# Applications and Amendments to Facility Operating Licenses Involving No Significant Hazards Considerations

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URL: https://www.frixlaw.com/law-library/documents/fr%3AX96-20522

## Record

- **Collection:** Federal Register
- **Document type:** Notice
- **Published:** May 22, 1996
- **Citation:** 61 FR 25696

## Text

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 Cool

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Source: Frix Law Library, https://www.frixlaw.com/law-library/documents/fr%3AX96-20522. Public record. Not legal advice.