# 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%3A94-4562

## Record

- **Collection:** Federal Register
- **Document type:** Uncategorized Document
- **Published:** March 2, 1994

## Text

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 February 5, 1994, through February 17, 1994.
The last biweekly notice was published on February 16, 1994 (59 FR
7685).

Notice Of Consideration Of Issuance Of Amendments To Facility
Operating Licenses, Proposed No Significant Hazards Consideration
Determination, And Opportunity For A Hearing

The Commission has made a proposed determination that the following
amendment requests involve no significant hazards consideration. Under
the Commission's regulations in 10 CFR 50.92, this means that operation
of the facility in accordance with the proposed amendment would not (1)
involve a significant increase in the probability or consequences of an
accident previously evaluated; or (2) create the possibility of a new
or different kind of accident from any accident previously evaluated;
or (3) involve a significant reduction in a margin of safety. The basis
for this proposed determination for each amendment request is shown
below.
The Commission is seeking public comments on this proposed
determination. Any comments received within 30 days after the date of
publication of this notice will be considered in making any final
determination.
Normally, the Commission will not issue the amendment until the
expiration of the 30-day notice period. However, should circumstances
change during the notice period such that failure to act in a timely
way would result, for example, in derating or shutdown of the facility,
the Commission may issue the license amendment before the expiration of
the 30-day notice period, provided that its final determination is that
the amendment involves no significant hazards consideration. The final
determination will consider all public and State comments received
before action is taken. Should the Commission take this action, it will
publish in the Federal Register a notice of issuance and provide for
opportunity for a hearing after issuance. The Commission expects that
the need to take this action will occur very infrequently.
Written comments may be submitted by mail to the Rules Review and
Directives Branch, Division of Freedom of Information and Publications
Services, Office of Administration, U.S. Nuclear Regulatory Commission,
Washington, DC 20555, and should cite the publication date and page
number of this Federal Register notice. Written comments may also be
delivered to Room P-223, Phillips Building, 7920 Norfolk Avenue,
Bethesda, 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 20555.
The filing of requests for a hearing and petitions for leave to
intervene is discussed below.
By April 1, 1994, 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 20555 and at the local
public document room for the particular facility involved. If a request
for a hearing or petition for leave to intervene is filed by the above
date, the Commission or an Atomic Safety and Licensing Board,
designated by the Commission or by the Chairman of the Atomic Safety
and Licensing Board Panel, will rule on the request and/or petition;
and the Secretary or the designated Atomic Safety and Licensing Board
will issue a notice of a hearing or an appropriate order.
As required by 10 CFR 2.714, a petition for leave to intervene
shall set forth with particularity the interest of the petitioner in
the proceeding, and how that interest may be affected by the results of
the proceeding. The petition should specifically explain the reasons
why intervention should be permitted with particular reference to the
following factors: (1) the nature of the petitioner's right under the
Act to be made a party to the proceeding; (2) the nature and extent of
the petitioner's property, financial, or other interest in the
proceeding; and (3) the possible effect of any order which may be
entered in the proceeding on the petitioner's interest. The petition
should also identify the specific aspect(s) of the subject matter of
the proceeding as to which petitioner wishes to intervene. Any person
who has filed a petition for leave to intervene or who has been
admitted as a party may amend the petition without requesting leave of
the Board up to 15 days prior to the first prehearing conference
scheduled in the proceeding, but such an amended petition must satisfy
the specificity requirements described above.
Not later than 15 days prior to the first prehearing conference
scheduled in the proceeding, a petitioner shall file a supplement to
the petition to intervene which must include a list of the contentions
which are sought to be litigated in the matter. Each contention must
consist of a specific statement of the issue of law or fact to be
raised or controverted. In addition, the petitioner shall provide a
brief explanation of the bases of the contention and a concise
statement of the alleged facts or expert opinion which support the
contention and on which the petitioner intends to rely in proving the
contention at the hearing. The petitioner must also provide references
to those specific sources and documents of which the petitioner is
aware and on which the petitioner intends to rely to establish those
facts or expert opinion. Petitioner must provide sufficient information
to show that a genuine dispute exists with the applicant on a material
issue of law or fact. Contentions shall be limited to matters within
the scope of the amendment under consideration. The contention must be
one which, if proven, would entitle the petitioner to relief. A
petitioner who fails to file such a supplement which satisfies these
requirements with respect to at least one contention will not be
permitted to participate as a party.
Those permitted to intervene become parties to the proceeding,
subject to any limitations in the order granting leave to intervene,
and have the opportunity to participate fully in the conduct of the
hearing, including the opportunity to present evidence and cross-
examine witnesses.
If a hearing is requested, the Commission will make a final
determination on the issue of no significant hazards consideration. The
final determination will serve to decide when the hearing is held.
If the final determination is that the amendment request involves
no significant hazards consideration, the Commission may issue the
amendment and make it immediately effective, notwithstanding the
request for a hearing. Any hearing held would take place after issuance
of the amendment.
If the final determination is that the amendment request involves a
significant hazards consideration, any hearing held would take place
before the issuance of any amendment.
A request for a hearing or a petition for leave to intervene must
be filed with the Secretary of the Commission, U.S. Nuclear Regulatory
Commission, Washington, DC 20555, 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 20555, by the
above date. Where petitions are filed during the last 10 days of the
notice period, it is requested that the petitioner promptly so inform
the Commission by a toll-free telephone call to Western Union at 1-
(800) 248-5100 (in Missouri 1-(800) 342-6700). The Western Union
operator should be given Datagram Identification Number N1023 and the
following message addressed to (Project Director): petitioner's name
and telephone number, date petition was mailed, plant name, and
publication date and page number of this Federal Register notice. A
copy of the petition should also be sent to the Office of the General
Counsel, U.S. Nuclear Regulatory Commission, Washington, DC 20555, and
to the attorney for the licensee.
Nontimely filings of petitions for leave to intervene, amended
petitions, supplemental petitions and/or requests for a hearing will
not be entertained absent a determination by the Commission, the
presiding officer or the Atomic Safety and Licensing Board that the
petition and/or request should be granted based upon a balancing of
factors specified in 10 CFR 2.714(a)(1)(i)-(v) and 2.714(d).
For further details with respect to this action, see the
application for amendment which is available for public inspection at
the Commission's Public Document Room, the Gelman Building, 2120 L
Street, NW., Washington, DC 20555, and at the local public document
room for the particular facility involved.

Arizona Public Service Company, et al., Docket Nos. STN 50-528, STN
50-529, and STN 50-530, Palo Verde Nuclear Generating Station, Unit
Nos. 1, 2, and 3, Maricopa County, Arizona

Date of amendment requests: January 13, 1994
Description of amendment requests: Request for NRC consent to the
indirect transfer of control of El Paso Electric Company's interest in
Operating License Nos. NPF-41, NPF-51, NPF-74 and to amend Operating
License Nos. NPF-51 and NPF-74 to delete provisions for El Paso
Electric Company's sale-leaseback arrangements.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensees have
provided their analysis about the issue of no significant hazards
consideration, which is presented below:
Standard 1 -- Involve a significant increase in the probability
or consequences of an accident previously evaluated.
This amendment request does not involve a significant increase
in the probability or consequences of an accident previously
evaluated because the proposed change is administrative in nature.
The proposed change deletes Sections 2.B.(7)(a) and (b) of License
No. NPF-51, and Sections 2.B.(6)(a) and (b) of License No. No. NPF-
74. These section describe the structure of the financing of El
Paso's interest in Palo Verde, specifically authorizing sale and
leaseback transactions. The proposed change does not affect the
assumptions used in the accident analyses, nor does the proposed
change result in changes to the physical configuration of the
facility, design parameters, technical specifications, or operation
and maintenance of the facility. Therefore, this amendment request
does not involve a significant increase in the probability or
consequences of an accident previously evaluated.
Standard 2 -- Create the possibility of a new or different kind
of accident from any accident previously analyzed.
This amendment request does not create the possibility of a new
or different kind of accident from any accident previously analyzed
because the proposed change is administrative in nature. The
proposed change deletes Sections 2.B.(7)(a) and (b) of License No.
NPF-51, and Sections 2.B.(6)(a) and (b) of License No. NPF-74. These
sections describe the structure of the financing of El Paso's
interest in Palo Verde, specifically authorizing sale and leaseback
transactions. The proposed change does not involve modifications to
any of the existing equipment nor does the change affect the
operation and maintenance of the facility. Therefore, this amendment
request does not create the possibility of a new or different kind
of accident not previously analyzed.
Standard 3 -- Involve a significant reduction in a margin of
safety.
This amendment request does not involve a significant reduction
in a margin of safety because it is administrative in nature. The
proposed change deletes Sections 2.B.(7)(a) and (b) of License No.
NPF-51, and Sections 2.B.(6)(a) and (b) of License No. NPF-74. These
sections describe the structure of the financing of El Paso's
interest in Palo Verde, specifically authorizing sale and leaseback
transactions. The proposed change does not involve changes to any
existing plant equipment or accident analyses that provide for or
establish margins of safety. There are no changes to the operation
or maintenance of the facility and the existing margins of safety
are not changed by the proposed change. Therefore, this amendment
request does not involve a signigicant reduction in a margin of
safety.
The NRC staff has reviewed the licensees' analysis and, based on
that review, it appears that the proposed license amendment reflects
only a change in the structure of the financing of El Paso's interest
in Palo Verde and the three standards of 50.92(c) are satisfied.
Therefore, the NRC staff proposes to determine that the amendment
requests involve no significant hazards consideration.
Local Public Document Room location: Phoenix Public Library, 12
East McDowell Road, Phoenix, Arizona 85004 Attorney for licensees:
Nancy C. Loftin, Esq., Corporate Secretary and Counsel, Arizona Public
Service Company, P.O. Box 53999, Mail Station 9068, Phoenix, Arizona
85072-3999
NRC Project Director: Theodore R. Quay

Carolina Power & Light Company, et al., Docket No. 50-400, Shearon
Harris Nuclear Power Plant, Unit 1, Wake and Chatham Counties,
North Carolina

Date of amendment request: February 4, 1994
Description of amendment request: The proposed amendment would
revise Technical Specification 3/4.6.4, Containment Systems Combustible
Gas Control, by eliminating the 12-hour channel check surveillance
requirement for the containment hydrogen monitoring system in
conformance with the new Standard Technical Specifications, NUREG-1431.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. The proposed amendment does not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
Final Safety Analysis Report [FSAR] section 6.2.5.2.3 states
that the Hydrogen Analyzer is only required to be functioning
(continuously indicating and recording hydrogen concentration)
within 30 minutes of safety injection initiation. The performance of
an analog operational test every 31 days and a channel calibration
test every 92 days verifies this operability. Based on this, the
monitors will be fully capable of performing their intended design
function following a safety injection initiation. Therefore, the
elimination of the 12-hour channel check would not increase the
probability or consequences of an accident previously evaluated.
2. The proposed amendment does not create the possibility of a
new or different kind of accident from any accident previously
evaluated.
The Hydrogen Monitors perform an ``indication'' function only,
[sic] to help ensure that hydrogen concentrations within containment
are maintained below flammable limits during a post-LOCA [loss-of-
coolant accident] condition. The proposed changes do not involve any
modifications or additions to plant equipment and the design and
operation of the plant will not be affected. Therefore, the
elimination of the 12-hour channel check does not affect any
parameters which relate to the margin of safety as defined in the
Technical Specifications or the FSAR. Therefore, the proposed
changes do not involve a significant reduction in a margin of
safety.
3. The proposed amendment does not involve a significant
reduction in the margin of safety.
The proposed elimination of the 12-hour channel check does not
affect any parameters which relate to the margin of safety as
defined in the Technical Specifications or the FSAR. Therefore, the
proposed changes do not involve a significant reduction in a margin
of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: Cameron Village Regional
Library, 1930 Clark Avenue, Raleigh, North Carolina 27605.
Attorney for licensee: R. E. Jones, General Counsel, Carolina Power
& Light Company, Post Office Box 1551, Raleigh, North Carolina 27602
NRC Project Director: S. Singh Bajwa

Commonwealth Edison Company, Docket Nos. 50-237 and 50-249, Dresden
Nuclear Power Station, Units 2 and 3, Grundy County, Illinois

Date of amendment request: March 26, 1993
Description of amendment request: The proposed amendment would
modify trip level settings for the Isolation Condenser and High
Pressure Core Injection (HPCI) System Steam lines to more conservative
values. In addition, the proposed amendment would revise the Emergency
Core Cooling System Low-Low Water Level initiation trip level setting
tolerance.
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:
Involve a significant increase in the probability or
consequences of an accident previously evaluated because:
HPCI Steamline High Flow Isolation Trip Level Setting
The purpose of the HPCI leak detection systems are to detect
breaks in the system piping. Normal steam flows within the system
can fluctuate in excess of 250% rated flow and exceed 500% rated
steam flow after experiencing a break. During the original licensing
of the plant, it was analytically determined by GE that three times
maximum steam flow (300%) is the optimum setpoint for the isolation
of HPCI. A 300% steam flow setpoint ensures that spurious trips are
avoided and that breaks in the piping are identified. Because the
HPCI High Steamline Flow Isolation setpoint is not assumed as an
accident precursor, the probability of any previously evaluated
accident is not increased by the changed setpoint.
The proposed changes to the setpoint allow a more accurate and
conservative value for 300% steam flow. The proposed change in
conjunction with a more conservative field setting ensures HPCI
isolation occurring between the range of 300% and 500% steam flow,
thus ensuring HPCI isolation in the event of a pipe break. Because
the HPCI high steamline flow setpoint will be maintained above
normally found operational values (272% steam flow) and below
expected conditions with a pipe break (500% steam flow), the
consequences of any previously evaluated accident are not increased
with the proposed setpoint change.
solation Condenser Steamline High Flow Isolation Trip Level
Setting
The purpose of the Isolation Condenser leak detection
instrumentation is to detect breaks in the system piping. Normal
steam flows within the system can fluctuate in excess of 250% rated
flow and exceed 500% rated steam flow after experiencing a break.
During the original licensing of the plant, it was analytically
determined by GE that three times rated steam flow (300%) is the
optimum setpoint for the isolation of the Isolation Condenser. A
300% steam flow Isolation setpoint ensures that spurious trips are
avoided and that breaks in the piping are identified. Because the
Isolation Condenser High Steamline Flow setpoint is not assumed as
an accident precursor, the probability of any previously evaluated
accident is not increased by the changed setpoint. The proposed
changes to the setpoint provide a more accurate and conservative
field setting for 300% steam flow.
The proposed changes in conjunction with a more conservative
field setting results in Isolation Condenser isolation occurring
between the range of 300% and 500% steam flow, thus ensuring
Isolation Condenser isolation in the event of a pipe break. Because
the Isolation Condenser High Steamline Flow Isolation setpoint will
be maintained above normally found operational values (272% steam
flow) and below expected conditions with a pipe break (500% steam
flow), the consequences of any previously evaluated accident are not
increased with the proposed setpoint change.
Reactor Low-Low Water Level Trip Level Setting Tolerance
The Low-Low Reactor Water Level trip setting is designed to
initiate ECCS when reactor water level is less than or equal to 444
inches above vessel zero. Top of active fuel (TAF) is defined as 360
inches above vessel zero. -59 inches is 84 inches above TAF. The
present trip setting tolerance (84 inches, + 4, - 0, above TAF) only
allows a deviation of 4 inches in the conservative direction. The
proposed change (greater than or equal to 84 inches) does not impose
a restriction on the limit toward the conservative direction.
Because a level switch trip level setting by itself is not assumed
as an accident precursor, the probability of any previously
evaluated accident is not increased by the changed setpoint.
The proposed change eliminates a restriction on the trip level
setting for Low-Low Reactor Water Level. Dresden proposes modifying
the acceptance limit of the Low-Low trip setting such that the
instrument field setting will not deviate below 84 inches.
Therefore, the actuation of appropriate ECCS are unchanged and the
consequences of any previously evaluated accident are not increased
with the proposed setpoint change.
Create the possibility of a new or different kind of accident
from any accident previously evaluated because:
HPCI Steamline High Flow Isolation Trip Level Setting
The purpose of the HPCI Steamline High Flow Isolation trip level
setting is to detect breaks in system piping and initiate isolation
of the system if breaks are discovered. Normal steam flows within
the system can fluctuate as high as 250% rated flow and exceed 500%
rated steam flow after experiencing a break. 300% steam flow has
been used as the setpoint to ensure that spurious trips are avoided
and that breaks in the piping are identified. The changes to the
HPCI High Steamline Flow setpoint ensure that isolation occurs at
300% rated steam flow (below 500% rated steam flow). The current
setpoint will also isolate below 500% rated steam flow. Because the
new setpoint continues to allow normal operational flexibility and
ensures isolation in the event of a pipe break, the proposed changes
do not create the possibility of a new or different kind of accident
than previously evaluated.
Isolation Condenser Steamline High Flow Isolation Trip Level
Setting
The purpose of the Isolation Condenser Steamline High Flow
Isolation trip level setting is to detect breaks in system piping
and initiate isolation of the system if breaks are discovered.
Normal steam flows within the system can fluctuate in excess of 250%
rated flow and exceed 500% rated steam flow after experiencing a
break. 300% steam flow has been used as the setpoint to ensure that
spurious trips are avoided and ensures that isolation occurs at 300%
rated steam flow (below 500% rated steam flow). The current setpoint
will also isolate below 500% rated steam flow. Because the new
setpoint continues to allow normal operational flexibility and
ensures isolation in the event of a pipe break, the proposed changes
do not create the possibility of a new or different kind of accident
than previously evaluated.
Reactor Low-Low Level Trip Level Setting Tolerance
The Reactor Low-Low Water Level trip setting is designed to
initiate the appropriate ECCS when Reactor Water Level is
decreasing. The proposed change to the setpoint only eliminates the
overly burdensome restriction within the setpoint tolerances. The
absolute low limit of 84 inches is unchanged, thus maintaining all
assumptions related to 84 inches (-59 inches indicated level) within
Dresden's Safety Analysis. The removal of the upper tolerance will
not increase the probability of inadvertent ECCS initiation since
the actual field setting will be at a reactor vessel level which has
not been reached in 40 + years of operation at Dresden Units 2 and
3. Therefore, the proposed changes do not create the possibility of
a new or different kind of accident than previously evaluated.
Involve a significant reduction in the margin of safety because:
High Pressure Coolant Injection Setpoint
The HPCI high steamline flow setpoint ensures that isolation
occurs at 300% maximum steam flow (below 500% rated steam flow). The
current Technical Specification setpoint will also allow isolation
below 500% rated steam flow but at a value greater than 300%. Thus,
the proposed setpoint isolates at a lower steam flow rate than the
current limit. Therefore, because isolation of HPCI would occur at a
lower steam flow rate during a pipe break, the proposed changes do
not involve a significant reduction in the margin of safety.
Isolation Condenser Steamline High Flow Isolation Trip Level
Setting
The Isolation Condenser High Steamline Flow Isolation Trip Level
setting ensures that isolation occurs at 300% rated steam flow
(below 500% rated steam flow). The current setpoint will also
isolate below 500% rated steam flow but at a value greater than
300%. Thus, the proposed setpoint isolates at a lower steam flow
rate than the current limit. Therefore, because isolation of the
Isolation Condenser would occur at a lower steam flow rate during a
pipe break, the proposed changes do not involve a significant
reduction in the margin of safety.
Reactor Low-Low Level Trip Level Setting Tolerance
The Reactor Low-Low Water Level trip setting tolerance ensures
the proper initiation of appropriate ECCS in the event of a loss of
inventory to the vessel. The proposed change to the setpoint only
eliminates the restriction within the setpoint tolerances. The
absolute low limit of 84 inches is unchanged, thus maintaining all
assumptions related to 84 inches (minus 59 inches indicated) within
Dresden's Safety Analysis. 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: Morris Public Library, 604
Liberty Street, Morris, Illinois 60450
Attorney for licensee: Michael I. Miller, Esquire; Sidley and
Austin, One First National Plaza, Chicago, Illinois 60690
NRC Project Director: James E. Dyer

Commonwealth Edison Company, Docket Nos. 50-254 and 50-265, Quad
Cities Nuclear Power Station, Units 1 and 2, Rock Island County,
Illinois

Date of amendment request: December 20, 1993
Description of amendment request: The proposed amendment would
revise a minimum critical power ratio (MCPR) safety limit from 1.06 to
1.07 based on General Electric Standard Application for Reactor Fuel II
(GESTAR II) NEDE-24011-P-A-10 for GE10 fuel design. The NRC staff has
previously reviewed and approved the GE10 fuel design.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
The proposed change does not involve a significant increase in
the probability or consequences of an accident previously evaluated
because:
The change is based on GE's generic rule licensing document
GESTAR II (NEDE-24011-P-A-10) which has conservatively addressed the
use of GE10 fuel in D-lattice cores with NRC approved methods and
therefore does not adversely affect the consequences of previously
evaluated accidents. The Safety Limit MCPR change does not affect
the probability of analyzed accidents because it does not adversely
impact any equipment important to safety. Increasing the Safety
Limit MCPR from 1.06 to 1.07 upon implementation of GE10 fuel for
Cycle 14 operation of Quad Cities Units 1 and 2 therefore does not
involve a significant increase in the probability or consequences of
any accident previously evaluated in the FSAR.
The proposed change does not create the possibility of a new or
different kind of accident from any accident previously evaluated
because:
The Safety Limit MCPR change results from the use of NRC
approved methods in GESTAR II NEDE-24011-P-A-10 for application to
GE10 fuel for Cycle 14 for Quad Cities Units 1 and 2. The Safety
Limit MCPR change does not result in any new interaction with
equipment related to the safe shutdown of the plant. The change does
not adversely impact equipment important to safety and, therefore
does not create the possibility of a new or different kind of
accident scenario. Therefore, the Safety Limit MCPR change from 1.06
to 1.07 in no way creates the possibility of a new or different kind
of accident scenario from any accident previously evaluated.
The proposed change does not involve a significant reduction in
a margin of safety because:
Since the GE10 design in a D-lattice core has a geometry between
C-lattice and D-lattice designs and the C-lattice design has a
higher, more restrictive safety limit MCPR that the D-lattice
design, the use of C-lattice safety limit MCPR for the GE10 design
is a conservative approach. The GE10 fuel design has been
generically analyzed with approved methods per GESTAR II NEDE-24011-
P-A-10 and the use of the 1.07 Safety Limit MCPR value has been
previously approved as conservative for application to GE10 fuel in
D-lattice plants such as Quad Cities. Therefore, the proposed change
to increase the Safety Limit MCPR from 1.06 to 1.07 maintains the
margin to safety relative to the current level.
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: Dixon Public Library, 221
Hennepin Avenue, Dixon, Illinois 61021
Attorney for licensee: Michael I. Miller, Esquire; Sidley and
Austin, One First National Plaza, Chicago, Illinois 60690
NRC Project Director: James E. Dyer

Consolidated Edison Company of New York, Docket No. 50-247, Indian
Point Nuclear Generating Unit No. 2, Westchester County, New York

Date of amendment request: December 10, 1993
Description of amendment request: The proposed amendment request
would revise the Technical Specifications to amend (1) Section 5.3.A
(Reactor Core) to allow the use of VANTAGE + fuel with ZIRLO cladding
and fuel with filler rods to allow fuel reconstitution, and (2) the
Basis to Section 2.1 (Safety Limit: Reactor Core) to allow the use of
departure from nucleate boiling (DNB) Correlations applicable to
VANTAGE + fuel.
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:
Consistent with the requirements of 10 CFR 50.92, the enclosed
application involves no significant hazards based on the following
information:
1. Does the proposed license amendment involve a significant
increase in the probability or consequences of an accident
previously evaluated?
Response:
Neither the probability nor the consequences of an accident
previously analyzed is increased due to the proposed changes. As
discussed in [Letter from Thadani to Tritch, ``Acceptance

for Referencing of Topical Report WCAP-12610, VANTAGE + Fuel
Assembly Reference Core Report'' (TAC No. 77258) July 1, 1991] the
fuel containing ZIRLO clad will meet all the same material and
mechanical design criteria as the Zircaloy clad fuel. The use of
approved Westinghouse Methodology for fuel assembly reconstitution
as documented in [Letter from Thadani to Tritch, ``Acceptance for
Referencing of Topical Report WCAP-13060-P, Westinghouse Fuel
Assembly Reconstitution Evaluation Methodology'' (TAC No. M821391),
March 30, 1993] will ensure that all criteria are met. The change to
the basis of Section 2.1 more accurately describes DNB methodology
and application.
2. Does the proposed license amendment create the possibility of
a new or different kind of accident from any previously evaluated?
Response:

The changes will not create the possibility of a new or
different kind of accident. The proposed changes involve approved
methodology which have been shown to meet design and safety
criteria. In addition, approved procedures will be used to implement
the changes.
Response:
3. Does the proposed amendment involve a significant reduction
in the margin of safety?
The proposed amendment does not involve a significant reduction
in the margin of safety. The changes involve the use of approved
methodology which meet design and safety criteria. The change to the
Section 2.1 basis is descriptive and will more accurately describe
the DNB methodology used in conjunction with the use of VANTAGE +
fuel.
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: White Plains Public Library,
100 Martine Avenue, White Plains, New York 10610.
Attorney for licensee: Brent L. Brandenburg, Esq., 4 Irving Place,
New York, New York 10003.
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: September 28, 1993
Description of amendment request: The requested amendments would
delete the portion of the 18-month surveillance requirement on the
autoclosure interlock (ACI) contained in TS 4.5.2.d associated with
verifying that the decay heat removal system suction isolation valves
automatically close on a reactor coolant system pressure signal. The
terms decay heat removal (ND) and residual heat removal (RHR) are used
interchangeably here. Also, an obsolete footnote to TS 4.5.2.e relating
to the completion of the first Unit 1 refueling outage is proposed to
be deleted.
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 requested amendments reference Westinghouse topical report
WCAP-11736-A, ``Residual Heat Removal System Autoclosure Interlock
Removal Report for the Westinghouse Owners Group'', for the general
justification and safety analysis for removing the ACI feature from
the Catawba ND suction isolation valves. This WCAP, which
specifically covers the Catawba Nuclear station, has been deemed an
acceptable reference by the NRC for use in making plant-specific
licensing submittals. Additional Catawba-specific information/
improvements and analyses, as required by the WCAP and associated
NRC safety evaluation, have been either completed or committed to,
thereby ensuring that the WCAP/SE conclusion that removal of the RHR
ACI produces a net safety benefit remains valid.
Criterion 1
The requested amendments will not involve a significant increase
in the probability or consequences of an accident previously
evaluated. The deletion of the RHR ACI was analyzed in the WCAP for
Callaway Nuclear Station in terms of (1) the frequency of an
interfacing LOCA, (2) the availability of the RHR system, and (3)
the effect on overpressure transients. Callaway is the WCAP's
reference plant for Catawba Units 1 and 2, and a Catawba-specific
Probabilistic Risk Assessment (PRA) review of the WCAP determined
that removal of the ND ACI at Catawba will not invalidate the basic
conclusions of the WCAP. Consequently, the following information
from the Callaway analysis is considered applicable to Catawba Units
1 and 2.
With the removal of the ACI and addition of a control room
alarm, the probabilistic risk analysis predicts a decrease in the
frequency of interfacing LOCAs from 1.52E-06/year to 1.16E-06/year,
a decrease of approximately 24%.
The availability of the RHR system was analyzed in three phases:
initiation, short term cooling, and long term cooling. The
probabilistic analysis indicated that deletion of the RHR ACI has no
impact on the failure probability for RHR initiation. During short
term cooling (72 hours after initiation), RHR ACI deletion decreased
the RHR failure probability by 12%, from 1.64E-02 to 1.44E-02. The
long term cooling RHR failure probability was calculated to decrease
by 70%, from 3.91E-02 to 1.17E-02.
Appendix D of the WCAP presents the analysis used to determine
the effect of removal of the ACI on overpressurization transients.
The analysis categorizes the types of initiating events, determines
their frequency of occurrence, and then identifies the consequences
of these occurrences both with and without the ACI feature. The
result is a list of overpressure consequence categories with
associated failure probabilities (reference the WCAP's Appendix D,
Tables D-14, -15, and -16). For the charging/safety injection event,
consequence frequencies increased on the order of 1.0E-12/shutdown
year. This is an insignificant increase, as the overall consequence
frequency of the charging/safety injection event is 1.25E-01.
Likewise, for the letdown isolation with RHR system operable case,
one frequency category was increased on the order of 1.0E-15. Again,
this is insignificant when compared with the total frequency of
these events of 1.25E-01. For the letdown isolation with RHR system
isolated event, the overall consequence frequency was reduced from
4.45E-01 to 2.22E-01. This occurs because many spurious closures of
the RHR isolation valves cause the isolation of letdown. Removing
the RHR ACI reduces the frequency of this event by approximately
50%. It is concluded that the removal of the RHR ACI circuitry has
an insignificant impact on the frequency of overpressurization
events at Callaway (and thus Catawba) Nuclear Station.
Criterion 2
The requested amendments will not create the possibility of a
new or different kind of accident from any accident previously
evaluated. The effect of an overpressure transient at cold shutdown
conditions will not be altered by removal of the ND ACI function.
With or without the ACI function, the ND system could be subject to
overpressrue for which the ND relief valves must be relied upon to
limit pressure to within ND design parameters. While it is true that
the ACI initiates an automatic closure of the ND suction/isolation
valves on high NC system pressure, overpressure protection of the ND
system is provided by the ND system relief valves and not by the
suction/isolation valves that isolate the ND system from the NC
system. (Refer to NUREG-0954, ``Safety Evaluation Report related to
the operation of Catawba Nuclear Station, Units 1 and 2,'' Section
5.4.4.3.)
The purpose of the ACI feature is to ensure that there is a
double barrier between the ND system and the NC system when the
plant is at normal operating conditions (i.e., heated and
pressurized) and not in the ND cooling mode. Thus, the ACI feature
serves to preclude conditions that could lead to a LOCA outside of
containment due to operator error. The safety function of the ACI is
not to isolate the ND system from the NC system when the ND system
is operating in the decay heat removal mode.
There are several methods to ensure that there is a double
barrier between the ND system and the NC system when the plant is at
normal operating conditions. First, plant operating procedures
instruct the operators to isolate the ND system during plant heatup.
Second, the alarm that will be installed as part of this change will
annunciate in the control room given an open or intermediate valve
position signal in conjunction with a high NC pressure signal. This
alarm will alert operators that any of the four suction/isolation
valves is (are) not fully closed and that double isolation has not
been achieved. In conjunction with this alarm, operators will be
trained using an annunciator response procedure to ensure that they
act to restore double isolation or return to a safe shutdown
condition. Third, the Open Permissive Interlock (OPI), which is not
being removed, will prevent the opening of the valves whenever NC
system pressure is greater than 385.5 psig.
Since relief valves prevent overpressurization of the ND system
during shutdown conditions and since several methods are in place to
ensure that the ND system is isolated from the NC system during
normal plant conditions, removal of the ACI will not create the
possibility of a new or different kind of accident from any accident
previously evaluated.
Criterion 3
The requested amendments will not involve a significant
reduction in a margin of safety. The ND ACI function is not a
consideration in a margin of safety in the basis for any technical
specification. Since the probabilistic analysis of the WCAP for
Callaway (which is applicable to Catawba as discussed above)
indicates that the availability of the RHR system is increased with
the removal of the ACI, overall safety will be increased.
In addition, similar amendments for other Westinghouse plants in
the past have been determined to not involve significant hazards
considerations.
Based upon the preceding analyses, Duke Power Company concludes
that the requested amendments do not involve a significant hazards
consideration.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: 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: Loren R. Plisco, Acting

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: January 27, 1994
Description of amendment request: The requested amendments delete
the verification that each upper and lower Containment Purge System
(VP) supply and exhaust valve actuates to its isolation position on a
High Relative Humidity (70%) isolation test signal and will
allow elimination of the humidity control function of the VP System
humidistats.
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
This TS [Technical Specification] amendment will not increase
the probability or consequences of an accident which has been
previously evaluated. No physical changes will be made to the plant
that would impact fuel handling inside containment, therefore, there
is no increase in the probability of an accident. Control wiring
changes that remove the humidistats from the [Containment Purge]
System control circuits will be the only physical change.
The heaters will be maintained providing additional margin over
analyzed conditions. For the reasons stated above, there will be no
increase in the consequences of an accident previously evaluated.
CRITERION 2
This proposed TS amendment does not create the possibility of a
new or different kind of accident from any accident previously
evaluated. This proposed TS amendment will not cause any physical
changes to the plant that will impact the handling of fuel inside
containment or changes to fuel handling procedures. Because the
plant will operate the same way it does now, this proposed amendment
does not create the possibility of any new or different accident
from any previously evaluated.
CRITERION 3
This proposed TS change will not cause a significant reduction
in the margin of safety. The test method use[d] to evaluate the
carbon after TS changes 90 ([Unit] 1) and 84
([Unit] 2) does not consider heater availability. However the
heaters will be tested and maintained per Technical specification
4.9.4.2.d.2. Therefore, the relative humidity of the air entering
the carbon adsorber is never expected to reach 95% [relative
humidity].
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: Loren R. Plisco, Acting

Entergy Operations, Inc., Docket No. 50-313, Arkansas Nuclear One,
Unit No. 1 (ANO-1), Pope County, Arkansas

Date of amendment request: January 13, 1994
Description of amendment request: This amendment revises the
specifications governing the reactor protection (RPS). It modifies the
use of the RPS channel bypass as specified by Technical Specification
(TS) 3.5.1.3 and revises a note with Table 3.5.1-1, to refer to a more
appropriate action.
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 RPS and EFIC [emergency feedwater initiation and control]
system provide accident mitigation features and are not considered
to be accident initiators. The accident mitigation features of the
plant are not affected by the proposed amendment. In any
configuration allowed by the revised specifications, the trip logic
instituted on the RPS is at least equivalent to the trip logic
instituted by placing a channel in channel or maintenance bypass.
The RPS remains single-failure proof with one channel in channel
bypass, manually tripped, or with an inoperable function unbypassed
in the untripped state. Therefore, upon receipt of an initiating
signal, a single failure will not prevent the proper actuation of
RPS. Should a channel of RPS contain an inoperable function
unbypassed in the untripped condition which does not affect an EFIC
channel, any channel of EFIC may be placed in maintenance bypass.
RPS and EFIC remain single-failure proof in this configuration.
Administrative controls are established to ensure that all
inoperable RPS functions are evaluated for continued operation in
the untripped state. Upon detection of a failed function in any
channel of RPS, the administratively controlled condition reporting
process evaluates the failure and its effect on other systems for
continued operability. The operator is informed of the continuing
status of inoperable functions through the use of Station Log
entries and Plant Status board entries. In addition, during
operation with an inoperable function in the untripped, unbypassed
condition, the remaining RPS channel key-lock channel bypass
switches will be ``Hold Carded'' (tagged) to prevent their operation
without prior management approval consistent with the requirements
of TS Table 3.5.1-1. Plant management maintains the responsibility
to approve continued operation with inoperable functions unbypassed
in the untripped state to ensure that the plant is operated in the
safest configuration with regard to the extent of the failure, and
the plant operating conditions. Prior to placing any channel of RPS
or EFIC in bypass, the operator checks the status of redundant
systems for operability and TS compliance and takes the proper
action as required by existing plant conditions, plant operating
procedures and TS.
The clarification to TS 3.5.1.3 which directs the operator to
the appropriate actions if multiple channels become inoperable, or
in the event of an inoperable channel or inoperable function
occurring concurrent with one channel in bypass is considered to be
administrative in nature. The change to Note 6 of Table 3.5.1-1
results in the correction of misleading information and directs the
Operator to place the plant in a safe mode depending on the system
which is affected by a failure, and is also considered to be
administrative in nature. The Bases changes add additional
information to clarify the specifications.
Therefore, this change does not involve a significant increase
in the probability or consequences of any accident previously
evaluated.
Criterion 2 - Does Not Create the Possibility of a New or
Different Kind of Accident from any Previously Evaluated.
The probability or consequences of equipment important to safety
malfunctioning will not be increased. In any configuration allowed
by the revised specifications, the trip logic instituted on the RPS
is at least equivalent to the trip logic instituted by placing a
channel in channel bypass. The RPS remains single-failure proof with
one channel in channel bypass, manually tripped, or with an
inoperable function unbypassed in the untripped state. Therefore,
upon receipt of an initiating signal, a single failure will not
prevent the proper actuation of RPS. Should a channel of RPS contain
an inoperable function unbypassed in the untripped condition which
does not affect an EFIC channel, any channel of EFIC may be placed
in maintenance bypass. RPS and EFIC remain single-failure proof in
this configuration.
The clarification to TS 3.5.1.3 which directs the operator to
the appropriate actions if multiple channels become inoperable, or
in the event of an inoperable channel or inoperable function
occurring concurrent with one channel in bypass is considered to be
administrative in nature. The change to Note 6 of Table 3.5.1-1 is
also considered to be administrative in nature, in that misleading
information in the specification has been corrected to an
appropriate requirement. The Bases changes add additional
information to clarify the specifications.
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 RPS and EFIC system have the same capabilities to mitigate
and/or prevent accidents as they had prior to this proposed change.
Allowing flexibility in the response to a function failure in one
channel of RPS allows placing the plant in the safest operating
condition for the existing plant conditions considering the extent
of the function failure. Operation of an RPS channel with an
inoperable function unbypassed in the untripped state results in
placing the inoperable function in a 2-out-of-3 trip logic
(equivalent to channel bypass) while the remainder of the RPS
functions remain in the normal 2-out-of-4 trip logic. The ANO-1 RPS
has been reviewed as a 3 channel system with one channel in bypass.
Implementing this change results in additional conservatism with
respect to any postulated single-failures.
Administrative controls are established to ensure that all
inoperable RPS functions are evaluated for continued operation in
the untripped state. Upon detection of a failed function in any
channel of RPS, the administratively controlled condition reporting
process evaluates the failure and its effect on other systems for
continued operability. The operator is informed of the continuing
status of inoperable functions through the use of Station Log
entries and Plant Status board entries. In addition, during
operation with an inoperable function in the untripped, bypassed
condition, the remaining RPS channel key-lock channel bypass
switches will be ``Hold Carded'' (tagged) to prevent their operation
without prior management approval consistent with the requirements
of TS Table 3.5.1-1. Plant management maintains the responsibility
to approve continued operation with inoperable functions unbypassed
in the untripped state to ensure that the plant is operated in the
safest configuration with regard to the extent of the failure, and
the plant operating conditions. Prior to placing any channel of RPS
or EFIC in bypass, the operator checks the status of redundant
systems for operability and TS compliance and takes the proper
action as required by existing plant conditions, plant operating
procedures and TS. Should a channel of RPS contain an inoperable
function unbypassed in the untripped condition which does not affect
an EFIC channel, any channel of EFIC may be placed in maintenance
bypass. RPS and EFIC remain single-failure proof in this
configuration.
The clarification of TS 3.5.1.3 which directs the operator to
the appropriate actions if multiple channels become inoperable, or
in the event of an inoperable channel or inoperable function
occurring concurrent with one channel in bypass is considered to be
administrative in nature. The change to Note 6 or Table 3.5.1-1
results in the correction of misleading information and directs the
Operator to place the plant in a safe mode depending on the system
which is affected by a failure, and is also considered to be
administrative in nature. The Bases changes add additional
information to clarify the specifications.
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, Arkansas 72801
Attorney for licensee: Nicholas S. Reynolds, Esquire, Winston and
Strawn, 1400 L Street, N.W., Washington, D.C. 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: January 13, 1994
Description of amendment request: This amendment requests the
removal of the interim technical specification limit on the number of
spent fuel assemblies that may be stored in the spent fuel pool at
Grand Gulf Nuclear Station.
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. No significant increase in the probability or consequences of
an accident previously evaluated results from this change.
The NRC approved the installation of high density spent fuel
storage racks in Amendment 17 to the Grand Gulf Nuclear Station
(GGNS) Operating License. This amendment also brought GGNS into
compliance with Standard Review Plan criteria which required
maintaining the spent fuel pool at less than or equal 140 deg.F. The
140 deg.F Technical Specifications (TS) limit remains in effect
thereby preventing operation at excessive temperatures.
The only outstanding question from Amendment 17, which resulted
in a 2324 assembly technical specification limit, was whether the
fuel pool cooling system could handle the heat load of a full fuel
pool without excessive reliance on residual heat removal for
extensive fuel pool cooling assist. Entergy Operations' proposed
solution to this question was accepted in the NRC's letter dated
July 30, 1992. The NRC accepted the solution pending submittal of
results from tests to verify the specified flows. These results were
submitted in a letter dated November 08, 1993.
With previous approval of the installation of the high density
spent fuel storage racks, the confirmation of adequate heat removal
capability, and the 140 deg.F TS temperature limit, removal of the
2324 limit to allow full use of the spent fuel pool would not cause
an increase in the probability or consequences of an accident
previously evaluated.
2. This change would not create the possibility of a new or
different kind of accident from any accident previously evaluated.
The additional heat load generated by a full spent fuel pool
(4348 assemblies) was evaluated. The evaluation concluded that full
use of the spent fuel pool storage spaces would not exceed the
temperature limits as are currently in place with the 2324 limit.
The NRC letter dated July 30, 1992 and Entergy Operations letter
dated November 08, 1993 resolved all outstanding heat removal
questions. Therefore, this change would not create the possibility
of a new or different kind of accident from any previously analyzed.
3. This change would not involve a significant reduction in a
margin of safety.
Entergy Operations demonstrated in their November 01, 1991
letter that the fuel pool temperature could be maintained at or
below 140*F as specified in TS 3/4.7.9. This letter also
demonstrated the ability to handle single active failures. Approval
of measures outlined in this letter was provided in a Safety
Evaluation Report contained in an NRC letter dated July 30, 1992.
Given the 140 deg.F maximum temperature requirement as contained
in TS 3/4.7.9 and compliance with single active failure criteria,
this change would not involve a significant reduction in a margin of
safety.
Based on the above evaluation in accordance with 10CFR50l92(c),
Entergy Operations, Inc. has concluded that operation in accordance
with the proposed amendment involves no significant hazards
considerations.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: Judge George W. Armstrong
Library, Post Office Box 1406, S. Commerce at Washington, Natchez,
Mississippi 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

Florida Power and Light Company, Docket Nos. 50-250 and 50-251,
Turkey Point Plant Units 3 and 4, Dade County, Florida

Date of amendment request: December 28, 1993
Description of amendment request: The proposed amendments would
revise the Technical Specifications (TS) for Turkey Point Units 3 and 4
to incorporate features for steam generator (SG) overfill protection.
Specifically, TS Tables 3.3-2, 3.3-3, 4.3-2 and the associated BASES
section would be revised to add SG Water Level-High-High protection
logic, instrumentation trip setpoints and surveillance requirements.
The proposed TS changes would be in accordance with NRC Generic Letter
(GL) 89-19, ``Safety Implication of Control Systems in LWR Nuclear
Power Plants.''
Basis for proposed no significant hazards consideration
determination: As a result of the technical resolution of USI A-47,
``Safety Implications of Control Systems in LWR Nuclear Power Plants,''
the Nuclear Regulatory Commission (NRC or the staff) concluded that all
Pressurized Water Reactors (PWR) plants should provide automatic SG
overfill protection. On September 20, 1989, the staff issued GL 89-19
and recommended that plant procedures and TS include provisions for
automatic SG overfill protection including surveillance requirements to
assure that automatic overfill protection is available to mitigate main
feedwater overfeed events during reactor power operation.
The licensee proposed TS changes in response to GL 89-19. No
physical changes to the plant would be required as a result of the
proposed license amendments.
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.
Including the SG Overfill protection requirements in the
Technical Specifications is not assumed in the initiation of any
analyzed event. These amendments will not increase the probability
or consequences of an accident previously evaluated since the SG
overfill event is not required or assumed for accident mitigation in
any Updated Final Safety Analysis Report (UFSAR) safety analyses
that comprise Turkey Point licensing basis. The additional
requirements for the SG overfill system helps ensure that continuous
addition of feedwater and carryover of excessive moisture to the
turbine, is prevented. As a result, equipment protection is improved
by the availability of this system function. As such, operation of
the facility in accordance with the proposed amendments would not
involve a significant increase in the probability or consequences of
an accident previously evaluated.
(2) Operation of the facility in accordance with the proposed
amendments would not create the possibility of a new or different
kind of accident from any accident previously evaluated.
The operation of the facility will not change as a result of the
proposed license amendments, since Turkey Point currently maintains
this protection logic. This change involves only the inclusion of
the systems requirements into the Technical Specifications. The
proposed change will not impose any new or unique requirements.
Therefore, operation of the facility in accordance with the proposed
amendments will not create the possibility of a new or different
kind of accident from any accident previously evaluated.
(3) Operation of the facility in accordance with the proposed
amendments would not involve a significant reduction in a margin of
safety.
The proposed change does not involve a significant reduction in
a margin of safety as the function, operation and testing of the
installed SG Overfill protection is not described in the UFSAR. In
addition, the SG overfill protection logic is not required or
assumed for accident mitigation in any of the safety analyses that
comprise the Turkey Point licensing basis. The proposed change
formalizes the existing design, operating and testing requirements
in the Technical Specifications. Therefore, the proposed change does
not involve a significant reduction in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: Florida International
University, University Park, Miami, Florida 33199
Attorney for licensee: Harold F. Reis, Esquire, Newman and Holtzer,
P.C., 1615 L Street, NW., Washington, DC 20036
NRC Project Director: Herbert N. Berkow

Georgia Power Company, Oglethorpe Power Corporation, Municipal
Electric Authority of Georgia, City of Dalton, Georgia, Docket Nos.
50-424 and 50-425, Vogtle Electric Generating Plant, Units 1 and 2,
Burke County, Georgia

Date of amendment request: December 30, 1993
Description of amendment request: The proposed change would allow a
one time extension of the allowable outage time for each residual heat
removal (RHR) pump from 3 to 7 days to allow modifications to the RHR
system while the plant is in Mode 1.
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 to the Technical Specifications does not
involve a significant increase in the probability or consequences of
an accident previously evaluated because the redundant train will
remain available to assure that the RHR will respond to an accident
as assumed in the accident analysis. A one time increase in the
allowable outage time from 3 to 7 days has been shown to have only a
small effect on the calculated frequency of core damage.
2. The proposed change to the Technical Specifications does not
create the possibility of a new or different kind of accident from
any accident previously evaluated because the change only results in
a one time increase of the allowable outage time. It does not result
in an operational condition different from that which has already
been considered by the Technical Specifications.
3. The proposed addition to the Technical Specifications does
not involve a significant reduction in a margin of safety because
the effects of increasing the allowed outage time on the calculated
core damage frequency has been evaluated and determined to be small.
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: Burke County Public Library,
412 Fourth Street, Waynesboro, Georgia 30830.
Attorney for licensee: Mr. Arthur H. Domby, Troutman Sanders,
NationsBank Plaza, Suite 5200, 600 Peachtree Street, NE., Atlanta,
Georgia 30308
NRC Project Director: Loren R. Plisco, Acting

IES Utilities Inc., Docket No. 50-331, Duane Arnold Energy Center,
Linn County, Iowa

Date of amendment request: December 22, 1993
Description of amendment request: The proposed amendment would make
editorial changes to correct typographical and administrative errors in
the Technical Specifications.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1) The proposed amendment does not involve a significant
increase in the probability or consequences of an accident
previously evaluated. The amendment would only correct
administrative and typographical errors. No physical changes to the
plant or to the operation of the plant would result from this
amendment.
2) The proposed amendment will not create the possibility of a
new or different kind of accident from any evaluated previously. The
amendment would only correct administrative and typographical
errors. No physical changes to the plant or to operation of the
plant would result from this amendment.
3) The proposed amendment will not reduce the margin of safety.
The amendment would only correct administrative and typographical
errors. No physical changes to the plant or to operation of the
plant would result from this amendment.
The NRC staff has reviewed the licensee's analysis and, based on
thisreview, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: Cedar Rapids Public Library,
500 First Street, S.E., Cedar Rapids, Iowa 52401
Attorney for licensee: Jack Newman, Esquire, Kathleen H. Shea,
Esquire, Newman and Holtzinger, 1615 L Street, NW., Washington, DC
20036
NRC Project Director: John N. Hannon

IES Utilities Inc., Docket No. 50-331, Duane Arnold Energy Center,
Linn County, Iowa

Date of amendment request: January 21, 1994
Description of amendment request: The proposed amendment would
change the name of the company licensed to own a share of and operate
the Duane Arnold Energy Center (DAEC) from Iowa Electric Light and
Power Company to IES Utilities Incorporated, wherever it is referenced
in the Operating License and Technical Specifications for DAEC. The
title of the position responsible for the management of the Nuclear
Division has also been changed to ``Vice President, Nuclear'' from
``Manager-Nuclear Division.''
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1) The proposed amendment does not involve a significant
increase in the probability or consequences of an accident
previously evaluated. No physical or operational changes to the DAEC
will result from changing the corporate name or the position title.
The DAEC will continue to be operated in the same manner with the
same organization. The position title change results from the
elimination of a layer of management. Formerly, the Manager-Nuclear
Division reported through the Vice President, Production to the
President of IELP. Now the Nuclear Division is headed by the Vice
President, Nuclear who reports directly to the President of the
corporation.
2) The proposed amendment does not create the possibility of a
new or different kind of accident from any previously evaluated. No
physical or operational changes will result. The title change
results from the elimination of a layer of management.
3) The proposed change will not reduce any margin of safety.
This change only revises the operating company name and changes a
title.
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: Cedar Rapids Public Library,
500 First Street, S.E., Cedar Rapids, Iowa 52401
Attorney for licensee: Jack Newman, Esquire, Kathleen H. Shea,
Esquire, Newman and Holtzinger, 1615 L Street, NW., Washington, DC
20036
NRC Project Director: John N. Hannon

Indiana Michigan Power Company, Docket Nos. 50-315 and 50-316,
Donald C. Cook Nuclear Plant, Unit Nos. 1 and 2, Berrien County,
Michigan

Date of amendment requests: January 17, 1994
Description of amendment requests: The proposed amendments would
change Technical Specification (TS) 3/4.1.3 for both units to increase
the limit for control rod misalignment at or below 85% rated thermal
power (RTP). The proposed changes would also increase the TS limit for
control rod 8misalignment about 85% RTP if there is sufficient margin
in the heat flux (FQ(Z)) and the nuclear enthalpy
(FNdelta H) hot channel factors.
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:
Per 10 CFR 50.92, a proposed amendment to an operating license
will not involve a significant hazards consideration if the proposed
amendment satisfies the following three criteria:
1. Does not involve a significant increase in the probability or
consequences of an accident previously analyzed,
2. Does not create the possibility of a new or different kind of
accident from any accident previously analyzed or evaluated, or
3. Does not involve a significant reduction in a margin of
safety.
Criteria 1 and 3
As seen in Attachment 4 [of the amendment request], sufficient
margin exists in power distribution at 85% RTP to allow for
increased misalignment. At 100% RTP, increased misalignment is
allowed only if there is adequate margin in the peaking factors.
Therefore, initial conditions remain unchanged from that assumed in
the safety analyses. As far as the dropped rod and rod ejection
accidents are concerned, the analyses were performed with
conservative assumptions to envelope the increased misalignment. It
should be noted that the power dependent insertion limit for Unit 1
will be changed in a conservative manner at the beginning of cycle
14. Based on these analyses, it is concluded that the proposed T/S
changes do not significantly increase the probability or
consequences of a previously analyzed accident or constitute a
significant reduction in the margin of safety.
Criterion 2
The proposed T/S changes will not result in physical changes to
the plant. Therefore, we believe that the proposed T/S changes will
not create the possibility of a new or different kind of accident
from any previously evaluated. Also, operation of the reactor with
possible deeper rod insertion will not create the possibility of a
new or different kind of accident.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment requests involve no significant hazards consideration.
Local Public Document Room location: Maud Preston Palenske Memorial
Library, 500 Market Street, St. Joseph, Michigan 49085
Attorney for licensee: Gerald Charnoff, Esq., Shaw, Pittman, Potts
and Trowbridge, 2300 N Street, NW, Washington, DC 20037
NRC Project Director: A. Randolph Blough, Acting

Niagara Mohawk Power Corporation, Docket No. 50-220, Nine Mile
Point Nuclear Station Unit No. 1, Oswego County, New York

Date of amendment request: January 21, 1994
Description of amendment request: The proposed amendment would
revise Technical Specification (TS) 4.6.3, (Emergency Power Sources) to
eliminate unnecessary testing of an operable emergency diesel generator
(EDG) when the redundant EDG becomes inoperable. Eliminating
unnecessary testing will potentially increase EDG reliability by
reducing the stresses caused by such testing. The licensee stated that
this proposed change is consistent with the guidance provided in NUREG-
1366, ``Improvements to Technical Specifications Surveillance
Requirements,'' and NUREG-1433, ``Improved Standard Technical
Specifications, General Electric Plants.''
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
The operation of Nine Mile Point Unit 1, in accordance with the
proposed amendment, will not involve a significant increase in the
probability or consequences of an accident previously evaluated.
Technical Specification 4.6.3.e requires that the operable
diesel-generator be manually started and operated at rated load for
a minimum time of one hour immediately and once per week thereafter
in the event any diesel-generator becomes inoperable.
Niagara Mohawk proposes to revise Technical Specification
4.6.3.e such that if a diesel-generator is declared inoperable due
to preplanned maintenance or testing or due to a support system
being inoperable, redundant diesel-generator testing would not be
required. Declaring a diesel-generator inoperable due to preplanned
maintenance or testing or due to a support system being inoperable
does not affect the reliability of the operable diesel-generator nor
does it in any way imply that a common cause failure exists.
The normally required Technical Specification surveillance
testing schedule demonstrates acceptable reliability and assures
that the operable diesel-generator is capable of performing its
intended safety function.
Niagara Mohawk proposes to add wording to Technical
Specification 4.6.3.e to permit an operator to evaluate a diesel-
generator failure to determine if a common cause failure exists
before requiring testing of the redundant diesel-generator. As noted
above, the intent of the additional diesel-generator testing is, in
part, to determine if a common cause failure exists. Once the
potential for a common cause failure has been examined and
dismissed, testing beyond the normal surveillance schedule is
excessive and does not contribute to improved diesel-generator
reliability. Within eight (8) hours, the determination that no
common cause failure exists is required to be completed or the
operable diesel-generator will be tested. Eight (8) hours is
consistent with the guidance provided in NUREG-1366, ``Improvements
to Technical Specifications Surveillance Requirements.''
Technical Specification 4.6.3.e requires that the operable
diesel-generator be operated at rated load (i.e., connected to
offsite power) to demonstrate its operability in the event any
diesel-generator becomes inoperable. As indicated in Information
Notice 84-69, when a diesel-generator is operated connected to
offsite or non-vital loads, the emergency power system is not
independent of disturbances on the non-vital and offsite power
systems. Therefore, diesel-generator availability is potentially
lessened by a demonstration of operability requiring connection of
the diesel-generator to offsite power sources. At a time when at
least one diesel-generator is already inoperable, this Surveillance
Requirement could add further risk to losing the remaining operable
diesel-generator. Therefore, Niagara Mohawk proposes that
Surveillance Requirement 4.6.3.e be changed such that a diesel-
generator does not have to be operated at rated load. These changes
will preclude offsite power source disturbances from affecting
diesel-generator reliability.
Existing Technical Specification 4.6.3.e requires that the
operable diesel-generator be started immediately in the event a
diesel-generator becomes inoperable. The requirement to immediately
test a diesel-generator is overly burdensome when compared to more
recent diesel-generator Technical Specification requirements. As
previously discussed, Niagara Mohawk proposes to add wording to
Technical Specification 4.6.3.e to give an operator eight (8) hours
to determine whether a common cause failure exists or to test the
operable diesel-generator when a diesel-generator is declared
inoperable for a reason other than an inoperable support system or
preplanned maintenance or testing. Eight (8) hours is consistent
with the guidance provided in NUREG-1366, ``Improvements to
Technical Specifications Surveillance Requirements.''
Existing Technical Specification 4.6.3.e requires that the
operable diesel-generator be tested immediately and once per week
thereafter. Technical Specification 3.6.3.c requires that an
inoperable diesel-generator be returned to an operable condition
within seven (7) days to meet the Limiting Condition for Operation.
Therefore, the requirement to test the operable diesel-generator
``once a week thereafter'' is not applicable. In addition, testing
the operable diesel-generator one time is adequate to confirm
operability of a diesel-generator. Repetitive testing following
initial confirmation of operability is unwarranted. Therefore,
Niagara Mohawk proposes to delete the requirement to test the
operable diesel-generator weekly following the initial test.
Because the proposed change does not affect the design or
performance of the diesel-generators nor adversely affect the
reliability of the diesel-generators, the change will not result in
an increase in the consequences of an accident previously evaluated
(i.e., Station Blackout analyses). Because this change does not
affect the probability of accident precursors, the proposed change
does not affect the probability of an accident previously evaluated.
The operation of Nine Mile Point Unit 1, in accordance with the
proposed amendment, will not create the possibility of a new or
different kind of accident from any accident previously evaluated
The proposed change to Technical Specification 4.6.3.e does not
introduce any new operating configurations or new accident
precursors and does not involve any physical alterations to plant
configurations which could initiate a new or different kind of
accident. The proposed change does not affect the design or
performance characteristics of the diesel-generators nor does the
change create the possibility of the loss of both diesel-generators
because common cause failure assessments will be performed. The
change will delete excessive diesel-generator testing and therefore
increase overall plant safety by increasing diesel-generator
reliability. Therefore, the proposed amendment will not create the
possibility of a new or different kind of accident from any
previously evaluated.
The operation of Nine Mile Point Unit 1, in accordance with the
proposed amendment, will not involve a significant reduction in a
margin of safety
The proposed change to Technical Specification 4.6.3.e will not
reduce the number of emergency power sources required by Technical
Specification Limiting Condition for Operation 3.6.3 or affect the
normal surveillance requirements as described in Technical
Specification 4.6.3. The normal surveillance tests demonstrate
acceptable reliability and assure that the operable diesel-generator
is capable of performing its intended function. The proposed change
to delete the excessive testing requirements does not affect the
design or performance of any diesel-generator and does not adversely
affect diesel-generator reliability. Eliminating unnecessary testing
will potentially increase diesel-generator reliability by reducing
the stresses caused by such testing. Therefore, the proposed change
does not involve a significant reduction in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: Reference and Documents
Department, Penfield Library, State University of New York, Oswego, New
York 13126.
Attorney for licensee: Mark J. Wetterhahn, Esquire, Winston &
Strawn, 1400 L Street, NW., Washington, DC 20005-3502.
NRC Project Director: Robert A. Capra

Northern States Power Company, Docket No. 50-263, Monticello
Nuclear Generating Plant, Wright County, Minnesota

Date of amendment request: November 30, 1993
Description of amendment request: The proposed amendment would
change sections 3.2/4.2, Protective Instrumentation, and 3.17/4.17,
Control Room Habitability, by deleting the requirements for a chlorine
detection system and revises the limiting conditions for operation for
the Control Room Ventilation System to be more consistent with Standard
Technical Specifications. Due to design changes at the Monticello
Nuclear Generating Plant, chlorine is no longer stored onsite as a
liquified gas and regulations requiring early warning of an onsite
chlorine release do not apply.
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. Concerning Deletion of Requirements for the Chlorine
Detection System
The proposed amendment will not involve a significant increase
in the probability or consequences of an accident previously
evaluated.
Postulated chemical releases of chlorine have been shown to be
such that incapacitation of the control room operators would not
occur within allowed time frames for the donning of protective
breathing equipment, or that the probability of a chlorine trucking
transportation accident which causes incapacitation of control room
operators with potential consequences of a radioactive release in
excess of 10 CFR 100 guidelines is well below the level of concern
as established in regulatory guidance. Therefore, this amendment
will not cause a significant increase in the probability or
consequences of an accident previously evaluated for the Monticello
plant.
The proposed amendment will not create the possibility of a new
or different kind of accident from any accident previously analyzed.
The performance of a new toxic chemical analysis for the
Monticello site has demonstrated that human detection may be relied
upon to detect chlorine toxic chemical releases. Operator protection
is established via the donning of protective breathing equipment.
The capability to manually isolate the control room with dampers is
retained. The ability of the operators to cope with a chlorine toxic
gas hazard remains consistent with the protection measures available
for other toxic chemicals stored onsite, stored in the vicinity of
the site, or transported near the plant site. The proposed amendment
will not create the possibility of a new or different kind of
accident.
The proposed amendment will not involve a significant reduction
in the margin of safety.
The performance of a new toxic chemical analysis for the
Monticello site has demonstrated that incapacitation of the control
room operators would not occur within allowed time frames for the
donning of protective breathing equipment and that a postulated
hazardous chemical release due to a trucking transportation accident
involving chlorine is of a sufficiently low probability of
occurrence that it need not be considered. The basis of the chlorine
detectors and associated Technical Specifications is to provide
protection against an accident scenario which has been demonstrated
to be of extremely low probability (a trucking transportation
accident involving chlorine within five miles of the plant),
therefore removal of the chlorine detectors from the plant design
and the associated Technical Specifications will not involve a
significant reduction in the margin of safety.
2. Concerning the Limiting Conditions for Operation for the
Control Room Ventilation System and Technical Specification Bases

The proposed amendment will not involve a significant increase
in the probability or consequences of an accident previously
evaluated.
The Control Room Ventilation system ensures that Main Control
Room habitability is maintained such that personnel and equipment
located in the control room can respond to mitigate the consequences
of an accident. The system does not contribute to the probability of
occurrence of any design basis accident. The operability
requirements as proposed for the revised specification 3.17.A ensure
that the Control Room Ventilation system is operable during plant
conditions for which significant radioactive releases are postulated
consistent with the Standard Technical Specification. The proposed
changes ensure the Control Room Ventilation system is restored to an
operable status or that actions are taken to minimize the importance
of the system function within time frames which take into
consideration the low probability of an event occurring which would
require Control Room Ventilation system function. Therefore, this
amendment will not cause a significant increase in the probability
or consequences of an accident previously evaluated for the
Monticello plant.
The proposed amendment will not create the possibility of a new
or different kind of accident from any accident previously analyzed.
The proposed changes to Technical Specifications 3.17.A do not
alter the function of the Control Room Ventilation system or its
interrelationships with other systems. The proposed changes provide
requirements to ensure the Control Room Ventilation system is
capable of performing its required function or that actions are
taken to minimize the potential for its function being required
consistent with regulatory guidance; therefore, this amendment will
not create the possibility of a new or different kind of accident
from any accident previously analyzed.
The proposed amendment will not involve a significant reduction
in the margin of safety.
The operability requirements as proposed for the revised
specification 3.17.A ensure that the Control Room Ventilation system
is operable during plant conditions for which significant
radioactive releases are postulated. The performance of a new toxic
chemical analysis for the Monticello site has demonstrated that a
postulated hazardous chemical release due to a trucking
transportation accident involving chlorine is of a sufficiently low
probability of occurrence that it need not be considered. As the
basis of the chlorine detectors and current operability requirements
for the control Room Ventilation system is to provide protection
against an accident scenario which has been demonstrated to be of
extremely low probability, the proposed revision to the Control Room
Ventilation operability requirements will not involve a significant
reduction in the margin of safety.
The proposed changes to Technical Specification 3.17.A ensure
that both trains of the Control Room Ventilation system are restored
to an operable status within a time frame which takes into
consideration the low probability of an event occurring requiring
Control Room Ventilation system function, the availability of the
redundant Control Room Ventilation train and the capability of the
safety related Emergency Filtration Train to pressurize the control
room without the Control Room Ventilation system. The proposed
changes provide requirements to ensure the Control Room Ventilation
system is capable of performing its required function or that
actions are taken to minimize the potential for its function to be
required consistent with regulatory guidance; therefore, the
proposed change will 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: Minneapolis Public Library,
Technology and Science Department, 300 Nicollet Mall, Minneapolis,
Minnesota 55401
Attorney for licensee: Gerald Charnoff, Esq., Shaw, Pittman, Potts
and Trowbridge, 2300 N Street, NW, Washington, DC 20037
NRC Project Director: L. B. Marsh

Northern States Power Company, Docket No. 50-263, Monticello
Nuclear Generating Plant, Wright County, Minnesota

Date of amendment request: January 3, 1994
Description of amendment request: The proposed amendment would
revise the requirements of Technical Specification 4.6.E.1.a, which
currently specifies that a minimum of seven safety/relief valves shall
be bench checked or replaced with a bench checked valve each refueling
outage. The proposed amendment would change this specification to
require the valves to be tested in accordance with the Section XI
Inservice Testing Requirements of the ASME Boiler and Pressure Vessel
Code. The proposed change is consistent with the Improved Standard
Technical Specifications, NUREG-1433.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
a. The proposed amendment will not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
The proposed amendment is limited to changes to the surveillance
testing requirements (bench checking or replacement) applicable to
the main steam system safety/relief valves. This surveillance
requirement is performed while the plant is in a cold shutdown
condition at a time when the safety/relief valves are not required
to be operable. The performance of this evolution is not an input or
consideration in any accident previously evaluated, thus the
proposed change will not increase the probability of any such
accident occurring. Current safety analyses conclude that the
pressure relief capabilities of the Safety Relief valves are
adequate assuming that one of the eight safety/relief valves fails
to open upon demand. The proposed change will not adversely affect
the reliability of the valves and will therefore not reduce the
conservatism of this assumption.
Similarly, the proposed amendment specifies testing requirements
consistent with accepted industry codes and regulatory guidance to
provide assurance that the valves will function as designed. The
amendment will not diminish the capability of the safety/relief
valves to perform as required during any accident previously
evaluated and will therefore not increase the consequences of any
such accident.
b. The proposed amendment will not create the possibility of a
new or different kind of accident from any accident previously
analyzed.
The proposed amendment does not involve any modification to
plant equipment or operating procedures, nor will it introduce any
new safety/relief valve failure modes that have not been previously
considered. The net result of the proposed amendment will be to
allow the plant staff the option of decreasing the frequency of
safety/relief valve testing to a level that has been acknowledged as
acceptable by the ASME Code and NUREG-1433. We therefore conclude
the proposed changes will not create the possibility of a new or
different kind of accident from any accident previously analyzed.
c. The proposed amendment will not involve a significant
reduction in the margin of safety.
The proposed amendment does not involve a decrease in the number
or capacity of safety/relief valves that are provided in the system,
nor does it involve any change in safety/relief valve setpoints,
operability requirements, or limiting conditions for operation.
Based on these considerations, we conclude the proposed amendment
will 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: Minneapolis Public Library,
Technology and Science Department, 300 Nicollet Mall, Minneapolis,
Minnesota 55401
Attorney for licensee: Gerald Charnoff, Esq., Shaw, Pittman, Potts
and Trowbridge, 2300 N Street, NW, Washington, DC 20037
NRC Project Director: L. B. Marsh

Northern States Power Company, Docket No. 50-263, Monticello
Nuclear Generating Plant, Wright County, Minnesota

Date of amendment request: January 4, 1994
Description of amendment request: The proposed amendment would
change Technical Specifications section 3.11, Reactor Fuel Assemblies,
by removing information concerning the analytical method to determine
average planar linear heat generation rate (APLHGR) and providing
reference to the presentation of the information in the Core Operating
Limits Report. In addition, this proposed amendment would change
section 6.7, Reporting Requirements, by revising the listing of
approved analytical methods for developing the Core Operating Limits
Report, and it would revise the Technical Specification Bases for
section 3.11 concerning the calculation methodology for MCPR [minimum
critical power ratio]. The proposed change to specification 3.11.A
would eliminate the duplication of requirements specified in
specification 6.7.A.7 and the Core Operating Limits Report for
establishing APLHGR limits.
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 amendment will not involve a significant increase
in the probability or consequences of an accident previously
evaluated.
The APLHGR limits originate from and are associated with LOCA
[loss-of-coolant accident] analyses. Standard exposure dependent
APLHGR limits are generated from LOCA analyses initiated from rated
power and flow conditions. For any allowable off power and off flow
condition the APLHGR limit is the smaller of the flow dependent or
power dependent limit. These limits are also used in the fuel
thermal-mechanical analysis and transient analysis. Flow dependent
APLHGR requirements will continue to be established based on
analysis and fuel type specific limits determined using NRC approved
methodologies to ensure that peak transient average planar heat
generation rate during these events is not increased above the fuel
design basis values. Power dependent APLHGR limits will continue to
be established based on analysis and fuel type specific limits
determined using NRC approved methodologies to ensure that peak
transient average planar heat generation rate during any transient
is not increased above the rated fuel design basis transient values.
The proposed amendment establishes appropriate controls to ensure
that the APLHGR limits will continue to be determined and
established using NRC approved methodology; therefore, this
amendment will not cause a significant increase in the probability
or consequences of an accident previously evaluated for the
Monticello plant.
The proposed amendment will not create the possibility of a new
or different kind of accident from any accident previously analyzed.
The proposed amendment does not involve any modifications to
plant equipment or operating procedures, nor will it introduce any
new failure modes. The proposed amendment ensures that cycle
specific APLHGR limits are determined and established using approved
methodologies and will not create the possibility of a new or
different kind of accident.
The proposed amendment will not involve a significant reduction
in the margin of safety.
The proposed amendment removes duplication which exists in the
Monticello Technical Specification for the identification of the
approved analytical methods for establishing the APLHGR core
operating limit. In addition the proposed amendment adds the NRC
approved Siemens' analytical method for the determination of APLHGR
limits based on LOCA/ECCS [emergency core cooling system] analyses.
Inclusion of the NRC approved Siemens' analytical method ensures
proper coordination of the methodology employed to establish the
APLHGR limiting condition for operation for each type of fuel as a
function of axial location and average planar exposure. APLHGR
limits will continue to be determined using NRC approved methodology
as established in specification 6.7.A.7.b. The established APLHGR
limits will be verified to be consistent with the accident analysis
contained in the Monticello Updated Safety Analysis Report. The
proposed amendment will 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: Minneapolis Public Library,
Technology and Science Department, 300 Nicollet Mall, Minneapolis,
Minnesota 55401
Attorney for licensee: Gerald Charnoff, Esq., Shaw, Pittman, Potts
and Trowbridge, 2300 N Street, NW, Washington, DC 20037
NRC Project Director: L. B. Marsh

Northern States Power Company, Docket Nos. 50-282 and 50-306,
Prairie Island Nuclear Generating Plant, Unit Nos. 1 and 2, Goodhue
County, Minnesota

Date of amendments request: September 21, 1992, as revised December
29, 1992, and November 24, 1993
Description of amendments requests: The proposed amendments would
change various Technical Specification (TS) sections and associated
Bases for surveillance test intervals and allowed outage times for the
engineered safety features and reactor protection system
instrumentation consistent with the NRC Staff position as documented in
NRC letters to the Westinghouse Owners Group.
The proposed license amendment request also updates operation modes
to be consistent with Westinghouse Standard Technical Specification
operational modes and also includes several editorial changes to the
Prairie Island TS that are unrelated to the changes described above.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. The proposed amendment will not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
The determination that the results of the proposed change are
within all acceptable criteria have been established in the SERs
prepared for WCAP-10271, WCAP-10271 Supplement 1, WCAP-10271
Supplement 2 and WCAP-10271 Supplement 2, Revision 1 issued by
References 1, 2, and 5 [of the November 24, 1993, application].
Implementation of the proposed changes is expected to result in an
acceptable increase in total Reactor Protection and Engineered
Safety Features Systems yearly unavailability. This increase, which
is primarily due to less frequent surveillance, results in a[n]
increase of similar magnitude in the probability of an Anticipated
Transient Without Scram (ATWS) and in the probability of core melt
resulting from an ATWS and also results in a small increase in core
damage frequency (CD) due to Engineered Safety Features
unavailability.
Implementation of the proposed changes is expected to result in
a significant reduction in the probability of core melt from
inadvertent reactor trips. This is a result of a reduction in the
number of inadvertent reactor trips (0.5 fewer inadvertent reactor
trips per unit per year) occurring during testing of Reactor
Protection System instrumentation. This reduction is primarily
attributable to less frequent surveillance.
The reduction in inadvertent core melt frequency is sufficiently
large to counter the increase in ATWS core melt probability
resulting in an overall reduction in total core melt probability.
The values determined by the Westinghouse Owners Group and
presented in the WCAP for the increase in core damage frequency were
verified by Brookhaven National Laboratory (BNL) as part of an audit
and sensitivity analyses for the NRC Staff. Based on the small value
of the increase compared to the range of uncertainty in the core
damage frequency, the increase is considered acceptable.
The changes of an editorial nature, including the change to
Standard Technical Specification format for the instrumentation
Technical Specifications and mode definitions, have no impact on the
severity or consequences of an accident previously evaluated.
The proposed changes do not result in an increase in the
severity or consequences of an accident previously evaluated.
Implementation of the proposed changes affects the probability of
failure of the Reactor Protection System and Engineered Safety
Features but does not alter the manner in which protection is
afforded nor the manner in which limiting criteria are established.
2. The proposed amendment will not create the possibility of a
new or different kind of accident from any accident previously
analyzed.
The proposed changes do not involve hardware changes and do not
result in a change in the manner in which the Reactor Protection
System and Engineered Safety Features provide plant protection. No
change is being made which alters the functioning of the Reactor
Protection System or Engineered Safety Features. Rather the
likelihood or probability of the Reactor Protection System or
Engineered Safety Features functioning properly is affected as
described above. Therefore the proposed changes do not create the
possibility of a new or different kind of accident from any accident
previously evaluated.
The changes of an editorial nature, including the change to
Standard Technical Specification format for the instrumentation
Technical Specifications and mode definitions does not create the
possibility of a new or different kind of accident from any
previously evaluated.
. The proposed amendment will not involve a significant
reduction in the margin of safety.
The proposed changes do not alter the manner in which safety
limits, limiting safety system setpoints or limiting conditions for
operation are determined. The impact of reduced testing other than
as addressed above is to allow a longer time interval over which
instrument uncertainties (e.g., drift) may act. Experience has shown
that the initial uncertainty assumptions are valid for reduced
testing.
Implementation of the proposed changes is expected to result in
an overall improvement in safety by:
a. Less frequent testing will result in less inadvertent reactor
trips and actuation of Engineered Safety Features components.
b. Higher quality repairs leading to improved equipment
reliability due to longer repair times.
c. Improvements in the effectiveness of the operating staff in
monitoring and controlling plant operation. This is due to less
frequent distraction of the operator and shift supervisor to attend
to instrumentation testing.
The changes of an editorial nature, including the change to
Standard Technical Specification format for the instrumentation
Technical Specifications and mode definitions [do] not lead to a
reduction in any margin of safety.
Local Public Document Room location: Minneapolis Public Library,
Technology and Science Department, 300 Nicollet Mall, Minneapolis,
Minnesota 55401
Attorney for licensee: Jay Silberg, Esq., Shaw, Pittman, Potts, and
Trowbridge, 2300 N Street, NW, Washington, DC 20037
NRC Project Director: L. B. Marsh

Omaha Public Power District, Docket No. 50-285, Fort Calhoun
Station,Unit No. 1, Washington County, Nebraska

Date of amendment request: December 28, 1993
Description of amendment request: The proposed amendment to the
Technical Specifications would revise the surveillance test frequency
from monthly to quarterly for several channel functional tests for
Reactor Protective System and Engineered Safety Feature Instrumentation
and Controls based on Generic Letter 93-05.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
The proposed change does not involve significant hazards
considerations because operation of Fort Calhoun Station Unit (FCS)
No. 1 in accordance with this change would not:
(1) Involve a significant increase in the probability or
consequences of an accident previously evaluated.
Increasing the surveillance test interval (STI) from monthly to
quarterly for the Reactor Protective System (RPS) and Engineered
Safety Features Actuation System (ESFAS) instrumentation has two
principal effects with opposing impacts on core melt risk. The first
impact is a slight increase in core melt frequency that results from
the increased unavailability of the instrumentation in question. The
unavailability of the tested instrumentation components is
translated to result in a failure of the reactor to trip, an
Anticipated Transient Without Scram (ATWS), or a failure of the
appropriate engineered safety features to actuate when required. The
opposing impact on core melt risk is the corresponding reduction in
core melt frequency that would result due to the reduced exposure of
the plant to test-induced transients. This results in a net decrease
in core melt frequency of approximately 4.1x10-8 per year.
Representative fault tree models were developed for FCS and the
corresponding changes in core melt frequency were quantified in
evaluations CEN-327-A and CEN-327-A, Supplement 1. The NRC issued a
Safety Evaluation Report (SER) which found that these evaluations
were acceptable for justifying the extensions in the STIs for the
RPS and ESFAS from 30 days to 90 days and that the RPS
unavailabilities resulting from extending the STIs were not
considered to be significant. Estimates of the reduction in scram
frequency from the reduction in test-induced scrams and the
corresponding reduction in core melt frequency were found
acceptable. STIs of 90 days were found to result in a net reduction
in core melt risk.
A plant specific calculation/setpoint drift analysis was
conducted, as required by the NRC SER, that analyzed the effect on
instrument drift of extending the RPS and ESF instrumentation and
controls functional STI from monthly to quarterly. The results
demonstrated that the observed changes in instrument uncertainties
for the extended STI do not exceed the current 30-day setpoint
assumptions. Therefore, it is unnecessary to change any setpoints to
accommodate the proposed extended STI.
Operation of the facility in accordance with this proposed
change, therefore, will not involve a significant increase in the
probability or consequences of an accident previously evaluated.
(2) Create the possibility of a new or different kind of
accident from any accident previously evaluated.
The proposed change does not involve any changes in equipment
and will not alter the manner in which the plant will be operated.
RPS and ESFAS setpoints will not be changed as the instrument
uncertainties resulting from the proposed STI (calculated using
actual plant data) are less than the instrument uncertainties
assumed for 30 days. Thus, this proposed change will not create the
possibility of a new or different kind of accident from any
previously evaluated.
(3) Involve a significant reduction in a margin of safety.
There are no changes to the equipment or plant operations. RPS
and ESFAS setpoints will not be changed as the instrument
uncertainties resulting from the proposed STI (calculated using
actual plant data) are less that the instrument uncertainties
assumed for 30 days.
Implementation of the proposed changes is expected to result in
an overall improvement in plant safety due to the fact that reduced
testing intervals will result in fewer inadvertent reactor trips and
less frequent actuation of ESFAS components. The conclusions of the
accident analyses in the FCS Updated Safety Analysis Report (USAR)
remain valid and the safety limits continue to be met. Thus, this
proposed change does not reduce a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: W. Dale Clark Library, 215
South 15th Street, Omaha, Nebraska 68102
Attorney for licensee: LeBoeuf, Lamb, Leiby, and MacRae, 1875
Connecticut Avenue, N.W., Washington, D.C. 20009-5728NRC Project
Director:
William D. Beckner

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: December 28, 1993
Description of amendment request: The proposed amendment to the
James A. FitzPatrick Technical Specifications (TSs) clarifies Limiting
Condition for Operation (LCO) 3.5.D.4. Amendment No. 179 to the TS
added LCO 3.5.D.4 to permit hydrostatic and leakage testing at
temperatures up to 300 deg.F without requiring certain equipment,
including the automatic depressurization system (ADS), to be operable.
However, LCO 3.5.D.4 can be mistakenly interpreted to require the ADS
be operable at temperatures less than 212 deg.F. Requiring the ADS to
be operable during hydrostatic and leakage testing with temperatures
below 212 deg.F was clearly not the intent of Amendment No. 179. The
proposed change will clarify LCO 3.5.D.4 to resolve this concern.
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 plant accident analyses are not affected by the proposed
Technical Specification change. Prior to implementation of Amendment
179, hydrostatic and leakage testing of the RCS was performed with
reactor coolant temperatures below 212 deg.F while the ADS was
inoperable. Amendment 179 revised the Technical Specifications in
anticipation of increased pressure temperature limits requiring
hydrostatic and leakage testing at or above 212 deg.F. Requiring the
ADS to be operable during hydrostatic or leakage testing with
temperatures below 212 deg.F was clearly not the intent of Amendment
179. The change will not increase the probability or consequences of
previously evaluated accidents.
2. create the possibility of a new or different kind of accident
from those previously evaluated.
The proposed change involves no modifications to hardware,
analyses, operations or procedures. The change clarifies LCO 3.5.D.4
to allow hydrostatic and leakage testing of the RCS below 300 deg.F
without requiring the ADS to be operable. The change is
administrative in nature since it only clarifies the intent of the
Technical Specifications as agreed to with the NRC and cannot create
a new or different kind of accident.
3. involve a significant reduction in the margin of safety.
The proposed change will not affect any plant safety margins.
The existing plant accident analyses are not affected by the
proposed change. This revision of LCO 3.5.D.4 is intended to clarify
that the ADS is not required to be operable during hydrostatic or
leakage testing of the RCS. This position is substantiated by the
NRC safety evaluation for Amendment 179 which acknowledges that
hydrostatic and leakage testing can not be performed without making
the ADS, and other systems, inoperable.
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: Robert A. Capra

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 31, 1994
Description of amendment request: The proposed amendment to the
James A. FitzPatrick Technical Specifications would revise the limiting
conditions for operation (LCO), surveillance requirements, and Bases
section for the main condenser steam jet air ejectors (SJAE). The
proposed changes correct a typographical error, clarify the modes of
operation during which the SJAE LCOs and surveillance requirements are
applicable, revise the action required upon entering a SJAE LCO, and
establish a threshold level below which there will be no requirement to
perform grab samples and isotopic analyses of SJAE effluent.
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 amendment revision involves no hardware changes, no
changes to the operation of any systems or components and no changes
to structures. The changes clarify the Technical Specifications by
specifying the modes of operation during which the LCOs and
Surveillance Requirements of Specification 3.5 are applicable. The
changes also include specific guidance for the operators to prevent
or minimize the release of radioactive gases to the environment.
These changes can not cause an increase in the probability of, nor
alter the consequences of, an accident previously evaluated.
The establishment of a threshold below which grab samples are
not required will alter procedures by allowing SJAE operation
without grab samples to determine effluent content at low levels of
radioactivity (i.e., less than 5,000 micro Ci/sec). This will not
affect the monitoring system's ability to detect, alarm, and isolate
the offgas system if the concentration of radioactive material in
the effluence reaches the appropriate setpoint.
The surveillance requirement for taking a grab sample after a
greater than 50% increase in release rate is intended to assist
operators in determining if there is any increase in fuel failure
during steady state operations. This would assure that routine
operational limits are maintained. The grab samples do not provide
any automatic protective function (e.g., MSIV [main steam isolation
valve] or Offgas System isolation) for mitigating an accident but
provide radionuclide concentration data.
The performance of SJAE effluent grab samples is not credited
towards detecting nor mitigating any design basis accidents since
spontaneous fuel failure is not a FSAR [Final Safety Analysis
Report] accident initiator but a consequence of an accident.
Therefore, the use of a 5,000 micro Ci/sec threshold, which is
approximately 1% of the trip setpoint, would not alter the
consequences or probabilities of established accident scenarios.
2. create the possibility of a new or different kind of accident
from those previously evaluated.
The proposed changes provide improved clarity concerning
applicability of the specifications and specific guidance for
preventing/mitigating the release of radioactive gases to the
environment. The proposed changes also provide guidance for limiting
the number of unnecessary grab samples.
These changes do not affect the manner in which the main
condenser steam jet air ejector is operated. The proposed changes to
the Technical Specifications reflect either established plant
practice (i.e., applicable modes or mitigation procedures) or new
surveillance guidelines to minimize unnecessary grab samples. In all
cases, the proposed changes have no affect on any parameters which
would be considered or used in an accident analysis. The changes,
therefore, do not pose a safety issue different from those analyzed
previously for the FSAR.
3. involve a significant reduction in the margin of safety.
The proposed changes to the Technical Specifications will not
alter the intent of the surveillance requirement to monitor for the
possibility of fuel failure. Considering the difference between the
proposed threshold value and the current alarm setpoint, a reduction
in grab samples during plant operation with low concentrations of
radioactivity in the primary coolant will not affect any plant
safety m

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Source: Frix Law Library, https://www.frixlaw.com/law-library/documents/fr%3A94-4562. Public record. Not legal advice.