation confirmed that, with the exception of
loss of fill-oil of Rosemount transmitters, the remaining TS-
required testing is sufficient to identify failure modes or
degradations in instrument response times and ensure that operation
of the applicable instrumentation is within acceptable limits. The
actions taken in response to NRC Bulletin 90-01 and Supplement 1 are
adequate to identify loss of fill-oil failures of Rosemount
transmitters. As a result, it has been concluded that plant and
system response to an initiating event will remain in compliance
with the assumptions of the safety analysis.
Further, although not explicitly evaluated, the proposed changes
will provide an improvement to plant safety and operation by the
following:
a. Reducing the time safety systems are unavailable,
b. Reducing the potential for safety system actuations,
c. Reducing plant shutdown risk,
d. Limiting radiation exposure to plant personnel, and
e. Eliminating the diversion of key personnel resources to
conduct unnecessary testing.
Therefore, LaSalle has concluded that this request will not
significantly reduce the margin of safety, and may actually cause an
increase 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
requested amendments involve no significant hazards consideration.
Local Public Document Room location: Jacobs Memorial Library,
Illinois Valley Community College, Oglesby, Illinois 61348.
Attorney for licensee: Michael I. Miller, Esquire; Sidley and
Austin, One First National Plaza, Chicago, Illinois 60603
NRC Project Director: Robert A. Capra
Duke Power Company, et al., Docket Nos. 50-413 and 50-414, Catawba
Nuclear Station, Units 1 and 2, York County, South Carolina
Date of amendment request: November 2, 1994
Description of amendment request: The proposed amendments would
delete the content of Appendix B, ``Environmental Protection Plan''
(nonradiological), and modify License Condition 2.C.(2) to delete that
portion which refers to the Environmental Protection Plan.
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 amendments would not involve a significant
increase in the probability or consequences of an accident
previously evaluated]:
Deletion of the Environmental Protection Plan and modifying
License Condition 2.C.(2) will have no impact on the probability or
consequences of an accident previously evaluated because the changes
will not have any impact upon the design or operation of any plant
systems or components.
2. [The proposed amendments would not create the possibility of
a new or different kind of accident from any accident previously
evaluated]:
The proposed revision will not create the possibility of a new
or different kind of accident from any previously evaluated because
the revision is administrative in nature and will not change the
types and amounts of effluent that will be released.
3. [The proposed amendments would not involve a significant
reduction in a margin of safety]:
[[Page 25703]]
The proposed revision will not reduce a margin of safety because
it is administrative in nature and will not [a]ffect the margin of
safety as defined in the basis for any Technical Specifications.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: York County Library, 138 East
Black Street, Rock Hill, South Carolina 29730
Attorney for licensee: Mr. Albert Carr, Duke Power Company, 422
South Church Street, Charlotte, North Carolina 28242
NRC Project Director: Herbert N. Berkow
Duquesne Light Company, et al., Docket No. 50-412, Beaver Valley
Power Station, Unit 2, Shippingport, Pennsylvania
Date of amendment request: April 29, 1996
Description of amendment request: The proposed amendment would
revise Technical Specification (TS) 5.3.1 to allow the use of ZIRCO as
an alternate zirconium-based fuel rod material and remove the word clad
since it has been eliminated from the text of the NRC's improved
Standard Technical Specifications (NUREG-1431). Limited substitution of
fuel rods by ZIRCO filler rods would also be permitted. The proposed
amendment would revise Note 2 on TS Table 3.9-1 to specify that the
maximum burnup in the peak fuel rod in a fuel assembly stored in Region
2 spent fuel racks should not exceed the NRC-approved limit for WCAP-
12610 rather than the current maximum burnup limit of 60 GWD/MTU.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. Does the change involve a significant increase in the
probability or consequences of an accident previously evaluated?
The methodologies used in the accident analyses remain
unchanged. The proposed changes do not change or alter the design
assumptions for the systems or components used to mitigate the
consequences of an accident. Use of ZIRLO fuel rod material does not
adversely affect fuel performance or impact nuclear design
methodology. Therefore, accident analysis results are not impacted.
The operating limits will not be changed and the analysis
methods to demonstrate operation within the limits will remain in
accordance with NRC approved methodologies. Other than the changes
to the fuel assemblies, there are no physical changes to the plant
associated with this technical specification change. A safety
analysis will continue to be performed for each cycle to demonstrate
compliance with all fuel safety design bases.
VANTAGE 5 fuel assemblies with ZIRLO fuel rods meet the same
fuel assembly and fuel rod design bases as other VANTAGE 5 fuel
assemblies. In addition, the 10 CFR 50.46 criteria are applied to
the ZIRLO fuel rods. The use of these fuel assemblies will not
result in a change to the reload design and safety analysis limits.
Since the original design criteria are met, the ZIRLO fuel rods will
not be an initiator for any new accident. The fuel rod material is
similar in chemical composition and has similar physical and
mechanical properties as Zircaloy-4. Thus, the fuel rod integrity is
maintained and the structural integrity of the fuel assembly is not
affected. ZIRLO improves corrosion performance and dimensional
stability. No concerns have been identified with respect to the use
of an assembly containing a combination of Zircaloy-4 and ZIRLO fuel
rods.
The dose predictions in the safety analyses are not sensitive to
the fuel rod material used; therefore, the radiological consequences
of accidents previously evaluated in the safety analysis remain
valid. A reload analysis is completed for each cycle, in accordance
with NRC approved methodologies. Therefore, the proposed change does
not involve a significant increase in the probability or
consequences of an accident previously evaluated.
2. Does the change create the possibility of a new or different
kind of accident from any accident previously evaluated.
VANTAGE 5 fuel assemblies with ZIRLO fuel rods satisfy the same
design bases as those used for other VANTAGE 5 fuel assemblies. All
design and performance criteria continue to be met and no new
failure mechanisms have been identified. The ZIRLO fuel rod material
offers improved corrosion resistance and structural integrity.
The proposed changes do not affect the design or operation of
any system or component in the plant. The safety functions of the
related structures, systems, or components are not changed in any
manner, nor is the reliability of any structure, system, or
component reduced. The changes do not affect the manner by which the
facility is operated and do not change any facility design feature,
structure, or system. No new or different type of equipment will be
installed. Since there is no change to the facility or operating
procedures, and the safety functions and reliability of structures,
systems, or components are not affected, the proposed changes do not
create the possibility of a new or different kind of accident from
any accident previously evaluated.
3. Does the change involve a significant reduction in a margin
of safety?
The use of Zircaloy-4, ZIRLO, or stainless steal filler rods in
fuel assemblies will not involve a significant reduction in the
margin of safety because analyses using NRC approved methodology
will be performed for each configuration to demonstrate continued
operation within the limits that assure acceptable plant response to
accidents and transients. These analyses will be performed using NRC
approved methods that have been approved for application to the fuel
configuration.
Use of ZIRLO as fuel rod material does not change the VANTAGE 5
reload design and safety analysis limits. The use of these fuel
assemblies will take into consideration the normal core operating
conditions allowed in the technical specifications. For each reload
core, the fuel assemblies will be evaluated using NRC approved
reload design methods, including consideration of the core physics
analysis peaking factors and core average linear heat rate effects.
Based on the above, it is concluded that the proposed license
amendment request does not result in a significant reduction in
margin with respect to plant safety as defined in the UFSAR [Updated
Final Safety Analysis Report] or any plant technical specification
BASES.
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: B. F. Jones Memorial Library,
663 Franklin Avenue, Aliquippa, Pennsylvania 1500l.
Attorney for licensee: Jay E. Silberg, Esquire, Shaw, Pittman,
Potts & Trowbridge, 2300 N Street, NW., Washington, DC 20037.
NRC Project Director: John F. Stolz
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 2, 1996
Description of amendment request: The proposed technical
specification amendments would extend the allowed outage times for
emergency diesel generators at Arkansas Nuclear One, Units 1 and 2 to 7
days with an additional, once per refueling cycle extension of 7 more
days for each machine.
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 emergency diesel generators (EDGs) are backup alternating
current power sources
[[Page 25704]]
designed to power essential safety systems in the event of a loss of
offsite power. The EDGs are not accident initiators in any accident
previously evaluated. Probabilistic Safety Analysis (PSA) methods
were utilized in order to fully evaluate the EDG allowed outage time
(AOT) extension proposed in this submittal. The results of these
analyses indicate there is not a significant increase in the
probability of an accident previously evaluated. Therefore, this
change does not involve an increase in the probability of an
accident previously evaluated.
The EDGs provide backup power to components that mitigate the
consequences of accidents. The current TSs allow for an EDG to be
removed from service for an AOT. The proposed amendment extends the
current AOT for an EDG. The proposed change does not allow any more
equipment to be removed from service at one time. The proposed
changes to the AOTs do not affect any of the assumptions used in
deterministic safety analysis. By extending the EDG AOT, the
consequences of an accident previously evaluated will remain
unchanged.
The proposed change removes redundant requirements associated
with an inoperable emergency power supply from the TS for the
pressurizer proportional heaters. The operability requirements for
emergency power supplies and actions to be taken if an EDG is
inoperable are already addressed in the ANO-2 TS 3.8.1.1.
The associated changes that remove the requirements to test the
EDGs if one or both offsite power supplies are inoperable, for an
inoperable station battery, for an inoperable component in the two
ESF electrical distribution systems, the accelerated testing
requirements of the EDGs, and the daily testing requirements for the
operable EDGs improve the reliability for the operable EDGs by
reducing the number of unnecessary starts and stops. By improving
the EDG reliability, this change will not increase the consequences
of the accidents previously evaluated.
The other changes in this submittal associated with the bases
are considered administrative in nature and have no effect on the
consequences of an accident previously evaluated.
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.
This proposed change does not alter the design, configuration,
or method of operation of the plant. Therefore, this change does not
create the possibility of a new or different kind of accident from
any previously evaluated.
Criterion 3 - Does Not Involve a Significant Reduction in the
Margin of Safety.
The proposed changes do not affect the Technical Specification
limiting conditions for operation or their bases which support the
deterministic analyses used to establish the margin of safety.
Calculations performed to analyze the change in risk based on
these changes produced acceptable values which are included in the
tables located in the description of changes section. These
calculated changes in risk fall well within that which is normally
considered acceptable. When the additional benefit of maintaining
the Emergency Diesel Generators available during shutdown cooling
operations associated with refueling outages in considered, the
overall change in risk is further reduced.
The remaining proposed changes are either associated with
increasing EDG reliability or considered administrative in nature.
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
Entergy Operations, Inc., et al., Docket No. 50-416, Grand Gulf
Nuclear Station, Unit 1, Claiborne County, Mississippi
Date of amendment request: November 20, 1995, as supplemented by
the letter dated December 15, 1995.
Description of amendment request: The licensee has proposed to
revise the Grand Gulf Nuclear Station (GGNS), Unit 1, Technical
Specifications (TSs) as follows for the drywell, the drywell airlock,
and the drywell isolation valves:
1. For the drywell in Limiting Condition of Operation (LCO)
3.6.5.1, the surveillance frequency interval for the drywell bypass
test in Surveillance Requirement (SR) 3.6.5.1.1 would be increased from
18 months to 10 years. For this interval change, an increased testing
frequency would be required if bypass performance degrades (i.e., the
leakage is greater than the limit for two consecutive tests) and the
application of SR 3.0.2, the allowance to extend the surveillance
interval by 25 percent, would be restricted to 12 months on the 10-year
interval. This includes deleting the Note in SR 3.6.5.1.1.
2. For the drywell airlock in LCO 3.6.5.2, the following changes
are requested: (a) the leak rate SR 3.6.5.2.2 would be transferred from
the airlock LCO (3.6.5.2) to SR 3.6.5.1.3 in the drywell LCO (3.6.5.1),
(b) the requirement in SR 3.6.5.2.2 for the air lock to meet a specific
overall leakage limit would be deleted, (c) the Note in SR 3.6.5.2.2
that stated that an inoperable air lock door does not invalidate the
previous air lock leakage test would be deleted, (d) the test pressure
for the air lock leakage test in SR 3.6.5.2.2 would be reduced from
11.5 psig to 3 psid, and (e) the surveillance frequency interval for
the air lock leakage and interlock testing, required in SRs 3.6.5.2.1
and 3.6.5.2.2, would be increased from 18 months to 24 months.
3. For the drywell airlock in LCO 3.6.5.2 and the drywell isolation
valves in LCO 3.6.5.3, the Action Notes, which identify that the
actions required by drywell LCO 3.6.5.1 must be taken when the drywell
bypass leakage limit is not met, would be deleted. Action C.1 of LCO
3.6.5.2 and its associated completion time would also be deleted.There
would also be changes to the Bases of the TSs for the above LCOs and
SRs, based on the proposed changes.
Basis for proposed no significant hazards consideration
determination: The amendment request dated November 20, 1995, applied
to both the Grand Gulf Nuclear Station (GGNS) and the River Bend
Station (RSB); however, not all of the proposed amendments apply to
GGNS. This Notice only discusses the amendment request for GGNS. The
reference below to proposed amendments which do not apply to GGNS are
marked by ``[....]''.
As required by 10 CFR 50.91(a), the licensee has provided its
analysis of the issue of no significant hazards consideration in its
application dated November 20, 1995, which is presented below:
Entergy Operations, Inc. proposes to change the current Grand
Gulf Nuclear Station (GGNS) [....] Technical Specifications. The
specific proposed changes are:
1. The Surveillance Frequency [interval] for the drywell bypass
test is changed [increased] from 18 months to 10 years with an
increased testing frequency required if performance degrades.
2. The following changes are requested for the drywell air lock
testing: (a) the leakage rate surveillance is moved from the air
lock Limiting Condition for Operation (LCO) to the drywell LCO, (b)
the requirement for the air lock to meet a specific overall leakage
limit is deleted, (c) the Note that an inoperable air lock door does
not invalidate the previous air lock leakage test is deleted, (d)
the GGNS test pressure for the air lock leakage test is changed
[reduced] from 11.5 psig to 3 psid, [...,] and ([e]) the
Surveillance Frequency [interval] for the air lock leakage test and
interlock test is changed [increased] from 18 months to 24 months.
3. The Actions Notes in the drywell air lock LCO and the drywell
isolation valve LCO that identifies that the Actions required
[[Page 25705]]
by the drywell LCO must be taken when the drywell bypass leakage
limit is not met is deleted. [Action C.1 of LCO 3.6.5.2 and its
associated completion time would also be deleted.]
[4. ...]
The Commission has provided standards for determining whether a
no significant hazards consideration exists as stated in 10 CFR
50.92(c). The proposed changes involve the withdrawal of operating
restrictions previously imposed because acceptable operation of the
Mark III primary containment design had not been demonstrated at the
time of licensing. As published in the Federal Register regarding no
significant hazards consideration criteria, granting of a relief,
based upon demonstration of acceptable operation from an operating
restriction that was imposed because acceptable operation had not
yet been demonstrated does not involve a significant hazards
consideration (Ref. 48 FR 14870). Furthermore, a proposed amendment
to an operating license involves no significant hazards
consideration if 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.
Entergy Operations, Inc. has evaluated the no significant
hazards consideration in its request for this license amendment,
even though the above-mentioned criterion is satisfied by this
proposal. In accordance with 10 CFR 50.91(a), Entergy Operations,
Inc. is providing the analysis of the proposed amendment against the
three standards in 10 CFR 50.92(c). A description of the no
significant hazards consideration determination follows:
I. The proposed change does not significantly increase the
probability or consequences of an accident previously evaluated.
The requested changes are either administrative changes which
clarify the format of the requirement or change the requirement to
match the design bases of the plant, a change which relocates the
requirement to the Technical Specification Bases, or a change in
[the] surveillance interval. Each of these types of change are
discussed below:
1. The administrative changes clarify the format of the
requirement or change the requirement to match the design bases of
the plant. Clarifying [the] administrative format of the Technical
Specifications does not result in any changes to the Technical
Specification requirements and, as a result, does not involve a
significant increase in the probability or consequences of an
accident previously evaluated. Also, changing the requirements of
the Technical Specifications to more closely match the design bases
of the plant will continue to assure that the plant will respond as
assumed in the accident analyses and, as a result, does not involve
a significant increase in the probability or consequences of an
accident previously evaluated.
2. The proposed changes relocate information to the Technical
Specification Bases. In the Technical Specifications Bases the
relocated information will be maintained in accordance with 10 CFR
50.59 and subject to the change control provisions in Chapter 5 of
Technical Specifications. Since any changes to the Technical
Specifications Bases will be evaluated per the requirements of 10
CFR 50.59, no increase (significant or insignificant) in the
probability or consequences of an accident previously evaluated will
be allowed. Therefore, this change does not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
3. The proposed changes in frequency for the drywell bypass
leakage and drywell air lock surveillances will continue to ensure
that no paths exist through passive drywell boundary components that
would permit gross leakage from the drywell to the primary
containment air space and result in bypassing the primary
containment pressure-suppression feature beyond the design basis
limit. The Mark III primary containment system satisfies General
Design Criterion 16 of Appendix A to 10 CFR Part 50. Maximum drywell
bypass leakage was determined previously by reviewing the full range
of postulated primary system break sizes. The limiting case was a
primary system small break loss of coolant accident (LOCA) and
yielded a design allowable drywell bypass leakage rate limit of
approximately 35,000 scfm for GGNS [....]. The Technical
Specifications acceptable limit for the bypass leakage following a
surveillance is less than 10% of this design basis value. The most
recent bypass leakage value was approximately 2.5% for GGNS [....]
of the design allowable leakage rate limit for the limiting event.
EOI [Entergy Operations, Inc.] is committed to maintaining
programmatic and oversight controls that ensure that drywell bypass
leakage remains a small fraction of the design allowable leakage
limit.
The drywell is typically exposed to essentially 0 psig during
normal plant operation and 3 psig during drywell bypass leak rate
testing. These pressures are considerably lower than the structural
integrity test pressure and are less likely to initiate a crack or
cause an existing crack to grow. Visual inspections of the
accessible drywell surfaces that have been performed since the
structural integrity tests have not revealed the presence of
additional cracking or other abnormalities. Therefore, additional
cracking of the drywell structure is not expected due to testing or
operation and, similar to the justification for the ten year 10 CFR
50 Appendix J Type A test interval, it is not considered credible
for the passive drywell structure to begin to leak sufficiently to
impact the design drywell bypass leakage limit.
The primary containment's ability to perform its safety function
is fairly insensitive to the amount of drywell leakage, thereby
providing a margin to loss of the drywell safety function that is
not normally available for systems. This insensitivity is
demonstrated by the extremely high limiting event design basis
allowable leakage for the drywell (e.g., 35,000 scfm for GGNS
[....]). The limiting leakage is almost an order of magnitude higher
for other events. Additionally, an even higher allowable leakage can
be realistically accommodated by the primary containment due to the
margins in the containment design. Because of the margins available,
it will take valves in multiple penetration flow paths leaking
excessively to cause the primary containment to fail as a result of
overpressurization, the probability that drywell isolation valve
leakage will result in primary containment failure due to excessive
drywell leakage is not considered significant and this drywell/
primary containment failure mode is not considered credible.
The proposed Technical Specification changes have no significant
impact on the GGNS Individual Plant Examination (IPE) [....]
conducted per NRC Generic Letter 88-20. The IPEs considered
overpressurization failure of primary containment as part of the
primary containment performance assessment. Due to the magnitude of
acceptable drywell leakage and the extremely low probabilities of
achieving such leakage, primary containment failure due to
preexisting excessive drywell leakage was considered a non
significant contributor to primary containment failure. Primary
containment overpressurization failure can occur with or without
preexisting excessive drywell leakage in a severe accident. This is
due to physical phenomena associated with potentially extreme
environmental conditions inside primary containment following a
severe accident. However, the calculated frequency of such extreme
conditions is very small. The proposed changes do not impact the IPE
evaluated phenomena causing primary containment overpressurization
failure nor significantly increase the probability that the drywell
has preexisting excessive leakage and therefore would not contribute
to these accident scenarios.
For the reasons discussed above, the proposed changes do not
have any significant risk impact to accidents previously evaluated
and do not significantly increase the consequences of an accident
previously evaluated. Additionally, drywell leakage is not the
initiator of any accident evaluated; therefore, changes in the
frequency of the surveillance for drywell leakage does not increase
the probability of any accident evaluated.
Therefore, the proposed changes do not significantly increase
the probability or consequences of an accident previously evaluated.
II. The proposed change does not create the possibility of a new
or different kind of accident from any accident previously
evaluated.
The requested changes are either administrative changes which
clarify the format of the requirement or change the requirement to
match the design bases of the plant, a change which relocates the
requirement to the Technical Specification Bases, or a change in
surveillance interval. Each of these types of change are discussed
below:
1. The administrative changes in the Technical Specification
requirements do not
[[Page 25706]]
involve a physical alteration of the plant (no new or different type
of equipment will be installed) nor does it change the methods
governing normal plant operation. Thus, this change does not create
the possibility of a new or different kind of accident from any
accident previously evaluated.
2. The proposed relocation of requirements does not involve a
physical alteration of the plant (no new or different type of
equipment will be installed) nor does it change the methods
governing normal plant operation. The proposed change will not
impose or eliminate any requirements. Adequate control of the
information will be maintained in the Technical Specification Bases.
Thus, the change proposed does not create the possibility of a new
or different kind of accident from any accident previously
evaluated.
3. The proposed change modifies the surveillance frequency for
drywell bypass leakage and drywell air lock surveillances. The
changes only impact the test frequency and do not result in any
change in the response of the equipment to an accident. The changes
do not alter equipment design or capabilities. The changes do not
present any new or additional failure mechanisms. The drywell is
passive in nature and the surveillance will continue to verify that
its integrity has not deteriorated. Therefore, the proposed change
does not create the possibility of a new or different kind of
accident from any accident previously evaluated.
Therefore, the proposed changes do not create the possibility of
a new or different kind of accident from any accident previously
evaluated.
III. The proposed change does not involve a significant
reduction in a margin of safety.
The requested changes are either administrative changes which
clarify the format of the requirement or change the requirement to
match the design bases of the plant, a change which relocates the
requirement to the Technical Specification Bases, or a change in
surveillance interval. Each of these types of changes are discussed
below:
1. The administrative changes in the Technical Specification
requirements do not involve a physical alteration of the plant (no
new or different type of equipment will be installed) nor does it
change the methods governing normal plant operation. Thus, this
change does not cause a significant reduction in the margin of
safety.
2. The relocation of requirements will not reduce a margin of
safety because it has no impact on any safety analysis assumptions.
In addition, the requirements to be transferred from the Technical
Specifications to the Technical Specifications Bases are the same as
the existing Technical Specifications. Since any future changes to
these requirements in the Technical Specifications Bases will be
evaluated per the requirements of 10 CFR 50.59, no reduction
(significant or insignificant) in a margin of safety will be
allowed.
3. The proposed change modifies the surveillance frequency for
drywell bypass leakage and associated air lock surveillances.
Reliability of drywell integrity is evidenced by the measured
leakage rate during past drywell bypass leakage surveillances.
Appropriate design basis assumptions will be upheld, even when
combined with the complementary bypass leakage surveillances as
proposed. Drywell integrity will continue to be tested by means of
the proposed periodic drywell bypass leakage test, performance of
the drywell air lock door latching and interlock mechanism
surveillance, and performance of additional surveillances including
exercising of drywell isolation valves. The combination of these
surveillances will provide adequate assurance that drywell bypass
leakage will not exceed the design basis limit. Margins of safety
would not be reduced unless leakage rates exceeded the design
allowable drywell bypass leakage limit. Therefore, the proposed
change does not cause a significant reduction in the margin of
safety.
Therefore, the proposed changes do not cause 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: Judge George W. Armstrong
Library, 220 S. Commerce Street, Natchez, MS 39120
Attorney for licensee: Nicholas S. Reynolds, Esquire, Winston and
Strawn, 1400 L Street, N.W., 12th Floor, Washington, DC 20005-3502
NRC Project Director: William D. Beckner
Entergy Operations Inc., Docket No. 50-382, Waterford Steam
Electric Station, Unit 3, St. Charles Parish, Louisiana
Date of amendment request: August 11, 1995, as supplemented by
letter dated February 12, 1996.
Description of amendment request: The proposed change will reduce
the minimum reactor coolant cold leg temperature from 544 Degrees F to
541 degrees F in Technical Specification Section 3.2.6, ``Reactor
Coolant Cold Leg Temperature.''
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
The proposed change involves a 3 deg.F reduction in the minimum
core inlet temperature. This change will not have any impact on the
probability of occurrence of any accident documented in the FSAR.
The impact of this change on the consequences of events
documented in the FSAR has been evaluated. The evaluation
demonstrated that most events are insensitive to the core inlet
temperature. The events that are impacted by lower core inlet
temperature are:
Loss of condenser vacuum (LOCV),
Part length CEA drop,
Single CEA withdrawal within deadband, and
CEA ejection.
The LOCV event has been reanalyzed for the upcoming Cycle (Cycle
8) and the results indicate that the peak RCS pressure remains below
the acceptable limit (110% of the design pressure, i.e., 2750 psia).
The reactivity anomaly events (remaining events) will be reanalyzed
as part of COLSS/CPC setpoint calculations. These calculations will
be performed prior to Cycle 8 startup and will address the impact of
the 3 deg.F reduction on the minimum core inlet temperature. The
CPC/COLSS databases and/or addressable constants will be modified,
as needed due to proposed change, prior to cycle startup.
A qualitative assessment of the impact of the proposed change on
the calculated LOCA blowdown loads that are applied to the major
NSSS components, their supports and the reactor vessel internals was
also performed. This assessment consisted of an evaluation of the
design margins on the major components and a determination of the
impact this lower temperature would have on those margins. The
evaluation concluded that the impact of a 3 deg.F cold leg
temperature reduction will be well within the current design
margins. Therefore, the proposed change will not involve a
significant increase in the probability or consequences of any
accident previously evaluated.
The proposed change to the minimum core inlet temperature does
not involve any change to any equipment or the manner in which the
plant will be operated. Since no hardware modifications or changes
in operation procedures will be made, the proposed change would not
create the possibility of a new or different kind of accident from
any accident previously evaluated. Therefore, the proposed change
will not create the possibility of a new or different kind of
accident from any accident previously evaluated.
The impact of the proposed change on the Waterford 3 FSAR
analyses have been evaluated. The evaluation showed that the events
that were impacted were important with respect to RCS pressure and
fuel thermal limits. One of the events that was impacted by the
proposed change was the LOCV event. This event was analyzed and the
results showed that the peak RCS pressure remained below the
acceptable limit. The impact of this change on other events
(reactivity anomaly events) will be evaluated as part of the COLSS/
CPC setpoint calculations and the COLSS/CPC databases and/or
addressable constants will modified as needed to account for any
adverse impact on the results of these events due to the proposed
change.
The impact of this change on the Linear Heat Generation Rate
limits which varies as a function of the cold leg temperature, is
accounted for by Technical Specification 3.2.1, ``Linear Heat
Rate''. The impact of this change on LOCA blowdown loads were
evaluated to be insignificant compared to the
[[Page 25707]]
current design margins. Therefore, the proposed change will not
involve a significant reduction in a margin of safety, specifically
fuel thermal limits and RCS pressure limit.
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 New Orleans
Library, Louisiana Collection, Lakefront, New Orleans, LA 70122
Attorney for licensee: N.S. Reynolds, Esq., Winston & Strawn 1400 L
Street N.W., Washington, D.C. 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: March 20, 1996, and April 23, 1996
Description of amendment request: The licensee proposed to change
the Turkey Point Units 3 and 4 Technical Specifications (TS) to
relocate the requirements for surveillance testing of the water level
and pressure channel instrumentation for the reactor coolant system
accumulators and clarify the remaining TS surveillance tests. These
amendments also modify the existing action statements of TS 3.5.1 for
accumulators to reflect the requirements of NUREG-1431 by requiring a
72-hour period to restore boron concentration if it is not within the
limits, and a 1-hour period to restore any other condition rendering
the accumulators inoperable.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below.
(1) Operation of the facility in accordance with the proposed
amendments would not involve a significant increase in the
probability or consequences of an accident previously evaluated.
The proposed amendments do not involve a significant increase in
the probability or consequences of an accident previously evaluated
because the proposed amendments conform to the guidance given in
Enclosure 1 of the NRC GL [Generic Letter] 93-05. The overall
functional capabilities of the Emergency Core Cooling System (ECCS)
accumulators will not be modified by the proposed change. This
amendment will not involve a significant increase in the probability
or consequences of an accident previously evaluated for the
following reasons:
1) The Water Level and Pressure Channel Instrumentation does not
perform a specific safety function, and merely provides an
indicating function. The instrumentation in no way affects the
capability of the accumulators to perform their respective safety
function.
2) The changes in most of the ACTION statements are more
restrictive than current TS requirements due to the one hour vice
four hour completion time, and therefore will not increase the
probability or consequences of a previously evaluated accident. If
one accumulator is inoperable for a reason other than boron
concentration, the accumulator must be returned to OPERABLE status
within 1 hour. In this condition, the required contents of three
accumulators cannot be assumed to reach the core during a Loss Of
Coolant Accident (LOCA). Due to the severity of the consequences
should a LOCA occur in these conditions, the 1 hour completion time
to open the valve, remove power to the valve, or restore the proper
water volume or nitrogen cover pressure ensures that prompt action
will be taken to return the inoperable accumulator to OPERABLE
status. The completion time minimizes the potential for exposure of
the plant to a LOCA under these conditions. The 1 hour requirement
for restoring a closed isolation valve is merely a clarification of
the existing ``immediate'' time requirement.
3) In the case of low-out-of-specification boron concentration
in one accumulator, it must be returned to within the limits within
72 hours. In this condition, ability to maintain subcriticality or
minimum boron precipitation time may be reduced. The boron in the
accumulators contributes to the assumption that the combined ECCS
water in the partially recovered core during the early reflooding
phase of a large break LOCA is sufficient to keep that portion of
the core subcritical. One accumulator below the minimum boron
concentration limit, however, will have no effect on available ECCS
water and an insignificant effect on core subcriticality during
reflood. Boiling of ECCS water in the core during reflood
concentrates boron in the saturated liquid that remains in the core.
In addition, current Turkey Point analysis demonstrate that the
accumulators discharge only a small amount following a large main
steam line break. Therefore, their impact on boron concentration in
the reactor coolant system is minor and not a design limiting event.
Thus, 72 hours is allowed to return the boron concentration to
within limits and does not increase 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 surveillance changes and clarifications, 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.
(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
system are unchanged by the proposed amendment. The modified
specifications which remove surveillance requirements from the TS to
plant procedures are consistent with the NRC GL 93-05 line-item
improvement guidance do not significantly reduce any of the margins
of safety even though the amount of surveillances is decreased. The
modification of the existing ACTION Statements do not have an
adverse on [sic] affect on the margin of safety for the following
reasons:
1) The SI [Safety Injection] Accumulator Water Level and
Pressure Channel instrumentation performs no safety function.
2) The changes in ACTION statements a) and b) are for the most
part more restrictive than existing TS requirements, the reason
being the removal of instrumentation requirements for operability.
3) In the case of low-out-of-specification boron concentration
in one accumulator, the requirement will be less restrictive, but
the low boron concentration in one accumulator will have no effect
on available ECCS water and an insignificant effect on core
subcriticality during reflood and therefore will not significantly
reduce the margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 50.92(c) are
satisfied.Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: 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
Illinois Power Company and Soyland Power Cooperative, Inc., Docket
No. 50-461, Clinton Power Station, Unit No. 1, DeWitt County,
Illinois
Date of amendment request: April 19, 1996
Description of amendment request: The proposed amendment would
include revisions to Technical Specification (TS) 3.3.6.1, ``Primary
Containment and Drywell Isolation Instrumentation; TS
3.3.6.2, ``Secondary Containment Isolation Instrumentation;
TS 3.3.7.1, ``Control Room Ventilation System
Instrumentation; TS 3.6.1.2, ``Primary Containment Air
Locks; TS 3.6.1.3,
[[Page 25708]]
``Primary Containment Isolation Valves; TS 3.6.4.1,
``Secondary Containment; TS 3.6.4.2, ``Secondary Containment
Isolation Dampers; TS 3.6.4.3, ``Standby Gas
Treatment; TS 3.7.3, ``Control Room Ventilation;
and TS 3.7.4, ``Control Room AC System.'' These TSs would be revised to
eliminate CORE ALTERATIONS as an applicable condition for which the
associated Limiting Conditions for Operation (LCO) must be met.
Consistent changes are also proposed for the associated ACTIONS in each
of these LCOs, to reflect the changes in the applicable conditions. The
intent of these proposed changes is to allow certain activities such as
control rod venting, which is considered a CORE ALTERATION in MODE 5,
to be performed without the requirements of the identified LCOs being
met.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration which is presented below:
1. The proposed changes eliminate CORE ALTERATIONS as an
applicable condition requiring operability of the primary and
secondary containment and control room ventilation system. As stated
in the BASES for the associated Technical Specifications,
operability of these systems is primarily required for mitigation of
the design basis accident - fuel handling accident (DBA-FHA) and
design basis accident - loss of coolant accident (DBA-LOCA). The
performance of CORE ALTERATIONS alone is neither a precursor to, nor
a condition during which these DBAs are postulated to occur. The
proposed changes only delete CORE ALTERATIONS as an applicable
condition for the affected Technical Specifications. All other
applicable MODES or specified conditions, including operations with
the potential for draining the reactor vessels (OPDRVs) and the
movement of irradiated fuel assemblies within the primary or
secondary containment, remain unchanged. Further, the limitations
placed on the handling of light loads are also unchanged. The
Technical Specifications (and the separate requirements imposed on
the handling of light loads) will thus continue to require that
systems or functions designed to mitigate design-basis/previously
evaluated accidents are OPERABLE during the relevant operating MODES
or conditions. On the basis of the above, it is concluded that the
requested amendment will not increase the probability or
consequences of any accident previously evaluated.
2. The proposed changes do not involve any modification to the
plant design or to the operation of plant systems (except to
determine when certain analyzed accident-mitigating systems or
features are required to be OPERABLE). The failure modes considered
for the proposed changes are the same as those previously
considered, therefore, it can be concluded that no new failure modes
will be created. On this basis, the proposed amendment will not
create the possibility of a new or different kind of accident from
any accident previously evaluated.
3. The changes being made to eliminate CORE ALTERATIONS as an
applicable condition for which certain LCOs must be met, do not
eliminate the requirements for operability of those systems or
features assumed to mitigate design-basis or analyzed accidents
during the applicable MODES when such systems or features are
assumed to be available for performing their mitigating function.
The safety margins assumed or established by the accident analyses
for those design-basis events (as described in the accident analyses
of the Clinton Power Station Updated Final Safety Analysis Report)
therefore remain unchanged. Further, the proposed changes do not
impact the controls imposed on the handling of light loads
(including unirradiated fuel assemblies) for ensuring that such
activities cannot result in an event that yields consequences more
severe than those calculated for the DBA-FHA. With respect to
reactivity concerns during refueling operations (MODE 5), all
systems or features required to be OPERABLE for precluding
inadvertent criticality and monitoring reactivity changes will
continue to be required OPERABLE as per the current Technical
Specification requirements. The deletion of CORE ALTERATIONS as an
applicable condition only applies to the noted systems which do not
contribute to precluding reactivity events. Based on the above, the
proposed changes do not involve a significant reduction in the
margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: Vespasian Warner Public
Library, 120 West Johnson Street, Clinton, Illinois 61727
Attorney for licensee: Leah Manning Stetener, Vice President,
General Counsel, and Corporate Secretary, 500 South 27th Street,
Decatur, Illinois 62525
NRC Project Director: Gail H. Marcus
Illinois Power Company and Soyland Power Cooperative, Inc., Docket
No. 50-461, Clinton Power Station, Unit No. 1, DeWitt County,
Illinois
Date of amendment request: May 1, 1996
Description of amendment request: The proposed amendment would
revise the Clinton Power Station (CPS) Operating License and 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 would be made to paragraph
2.D of the Operating License; TS Section 1.1, ``Definitions;'' TS
3.6.1.1, ``Primary Containment;'' TS 3.6.1.1, ``Primary Containment Air
Locks;'' TS 3.6.1.3, ``Primary Containment Isolation Valves (PCIVs);''
and TS Section 5.5, ``Programs and Manuals.''
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 implements new Option B of 10 CFR 50
Appendix J for performance-based primary containment leakage
testing. The proposed change does not involve a change to the plant
design or operation. As a result, the proposed change does not
affect any parameters or conditions that contribute to the
initiation of any accidents previously evaluated. Thus, the proposed
change cannot increase the probability of any accident previously
evaluated.
The proposed change potentially affects the leak-tight integrity
of the primary containment structure which is designed to mitigate
the consequences of a loss-of-coolant accident (LOCA) by limiting
the release of fission products contained in the post-LOCA primary
containment atmosphere. Functional integrity of the primary
containment must be maintained during and following the peak
transient pressures and temperatures that may result from a LOCA.
Because the proposed change does not alter the plant design,
including the primary containment and primary containment
penetrations, and because it only affects the frequency of measuring
Type A, B, and C leakage without changing the acceptance criteria
for the Type A, B, and C leakage rate tests, the proposed change
does not directly result in an increase in the primary containment
leakage. However, decreasing the test frequency can increase the
probability that an increase in primary containment leakage could go
undetected for an extended period of time. To minimize that
probability, test intervals will be established based on the
performance history of components being tested.
NUREG-1493, ``Performance-Based Containment Leak-Test Program,''
provides the technical basis for the NRC's rulemaking to revise
primary containment leakage testing requirements for nuclear power
reactors in 10 CFR 50, Appendix J. NUREG-1493 documents the NRC's
determination that the effect of primary containment leakage on
overall accident risk is minimal since risk is dominated by accident
sequences that result in failure of bypass of primary containment.
NUREG-1493 also documents that increasing the Type A leakage test
intervals would have a minimal impact on public risk, and that Type
B and C tests can identify the vast majority (greater than ninety
five percent) of all leakage paths. Therefore, performance-based
alternatives to current local leakage-testing requirements are
feasible without significant risk impacts.
[[Page 25709]]
Based on the above, IP has concluded that the proposed change
will not result in a significant increase in the probability or
consequences of any accident previously evaluated.
2. The proposed change does not involve a change to the plant
design or operation. As a result, the proposed change does not
affect any of the parameters or conditions that could contribute to
initiation of any accidents. This change involves the reduction of
Type A, B, and C test frequency. Except for the method of defining
the test frequency, the methods for performing the actual tests are
not changed. No new accident modes are created by extending the
testing intervals. No safety-related equipment or safety functions
are altered as a result of this change. Thus, extending the test
frequency has no influence on, nor does it contribute to the
possibility of a new or different kind of accident or malfunction
from those previously analyzed.
Based on the above, IP has concluded that the proposed change
will not create the possibility of a new or different kind of
accident not previously evaluated.
3. The request does not involve a significant reduction in a
margin to safety. The proposed change only affects the frequency of
the Type A, B, and C testing. Except for the method of defining the
test frequency, the methods for performing the actual tests are not
changed. However, the proposed change can increase the probability
that an increase in primary containment leakage could go undetected
for an extended period of time. NUREG-1493 has determined that under
several different accident scenarios, the increased risk of
radioactivity release from primary containment is negligible with
the implementation of these proposed changes.
The margin of safety that has the potential of being impacted by
the proposed change involves the offsite dose consequences of
postulated accidents which are directly related to the rate of
primary containment leakage. The primary containment isolation
system is designed to limit leakage to La, which is defined by
the CPS Technical Specifications to be 0.65% of primary containment
air weight per day at the calculated peak containment internal
pressure for the design basis loss of coolant accident (Pa).
The limitation on the rate of primary containment leakage is
designed to ensure that the total leakage volume will not exceed the
value assumed in the accident analyses at the peak accident pressure
(Pa). The margin of safety for the offsite dose consequences of
postulated accidents directly related to the primary containment
leakage rate is maintained by continuing to meet the 1.0 La
acceptance criteria. The La value is not being modified by this
proposed change.
Except for the method of defining the test frequency, no change
in the method of testing is being proposed. The Type A, B, and C
tests will continue to be done at full pressure (Pa) or
greater. Other programs are in place to ensure that proper
maintenance and repairs are performed during the service life of the
primary containment and systems and components penetrating the
primary containment.
As a result, IP has concluded that the proposed change will not
result in a significant reduction in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: Vespasian Warner Public
Library, 120 West Johnson Street, Clinton, Illinois 61727
Attorney for licensee: Leah Manning Stetener, Vice President,
General Counsel, and Corporate Secretary, 500 South 27th Street,
Decatur, Illinois 62525
NRC Project Director: Gail H. Marcus
Power Authority of the State of New York, Docket No. 50-333, James
A. FitzPatrick Nuclear Power Plant, Oswego County, New York
Date of amendment request: January 25, 1996
Description of amendment request: The amendment proposes to extend
instrumentation and miscellaneous surveillance test intervals (STI) to
support 24-month operating cycles. Additionally, this application
proposes: (1) to revise the Trip Level Settings for Emergency Bus Loss
of Voltage and Degraded Voltage Instrumentation, (2) to revise the
Reactor Protection System (RPS) Normal Supply Electrical Protection
Assembly (EPA) Undervoltage Trip Setpoint, and (3) to make editorial
revisions, clarification and Bases changes.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
Operation of the FitzPatrick plant in accordance with the
proposed Amendment would not involve a significant hazards
consideration as defined in 10 CFR 50.92, since it would not:
1. involve a significant increase in the probability or
consequences of an accident previously evaluated.
The proposed STI changes evaluated in Section IV.A do not
involve any physical changes to the plant, do not alter the way
these systems function, and will not degrade the performance of the
plant safety systems. Proposed instrument setpoint changes ensure
that plant safety limits are not exceeded due to instrument drift
predicted for the longer calibration interval. The type of testing
and the corrective actions required if the subject surveillances
fail remains the same. The proposed changes do not adversely affect
the reliability of these systems or affect the ability of the
systems to meet their design objectives. A historical review of
surveillance test results supports these conclusions.
The Trip Level Setpoint changes evaluated in Section IV.B ensure
that the related systems perform as assumed in the transient and
accident analysis by ensuring that plant safety limits are not
exceeded due to instrument drift predicted for the longer
calibration interval. The changes do not alter the system function,
and will not degrade the performance of plant safety systems. The
proposed Trip Level Setting changes do not adversely affect the
reliability of these systems or adversely affect the ability of
these systems to meet their design objectives.
The editorial, clarification and Bases changes evaluated in
Section IV.C propose enhancements that clarify the Technical
Specifications requirements and are editorial in nature. These
changes do not alter any Technical Specification requirement, do not
involve physical changes to the plant, or alter any operational
setpoints. There are no safety implications in these proposed
changes.
2. create the possibility of a new or different kind of accident
from any accident previously evaluated.
The proposed STI changes evaluated in Section IV.A do not modify
the design or operation of the plant, therefore, no new failure
modes are introduced. Proposed instrument setpoint changes ensure
that plant safety limits are not exceeded due to instrument drift
resulting from the longer calibration interval. No changes are
proposed to the type and method of testing performed, only to the
length of the surveillance test interval. Past equipment performance
and on-line testing indicate that longer test intervals will not
degrade these systems. A historical review of surveillance test
results supports these conclusions.
The Trip Level Setpoint changes evaluated in Section IV.B ensure
that the related systems perform as assumed in the transient and
accident analysis by ensuring that plant safety limits are not
exceeded due to instrument drift predicted for the longer
calibration interval. The changes do not alter the system function,
introduce any new failure modes, and will not degrade the
performance of plant safety systems. The proposed Trip Level Setting
changes do not adversely affect the reliability of these systems or
adversely affect the ability of these systems to meet their design
objectives.
The editorial, clarification and Bases changes evaluated in
Section IV.C propose enhancements that clarify the Technical
Specifications requirements and are editorial in nature. These
changes do not alter any Technical Specification requirement, do not
involve physical changes to the plant, or alter any operational
setpoints. There are no safety implications in these proposed
changes.
3. involve a significant reduction in a margin of safety.
Although the proposed STI changes evaluated in Section IV.A will
result in an increase in the interval between surveillance tests,
the impact on system reliability is minimal. This is based on more
frequent on-line testing and the redundant design of the evaluated
systems. A review of past surveillance history has shown no evidence
[[Page 25710]]
of failures which would significantly impact the reliability of
these systems. Operation of the plant remains unchanged by these
proposed STI extensions. The assumptions in the Plant Licensing
Basis are not adversely impacted. Therefore, the proposed changes do
not result in a significant reduction in the margin of safety.
The Trip Level Setpoint changes evaluated in Section IV.B ensure
that the related systems perform as assumed in the transient and
accident analysis by ensuring that plant safety limits are not
exceeded due to instrument drift predicted for the longer
calibration interval. The changes do not alter the system function,
introduce any new failure modes, and will not degrade the
performance of plant safety systems. The proposed Trip Level Setting
changes do not adversely affect the reliability of these systems or
adversely affect the ability of these systems to meet their design
objectives.
The editorial, clarification and Bases changes evaluated in
Section IV.C propose enhancements that clarify the Technical
Specifications requirements and are editorial in nature. These
changes do not alter any Technical Specification requirement, do not
involve physical changes to the plant, or alter any operational
setpoints. There are no safety implications in these proposed
changes.
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: Reference and Documents
Department, Penfield Library, State University of New York, Oswego, New
York 13126.
Attorney for licensee: Mr. Charles M. Pratt, 1633 Broadway, New
York, New York 10019.
NRC Project Director: Susan Frant Shankman, Acting
Power Authority of the State of New York, Docket No. 50-333, James
A. FitzPatrick Nuclear Power Plant, Oswego County, New York
Date of amendment request: April 24, 1996
Description of amendment request: This amendment proposes to
relocate Technical Specification (TS) 3.11.B/4.11.B ``Crescent Area
Ventilation'' and associated Bases from the TS to an Authority
controlled procedure.
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:
Operation of the FitzPatrick plant in accordance with the
proposed Amendment will not involve a significant hazards
consideration as defined in 10 CFR 50.92, based on the following:
(1) These changes do not involve a significant increase in the
probability or consequences of an accident previously evaluated
because:
No modifications, no changes to operating procedure
requirements, and no reduction in equipment reliability are being
made as a result of these changes. Operating limitations will
continue to be imposed, and required surveillance will continue to
be performed in accordance with regulations, and written procedures
and instructions that are auditable by the [Nuclear Regulatory
Commission] NRC. Crescent Area Ventilation operability and testing
requirements will continue to be an integral part of FitzPatrick
plant operation.
Although future changes to the Crescent Area Ventilation system
will no longer be controlled by 10 CFR 50.90, proposed changes will
be evaluated under 10 CFR 50.59 and plant procedures. Programmatic
controls will continue to assure that Crescent Area Ventilation
system changes will not adversely affect [Emergency Core Cooling
System] ECCS or [Reactor Core Isolation Cooling] RCIC system
operability. As such, there is no significant increase in the
probability or consequences of an accident previously evaluated.
(2) These changes do not create the possibility of a new or
different type of accident previously evaluated because:
No modifications, no changes to operating procedure
requirements, and no reduction in equipment reliability are being
made as a result of these changes. Compliance with Crescent Area
Ventilation system operability and surveillance requirements will be
assured by maintaining them in an Authority controlled procedure.
Changes to the Crescent Area Ventilation system will be subject to
the requirements of 10 CFR 50.59. Therefore, the proposed changes do
not introduce any failure mechanism of a different type than those
previously evaluated since there are no changes being made to the
facility and do not create the possibility of a new or different
type of accident previously evaluated.
(3) The proposed amendment does not involve a reduction in a
margin of safety because:
The Crescent Area Ventilation system supports Core Spray, [Low
Pressure Coolant Injection] LPCI mode of [Residual Heat Removal]
RHR, containment cooling mode of RHR, [High Pressure Coolant
Injection] HPCI, and RCIC operability, and Crescent Area Ventilation
system inoperability does affect these systems. As a result, the
requirement for Crescent Area Ventilation to be operable for these
systems to be considered operable is implicit in TS Sections 3.5.A,
3.5.B, 3.5.C, 3.5.E, and the definition of OPERABLE contained in TS
Section 1.0.J. Therefore, the proposed changes do not involve a
significant reduction in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: Reference and Documents
Department, Penfield Library, State University of New York, Oswego, New
York 13126.
Attorney for licensee: Mr. Charles M. Pratt, 1633 Broadway, New
York, New York 10019.
NRC Project Director: Susan Frant Shankman, Acting
Public Service Electric & Gas Company, Docket No. 50-311, Salem
Nuclear Generating Station, Unit No. 2, Salem County, New Jersey
Date of amendment request: May 7, 1996
Description of amendment request: The proposed amendment involves a
one-time change to Technical Specification (TS) 3/4.7.6, ``Control Room
Emergency Air Conditioning System.'' The change would permit refueling
of Salem, Unit 2, with the Control Room Emergency Air Conditioning
System (CREACS) inoperable in Modes 5 and 6. The change will expire
after the completion of the Control Room and CREACS upgrade, which is
currently in progress, and the restart and entry into Mode 4 of Unit 2
from the current outage.
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 CREACS is not an accident initiator. CREACS functions post-
accident to provide cooling for Control Room equipment and
habitability for operations personnel. Therefore, CREACS has no
influence on the probability of any of the previously evaluated
accidents or the other events evaluated as listed below.
Event
Fuel Handling Accident (Salem)
Waste Gas or Volume Control Tank Failures
Uncontrolled Boron Dilution
Loss of Offsite Power
Fuel Handling Accident (Hope Creek)
Liquid and Gaseous Waste Releases (Hope Creek)
Loss of Coolant Accident (LOCA) (Hope Creek)
Chemical Storage
Barge Collision
Control Room Internal and External Fire
Loss of Spent Fuel Pool Cooling
Loss of Decay Heat Removal
The Control Area Air Conditioning System (CAACS) and other
measures will be
[[Page 25711]]
available to maintain Control Room Envelope (CRE) ambient
temperatures and habitability.
The proposed one-time change does not impact the consequences of
an accident previously evaluated based on the following discussions.
The fuel has decayed to such low levels for more than six months
that doses associated with the fuel handling accident are well
within the limits of GDC [General Design Criteria] 19. There is
insufficient activity remaining in either gaseous waste storage or
liquid waste storage to force a Control Room evacuation. In the
event of a Loss of Offsite Power (LOOP), uncontrolled boron dilution
event, loss of spent fuel pool cooling or loss of decay heat
removal, CREACS is not required in Modes 5 or 6 to mitigate the
consequences of this event and CRE habitability will be maintained.
For a Hope Creek fuel handling accident, gaseous radwaste
release of LOCA, dose to Salem Control Room personnel will not
exceed GDC 19 limits. PSE&G [Public Service Electric & Gas] will
maintain the CAACS [Control Area Air Conditioning System] outside
air intakes either isolated or capable of being isolated in the
event of a Hope Creek LOCA. The Hope Creek Event Classification
Guide (ECG) requires notification of the Salem Control Room in the
event of an emergency that has the potential to result in a
radioactive release. The Salem Control Room will isolate the outside
air intakes if isolation has not already been accomplished.
For the other events evaluated, the need for evacuation is not
considered credible for any event with the exception of an internal
or external fire. However, the possibility of evacuation of the CRE
in the event of an internal or external fire would be no different
whether or not CREACS is operating. In the event of an internal
fire, CAACS will remain in operation to provide purging of the CRE.
For the case of a possible external fire, the need for evacuation is
not considered credible because of the short duration of the CREACS
outage and improbability of the factors which are necessary to
require an evacuation of the Control Room (i.e. wind direction, wind
speed, amount of smoke). If an external fire is detected, operator
action will be taken to isolate the CRE from outside air while CAACS
remains available. In the unlikely event that the Control Room would
become uninhabitable due to smoke in the atmosphere, evacuation
procedures would be followed as in the case of the internal fire.
The one chemical storage type event which might impact the
Control Room, rupture of an ammonium hydroxide tanker, is precluded
by administrative controls such that no ammonium hydroxide tanker
deliveries will be allowed during the system upgrade period.
The CAACS will maintain the current design function and TS Bases
requirements of the CREACS that the ambient air temperature does not
exceed the allowable temperature for continuous duty rating for
equipment and instrumentation cooled by the system for the combined
CRE. The CAACS will be maintained functional while modification to
the CREACS is ongoing to provide cooling during normal operation and
under postulated accident conditions. Should the temperature in the
CRE exceed allowable levels (85 Degrees F), administrative controls
will be in place to require restoration of the temperature to within
acceptable levels using CAACS, and prevent any Core Alteration
activities or positive reactivity changes until the temperature is
restored to acceptable levels.
Therefore, the proposed one-time TS change does not involve a
significant increase in the probability or consequences of an
accident previously evaluated.
2. The proposed change does not create the possibility of a new
or different kind of accident from any accident previously
evaluated.
The CREACS is not an accident initiator. CREACS functions post-
accident to provide cooling for Control Room equipment and
habitability for operations personnel. Therefore, CREACS
inoperability during Modes 5 and 6 will not result in the creation
of a new or different kind of accident from any accident previously
evaluated. All pertinent accidents have been assessed and no other
scenarios dealing with fuel movement, or the need for an operable
CREACS in Mode 5 or 6, have been deemed credible.
Therefore, the proposed one-time 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 one-time change does not significantly reduce the
margin of safety as defined in the Bases for the TS because (1)
there is no credible event as analyzed in Salem UFSAR [updated final
safety analysis report] Chapter 15 which can cause an unacceptable
environment in the CRE since the fuel has been decaying for at least
six months, (2) fuel movement inside the Fuel Handling Building
(FHB) is restricted in accordance with plant TS unless FHB
ventilation is operable, (3) dose to Salem control room personnel
from a potential Hope Creek fuel handling accident, gaseous radwaste
release or Loss of Coolant Accident will not exceed GDC 19 limits
(4) the one event which might impact the Control Room, rupture of an
ammonium hydroxide tanker, is precluded by administrative controls
such that no ammonium hydroxide tanker deliveries will be allowed
during the CREACS upgrade period, and (5) in the unlikely event that
Control Room evacuation is required, there is no impact on operator
ability to mitigate the consequences of an accident in the current
plant configuration.
Therefore, the proposed one-time TS 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: Salem Free Public library, 112
West Broadway, Salem, New Jersey 08079
Attorney for licensee: Mark J. Wetterhahn, Esquire, Winston and
Strawn, 1400 L Street, NW, Washington, DC 20005-3502
NRC Project Director: John F. Stolz
Southern Nuclear Operating Company, Inc., Docket No. 50-364, Joseph
M. Farley Nuclear Plant, Unit 2, Houston County, Alabama
Date of amendment request: March 29, 1996
Description of amendment request: The proposed amendment would
revise Technical Specification 3/4.4.6 ``Steam Generators'' and its
associated Bases. Specifically, the steam generator repair limit would
be modified to clarify that the appropriate method for determining
serviceability for tubes with outside diameter stress corrosion
cracking at the tube support plate is by a methodology that more
reliably assesses structural integrity. This amendment request is in
accordance with NRC's Generic Letter 95-05, ``Voltage-Based Repair
Criteria for Westinghouse Steam Generator Tubes Affected by Outside
Diameter Stress Corrosion Cracking.''
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:
Operation of Farley units in accordance with the proposed
license amendment does not involve a significant increase in the
probability or consequences of an accident previously evaluated.
Testing of model boiler specimens for free standing tubes at
room temperature conditions shows burst pressures as high as
approximately 5000 psi for indications of outer diameter stress
corrosion cracking with voltage measurements as high as 26.5 volts.
Burst testing performed on pulled tubes, including tubes pulled from
Farley Unit 2, with up to 7.5 volt indications show burst pressures
in excess of 5300 psi at room temperature. As stated earlier, tube
burst criteria are inherently satisfied during normal operating
conditions by the presence of the tube support plate. Furthermore,
correcting for the effects of temperature on material properties and
minimum strength levels (as the burst testing was done at room
temperature), tube burst capability significantly exceeds the R.G.
[Regulatory Guide] 1.121 criterion requiring the maintenance of a
margin of 1.43 times the steam line break pressure differential on
tube burst if through-wall cracks are present without regard to the
presence of the tube support plate. Considering the existing data
base, this criterion is satisfied with bobbin coil indications with
signal amplitudes over twice the 2.0 volt voltage-based repair
criteria, regardless of the indicated depth measurement. This
structural limit is based on a lower 95% confidence level limit of
the
[[Page 25712]]
data at operating temperatures. The 2.0 volt criterion provides a
conservative margin of safety to the structural limit considering
expected growth rates of outside diameter stress corrosion cracking
at Farley. Alternate crack morphologies can correspond to a voltage
so that a unique crack length is not defined by a burst pressure to
voltage correlation. However, relative to expected leakage during
normal operating conditions, no field leakage has been reported from
tubes with indications with a voltage level of under 7.7 volts for a
3/4 inch tube with a 10 volt correlation to 7/8 inch tubing (as
compared to the 2.0 volt proposed voltage-based tube repair limit).
Thus, the proposed amendment does not involve a significant increase
in the probability or consequences of an accident.
Relative to the expected leakage during accident condition
loadings, the accidents that are affected by primary-to-secondary
leakage and steam release to the environment are Loss of External
Electrical Load and/or Turbine Trip, Loss of All AC Power to Station
Auxiliaries, Major Secondary System Pipe Failure, Steam Generator
Tube Rupture, Reactor Coolant Pump Locked Rotor, and Rupture of a
Control Rod Drive Mechanism Housing. Of these, the Major Secondary
System Pipe Failure is the most limiting for Farley in considering
the potential for off-site doses. The offsite dose analyses for the
other events which model primary-to secondary leakage and steam
releases from the secondary side to the environment assume that the
secondary side remains intact. The steam generator tubes are not
subjected to a sustained increase in differential pressure, as is
the case following a steam line break event. This increase in
differential pressure is responsible for the postulated increase in
leakage and associated offsite doses following a steam line break
event. In addition, the steam line break event results in a bypass
of containment for steam generator leakage. Upon implementation of
the voltage-based repair criteria, it must be verified that the
expected distributions of cracking indications at the tube support
plate intersections are such that primary-to-secondary leakage would
result in site boundary dose within the current licensing basis.
Data indicate that a threshold voltage of 2.8 volts could result in
through-wall cracks long enough to leak at steam line break
conditions. Application of the proposed repair criteria requires
that the current distribution of a number of indications versus
voltage be obtained during the refueling outages. The current
voltage is then combined with the rate of change in voltage
measurement and a voltage measurement uncertainty to establish an
end of cycle voltage distribution and, thus, leak rate during steam
line break pressure differential. The leak rate during a steam line
break is further increased by a factor related to the probability of
detection of the flaws. If it is found that the potential steam line
break leakage for degraded intersections planned to be left in
service coupled with the reduced allowable specific activity levels
result in radiological consequences outside the current licensing
basis, then additional tubes will be plugged or repaired to reduce
steam line break leakage potential to within the acceptance limit.
Thus, the consequences of the most limiting design basis accident
are constrained to present licensing basis limits.
2) The proposed license amendment does not create the
possibility of a new or different kind of accident from any accident
previously evaluated.
Implementation of the proposed voltage-based tube support plate
elevation steam generator tube repair criteria does not introduce
any significant changes to the plant design basis. Use of the
criteria does not provide a mechanism that could result in an
accident outside of the region of the tube support plate elevations.
Neither a single or multiple tube rupture event would be expected in
a steam generator in which the repair criteria have been applied
during all plant conditions. The bobbin probe signal amplitude
repair criteria are established such that operational leakage or
excessive leakage during a postulated steam line break condition is
not anticipated. Southern Nuclear has previously implemented a
maximum leakage limit of 150 gpd per steam generator. The R.G. 1.121
criterion for establishing operational leakage limits that require
plant shutdown are based upon leak-before-break considerations to
detect a free span crack before potential tube rupture. The 150 gpd
limit provides for leakage detection and plant shutdown in the event
of the occurrence of an unexpected single crack resulting in leakage
that is associated with the longest permissible crack length. R.G.
1.121 acceptance criteria for establishing operating leakage limits
are based on leak-before-break considerations such that plant
shutdown is initiated if the leakage associated with the longest
permissible crack is exceeded. The longest permissible crack is the
length that provides a factor of safety of 1.43 against bursting at
steam line break pressure differential. A voltage amplitude of
approximately 9 volts for typical outside diameter stress corrosion
cracking corresponds to meeting this tube burst requirement at the
95% prediction interval on the burst correlation. Alternate crack
morphologies can correspond to a voltage so that a unique crack
length is not defined by the burst pressure versus voltage
correlation. Consequently, a typical burst pressure versus through-
wall crack length correlation is used below to define the ``longest
permissible crack'' for evaluating operating leakage limits.
The single through-wall crack lengths that result in tube burst
at 1.43 times steam line break pressure differential and steam line
break conditions are about 0.54 inch and 0.84 inch, respectively.
Normal leakage for these crack lengths would range from about 0.4
gallons per minute to 4.5 gallons per minute, respectively, while
lower 95% confidence level leak rates would range from about 0.06
gallons per minute to 0.6 gallons per minute, respectively.
An operating leak rate of 150 gpd per steam generator has been
implemented. This leakage limit provides for detection of 0.4 inch
long cracks at nominal leak rates and 0.6 inch long cracks at the
lower 95% confidence level leak rates. Thus, the 150 gpd limit
provides for plant shutdown prior to reaching critical crack lengths
for steam line break conditions at leak rates less than a lower 95%
confidence level and for three times normal operating pressure
differential at less than nominal leak rates.
Considering the above, the implementation of voltage-based
plugging criteria will not create the possibility of a new or
different kind of accident from any previously evaluated.
3) The proposed license amendment does not involve a significant
reduction in margin of safety.
The use of the voltage-based tube support plate elevation repair
criteria is demonstrated to maintain steam generator tube integrity
commensurate with the requirements of Generic Letter 95-05 and R.G.
1.121. R.G. 1.121 describes a method acceptable to the NRC staff for
meeting GDC [Generic Design Criteria] 2, 14, 15, 31, and 32 by
reducing the probability of the consequences of steam generator tube
rupture. This is accomplished by determining the limiting conditions
of degradation of steam generator tubing, as established by
inservice inspection, for which tubes with unacceptable cracking
should be removed from service. Upon implementation of the criteria,
even under the worst case conditions, the occurrence of outside
diameter stress corrosion cracking at the tube support plate
elevations is not expected to lead to a steam generator tube rupture
event during normal or faulted plant conditions. The most limiting
effect would be a possible increase in leakage during a steam line
break event. Excessive leakage during a steam line break event,
however, is precluded by verifying that, once the criteria are
applied, the expected end of cycle distribution of crack indications
at the tube support plate elevations would result in minimal, and
acceptable primary to secondary leakage during the event and, hence,
help to demonstrate radiological conditions are less than an
appropriate fraction of the 10 CFR [Part] 100 guideline.
The margin to burst for the tubes using the voltage-based repair
criteria is comparable to that currently provided by existing
Technical Specifications.
In addressing the combined effects of LOCA [loss-of-coolant
accident] + SSE [safe-shutdown earthquake] on the steam generator
component (as required by GDC 2), it has been determined that tube
collapse may occur in the steam generators at some plants. This is
the case as the tube support plates may become deformed as a result
of lateral loads at the wedge supports at the periphery of the plate
due to either the LOCA rarefaction wave and/or SSE loadings. Then,
the resulting pressure differential on the deformed tubes may cause
some of the tubes to collapse.
There are two issues associated with steam generator tube
collapse. First, the collapse of steam generator tubing reduces the
RCS [reactor coolant system] flow area through the tubes. The
reduction in flow area increases the resistance to flow of steam
from the core during a LOCA which, in turn, may potentially increase
Peak Clad Temperature (PCT). Second, there is a potential the
partial through-wall cracks in tubes could progress to through-wall
cracks during tube deformation or collapse or that short through-
[[Page 25713]]
wall indications would leak at significantly higher leak rates than
included in the leak rate assessments.
Consequently, a detailed leak-before-break analysis was
performed and it was concluded that the leak-before-break
methodology (as permitted by GDC 4) is applicable to the Farley
reactor coolant system primary loops and, thus, the probability of
breaks in the primary loop piping is sufficiently low that they need
not be considered in the structural design basis of the plant.
Excluding breaks in the RCS primary loops, the LOCA loads from the
large branch line breaks were analyzed at Farley and were found to
be of insufficient magnitude to result in steam generator tube
collapse or significant deformation.
Regardless of whether or not leak-before-break is applied to the
primary loop piping at Farley, any flow area reduction is expected
to be minimal (much less than 1%) and PCT margin is available to
account for this potential effect. Based on analyses' results, no
tubes near wedge locations are expected to collapse or deform to the
degree that secondary to primary in-leakage would be increased over
current expected levels. For all other steam generator tubes, the
possibility of secondary-to-primary leakage in the event of a LOCA +
SSE event is not significant. In actuality, the amount of secondary-
to-primary leakage in the event of a LOCA + SSE is expected to be
less than that originally allowed, i.e., 500 gpd per steam
generator. Furthermore, secondary-to-primary in-leakage would be
less than primary-to-secondary leakage for the same pressure
differential since the cracks would tend to tighten under a
secondary-to-primary pressure differential. Also, the presence of
the tube support plate is expected to reduce the amount of in-
leakage.
Addressing the R.G. 1.83 considerations, implementation of the
tube repair criteria is supplemented by 100% inspection requirements
at the tube support plate elevations having outside diameter stress
corrosion cracking indications, reduced operating leakage limits,
eddy current inspection guidelines to provide consistency in voltage
normalization, and rotating probe inspection requirements for the
larger indications left in service to characterize the principle
degradation mechanism as outside diameter stress corrosion cracking.
As noted previously, implementation of the tube support plate
elevation repair criteria will decrease the number of tubes that
must be taken out of service with tube plugs or repaired. The
installation of steam generator tube plugs or tube sleeves would
reduce the RCS flow margin, thus implementation of the voltage-based
repair criteria will maintain the margin of flow that would
otherwise be reduced through increased tube plugging or sleeving.
Considering the above, it is concluded that the proposed change
does not result in a significant reduction in margin with respect to
plant safety as defined in the Final Safety Analysis Report or any
bases of the plant Technical Specifications.
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: Houston-Love Memorial Library,
212 W. Burdeshaw Street, Post Office Box 1369, Dothan, Alabama 36302
Attorney for licensee: M. Stanford Blanton, Esq., Balch and
Bingham, Post Office Box 306, 1710 Sixth Avenue North, Birmingham,
Alabama 35201
NRC Project Director: Herbert N. Berkow
Southern Nuclear Operating Company, Inc., Docket No. 50-364, Joseph
M. Farley Nuclear Plant, Unit 2, Houston County, Alabama
Date of amendment request: April 22, 1996
Description of amendment request: The proposed amendment would
implement a new F* criterion based on maintaining existing safety
margins for steam generator tube structural integrity concurrent with
allowance for NDE (nondestructive examination) eddy current
uncertainty.
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 retains the existing margin in the F*
distance used to meet regulatory guidance of draft Regulatory Guide
1.121 and only changes the amount of assumed NDE eddy current
uncertainty based on the type of eddy current technology utilized in
the inspection. Therefore, there is no 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. WCAP
11306, Revision 2, ``Tubesheet Region Plugging Criterion for the
Alabama Power Company Farley Nuclear Station Unit 2 Steam
Generators,'' provides adequate basis for the F* distance proposed
of 1.54 plus allowance for eddy current uncertainty measurement.
Since the value of 1.54 inches was used in the analysis no new or
different kind of accident from any accident previously evaluated
will be created.
3. The proposed change does not involve a significant reduction
in a margin safety. Since the value of 1.54 inches already is used
in the steam generator tube pull out analysis, there is no
significant change to a margin 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: Houston-Love Memorial Library,
212 W. Burdeshaw Street, Post Office Box 1369, Dothan, Alabama 36302
Attorney for licensee: M. Stanford Blanton, Esq., Balch and
Bingham, Post Office Box 306, 1710 Sixth Avenue North, Birmingham,
Alabama 35201
NRC Project Director: Herbert N. Berkow
Union Electric Company, Docket No. 50-483, Callaway Plant, Unit 1,
Callaway County, MissouriDate of application request: February 23,
1996, as supplemented by letter dated April 24, 1996.
Description of amendment request: The amendment would add a
footnote in the license for Callaway Plant, Unit No. 1 to indicate that
Union Electric Company has entered into a merger agreement with CIPSCO
Incorporated which provides for Union Electric Company to become a
wholly-owned operating company of Ameren Corporation, a registered
public utility holding company under the Public Utility Holding Company
Act of 1935, as amended. After the merger, Union Electric Company would
continue to own and operate the Callaway Plant as an operating company
subsidiary of Ameren Corporation.
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 does not affect accident initiators or
assumptions. The radiological consequences of any accident
previously evaluated remain unchanged. The change is an
administrative change to reflect Union Electric's status as an
operating company subsidiary of Ameren.
2. The proposed change does not create the possibility of a new
or different kind of accident from any accident previously
evaluated.
The proposed change does not reduce the margin of safety assumed
in any accident analysis or affect any safety limits. The change is
administrative and reflects Union Electric's status as an operating
company subsidiary of Ameren.
3. The proposed change does not involve a significant reduction
in a margin of safety.
The proposed change does not reduce the margin of safety assumed
in any accident
[[Page 25714]]
analysis or affect any safety limits. The change is administrative
and reflects Union Electric's status as an operating company
subsidiary of Ameren.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: Callaway County Public
Library, 710 Court Street, Fulton, Missouri 65251
Attorney for licensee: Gerald Charnoff, Esq., Shaw, Pittman, Potts
& Trowbridge, 2300 N Street, N.W., Washington, D.C. 20037
NRC Project Director: William H. Bateman
Wisconsin Public Service Corporation, Docket No. 50-305, Kewaunee
Nuclear Power Plant, Kewaunee County, Wisconsin
Date of amendment request: April 30, 1996
Description of amendment request: The proposed amendment would
revise Kewaunee Nuclear Power Plant (KNPP) Technical Specification (TS)
3.1.b.1, its associated bases, and Figure TS 3.1-4 by extending the low
temperature overpressure protection (LTOP) requirements through the end
of operating cycle 33 or 33.41 effective full power years. The only
technical change being proposed is the substitution of end of life
fluence for the end of operating cycle 21 fluence.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
The proposed change was reviewed in accordance with the
provisions of 10 CFR 50.92 to show no significant hazards exist. The
proposed change will not:
1. Involve a significant increase in the probability or
consequences of an accident previously evaluated.
The LTOP setpoint and revised P/T [pressure/temperature] limits
reflected in proposed Figure TS 3.1-4 ensure that the Appendix G
pressure/temperature limits are not exceeded, and therefore, help
ensure that RCS integrity is maintained. The changes do not modify
the reactor coolant system pressure boundary, nor make any physical
changes to the facility design, material, construction standards, or
setpoints. The LTOP valve setpoint remains set at 500 psi. The LTOP
enabling temperature based on Figure TS 3.1-2 is 338 deg.F and is
more conservative than a value of 303 deg. Figure TS 3.1-4. The LTOP
enabling temperature based on Figure TS 3.1-2 remains unchanged by
this PA [proposed amendment]. The probability of a LTOP event
occurring is independent of the pressure-temperature limits for the
RCS pressure boundary. Therefore, the probability of a LTOP event
occurring remains unchanged.
The calculation of pressure temperature limits in accordance
with approved regulatory methods provides assurance that reactor
pressure vessel fracture toughness requirements are met and the
integrity of the RCS [reactor coolant system] pressure boundary is
maintained. Similar methodology was used in calculations to support
approved amendment 120 to the Kewaunee Technical Specifications
dated April 26, 1995. The material property basis, including
chemistry factor and initial reference temperature for the
unirradiated material (RTNDT), used for this PA is the same as
that used in the current TS. The only technical change being made in
this PA is the use of end of life fluence.
The use of predicted fluence values through the end of operating
cycle 33 is appropriately considered within the calculations in
accordance with standard industry methodology previously docketed
under WCAP 13227 and WCAP 14279. The neutron exposure projections
utilized for calculation of the reference temperature were
multiplied by a factor of 1.11 to adjust for biases observed between
cycle specific calculations and the results of neutron dosimetry for
the four surveillance capsules removed from the KNPP reactor. The
factor of 1.11 was derived by taking the average of the measured to
calculation (M/C) flux ratios obtained from the dosimetry results of
capsules V, R, P, and S removed from the KNPP reactor vessel. The
resulting effect of using predicted fluence values through the end
of cycle 33 instead of cycle 21 is to require the plant to evaluate
LTOP transients to more limiting requirements. The proposed PT
limits are shifted to a lower pressure and higher temperature, which
is more conservative.
The changes do not adversely affect the integrity of the RCS
such that its function in the control of radiological consequences
is affected. In addition, the changes do not affect any fission
barrier. The changes do not degrade or prevent the response of the
LTOP relief valve or other safety related system to accidents
described in Chapter 14 of the USAR. In addition, the changes do not
alter any assumption previously made in the radiological
consequences evaluations nor affect the mitigation of the
radiological consequences of an accident described in the USAR.
Therefore, the consequences of an accident previously evaluated in
the USAR will not be increased.
Thus, the operation of KNPP Unit 1 in accordance with the PA
does not involve a significant increase in the probability or
consequences of any accident previously evaluated.
2. Create the possibility of a new or different type of accident
from an accident previously evaluated.
The Appendix G pressure temperature limitations were prepared
using methods derived from the ASME Boiler and Pressure Vessel Code
and the criteria set forth in NRC Regulatory Standard Review Plan
5.3.2. The changes do not cause the initiation of any accident nor
create any new credible limiting failure for safety-related systems
and components. The changes do not result in any event previously
deemed incredible being made credible. As such, it does not create
the possibility of an accident different than any evaluated in the
USAR.
The changes do not have any effect on the ability of the safety-
related systems to perform their intended safety functions. The
changes do not create failure modes that could adversely impact
safety-related equipment. Therefore, it will not create the
possibility of a malfunction of equipment important to safety
different than previously evaluated in the USAR. Thus, the PA does
not create the possibility of a new or different kind of accident
from any accident previously evaluated.
The use of Paragraph (c)(2)(ii)(A) of 10 CFR 50.61, initial
reference temperature of -50 deg.F, and the fluence values through
EOC [end of cycle] 33 does not modify the reactor coolant system
pressure boundary, nor make any physical changes to the LTOP
setpoint or system design. Proposed Figure TS 3.1-4 was prepared in
accordance with regulatory requirements and requires evaluation of
LTOP events to more limiting requirements of neutron exposure
projections of 33.41 EFPY instead of 18.40 EFPY.
Therefore, the PA does not create the possibility of a new or
different type of accident from any accident previously evaluated.
3. Involve a significant reduction in the margin of safety.
The Appendix G pressure temperature limitations were prepared
using methods derived from the ASME Boiler and Pressure Vessel Code
and the criteria set forth in NRC Regulatory Standard Review Plan
5.3.2. These documents along with the calculational limitations
specified in 10 CFR 50.61 are an acceptable method for implementing
the requirements of 10 CFR 50 Appendices G and H. Inherent
conservatism in the P/T limits resulting from these documents
include:
a. An assumed defect in the reactor vessel wall with a depth
equal to 1/4 of the thickness of the vessel wall (1/4T) and a length
equal to 1-1/2 times the thickness of the vessel wall.
b. Assumed reference flaw oriented in both longitudinal and
circumferential directions and limiting material property. At KNPP,
the only weld in the core region is oriented in the circumferential
direction.
c. A factor of safety of 2 is applied to the membrane stress
intensity factor.
d. The limiting toughness is based upon a reference value
(KIR) which is a lower bound on the dynamic crack initiation or
arrest toughness.
e. A 2-sigma margin term is applied in determining the adjusted
reference temperature (ART) that is used to calculate the limiting
toughness.
Similar methodology was used in calculations to support approved
amendment 120 dated April 26, 1995. Beyond the conservatism
described above, WPSC
[[Page 25715]]
[Wisconsin Public Service Corporation] has incorporated the
following additional margin in preparing this PA:
a. The neutron exposure projections were multiplied by a factor
of 1.11 to adjust for biases observed between cycle specific
calculations and the results of neutron dosimetry for the four
surveillance capsules removed from the KNPP reactor. The factor of
1.11 was derived by taking the average of the measured to
calculation (M/C) flux ratios obtained from the dosimetry results of
capsules V, R, P, and S removed from the KNPP reactor vessel.
b. The calculated material-specific chemistry factor value is
191.27 and is based on KNPP surveillance capsule data from capsules
V, R, and P. Utilization of KNPP's most recent surveillance capsule
data from capsule S results in chemistry factor value of 190.6.
Consistent with calculation C10689, Revision 1 the value used for
chemistry factor in this PA remains 191.27, which is conservative.
c. The LTOP enabling temperature based on Figure TS 3.1-2 is
338 deg.F and is more conservative than a value of 303 deg.F which
is supported by proposed Figure TS 3.1-4. The LTOP enabling
temperature based on Figure TS 3.1-2 remains unchanged by this PA.
d. The reactor coolant pump starting restrictions of TS
3.1.a.1.c remain in place.
An alternative methodology to the safety margins required by
Appendix G to 10 CFR Part 50 has been developed by the ASME Working
Group on Operating Plant Criteria. This methodology is contained in
ASME Code Case N-514. The Code Case N-514 provides criteria to
determine pressure limits during LTOP events that avoid certain
unnecessary operational restrictions, provide adequate margins
against failure of the reactor pressure vessel, and reduce the
potential for unnecessary activation of the relief valve used for
LTOP. Specifically, the ASME Code Case N-514 allows determination of
the setpoint for LTOP events such that the maximum pressure in the
vessel would not exceed 110% of the P/T limits of the existing ASME
Appendix G; and redefines the enabling temperature as a coolant
temperature less than 200 deg.F or a reactor vessel metal
temperature less than RTNDT + 50 deg.F greater. Code Case N-
514, ``Low Temperature Overpressure Protection,'' has been approved
by the ASME Code Committee but not yet approved for use in
Regulatory Guide 1.147. The content of this code case has been
incorporated into Appendix G of Section XI of the ASME Code and
published in the 1993 Addenda to Section XI. It is expected that
when the NRC revises 10 CFR 50.55a, it will endorse the 1993 Addenda
and Appendix G of Section XI into the regulations. As stated above,
this PA utilizes Appendix G limits and an enabling temperature
corresponding to a reactor vessel metal temperature less than
RTNDT + 90 deg.F, which is more conservative than the
alternative methodology contained in Code Case N-514.
The revised calculations meet the NRC acceptance criteria for
the LTOP setpoint and system design as described in NRC Safety
Evaluation Report (SER) dated September 6, 1995 which concluded that
``the spectrum of postulated pressure transients would be
mitigated...such that the temperature pressure limits of Appendix G
to 10 CFR 50 are maintained.''
Utilization of methodology set forth in the ASME Boiler and
Pressure Vessel Code, NRC Regulatory Standard Review Plan 5.3.2, 10
CFR 50.61, and 10 CFR 50 Appendices G and H with the above
additional margins ensures that proper limits and safety factors are
maintained. Thus, the PA 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: University of Wisconsin,
Cofrin Library, 2420 Nicolet Drive, Green Bay, Wisconsin 54311-7001
Attorney for licensee: Bradley D. Jackson, Esq., Foley and Lardner,
P. O. Box 1497, Madison, Wisconsin 53701-1497
NRC Project Director: Gail H. Marcus
Wisconsin Public Service Corporation, Docket No. 50-305, Kewaunee
Nuclear Power Plant, Kewaunee County, Wisconsin
Date of amendment request: May 1, 1996
Description of amendment request: The proposed amendment would
revise Kewaunee Nuclear Power Plant (KNPP) Technical Specification (TS)
4.2.b, ``Steam Generator Tubes,'' its associated bases, and Figure TS
4.2-1 by redefining the pressure boundary for Westinghouse mechanical
hybrid expansion joint (HEJ) steam generator (SG) tube sleeves. The
proposed amendment supersedes in its entirety a previously submitted
proposed amendment dated October 6, 1995, which was published in the
Federal Register on November 8, 1995 (60 FR 56372).
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
This proposed change was reviewed in accordance with the
provisions of 10 CFR 50.92 to show no significant hazards exist.
1. Operation of the KNPP in accordance with the proposed license
amendment does not involve a significant increase in the probability
or consequences of an accident previously evaluated.
Mechanical testing shows inherent structural integrity of the
HEJ [hybrid expansion joint] upper joint such that the tube rupture
capability recommendations of RG [Regulatory Guide] 1.121 are met,
even for instances of 100-percent throughwall, 360 degree
degradation in the HRLT [hardroll lower transition] region.
Structural test results are documented in WCAPs-14157, -14157
Addendum 1, -14446 and -14641. Based on this test data, the
structural recommendations of RG 1.121 are satisfied when there is a
difference of at least 0.003 inch, between the maximum hardroll
diameter of the sleeve, and the diameter at the elevation of the PTI
[parent tube indication] center line; i.e. there is an interference
lip of 0.003 inch or more. The proposed pressure boundary will allow
PTIs located such that there is a minimum diameter change of 0.003
inch (not including an allowance for measurement uncertainty)
between the maximum point of the sleeve hardroll, and the diameter
at the elevation of the PTI peak amplitude to remain in service.
Based on the high degree of structural integrity of the HEJ upper
joint, it can be concluded that application of the revised pressure
boundary criteria will not result in an increased probability of an
accident previously evaluated.
Each sleeved tube with a PTI located in the HRLT such that there
is a change in diameter of 0.003 inch to 0.013 inch, will be
assigned a conservatively bounding primary-to-secondary SLB [steam
line break] leakage value of 0.025 gpm per indication. Indications
located such that there is a change in diameter of greater than
0.013 inch will not contribute to the SLB leakage. The total number
of indications remaining in service will be limited such that the
primary-to-secondary leakage during a postulated SLB will not exceed
a small fraction of the 10 CFR Part 100 guidelines. For KNPP this
has been calculated to be 34.0 gpm for the faulted loop. Therefore,
it can be concluded that application of the revised pressure
boundary criteria will not increase the consequences of an accident
previously evaluated.
2. The proposed license amendment request does not create the
possibility of a new or different kind of accident from any accident
previously evaluated.
Implementation of the revised pressure boundary will not
introduce a change to the design basis or operation of the plant.
Mechanical testing of degraded sleeve joints supports the
conclusions that the joint retains structural integrity (tube burst)
capability consistent with RG 1.121, and leakage integrity with
regards to a small fraction of the 10 CFR Part 100 guidelines. As
with the initial installation of the sleeves, implementation of the
relocated pressure boundary does not interact with other portions of
the reactor coolant system. Any hypothetical accident as a result of
potential PTIs is bounded by the existing tube rupture accident
analysis. Neither the sleeve design nor implementation of the
redefined pressure boundary affects any other component or location
of the tube outside of the immediate area repaired. Therefore
application of the revised pressure boundary criteria will not
create the possibility of a new or different kind of accident from
any accident previously evaluated.
3. The proposed license amendment does not involve a significant
reduction in the margin of safety.
[[Page 25716]]
The safety factors used in establishment of the HEJ sleeved tube
pressure boundary are consistent with the safety factors in the ASME
Boiler and Pressure Vessel Code used in SG [steam generator] design.
Based on the sleeve-to-tube geometry, it is unrealistic to consider
that application of the revised pressure boundary could result in
single tube leak rates exceeding the normal makeup capacity during
normal operating conditions. The pressure boundary developed in
WCAPs-14446 and -14641 have been developed using the methodology of
RG 1.121. The performance characteristics of the postulated degraded
parent tubes of HEJ sleeve/tube joints have been verified by testing
to retain structural integrity and preclude significant leakage
during normal and postulated accident conditions. Testing indicates
that postulated circumferentially separated tubes which the pressure
boundary [addresses] would not experience axial displacement during
either normal operation or SLB conditions. The existing offsite dose
evaluation performed for KNPP in support of the voltage based repair
criteria for axial ODSCC [outside diameter stress corrosion
cracking] at TSP [tube support plate] intersections established a
faulted loop primary to secondary leak rate of 34.0 gpm. Following
implementation of the criteria, postulated leakage from all sources
must not exceed 34.0 gpm in the faulted loop. Maintenance of this
limit will ensure that offsite doses would not exceed the currently
accepted limit of a small fraction of the 10 CFR Part 100
guidelines. The pressure boundary definition uses a conservatively
established ``per indication'' leak rate for estimation of SLB
leakage. This leak rate is applied to all indications left in
service within the HRLT, regardless of indications length and
throughwall extent. Application of the revised pressure boundary
criteria will not result in a significant reduction in the margin of
safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: University of Wisconsin,
Cofrin Library, 2420 Nicolet Drive, Green Bay, Wisconsin 54311-7001.
Attorney for licensee: Bradley D. Jackson, Esq., Foley and Lardner,
P. O. Box 1497, Madison, Wisconsin 53701-1497
NRC Project Director: Gail H. Marcus
Wolf Creek Nuclear Operating Corporation, Docket No. 50-482, Wolf
Creek Generating Station, Coffey County, Kansas
Date of amendment request: July 29, 1994, as superseded by letter
dated September 15, 1995, and supplements dated March 8, 1996, and
April 18, 1996
Description of amendment request: The proposed amendment revises TS
3/4.8.1 and its associated Bases to improve overall emergency diesel
generator reliability and availability.
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.
These proposed changes do not involve a change in the
operational limits or physical design of the emergency power system.
Emergency diesel generator operability and reliability will continue
to be assured while minimizing the number of required emergency
diesel generator starts. Also, emergency diesel generator
reliability will be enhanced by minimizing severe test conditions
which can lead to premature failures.
2. The proposed change does not create the possibility of a new
or different kind of accident from any accident previously
evaluated.
These proposed changes do not involve a change in the
operational limits or physical design of the emergency power system.
The performance capability of the emergency diesel generator will
not be affected. Emergency diesel generator reliability and
availability will be improved by the implementation of the proposed
changes. There is no actual impact on any accident analysis.
3. The proposed change does not involve a significant reduction
in a margin of safety.
These proposed change do not involve a change in the operational
limits or physical design of the emergency power system. The
performance capability of the emergency diesel generator will not be
affected. Emergency diesel generator reliability and availability
will be improved by the implementation of the proposed changes. No
margin of safety is 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
amendment request involves no significant hazards consideration.
Local Public Document Room locations: Emporia State University,
William Allen White Library, 1200 Commercial Street, Emporia, Kansas
66801 and Washburn University School of Law Library, Topeka, Kansas
66621
Attorney for licensee: Jay Silberg, Esq., Shaw, Pittman, Potts and
Trowbridge, 2300 N Street, N.W., Washington, D.C. 20037
NRC Project Director: William H. Bateman
Wolf Creek Nuclear Operating Corporation, Docket No. 50-482, Wolf
Creek Generating Station, Coffey County, Kansas
Date of amendment request: May 1, 1996
Description of amendment request: This license amendment request
proposes to revise Section 6.0 of the technical specifications to
reflect position title changes within the Wolf Creek Nuclear Operating
Corporation (WCNOC) organization.
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 does not involve a significant increase in
the probability of consequences of an accident previously evaluated.
These changes involve administrative changes to the WCNOC
organization and to the position qualification of plant personnel.
2. The proposed change does not create the possibility of a new
or different kind of accident from any accident previously
evaluated.
The proposed change does not create the possibility of a new or
different kind of accident from any accident previously evaluated.
This change is administrative in nature and does not involve a
change to the installed plant systems or the overall operating
philosophy of Wolf Creek Generating Station.
3. The proposed change does not involve a significant reduction
in a margin of safety.
The proposed change does not involve a significant reduction in
a margin of safety. This change does not involve any changes in
overall organizational commitments. A position title change alone
does not reduce the margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room locations: Emporia State University,
William Allen White Library, 1200 Commercial Street, Emporia, Kansas
66801 and Washburn University School of Law Library, Topeka, Kansas
66621
Attorney for licensee: Jay Silberg, Esq., Shaw, Pittman, Potts and
Trowbridge, 2300 N Street, N.W., Washington, D.C. 20037
NRC Project Director: William H. Bateman
[[Page 25717]]
Previously Published Notices Of Consideration Of Issuance Of
Amendments To Facility Operating Licenses, Proposed No Significant
Hazards Consideration Determination, And Opportunity For A Hearing
The following notices were previously published as separate
individual notices. The notice content was the same as above. They were
published as individual notices either because time did not allow the
Commission to wait for this biweekly notice or because the action
involved exigent circumstances. They are repeated here because the
biweekly notice lists all amendments issued or proposed to be issued
involving no significant hazards consideration.
For details, see the individual notice in the Federal Register on
the day and page cited. This notice does not extend the notice period
of the original notice.
Philadelphia Electric Company, Docket Nos. 50-352 and 50-353,
Limerick Generating Station, Units 1 and 2, Montgomery County,
Pennsylvania
Date of amendment request: April 25, 1996
Brief description of amendment request: The amendment relocates the
technical specification (TS) Traversing In-Core Probe System Limiting
Condition for Operation 3/4.3.7.7 and its Bases 3/4.3.7.7 to the
Technical Requirements Manual, and modifies Note (f) of TS Table
4.3.1.1-1.
Date of publication of individual notice in Federal Register: May
8, 1996 (61 FR 20840)
Expiration date of individual notice: June 7, 1996
Local Public Document Room location: Pottstown Public Library, 500
High Street, Pottstown, Pennsylvania 19464.
The Cleveland Electric Illuminating Company, Centerior Service
Company, Duquesne Light Company, Ohio Edison Company, OES Nuclear,
Inc., Pennsylvania Power Company, Toledo Edison Company, Docket No.
50-440, Perry Nuclear Power Plant, Unit No. 1, Lake County, Ohio
Date of application for amendment: April 26, 1996
Brief description of amendment request: The proposed amendment
would correct minor technical and administrative errors in the Improved
Technical Specifications prior to its implementation.
Date of individual notice in Federal Register: May 9, 1996 (61 FR
21213)
Expiration date of individual notice: June 10, 1996
Local Public Document Room location: Perry Public Library, 3753
Main Street, Perry, Ohio
Notice Of Issuance Of Amendments To Facility Operating Licenses
During the period since publication of the last biweekly notice,
the Commission has issued the following amendments. The Commission has
determined for each of these amendments that the application complies
with the standards and requirements of the Atomic Energy Act of 1954,
as amended (the Act), and the Commission's rules and regulations. The
Commission has made appropriate findings as required by the Act and the
Commission's rules and regulations in 10 CFR Chapter I, which are set
forth in the license amendment.
Notice of Consideration of Issuance of Amendment to Facility
Operating License, Proposed No Significant Hazards Consideration
Determination, and Opportunity for A Hearing in connection with these
actions was published in the Federal Register as indicated.
Unless otherwise indicated, the Commission has determined that
these amendments satisfy the criteria for categorical exclusion in
accordance with 10 CFR 51.22. Therefore, pursuant to 10 CFR 51.22(b),
no environmental impact statement or environmental assessment need be
prepared for these amendments. If the Commission has prepared an
environmental assessment under the special circumstances provision in
10 CFR 51.12(b) and has made a determination based on that assessment,
it is so indicated.
For further details with respect to the action see (1) the
applications for amendment, (2) the amendment, and (3) the Commission's
related letter, Safety Evaluation and/or Environmental Assessment as
indicated. All of these items are available for public inspection at
the Commission's Public Document Room, the Gelman Building, 2120 L
Street, NW., Washington, DC, and at the local public document rooms for
the particular facilities involved.
Arizona Public Service Company, et al., Docket Nos. STN 50-528, STN
50-529, and STN 50-530, Palo Verde Nuclear Generating Station,
Units 1, 2, and 3, Maricopa County, Arizona
Date of application for amendments: February 1, 1996
Brief description of amendments: These amendments revised (1)
Technical Specifications (TS) 3/4.1.1.1, 6.9.1.9, and 6.9.1.10 to
relocate the shutdown margin (reactor trip breakers open) to the Core
Operating Limits Report; (2) TS 3/4.3.2 (Tables 3.3-3 and 3.3-4) to
specify an additional restriction for the allowed low-pressurizer-
pressure trip setpoint when reducing reactor coolant (RCS) system
pressure in Mode 3; (3) TS Section 2.2.1 (Table 2.2-1) to make it
consistent with the footnote in TS Tables 3.3-3 and 3.3-4; and (4) TS
Sections 3/4.5.2 and 3/4.5.3 to require two emergency core cooling
system subsystems to be operable in Mode 3 whenever the RCS cold-leg
temperature is equal to or above 485 deg.F. The Table of Contents and
the Bases are also revised to reflect these changes.
Date of issuance: April 30, 1996
Effective date: April 30, 1996, to be implemented within 45 days of
issuance
Amendment Nos.: Unit 1 - 106; Unit 2 - 98; Unit 3 - 78
Facility Operating License Nos. NPF-41, NPF-51, and NPF-74: The
amendments revised the Technical Specifications.
Date of initial notice in Federal Register: March 27, 1996 (61 FR
13522) The Commission's related evaluation of the amendments is
contained in a Safety Evaluation dated April 30, 1996.No significant
hazards consideration comments received: No.
Local Public Document Room location: Phoenix Public Library, 1221
N. Central Avenue, Phoenix, Arizona 85004
Duquesne Light Company, et al., Docket Nos. 50-334 and 50-412,
Beaver Valley Power Station, Unit Nos. 1 and 2, Shippingport,
Pennsylvania
Date of application for amendments: December 27, 1995
Brief description of amendments: These amendments modify Tables
3.3-11 and 4.3-7 of Beaver Valley Power Station, Unit Nos. 1 and 2
(BVPS-1 and BVPS-2) Technical Specification 3.3.3.8 (Accident
Monitoring Instrumentation) such that only one valve position
indication system for the power-operated relief valves and safety
valves is required to be operable. Minor editorial changes to BVPS-1 TS
3.3.3.8 and its associated Action Statements are also being made. These
changes make the requirements of TS 3.3.3.8 consistent with the NRC's
Improved Standard Technical Specifications (NUREG-1431, Revision 1) and
with the guidance of Regulatory Guide 1.97, NUREG-0578, and NUREG-0737.
Date of issuance: May 1, 1996
Effective date: As of the date of issuance, to be implemented
within 60 days.
Amendment Nos.: 199 and 81
[[Page 25718]]
Facility Operating License Nos. DPR-66 and NPF-73: Amendments
revised the Technical Specifications.
Date of initial notice in Federal Register: January 31, 1996 (61 FR
3499) The Commission's related evaluation of the amendments is
contained in a Safety Evaluation dated May 1, 1996No significant
hazards consideration comments received: No.
Local Public Document Room location: B. F. Jones Memorial Library,
663 Franklin Avenue, Aliquippa, Pennsylvania 15001.
Duquesne Light Company, et al., Docket No. 50-334, Beaver Valley
Power Station, Unit No. 1, Shippingport, Pennsylvania
Date of application for amendment: February 12, 1996
Brief description of amendment: The amendment revises Technical
Specification (TS) 4.6.2.2.d to delete the reference to the specific
test acceptance criteria for the Containment Recirculation Spray Pumps
and replaces the specific test acceptance criteria with reference to
the requirements of the Inservice Testing (IST) Program. In addition,
the 18-month test frequency is replaced with the test frequency
requirements specified in the IST Program. The amendment also revises
the Bases for TS 4.6.2.2.d to describe this revision to TS 4.6.2.2.d.
Date of issuance: May 7, 1996
Effective date: As of the date of issuance, to be implemented
within 60 days.
Amendment No: 200
Facility Operating License No. DPR-66. Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: March 13, 1996 (61 FR
10393) The Commission's related evaluation of the amendment is
contained in a Safety Evaluation dated May 7, 1996 No significant
hazards consideration comments received: No
Local Public Document Room location: B. F. Jones Memorial Library,
663 Franklin Avenue, Aliquippa, PA 15001
Florida Power Corporation, et al., Docket No. 50-302, Crystal River
Nuclear Generating Plant, Unit No. 3, Citrus County, Florida
Date of application for amendments: March 21, 1996 as supplemented
April 8, 15, and 18, 1996.
Description of amendment request: The proposed amendment provides
for interim repair criteria for volumetric intergranular attack (IGA)
indications in the once-through-steam generators (OTSG). The interim
repair criteria is based on bobbin coil voltage response and motorized
rotating pancake coil probe dimensional measurements. The amendment
would be applicable for IGA indications within the region below the
first tube support plate and the secondary face of the lower tubesheet
(first span) of the OTSG and for one cycle only until Refuel 11.
Date of issuance: April 30, 1996
Effective date: April 30, 1996Amendment Nos. 154
Facility Operating License No. DPR-72: Amendment revised the
Technical Specifications. Public comments requested as to proposed no
significant hazards consideration: Yes (61 FR 13888). That notice
provided an opportunity to submit comments on the Commission's proposed
no significant hazards consideration determination. No comments have
been received. The notice also provided for an opportunity to request a
hearing by April 29, 1996, but indicated that if the Commission makes a
final no significant hazards consideration determination any such
hearing would take place after issuance of amendment. The Commission's
related evaluation of this amendment is contained in a Safety
Evaluation dated April 30, 1996
Local Public Document Room location: Coastal Region Library, 8619
W. Crystal Street, Crystal River, Florida 32629
Indiana Michigan Power Company, Docket No. 50-316, Donald C. Cook,
Nuclear Plant, Unit No. 2, Berrien County, Michigan
Date of application for amendment: March 12, 1996 (AEP:NRC:1248)
Brief description of amendment: The amendment removes the technical
specifications related to shutdown and control rod position indication
while in shutdown modes 3, 4, and 5.
Date of issuance: May 2, 1996
Effective date: May 2, 1996, with full implementation within 45
days
Amendment No.: 194
Facility Operating License No. DPR-74. Amendment revises the
Technical Specifications.
Date of initial notice in Federal Register: March 27, 1996 (61 FR
13527) The Commission's related evaluation of the amendment is
contained in a Safety Evaluation dated May 2, 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
Nebraska Public Power District, Docket No. 50-298, Cooper Nuclear
Station, Nemaha County, Nebraska
Date of amendment request: May 5, 1995 and July 14, 1995,
supplemented by letter dated March 5, 1996
Brief description of amendment: The amendment revised the Technical
Specifications to 1) verify that the redundant diesel generator is
operable upon the loss of one diesel generator, and implement
provisions to verify that the operable diesel generator does not have a
common cause failure; 2) incorporate provisions to allow a modified
start for the diesel generators; and 3) remove the requirement that the
reactor power level be reduced to 25% of rated power upon loss of both
diesel generator units or both incoming power sources (start-up and
emergency transformers). In addition, the period of time allowed for
continued reactor operation with both diesels inoperable was reduced
from 24 to two hours.
Date of issuance: April 29, 1996
Effective date: April 29, 1996
Amendment No.: 175
Facility Operating License No. DPR-46: Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: September 27, 1995 (60
FR 49939) The Commission's related evaluation of the amendment is
contained in a Safety Evaluation dated April 29, 1996.No significant
hazards consideration comments received: No.
Local Public Document Room location: Auburn Memorial Library, 1810
Courthouse Avenue, Auburn, NE 68305.
North Atlantic Energy Service Corporation, Docket No. 50-443,
Seabrook Station, Unit No. 1, Rockingham County, New Hampshire
Date of amendment request: September 22, 1995
Description of amendment request: The amendment changes the ACTION
specified in Table 3.3-3, Engineered Safety Features Actuation System
Instrumentation, from ACTION 18 to ACTION 15 for Functional Unit 8.b,
Automatic Switchover to Containment Sump - RWST Level Low-Low.
Date of issuance: May 7, 1996,
Effective date: As of the date of issuance, to be implemented
within 60 days.
Amendment No.: 47
Facility Operating License No. NPF-86. Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: December 6, 1995 (60 FR
62493) The Commission's related
[[Page 25719]]
evaluation of the amendment is contained in a Safety Evaluation dated
May 7, 1996. No significant hazards consideration comments received:
No.
Local Public Document Room location: Exeter Public Library,
Founders Park, Exeter, NH 03833.
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: May 26, 1995, as supplemented
October 20, 1995, and May 3, 1996.
Brief description of amendment: The amendment modifies Technical
Specification (TS) 3.8.1.2, ``Electrical Power Systems, Shutdown,'' TS
3.8.2.2, ``Electrical Power Systems, A.C. Distribution - Shutdown,''
and TS 3.8.2.4, ``Electrical Power Systems, D.C. Distribution -
Shutdown,'' to provide operational flexibility as well as consistency
between action statements and to eliminate certain surveillance
requirements that are not applicable in Mode 5 or 6.
The proposed changes relating to TS 3.8.1.1, ``Electrical Power
Systems, A.C. Sources, Operating,'' are not included in this amendment
since this portion of the TS change is still under review by the staff
and will be addressed at a later date.
Date of issuance: May 6, 1996
Effective date: As of the date of issuance, to be implemented
within 30 days.
Amendment No.: 197
Facility Operating License No. DPR-65. Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: December 6, 1995 (60 FR
62493) The October 20, 1995, letter formally withdrew the need for
exigent handling of the May 26, 1995, request and requested an
additional change to TS 3.8.2.4. The May 3, 1996, letter withdrew a
portion of the initial request which did not affect the initial
proposed no significant hazards consideration. The Commission's related
evaluation of the amendment is contained in a Safety Evaluation dated
May 6, 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.
Power Authority of the State of New York, Docket No. 50-286, Indian
Point Nuclear Generating Unit No. 3, Westchester County, New York
Date of application for amendment: March 14, 1996
Brief description of amendment: The amendment allows a one-time
extension of the intervals for the pressurizer safety valve setpoint
and snubber functional testing that is due in May 1996.
Date of issuance: May 3, 1996
Effective date: As of the date of issuance to be implemented within
30 days.
Amendment No.: 165
Facility Operating License No. DPR-26: Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: April 3, 1996, (61 FR
14835) The Commission's related evaluation of the amendment is
contained in a Safety Evaluation dated May 3, 1996.No significant
hazards consideration comments received: No
Local Public Document Room location: White Plains Public Library,
100 Martine Avenue, White Plains, New York 10610.
Public Service Electric & Gas Company, Docket Nos. 50-272 and 50-
311, Salem Nuclear Generating Station, Unit Nos. 1 and 2, Salem
County, New Jersey
Date of application for amendments: January 4, 1996
Brief description of amendments: The amendments change Technical
Specification 3/4.8.2.5, ``28-Volt D.C. Distribution - Operating.'' The
amendment for Unit 1 makes Unit 1 requirements similar to Unit 2 by
defining the specific battery chargers that are required for each train
and by restricting the use of the backup battery charger to 7 days. The
amendments for both units also require that the 28-Volt DC bus be
energized for that bus to be OPERABLE.
Date of issuance: April 29, 1996
Effective date: Both units, as of date of issuance, to be
implemented within 60 days.Amendment Nos. 182 and 163
Facility Operating License Nos. DPR-70 and DPR-75. The amendments
revised the Technical Specifications.
Date of initial notice in Federal Register: February 14, 1996 (61
FR 5818) The Commission's related evaluation of the amendments is
contained in a Safety Evaluation dated April 29, 1996No significant
hazards consideration comments received: No
Local Public Document Room location: Salem Free Public Library, 112
West Broadway, Salem, New Jersey 08079
Union Electric Company, Docket No. 50-483, Callaway Plant, Unit 1,
Callaway County, Missouri
Date of application for amendment: September 6, 1995, as
supplemented by letters dated January 30, 1996, March 27, 1996, and
April 2, 1996.
Brief description of amendment: The amendment revises TS 5.3.1 to
reflect a change in the maximum initial enrichment for reload fuel,
subject to the integral fuel burnable absorber (IFBA) requirements, and
a change in the maximum fuel enrichment not requiring IFBAs. The
amendment also changes the maximum reference kinfinity in TS
5.6.1.1 for fuel storage in Region 1 of the spent fuel pool and revises
TS Figure 3.9-1 to reflect a change to the maximum initial enrichment
for fuel stored in Region 2 of the spent fuel pool.
Date of issuance: April 30, 1996
Effective date: April 30, 1996, to be implemented within 30 days
from the date of issuance.
Amendment No.: 109
Facility Operating License No. NPF-30: The amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: November 8, 1995 (60 FR
56372). The January 30, 1996, March 27, 1996, and April 2, 1996,
supplemental letters provided additional clarifying information and did
not change the original no significant hazards consideration
determination. The Commission's related evaluation of the amendment is
contained in a Safety Evaluation dated April 30, 1996.No significant
hazards consideration comments received: No.
Local Public Document Room location: Callaway County Public
Library, 710 Court Street, Fulton, Missouri 65251.
Union Electric Company, Docket No. 50-483, Callaway Plant, Unit 1,
Callaway County, Missouri
Date of application for amendment: February 9, 1996
Brief description of amendment: The amendment revised Technical
Specification 5.3.1 to allow the use of ZIRLO clad fuel rods and ZIRLO
filler rods.
Date of issuance: April 30, 1996
Effective date: April 30, 1996, to be implemented within 30 days of
issuance.
Amendment No.: 110
Facility Operating License No. NPF-30: The amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: February 28, 1996 (61
FR 7558) The Commission's related
[[Page 25720]]
evaluation of the amendment is contained in a Safety Evaluation dated
April 30, 1996.No significant hazards consideration comments received:
No.
Local Public Document Room location: Callaway County Public
Library, 710 Court Street, Fulton, Missouri 65251.
Virginia Electric and Power Company, Docket Nos. 50-280 and 50-281,
Surry Power Station, Unit Nos. 1 and 2, Surry County, Virginia.
Date of application for amendments: January 30, 1996
Brief description of amendments: These amendments modify the
Technical Specifications requirements for the sampling of the reactor
coolant for dissolved oxygen chlorides and fluorides.
Date of issuance: 209 and 209
Effective date: April 29, 1996
Amendment Nos. 209 and 209
Facility Operating License Nos. DPR-32 and DPR-37: Amendments
revised the Technical Specifications.
Date of initial notice in Federal Register: March 27, 1996 (61 FR
13533) The Commission's related evaluation of the amendment is
contained in a Safety Evaluation dated April 29, 1996.No significant
hazards consideration comments received: No
Local Public Document Room location: Swem Library, College of
William and Mary, Williamsburg, Virginia 23185
Washington Public Power Supply System, Docket No. 50-397, Nuclear
Project No. 2, Benton County, Washington
Date of application for amendment: January 19, 1996, as
supplemented by letter dated March 19, 1996.
Brief description of amendment: The amendment modifies the
Technical Specifications for leak tests of containment isolation
valves. The amendment replaces the current specified surveillance
intervals for containment leak testing with new surveillance
requirements to conduct containment leak testing according to a
performance-based containment leak test program.
Date of issuance: May 8, 1996
Effective date: May 8, 1996, to be implemented within 30 days of
issuance.
Amendment No.: 144
Facility Operating License No. NPF-21: The amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: February 14, 1996 (61
FR 5820) The March 19, 1996, supplemental letter provided additional
clarifying information and did not change the initial no significant
hazards consideration determination. The Commission's related
evaluation of the amendment is contained in a Safety Evaluation dated
May 8, 1996.No significant hazards consideration comments received: No.
Local Public Document Room location: Richland Public Library, 955
Northgate Street, Richland, Washington 99352
Dated at Rockville, Maryland, this 15th day of May 1996.
For the Nuclear Regulatory Commission
Steven A. Varga,
Director, Division of Reactor Projects - I/II, Office of Nuclear
Reactor Regulation
[Doc. 96-12691 Filed 5-21-96; 8:45 am]
BILLING CODE 7590-01-F
