# Biweekly Notice; 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%3A98-11911

## Record

- **Collection:** Federal Register
- **Document type:** Notice
- **Published:** May 6, 1998
- **Citation:** 63 FR 25101

## 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. Pub. L. 97-415 revised section 189 of the Atomic
Energy Act of 1954, as amended (the Act), to require the Commission to
publish notice of any amendments issued, or proposed to be issued,
under a new provision of section 189 of the Act. This provision grants
the Commission the authority to issue and make immediately effective
any amendment to an operating license upon a determination by the
Commission that such amendment involves no significant hazards
consideration, notwithstanding the pendency before the Commission of a
request for a hearing from any person.
This biweekly notice includes all notices of amendments issued, or
proposed to be issued from April 10 through April 24, 1998. The last
biweekly notice was published on April 22, 1998 (63 FR 19964).

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.

[[Page 25102]]

However, should circumstances change during the notice period such that
failure to act in a timely way would result, for example, in derating
or shutdown of the facility, the Commission may issue the license
amendment before the expiration of the 30-day notice period, provided
that its final determination is that the amendment involves no
significant hazards consideration. The final determination will
consider all public and State comments received before action is taken.
Should the Commission take this action, it will publish in the Federal
Register a notice of issuance and provide for opportunity for a hearing
after issuance. The Commission expects that the need to take this
action will occur very infrequently.
Written comments may be submitted by mail to the Chief, Rules and
Directives Branch, Division of Administration Services, Office of
Administration, U.S. Nuclear Regulatory Commission, Washington, DC
20555-0001, and should cite the publication date and page number of
this Federal Register notice. Written comments may also be delivered to
Room 6D22, Two White Flint North, 11545 Rockville Pike, Rockville,
Maryland from 7:30 a.m. to 4:15 p.m. Federal workdays. Copies of
written comments received may be examined at the NRC Public Document
Room, the Gelman Building, 2120 L Street, NW., Washington, DC. The
filing of requests for a hearing and petitions for leave to intervene
is discussed below.
By June 5, 1998, the licensee may file a request for a hearing with
respect to issuance of the amendment to the subject facility operating
license and any person whose interest may be affected by this
proceeding and who wishes to participate as a party in the proceeding
must file a written request for a hearing and a petition for leave to
intervene. Requests for a hearing and a petition for leave to intervene
shall be filed in accordance with the Commission's ``Rules of Practice
for Domestic Licensing Proceedings'' in 10 CFR part 2. Interested
persons should consult a current copy of 10 CFR 2.714 which is
available at the Commission's Public Document Room, the Gelman
Building, 2120 L Street, NW., Washington, DC and at the local public
document room for the particular facility involved. If a request for a
hearing or petition for leave to intervene is filed by the above date,
the Commission or an Atomic Safety and Licensing Board, designated by
the Commission or by the Chairman of the Atomic Safety and Licensing
Board Panel, will rule on the request and/or petition; and the
Secretary or the designated Atomic Safety and Licensing Board will
issue a notice of a hearing or an appropriate order.
As required by 10 CFR 2.714, a petition for leave to intervene
shall set forth with particularity the interest of the petitioner in
the proceeding, and how that interest may be affected by the results of
the proceeding. The petition should specifically explain the reasons
why intervention should be permitted with particular reference to the
following factors: (1) The nature of the petitioner's right under the
Act to be made a party to the proceeding; (2) the nature and extent of
the petitioner's property, financial, or other interest in the
proceeding; and (3) the possible effect of any order which may be
entered in the proceeding on the petitioner's interest. The petition
should also identify the specific aspect(s) of the subject matter of
the proceeding as to which petitioner wishes to intervene. Any person
who has filed a petition for leave to intervene or who has been
admitted as a party may amend the petition without requesting leave of
the Board up to 15 days prior to the first prehearing conference
scheduled in the proceeding, but such an amended petition must satisfy
the specificity requirements described above.
Not later than 15 days prior to the first prehearing conference
scheduled in the proceeding, a petitioner shall file a supplement to
the petition to intervene which must include a list of the contentions
which are sought to be litigated in the matter. Each contention must
consist of a specific statement of the issue of law or fact to be
raised or controverted. In addition, the petitioner shall provide a
brief explanation of the bases of the contention and a concise
statement of the alleged facts or expert opinion which support the
contention and on which the petitioner intends to rely in proving the
contention at the hearing. The petitioner must also provide references
to those specific sources and documents of which the petitioner is
aware and on which the petitioner intends to rely to establish those
facts or expert opinion. Petitioner must provide sufficient information
to show that a genuine dispute exists with the applicant on a material
issue of law or fact. Contentions shall be limited to matters within
the scope of the amendment under consideration. The contention must be
one which, if proven, would entitle the petitioner to relief. A
petitioner who fails to file such a supplement which satisfies these
requirements with respect to at least one contention will not be
permitted to participate as a party.
Those permitted to intervene become parties to the proceeding,
subject to any limitations in the order granting leave to intervene,
and have the opportunity to participate fully in the conduct of the
hearing, including the opportunity to present evidence and cross-
examine witnesses.
If a hearing is requested, the Commission will make a final
determination on the issue of no significant hazards consideration. The
final determination will serve to decide when the hearing is held.
If the final determination is that the amendment request involves
no significant hazards consideration, the Commission may issue the
amendment and make it immediately effective, notwithstanding the
request for a hearing. Any hearing held would take place after issuance
of the amendment.
If the final determination is that the amendment request involves a
significant hazards consideration, any hearing held would take place
before the issuance of any amendment.
A request for a hearing or a petition for leave to intervene must
be filed with the Secretary of the Commission, U.S. Nuclear Regulatory
Commission, Washington, DC 20555-0001, Attention: Rulemakings and
Adjudications Staff, or may be delivered to the Commission's Public
Document Room, the Gelman Building, 2120 L Street, NW., Washington DC,
by the above date. A copy of the petition should also be sent to the
Office of the General Counsel, U.S. Nuclear Regulatory Commission,
Washington, DC 20555-0001, and to the attorney for the licensee.
Nontimely filings of petitions for leave to intervene, amended
petitions, supplemental petitions and/or requests for a hearing will
not be entertained absent a determination by the Commission, the
presiding officer or the Atomic Safety and Licensing Board that the
petition and/or request should be granted based upon a balancing of
factors specified in 10 CFR 2.714(a)(1)(i)-(v) and 2.714(d).
For further details with respect to this action, see the
application for amendment which is available for public inspection at
the Commission's Public Document Room, the Gelman Building, 2120 L
Street, NW., Washington, DC, and at the local public document room for
the particular facility involved.

[[Page 25103]]

Carolina Power & Light Company, et al., Docket Nos. 50-325 and 50-324,
Brunswick Steam Electric Plant, Units 1 and 2, Brunswick County, North
Carolina

Date of amendment request: November 1, 1996, as supplemented by
letters dated October 13, 1997, February 26, 1998, and March 13, 1998.
Description of amendment request: Associated with a Carolina Power
& Light Company (the licensee) application to convert from the Current
Technical Specifications (CTS) for the Brunswick Steam Electric Plant,
Units 1 and 2, to Improved Technical Specifications (ITS), as contained
in Revision 1 of NUREG-1433, ``Standard Technical Specification General
Electric Plants, BWR/4,'' the licensee proposed removing a restriction
on a surveillance test described below.
CTS 4.8.1.1.1.b requires that the offsite electrical power circuits
be demonstrated OPERABLE, at least once per 18 months during shut down,
by manually transferring the unit power supply from the normal circuit
to the alternate circuit. As proposed, ITS SR 3.8.1.8.b will not
contain the restriction to perform the Surveillance ``during
shutdown.'' Currently, this test is performed by momentarily
paralleling the 230 kV offsite alternating current (AC) power sources.
The licensee has stated that paralleling offsite AC power sources is a
controlled evolution and the increased risk associated with the
performance of this test while the unit is at power is not significant
for the following reasons: (1) the frequency and voltages are verified
to be within the required range prior to paralleling the two offsite AC
power sources; (2) breaker interlocks ensure that the alternate circuit
is connected to the load prior to opening the preferred circuit; (3)
the test does not result in de-energization of any 4.16 kV emergency
bus and the potential for electrical perturbations on the grid system
is the same whether performing the transfer while the unit is at power
or while shutdown; and (4) operating history indicates that
transferring offsite AC power sources while the units were in
Operational Conditions 1 (power operation) or 2 (startup) has been
performed satisfactorily without electrical distribution system
perturbations. The licensee has further pointed out that Generic Letter
91-04, ``Changes in Technical Specifications to Accommodate a 24-Month
Fuel Cycle,'' states that licensees may omit the Technical
Specification qualification that a refueling interval surveillance is
to be performed ``during shutdown.'' Therefore, consistent with the
guidance provided in Generic Letter 91-04, the licensee proposed
deletion of the requirement to perform this Surveillance ``during
shutdown'' as part of the conversion from CTS 4.8.1.1.1.b to ITS SR
3.8.1.8.b.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:

1. Does the change involve a significant increase in the
probability or consequences of an accident previously evaluated?
This change would remove a specific restriction to perform the
verification of the manual transfer of the unit power supply from
the normal circuit to the alternate circuit ``during shutdown.'' The
transfer of the unit power supply from the normal circuit to the
alternate circuit is not an initiator of any previously analyzed
accident. Therefore, this change does not significantly increase the
frequency of such accidents. Currently, this test is performed by
momentarily paralleling the 230 kV offsite AC power sources.
Paralleling offsite AC power sources is a controlled evolution and
the increased risk associated with the performance of this test
while the unit is at power is not significant for the following
reasons: (1) The frequency and voltages are verified to be within
the required range prior to paralleling the two offsite AC power
sources; (2) breaker interlocks ensure that the alternate circuit is
connected to the load prior to opening the preferred circuit; (3)
the test does not result in de-energization of any 4.16 kV emergency
bus and the potential for electrical perturbations on the grid
system is the same whether performing the transfer while the unit is
at power or while shutdown; and (4) operating history indicates that
transferring offsite AC power sources while the units were in MODE
(Operational Condition) 1 or 2 has been performed satisfactorily
without electrical distribution system perturbations. The
appropriate plant conditions for performance of the Surveillance
will continue to be controlled to assure the potential consequences
are not significantly increased. This control method has been
previously determined to be acceptable as indicated in Generic
Letter 91-04. Therefore, this change does not significantly increase
the consequences of any previously analyzed accident.
2. Does the change create the possibility of a new or different
kind of accident from any accident previously evaluated?
This change removes a specific restriction on the plant
conditions for performing a Surveillance, but does not change the
method of performance. The appropriate plant conditions for
performance of the Surveillance will continue to be controlled to
assure the possibility for a new or different kind of accident are
not created. This control method has been previously determined to
be acceptable as indicated in Generic Letter 91-04. Therefore, this
change does not create the possibility of a new or different kind of
accident from any previously analyzed accident.
3. Does this change involve a significant reduction in a margin
of safety?
The margin of safety considered in determining the appropriate
plant conditions for performing the Surveillance will continue to be
controlled to assure that there is no significant reduction. This
control method has been previously determined to be acceptable as
indicated in Generic Letter 91-04. Therefore, the 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: University of North Carolina at
Wilmington, William Madison Randall Library, 601 S. College Road,
Wilmington, North Carolina 28403-3297
Attorney for licensee: William D. Johnson, Vice President and Senior
Counsel, Carolina Power & Light Company, Post Office Box 1551, Raleigh,
North Carolina 27602
NRC Project Director: Pao-Tsin Kuo

Carolina Power & Light Company, et al., Docket Nos. 50-325 and 50-324,
Brunswick Steam Electric Plant, Units 1 and 2, Brunswick County, North
Carolina

Date of amendment request: April 3, 1998.
Description of amendment request: The Carolina Power & Light
Company, licensee for the Brunswick Steam Electric Plant (BSEP), Unit
Nos. 1 and 2, proposed amendments to the Technical Specifications (TS)
to change the specified total volume of the condensate storage tank
(CST) from 150,000 gallons to 228,200 gallons. During a recent review
of industry operating experience, the licensee determined that
information contained in TS 3.5.3.1, Core Spray System (CSS), and the
associated bases regarding water inventory in the CST was incorrect.
Specifically, the minimum CST volume requirement contained in TS
3.5.3.1 would not assure the availability of 50,000 gallons of water
for the CSS, as indicated in TS Bases section 3/4.5.3.1 for the CSS.
The licensee has concluded that the proposed license amendments do
not involve a Significant Hazards Consideration. In support of this
determination, an evaluation of each of the three standards set forth
in 10 CFR 50.92 is provided below.

[[Page 25104]]

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 license amendments do not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
The proposed TS change revises the minimum CST [Condensate
Storage Tank] water volume required for OPERABILITY of the Core
Spray system (CSS) in OPERATIONAL CONDITIONS 4 AND 5 when the
suppression pool is inoperable. The proposed change does not alter
the operation of any plant system or component; does not involve a
physical modification to any structure, system, or component; and
does not affect an initiator to any accident previously evaluated.
The minimum CST water level is being increased to assure the
availability of 50,000 gallons of water for use by the CSS.
Therefore, the proposed license amendments do not involve an
increase in the probability or consequences of an accident
previously evaluated.
2. The proposed license amendments will not create the
possibility of a new or different kind of accident from any accident
previously evaluated. This proposed TS change revises the minimum
CST water volume required for OPERABILITY of the CSS in OPERATIONAL
CONDITIONS 4 and 5 when the suppression pool is inoperable. The
proposed change does not alter the operation of any plant system or
component; does not involve a physical modification to any
structure, system, or component; and does not affect an initiator to
any accident previously evaluated. The proposed change does not add
or modify equipment or components related to the CSS and will,
therefore, not create new failure modes or common failure modes. The
minimum CST water level is being increased to assure the
availability of 50,000 gallons of water for use by the CSS.
Therefore, the proposed license amendments do not create the
possibility of a new or different kind of accident from any accident
previously evaluated.
3. The proposed license amendments do not involve a significant
reduction in a margin of safety. The proposed license amendments
increase the minimum CST water level to assure the availability of
50,000 gallons of water for use by the CSS. These volumes ensure the
validity of existing analyses, and ensure that the existing TS Bases
are satisfied. The proposed change does not involve a physical
modification to any structure, system, or component, and does not
modify the operation of any existing equipment. Therefore, the
proposed license amendments do not involve a reduction in a margin
of safety.

The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendments request involves no significant hazards consideration.

Local Public Document Room location: University of North Carolina at
Wilmington, William Madison Randall Library, 601 S. College Road,
Wilmington, North Carolina 28403-3297
Attorney for licensee: William D. Johnson, Vice President and Senior
Counsel, Carolina Power & Light Company, Post Office Box 1551, Raleigh,
North Carolina 27602
NRC Project Director: Pao-Tsin Kuo (Acting)

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: March 31, 1998.
Description of amendment request: Unreviewed Safety Question
involving use of Station Blackout (SBO) diesel generators (DGs) and use
of a mobile safe shutdown (SSD) battery cart in the 10 CFR part 50,
appendix R, Safe Shutdown Safety Analysis.
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 licensee has provided a separate no significant hazards
consideration determination for the SBO DGs and the battery cart under
this amendment request. The following is the determination for the SBO
DGs:

(1) No significant increase in the probability or consequences
of an accident previously evaluated is involved because of the
following:
Two types of previously evaluated accidents are relevant to this
criterion: (1) A fire; (2) other accident evaluated in the UFSAR.
For these previously evaluated accidents, the change would not
result in an increase in either their probabilities of occurrence or
the consequences of their occurrence, for the following reasons.
The use of the SBO DGs in lieu of the [Emergency Diesel
Generators] EDGs does not change the probability or consequences of
a fire. The likelihood of a fire is unchanged. Use of the SBO DGs
does not significantly change the fire loading nor introduce
significant new ignition sources. The consequences of a fire are
unchanged because use of the SBO DGs continues to support the
station's ability to achieve and maintain shutdown in the event of a
fire.
Use of the SBO DGs for non-fire purposes is unchanged by use of
the SBO DGs for post-fire safe shutdown in the event of a fire in
areas requiring alternate shutdown capability. Accordingly there is
no change in the probability or consequences of a previously
evaluated accident involving the SBO DGs. Similarly, there is no
change to the probability or consequences of other accidents that
have been previously evaluated because they are independent of this
change in use of the SBO DGs.
(2) The possibility of a new or different kind of accident from
any accident previously evaluated is not created because:
The proposed change does not create the possibility of a new or
different kind of accident from that previously evaluated for Quad
Station. Although the SBO DGs will be used for a new function, there
is no significant change in the operation of the SBOs for a non-fire
event. Moreover, the overall use of the SBO DGs as an AC power
source is not significantly different from the use of the EDGs. The
SBO DGs buses provide power to the same buses that are powered from
the EDGs. No new modes of operation are introduced by the proposed
changes. The use of the SBO DGs provides a slightly different but
effective method for achieving and maintaining post-fire safe
shutdown for areas requiring alternate shutdown capability. As such,
the proposed change does not create the possibility of a new or
different kind of accident.
(3) No significant reduction in the margin of safety is involved
because:
A change in the fire protection program does not result in a
significant reduction in the margin of safety if the change does not
result in a significant adverse impact on the plant's ability to
achieve and maintain safe shutdown in the event of a fire. The
proposed use of the SBO DGs instead of the EDGs to achieve and
maintain safe shutdown within 72 hours change does not significantly
affect the capability or reliability of the equipment assumed to
operate in the safety analysis.
The demonstrated capability and reliability of the SBO and EDGs
are not significantly different. Indeed, the SBO DGs represent a
safety improvement due to their physical separation from the
postulated fire areas, and the operational benefits provided by
their greater capacity. Any narrow reduction in margin associated
with the need to manually start the SBO DGs is offset by the
reduction in manual actions necessary to reduce electrical loads
powered from the EDGs. The lack of Class 1E qualification for the
SBO DGs is not significant from a safety perspective because the
demonstrated reliability of the SBO DGs is comparable to the
reliability of the EDGs. The lack of seismic qualification and
single failure protection do not constitute a significant reduction
in margin since neither of these attributes is required by Appendix
R. Accordingly, the Commission has already determined that these
attributes are not part of the Appendix R acceptance criterion. Any
reduction in margin associated with the greater fuel consumption
rate of the SBO DGs is partially offset by the increased flexibility
in powering equipment to achieve and maintain post fire safe
shutdown. Additionally, onsite fuel storage and manual transfer
capabilities provide for at least 72 hours of SBO DG operation.
Within 72 hours, deliveries of diesel fuel from offsite supplies is
expected. Therefore, the use of the SBO DGs as an onsite AC power
source for

[[Page 25105]]

equipment necessary to achieve and maintain post-fire safe shutdown
in areas requiring alternate capabilities does not involve a
significant reduction in margin.

The licensee has evaluated the use of the mobile SSD battery cart to
provide the power source for the Automatic Depressurization System
(ADS) valves under certain scenarios where the valves are needed to
achieve cold shutdown and determined that it does not involve a
significant hazards consideration for the reasons discussed below.

(1) No significant increase in the probability or consequences
of an accident previously evaluated is involved.
The accident previously evaluated is the postulated fire
requiring alternate shutdown capability. The probability of a
previously evaluated fire is not increased significantly because the
mobile SSD batteries do not create significant new ignition sources
or any other fire initiators. The consequences of a previously
evaluated fire are not increased significantly because the mobile
SSD batteries do not significantly increase the fire loading in the
plant, do not interfere with the plant's ability to extinguish a
fire, and are fully capable of fulfilling the designed safety
function.
The associated systems related to this proposed change are not
affected in a way that could impact the initiation of any accident
sequence for the Quad Cities Station. No modes of operation are
introduced by the proposed change such that adverse consequences
result.
The probability of an accident involving the use of the mobile
SSD batteries would not be increased significantly by this proposed
use because the use is not significantly different from the
alternative manual attachment of a power source to the ADS valves.
The consequences of an accident involving the use of the mobile
SSD batteries are not increased because the only significant
consequences would be a delay in achieving cold shutdown and that
would have no different consequences than would a delay due to an
accident related to the currently used manual power source.
(2) The possibility of a new or different kind of accident from
any accident previously evaluated is not created.
The proposed change for the Quad Cities Station does not create
the possibility of a new or different kind of accident from that
previously evaluated. Because the mobile SSD batteries simply
provide a different form of manually connecting a source of power to
the ADS valves, the use of the mobile SSD batteries does not present
new or different kinds of accidents related to such manual actions.
Finally, because no new modes of operation are introduced by the
proposed change, the change does not create the possibility of a new
or different kind of accident that could be related to new modes of
operation.
(3) No significant reduction in the margin of safety is
involved.
The analytic framework for determining the extent to which a
proposed change affects the margin of safety has been discussed
above and, so will not be repeated here. In this case, a review of
the proposed changes shows that they will not have an adverse impact
on the ability to achieve and maintain safe shutdown. Several
features associated with the use of the mobile SSD batteries show,
as discussed above, that it provides an effective method for
achieving and maintaining safe shutdown following a fire. In
particular, use of the mobile SSD batteries reduces the overall
complexity of the cold shutdown repairs required to supply power to
the ADS valves and is familiar to plant personnel from their
training on its use for other purposes.
Design calculations regarding capabilities of the mobile SSD
batteries show they will be capable in fulfilling their intended
safety function for their design basis Appendix R scenario.
Reliability of the mobile SSD batteries will be maintained by
augmented quality standards. This will entail the conduct of
appropriate maintenance and surveillance which is designed to ensure
that the mobile batteries will function as intended. Reliability of
this power source is further enhanced by the circumstance that there
are two mobile SSD batteries, thus permitting one to act as a backup
to the other.
Under these circumstances, the margin of safety for achieving
cold shutdown using the ADS valves is not reduced significantly, if
at all, by the use of non-safety related mobile SSD batteries to
power the ADS valves. Although safety-related station batteries had
previously been used in this function, the method for attaching
those batteries was more prone to human error than the method which
has been developed for the mobile SSD batteries. Moreover,
substantial steps have been taken to provide a high level of
reliability for the mobile SSD batteries. Overall, therefore, the
ability to achieve and maintain safe shutdown in the event of a fire
has not been reduced by this change in the source of power to the
ADS valves.

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 are
satisfied. Therefore, the NRC staff proposes to determine that the
requested amendments involve no significant hazards consideration.

Local Public Document Room location: Dixon Public Library, 221 Hennepin
Avenue, Dixon, Illinois 61021
Attorney for licensee: Michael I. Miller, Esquire; Sidley and Austin,
One First National Plaza, Chicago, Illinois 60603
NRC Project Director: Stuart A. Richards

Commonwealth Edison Company, Docket Nos. 50-295 and 50-304, Zion
Nuclear Power Station, Units 1 and 2, Lake County, Illinois

Date of amendment request: March 30, 1998.
Description of amendment request: The proposed amendments would
restore the Zion Custom Technical Specifications (CTS) that had been
replaced with Improved Technical Specification by a previous amendment
and would reinstate License Conditions that were deleted by that
previous amendment. The proposed amendment would also modify the CTS to
allow the use of Certified Fuel Handlers to satisfy shift staffing
requirements and would change management titles and responsibilities to
reflect the permanently shutdown organization.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
Does the change involve a significant increase in the
probability or consequences of an accident previously evaluated?
With a plant permanently shutdown and defueled the spectrum of
accidents and events that remain credible is significantly reduced.
As discussed below the proposed changes do not affect the
probability or consequences of any accidents that do remain
credible.
The restoration of the CTS which were replaced with the ITS by
Amendments 178/165 cannot increase the probability or consequences
of any event or accident because the amendment was never
implemented. The CTS have been maintained as the legally binding
Technical Specifications in effect at Zion Station. The
reinstatement of the five License Conditions deleted by Amendments
178/165 is an administrative change in that the requirements
contained in the License Conditions had been relocated elsewhere and
are now being restored exactly as they were before the amendment was
issued. Since the actual requirements have not changed there can be
no change in the probability or consequences of any accident or
event.
The changes in management titles and responsibilities will not
increase the probability or consequences of any accident or event
because these changes are administrative and will not result in any
decrease in the quality of management applied to Zion Station. The
changes are commensurate with the significant reduction in site
activities, site staffing, and risk to public health and safety that
occurs when an operational nuclear power plant transitions to a
permanently shutdown and defueled plant. Responsible individuals
will have the authority to commit the personnel and resources
necessary to fulfill their obligations for safe storage and handling
of nuclear fuel. The change of position designations will have no
effect on the frequency of occurrence of accident or event
initiators, or on their consequences.
The changes to allow use of Certified Fuel Handlers in lieu of
personnel licensed in accordance with 10 CFR part 55 will not
increase the probability or consequences of an accident or event
because the Certified Fuel Handler Training and Retraining program
(which will be approved by the

[[Page 25106]]

NRC) has been developed using a Systems Approach to Training as
defined in 10 CFR 55.4. This approach provides assurance that the
Certified Fuel Handlers have the knowledge, skills, and abilities
that are commensurate with the tasks to be performed (i.e., the
proper monitoring, handling, storage, and cooling of nuclear fuel).
Therefore the frequency of occurrence of accident or event
initiators is not increased and the consequences of the accidents or
events are unaffected.
The changes in shift staffing numbers and crew composition will
not increase the probability or consequences of an accident or
event. These staffing changes are commensurate with the quantity,
complexity, and hazard level of the activities required for storage
and handling of nuclear fuel. The elimination of the Shift Control
Room Engineer does not affect any accident or event initiator or
consequence since the previous specification would not have required
that the position be manned with both units shut down. The
elimination of the requirement for a Radiation Protection Person on
shift will have no effect on the frequency of occurrence of
accidents or events, nor on the consequences of the accident or
event.
The changes in verbiage to eliminate any implication that units
are operational will not increase the probability or consequences of
an accident or event because they are largely editorial changes and
do not increase the frequency of occurrence of [or] event
initiators, nor do they increase the consequences.
Therefore this proposed amendment does not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
Does the change create the possibility of a new or different
kind of accident from any accident previously evaluated?
The changes proposed by this amendment do not involve new
structures, systems, or components, or the use of existing
structures, systems, or components in a new manner. Consequently no
new failure mechanisms are introduced. The design and operation of
structures, systems, or components is unaffected by:
The restoration of CTS,
The reinstatement of the five License Conditions deleted by
Amendments 178/165,
The changes in management titles and responsibilities,
The changes to allow use of Certified Fuel Handlers in lieu of
10 CFR [Part] 55 licensed personnel,
The changes in shift staffing numbers and crew composition, or
The changes in verbiage to eliminate any implication that units
are operational.
Therefore, the proposed changes do not create the possibility of
a new or different kind of accident from any previously evaluated.
Does the change involve a significant reduction in a margin of
safety?
One of the License Conditions that would be reinstated by this
amendment establishes limits that help ensure that the assumptions
of the fuel handling accident analysis remain valid. License
Condition 2.C.(7).b limits the weight of loads carried over fuel
stored in the spent fuel pool to the weight of a single fuel
assembly plus the tool for moving that assembly. This weight limit
ensures that the number of fuel rods broken in a fuel handling
accident does not exceed the maximum number of fuel rods assumed to
break in the accident analysis. Consequently, this change continues
to provide assurance that the margin of safety involving the number
of fuel rods broken in the accident will not be reduced.
Therefore, these changes do not involve a significant reduction
in the margin of safety.

The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
requested amendments involve no significant hazards consideration.

Local Public Document Room location: Waukegan Public Library, 128 N.
County Street, Waukegan, Illinois 60085
Attorney for licensee: Michael I. Miller, Esquire; Sidley and Austin,
One First National Plaza, Chicago, Illinois 60603
NRC Project Director: Stuart A. Richards

Duke Energy Corporation (DEC), et al., Docket Nos. 50-413 and 50-414,
Catawba Nuclear Station, Units 1 and 2, York County, South Carolina

Date of amendment request: May 27, 1997, as supplemented by a
letter dated April 20, 1998.
Description of amendment request: The proposed amendments would
revise the Technical Specifications (TS) of each unit to conform with
NUREG-1431, Revision 1, ``Standard Technical Specifications--
Westinghouse Plants.'' The Commission had previously issued a Notice of
Consideration of Issuance of Amendments in the Federal Register on July
14, 1997 (62 FR 37628) covering all the proposed changes that were
indeed within the scope of NUREG-1431. In DEC's May 27, 1997,
submittal, there are proposed changes that are beyond the scope of
NUREG-1431, which were thus not covered by the staff's July 14, 1997,
notice. The following descriptions and no significant hazard analyses
cover only those beyond-scope changes. Associated with each change are
administrative/editorial changes such that the new or revised
requirements would fit into the format of NUREG-1431.
1. This proposed change affects the surveillance requirement
currently contained in Sections 4.6.6.1 and 4.6.6.2, regarding the
containment valve injection water system. The requirement to assure
adequate capacity to maintain system pressure for at least 30 days
would be deleted, the required system pressure of 16.2 pounds per
square inch gauge (psig) would be replaced with a surge tank pressure
of 36.4 psig, and the system would be tested at lower pressures and
more restrictive leak rates.
2. Section 3.9.2.1, regarding the boron dilution mitigating system,
currently requires both trains to be operable in Mode 6 (refueling).
DEC proposed to add a note stating that the system may be blocked
during core reloading until two assemblies are loaded into the core.
Adequate shutdown margin will continue to be controlled and verified by
other specifications. This blocking would prevent inadvertent actuation
of the system, which could distract the operating personnel, but would
not diminish the monitoring function of the system.
3. DEC proposed to change the definition of `dose equivalent
iodine-131.' Subsequently, this proposed change was withdrawn by letter
dated April 20, 1998.
4. DEC proposed to change Section 3.3.3.6 regarding accident
monitoring instrumentation. Specifically, the change would (a) increase
the time allowed to return the required number of channels to operable;
and (b) permit continued operation if one channel is inoperable given
certain conditions are met, instead of requiring shutdown.
5. DEC proposed to change Section 4.6.4.1 regarding surveillance
requirements for the hydrogen monitors (combustible gas control).
Specifically, this would eliminate the channel operational test, and
extend the channel check frequency from once per 12 hours to once per
31 days.
6. DEC proposed to change Section 3.4.6.1 regarding reactor coolant
leakage detection systems; a system comprising diverse instruments such
as gaseous radioactivity monitoring, containment floor and equipment
sump monitoring, etc. In addition to the instruments specified by this
section, the plant has other installed instruments such as monitors for
humidity, temperature, etc., which can provide indication for reactor
coolant leakage. Currently, this specification allows operation up to
30 days if the containment floor and equipment sump monitoring system
is inoperable. The change would impose a requirement to perform a
precision water balance of the reactor coolant system every 24 hours
during this period. The change would also reduce the number of monitors
required operable provided compensatory measures are performed or
diverse instruments continue to be available.

[[Page 25107]]

7. DEC proposed to change Section 4.5.4.b, which currently requires
verification of the refueling water storage tank temperature to be
within the allowed range once per 24 hours if the outside air
temperature is less than 70 degrees or greater than 100 degrees
Fahrenheit. The proposed change would simply require that the tank
temperature be verified within range every 24 hours regardless of
outside air temperature.
8. DEC proposed to revise Table 3.7-1, which imposes limits on the
maximum allowable power range neutron flux high setpoint for various
numbers of inoperable safety valves on any operating steam generator.
The revision would reduce the setpoints, making them more conservative.
9. Section 3.7.6, regarding the condensate storage system,
currently only exists in the Unit 2 TS. DEC proposed to impose these
requirements also on Unit 1.
10. Several electrical busses and inverters currently covered by
Section 3.8.3.1 are qualified by a footnote, which specifies the
conditions under which the inverter may be disconnected from its direct
current source. DEC proposed to delete this footnote because it is not
needed.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analyses of the issue of no significant hazards
consideration for each of the above proposed changes. The NRC staff has
reviewed the licensee's analyses against the standards of 10 CFR
50.92(c). The NRC staff's analysis is presented below.
1. Will the change involve a significant increase in the
probability or consequences of an accident previously evaluated?
For changes 1, 2, 4, 5, 6, 7, 8, 9, and 10, the answer is ``no.''
The proposed changes will not affect the safety function of the subject
systems. There will be no direct effect on the design or operation of
any plant structures, systems, or components. No previously analyzed
accidents were initiated by the functions of these systems, and the
systems were not factors in the consequences of previously analyzed
accidents. Therefore, the proposed changes will have no impact on the
consequences or probabilities of any previously evaluated accidents.
2. Will the change create the possibility of a new or different
kind of accident from any accident previously evaluated?
For changes 1, 2, 4, 5, 6, 7, 8, 9, and 10, the answer is ``no.''
The proposed changes would not lead to any hardware or operating
procedure change. Hence, no new equipment failure modes or accidents
from those previously evaluated will be created.
3. Will the change involve a significant reduction in a margin of
safety?
For changes 1, 2, 4, 5, 6, 7, 8, 9, and 10, the answer is ``no.''
Margin of safety is associated with confidence in the design and
operation of the plant. The proposed changes to the TS do not involve
any change to plant design, operation, or analysis. Thus, the margin of
safety previously analyzed and evaluated is maintained.
Based on this analysis, it appears that the three standards of 10
CFR 50.92(c) are satisfied for each of the proposed changes. 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
Attorney for licensee: Mr. Paul R. Newton, Legal Department (PB05E),
Duke Energy Corporation, 422 South Church Street, Charlotte, North
Carolina
NRC Project Director: Herbert N. Berkow

Duke Energy Corporation (DEC), et al., Docket Nos. 50-413 and 50-414,
Catawba Nuclear Station, Units 1 and 2, York County, South Carolina

Date of amendment request: April 8, 1998.
Description of amendment request: The proposed amendments would
revise Section 3.6.5.1 and 4.6.5.1 of the Technical Specifications (TS)
of each unit to relax ice condenser stored ice weight requirements by
approximately 6 percent. The proposed change is based mainly on DEC's
gathered data showing lower sublimation rate than originally
anticipated.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analyses of the issue of no significant hazards
consideration for the proposed changes. The NRC staff has reviewed the
licensee's analyses against the standards of 10 CFR 50.92(c). The NRC
staff's analysis is presented below.
1. Will the changes involve a significant increase in the
probability or consequences of an accident previously evaluated?
No. The proposed changes will not affect the safety function of the
ice condenser in that there will be no changes to the design or
operation of any plant structures, systems, or components. No
previously analyzed accidents were initiated by the functions of the
ice condenser, and the ice condenser will remain fully capable of
performing its design accident mitigation function. Therefore, the
proposed changes will have no impact on the consequences or
probabilities of any previously evaluated accidents.
2. Will the changes create the possibility of a new or difference
kind of accident from any accident previously evaluated?
No. The proposed changes would not lead to any hardware or
operating procedure change. Reducing the required ice weight will not
have any impact on other plant systems that were assumed to be accident
initiators. Hence, no new equipment failure modes or accidents from
those previously evaluated will be created.
3. Will the changes involve a significant reduction in a margin of
safety? No. Margin of safety is associated with confidence in the
design and operation of the plant; specifically, the ability of the
fission product barriers to perform their design functions during and
following an accident. The proposed changes regarding required ice
weight do not involve any change to plant design, operation, or
analysis. Thus, the margin of safety previously analyzed and evaluated
is maintained.
Based on this analysis, it appears that the three standards of 10
CFR 50.92(c) are satisfied for the proposed changes. 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
Attorney for licensee: Mr. Paul R. Newton, Legal Department (PB05E),
Duke Energy Corporation, 422 South Church Street, Charlotte, North
Carolina
NRC Project Director: Herbert N. Berkow

Duke Energy Corporation (DEC), Docket Nos. 50-369 and 50-370, McGuire
Nuclear Station, Units 1 and 2, Mecklenburg County, North Carolina

Date of amendment request: May 27, 1997.
Description of amendment request: The proposed changes would lower
the minimum required diesel generator (DG) air start receiver pressure
from 220 per square inch gauge (psig) to 210 psig with a monthly
verification, and would include an allowed outage time of 48 hours for
a degraded air receiver provided the redundant air receiver is
maintained at equal to or greater than 210 psig. These proposed changes
are associated with DEC's application to convert to the Improved
Technical

[[Page 25108]]

Specifications. Also, they are considered less restrictive requirements
because of the lower required minimum pressure and the allowance of
continued operation with a degraded starting air system.
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 for each change, which is presented below:

1. (Do the changes) involve a significant increase in the
probability or consequence of an accident previously evaluated?
The proposed changes provide Actions for degraded capabilities
of the diesel starting air subsystems for the DG. The proposed
Actions establish limits for the DG starting air subsystems of 210
psig, (are) allowed to decrease below the required value for 48
hours(, and are verified every 31 days.) The Completion Times are
based on the amount of capability remaining, and the time needed to
correct any deficient condition. If the Completion Times are
exceeded, the specification requires the associated DG to be
declared inoperable immediately, consistent with the current TS
(technical specifications). Since the new Actions continue to assure
that the associated DG remains capable of performing its design
safety function, the proposed (changes do) not significantly affect
the probability or consequences of an accident previously evaluated.
2. (Do the changes) create the possibility of a new or different
kind of accident from any accident previously evaluated?
The proposed (changes do) not permit operation in a new or
different mode, or permit the installation of a new or different
type of equipment. The proposed changes provide Actions for degraded
capabilities of the DG starting air subsystems. The proposed Actions
establish Conditions, Required Actions, and Completion Times to be
entered when in a degraded condition. The DG remains capable of
performing its design safety function. Therefore, the proposed
(changes do) not create the possibility of a new or different kind
of accident from those previously evaluated.
3. (Do these changes) involve a significant reduction in a
margin of safety?
The proposed (changes do) not significantly increase the
probability or consequences of an accident previously evaluated. The
changes provide assurance that timely action will be initiated to
restore DG starting air subsystem when inoperabilities exist,
without unnecessarily forcing plant shutdown. Based on the limit for
the starting air subsystem for the DG, the limited time allowed is
acceptable to restore the parameter to within the requirements
without unnecessary plant shutdown. Therefore, (these 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: J. Murrey Atkins Library,
University of North Carolina at Charlotte, 9201 University City
Boulevard, Charlotte, North Carolina
Attorney for licensee: Mr. Albert Carr, Duke Energy Corporation, 422
South Church Street, Charlotte, North Carolina
NRC Project Director: Herbert N. Berkow

Duke Energy Corporation (DEC), Docket Nos. 50-369 and 50-370, McGuire
Nuclear Station, Units 1 and 2, Mecklenburg County, North Carolina

Date of amendment request: May 27, 1997.
Description of amendment request: The two proposed changes are
associated with DEC's application to convert to the Improved Technical
Specifications and are considered as administrative changes. The first
change would delete a current requirement to only verify the refueling
water storage tank temperature once every 24 hours if the outside air
temperature is less than 70 degrees or greater than 100 degrees
Fahrenheit, and would require that the tank temperature be verified
within range every 24 hours regardless of the outside air temperature
value. The second change would delete the current requirement that 32
of 33 hydrogen igniters be operable on each train, and would require
that 34 igniters per train to be operable. The actual design contains
35 igniters per train. This change would correct an inadvertent error
in the current Technical Specifications (TS). The number of igniters
was increased to 35 after the first refueling outage of each unit. This
change would correct the TS to reflect the requirements stated in
Safety Evaluation Report Supplement 7.
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 for each of the above proposed changes. The NRC staff has
reviewed the licensee's analyses against the standards of 10 CFR
50.92(c). The NRC staff's analysis is presented below:
1. Will the change involve a significant increase in the
probability or consequences of an accident previously evaluated?
The proposed changes will not affect the safety function of the
subject systems. There will be no direct effect on the design or
operation of any plant structures, systems, or components. No
previously analyzed accidents were initiated by the functions of these
systems, and the systems were not factors in the consequences of
previously analyzed accidents. Therefore, the proposed changes will
have no impact on the consequences or probabilities of any previously
evaluated accidents.
2. Will the change create the possibility of a new or different
kind of accident from any accident previously evaluated?
The proposed changes would not lead to any hardware or operating
procedure change. Hence, no new equipment failure modes or accidents
from those previously evaluated will be created.
3. Will the change involve a significant reduction in a margin of
safety?
Margin of safety is associated with confidence in the design and
operation of the plant. The proposed changes to the TS do not involve
any change to plant design, operation, or analysis. Thus, the margin of
safety previously analyzed and evaluated is maintained.
Based on this analysis, it appears that the three standards of 10
CFR 50.92(c) are satisfied for each of the proposed changes. Therefore,
the NRC staff proposes to determine that the amendment request involves
no significant hazards consideration.

Local Public Document Room location: J. Murrey Atkins Library,
University of North Carolina at Charlotte, 9201 University City
Boulevard, North Carolina
Attorney for licensee: Mr. Albert Carr, Duke Energy Corporation, 422
South Church Street, Charlotte, North Carolina
NRC Project Director: Herbert N. Berkow

Duke Energy Corporation, Docket Nos. 50-369 and 50-370, McGuire Nuclear
Station, Units 1 and 2, Mecklenburg County, North Carolina

Date of amendment request: May 27, 1997.
Description of amendment request: The proposed change would allow
two charging pumps or safety injection pumps capable of injecting into
the Reactor Coolant System (RCS) when the RCS is depressurized and an
RCS vent of at least 4.5 square inches is established. This proposed
change is associated with the licensee's application to convert to the
Improved Technical Specifications and results in a requirement less
restrictive than the current requirement.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the

[[Page 25109]]

issue of no significant hazards consideration for each change, which is
presented below:

1. Does the change involve a significant increase in the
probability or consequence of an accident previously evaluated?
The proposed change will provide an additional alternative for
low temperature (overpressure) relief capacity when two charging
pumps or safety injection pumps are capable of injecting into the
RCS. The low temperature (overpressure) protection is not considered
to be an initiator of any analyzed event, therefore, the proposed
change does not increase the probability of a previously analyzed
event.
The proposed change provides an equivalent vent size to the
existing two open PORVs (power-operated relief valves). Therefore,
this change does not involve a significant increase in the
probability or consequences of an accident previously evaluated.
2. Does the change create the possibility of a new or different
kind of accident from any accident previously evaluated?
The proposed change does not necessitate a physical alteration
of the plant (no new or different type of equipment will be
installed) or changes in the manner in which the plant is operated.
The proposed change adds an additional alternative to overpressure
protection equivalent to the current requirements. Therefore, the
proposed change will not create the possibility of a new or
different kind of accident than any previously evaluated.
3. Does this change involve a significant reduction in a margin
of safety?
As described above, the proposed change adds an additional
alternative to overpressure protection equivalent to the current
requirements. The inclusion of additional alternatives provides the
operating staff with additional flexibility in meeting low
temperature overpressure protection requirements. Therefore, the
change does not involve a significant reduction in a margin of
safety.

The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.

Local Public Document Room location: J. Murrey Atkins Library,
University of North Carolina at Charlotte, 9201 University City
Boulevard, Charlotte, North Carolina
Attorney for licensee: Mr. Albert Carr, Duke Energy Corporation, 422
South Church Street, Charlotte, North Carolina
NRC Project Director: Herbert N. Berkow

Entergy Operations Inc., Docket No. 50-382, Waterford Steam Electric
Station, Unit 3, St. Charles Parish, Louisiana

Date of amendment request: March 25, 1998
Description of amendment request: Revise Technical Specification
(TS) 3.9.8.1, ``Shutdown Coolant and Coolant Circulation High Water
Level,'' and TS 3.9.8.2, ``Shutdown Cooling and Coolant Circulation Low
Water Level,'' to change the minimum water level above the fuel
assemblies seated in the reactor vessel at which the Shutdown Cooling
(SDC) System is required to be maintained operable, or be in operation.
In addition, TS 3.8.1.2, ``Electric Power Systems, A.C. Sources,
Shutdown,'' and Technical Specification Bases 3/4.9.8, ``Shutdown
Cooling and Coolant Circulation,'' have been changed to make the
wording consistent with TS 3.9.8.1 and TS 3.9.8.2.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:

1. Will operation of the facility in accordance with this
proposed change involve a significant increase in the probability or
consequence of any accident?
Response: No.
The operation of the facility in accordance with this change
does not involve an increase in the probability of any accident.
Changing the water level at which the Shutdown Cooling (SDC)
System is required to be maintained operable or be in operation will
not increase the probability or consequences of an accident. The
design, operation, or configuration of the SDC system will not be
changed.
At least one shutdown cooling train will be in operation to
ensure sufficient cooling capacity is available to remove decay heat
and maintain the water in the reactor pressure vessel below 140
degree F as required during the refueling mode.
At least one shutdown cooling train will be in operation to
ensure sufficient coolant circulation is maintained through the
reactor core to minimize the effects of a boron dilution incident
and prevent boron stratification. Technical Specification 3.9.10.1,
``Refueling Operations Water Level--Reactor Vessel Fuel
Assemblies,'' will be complied with, and therefore, the assumptions
related to iodine removal and the fuel handling accident will be
preserved.
Sufficient time, approximately 1.00 hours, will be available to
the operators to initiate compensatory measures to preclude the
initiation of core boiling in the unlikely event SDC should be loss
[lost].
2. Will operation of the facility in accordance with this
proposed change create the possibility of a new or different kind of
accident from any accident previously evaluated?
Response: No.
The operation of the facility in accordance with this proposed
change will not create the possibility of a new or different kind of
accident from any accident previously evaluated.
The proposed change will not affect the design, configuration,
or operation of the SDC system, and therefore there are no new modes
of failure introduced.
3. Will operation of the facility in accordance with this
proposed change involve a significant reduction in a margin of
safety?
Response: No.
Operation of the facility in accordance with this proposed
change will not involve a significant reduction in a margin of
safety.
The calculation of the time to the initiation of boiling based
on 23 feet above the top of the fuel seated in the reactor vessel,
at four days after shutdown, demonstrates there is significant time
available, approximately 1.00 hour, to the operators within which to
take compensatory measures to preclude the initiation of boiling.
The calculation shows that based on 23 feet of water above the
reactor flange there is 2.04 hours to the initiation of boiling.
Although there is a reduction in the time to the initiation of
boiling, compensatory measures could be taken within a few minutes
to restore SDC, and thus, there is still a significant margin
available to the operators within which to preclude the initiation
of boiling. Thus, the margin of safety is not significantly reduced.
The time to core uncovery was determined to be 27.74 hours based
on four days after shutdown and water level twenty-three (23) feet
above the fuel assemblies seated in the reactor vessel.

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 are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.

Local Public Document Room Location: University of New Orleans Library,
Louisiana Collection, Lakefront, New Orleans, LA 70122
Attorney for licensee: N.S. Reynolds, Esq., Winston & Strawn, 1400 L
Street N.W., Washington DC 20005-3502
NRC Project Director: John N. Hannon

Florida Power Corporation, et al., Docket No. 50-302, Crystal River
Nuclear Generating Plant, Unit No. 3, Citrus County, Florida

Date of amendment request: March 20, 1998.
Description of amendment request: The proposed amendment requests
editorial changes to the Improved Technical Specifications (ITS) Safety
Limits and Administrative Controls to replace the titles of the Senior
Vice President, Nuclear Operations (SVPNO) and the Vice President,
Nuclear Production (VPNP) with the position of Chief Nuclear Officer
(CNO). The CNO combines the duties of the SVPNO and VPNP as currently
described in ITS and is required to be an officer of the company. The
proposed change is

[[Page 25110]]

intended to allow upgrading the position of the corporate officer
responsible for overall nuclear operations without limiting the title.
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.

Does not involve a significant increase in the probability or
consequences of an accident previously evaluated.
The proposed amendment does not involve a significant increase
in the probability or consequences of an accident previously
evaluated because the deletion and updating of individual titles
does not affect plant operation. No design basis accidents are
affected by the proposed administrative and editorial changes and,
as such, there are no physical changes to the facility or its
operation.
Does not create the possibility of a new or different kind of
accident from any accident previously evaluated.
The proposed ITS changes are administrative and editorial in
nature. No changes to the facility structures, systems and
components or their operation will result. The design and design
basis of the facility remain unchanged. The plant safety analyses
remain current and accurate. No new or different failure mechanisms
are introduced. Therefore, the possibility of a new or different
kind of accident from any accident previously evaluated is not
introduced.
Does not involve a significant reduction in the margin of
safety.
The proposed ITS changes are administrative and editorial in
nature. The proposed safety margins established through the design
and facility license including the Improved Technical Specifications
remain unchanged. In addition, the proposed amendment ensures
continued emphasis and assignment of responsibility for overall
nuclear safety. Therefore, all margins of safety are maintained.

The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of Sec. 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: Coastal Region Library, 8619 W.
Crystal Street, Crystal River, Florida 34428
Attorney for licensee: R. Alexander Glenn, General Counsel, Florida
Power Corporation, MAC-A5A, P.O. Box 14042, St. Petersburg, Florida
33733-4042
NRC Project Director: Frederick J. Hebdon

Florida Power Corporation, et al., Docket No. 50-302, Crystal River
Nuclear Generating Plant, Unit No. 3, Citrus County, Florida

Date of amendment request: March 20, 1998.
Description of amendment request: The proposed amendment would
change the Inservice Inspection Program described in Improved Technical
Specification (ITS) 5.6.2.8.c. This ITS currently states that the
reactor coolant pump (RCP) motor flywheels will be inspected during the
``Spring 1998 refueling outage,'' which would have been refueling
outage 11. Due to a recent 17-month extended outage, refueling outage
11 has been deferred until Fall 1999. The proposed change is intended
to accurately reflect the new refueling outage 11 schedule.
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 will not significantly increase the
probability or consequences of an accident previously evaluated.
The safety function of the RCP flywheels is to provide a
coastdown period during which the RCPs would continue to provide
reactor coolant flow to the reactor after loss of power to the RCPs.
The maximum loading on the RCP motor flywheel results from overspeed
following a large loss of coolant accident (LOCA). The estimated
maximum obtainable speed in the event of a Reactor Coolant System
piping break was established conservatively. The proposed one-time
editorial change to remove the words ``Spring 1998 refueling
outage'' and replace them with ``to coincide with Refueling Outage
11R'' does not affect that analysis. The proposed change in dates is
editorial in that it merely reflects the new date for cycle 11. The
usage time for the flywheels is bounded by the original estimates.
The proposed editorial change does not affect the amount of
radioactive material available for release or modify any systems
used for mitigation of such releases during accident conditions.
Therefore, the proposed change does not involve a significant
increase in the probability or consequences of any accident
previously evaluated.
The proposed change will not create the possibility of a new or
different kind of accident from any accident previously evaluated.
The proposed editorial change will not change the design,
configuration, or method of operation of the plant. Therefore, the
proposed change will not create the possibility of a new or
different kind of accident from any previously evaluated.
The proposed change will not involve a significant reduction to
any margin of safety.
The proposed Amendment is an editorial change to reflect that
CR-3's operating cycle is not ending in spring 1998, but in fall
1999. The proposed change does not affect the methods of inspection
or its acceptance criteria. Therefore, the margins of safety defined
in RG [Regulatory Guide] 1.14 are not changed.

The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of Sec. 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: Coastal Region Library, 8619 W.
Crystal Street, Crystal River, Florida 34428
Attorney for licensee: R. Alexander Glenn, General Counsel, Florida
Power Corporation, MAC-A5A, P.O. Box 14042, St. Petersburg, Florida
33733-4042
NRC Project Director: Frederick J. Hebdon

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

Date of amendment request: April 15, 1998.
Description of amendment request: The proposed amendment would
update the existing pressure-temperature curves with new curves with
values from 18 to 32 effective full power years based on the testing
and analysis of reactor pressure vessel surveillance materials.
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 pressure-temperature limits are not
derived from Design Basis Accident (DBA) analyses. They are
prescribed by the ASME B&PV Code and 10 CFR part 50 appendices G and
H as restrictions on normal operation to avoid encountering
pressure, temperature, and temperature rate of change conditions
that might cause undetected flaws to propagate and cause nonductile
failure of the reactor coolant pressure boundary.
(2) The proposed amendment will not create the possibility of a
new or different kind of accident from any accident previously
evaluated. The amendment will merely update the pressure-temperature
curves (and associated SRs and Bases) already existing in the plant
Improved Technical Specifications to provide limits from 18 to 32
EFPY of operation, which are based upon evaluation and analysis of
actual in-vessel material specimens, per 10 CFR part

[[Page 25111]]

50, appendices G and H. The pressure-temperature curves are
established to the requirements of 10 CFR part 50, appendix G to
assure that brittle fracture of the reactor vessel is prevented.
(3) The proposed amendment will not involve a significant
reduction in a margin of safety. 10 CFR part 50, appendix G
specifies fracture toughness requirements to provide adequate
margins of safety during operation over the service lifetime. The
values of adjusted reference temperature and upper shelf energy
determined as a result of the 10 CFR part 50, appendices G and H
analysis are expected to remain within the limits of Regulatory
Guide 1.99, Revision 2 and appendix G of 10 CFR part 50 (less than
200 deg. F and greater than 50 ft-lbs respectively) for at least 32
EFPY of operation.

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, SE., Cedar Rapids, IA 52401
Attorney for licensee: Jack Newman, Al Gutterman, Morgan, Lewis &
Bockius, 1800 M Street, NW., Washington, DC 20036-5869
Acting NRC Project Director: Richard P. Savio

Nebraska Public Power District, Docket No. 50-298, Cooper Nuclear
Station, Nemaha County, Nebraska

Date of amendment request: March 27, 1997.
Description of amendment request: The proposed amendment, included
as part of the proposed conversion from the current Technical
Specifications (TS) to improved TS, would establish Allowable Values
for the instrumentation included in Section 3.3, as a result of the
plant-specific application of the General Electric Instrument Setpoint
Methodology to the Cooper Nuclear Station (CNS).
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:

1. Does the change involve a significant increase in the
probability or consequences of an accident previously evaluated?
The proposed change in selected Allowable Values for the
instrumentation included in proposed Section 3.3 of the Technical
Specifications is the result of application of the CNS
instrumentation setpoint methodology. This methodology incorporates
the guidance of ISA Recommended Practice ISA-RP67.04, Part II,
``Methodologies for the Determination of Setpoints for Nuclear
Safety-Related Instrumentation,'' September 1994. Application of
this methodology results in instrumentation selected Allowable
Values which more accurately reflect total instrumentation loop
accuracy as well as that of test equipment and setpoint drift
between Surveillances. The proposed change will not result in any
hardware changes. The instrumentation included in proposed Section
3.3 of the Technical Specifications is not assumed to be an
initiator of any analyzed event. Existing operating margin between
plant conditions and actual plant setpoints is not significantly
reduced due to this change. As a result, the proposed change will
not result in unnecessary plant transients.
The role of the proposed Section 3.3 instrumentation is in
mitigating and thereby limiting the consequences of accidents. The
Allowable Values have been developed to ensure that the design and
safety analysis limits will be satisfied. The methodology used for
the development of the Allowable Values ensures the affected
instrumentation remains capable of mitigating design basis events as
described in the safety analyses and that the results and
consequences described in the safety analyses remain bounding.
Additionally, the proposed change does not alter the plant's ability
to detect and mitigate events. Therefore, this change does not
involve a significant increase in the probability or consequences of
an accident previously evaluated.
2. Does the change create the possibility of a new or different
kind of accident from any accident previously evaluated?
The proposed change is the result of application of the CNS
instrumentation setpoint methodology and do not create the
possibility of a new or different kind of accident from any accident
previously evaluated. This is based on the fact that the method and
manner of plant operation is unchanged. The use of the proposed
Allowable Values does not impact safe operation of CNS in that the
safety analysis limits will be maintained. The proposed Allowable
Values involve no system additions or physical modifications to
systems in the station.
These Allowable Values were developed using a methodology to
ensure the affected instrumentation remains capable of mitigating
accidents and transients. Plant equipment will not be operated in a
manner different from previous operation, except that setpoints may
be changed. Since operational methods remain unchanged and the
operating parameters have been evaluated to maintain the station
within existing design basis criteria, no different type of failure
or accident is created.
3. Does this change involve a significant reduction in a margin
of safety?
The proposed change does not involve a reduction in a margin of
safety. The proposed changes have been developed using a methodology
to ensure safety analysis limits are not exceeded. As such, this
proposed change does not involve a significant reduction in a margin
of safety.

The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.

Local Public Document Room location: Auburn Memorial Library, 1810
Courthouse Avenue, Auburn, NE 68305
Attorney for licensee: Mr. John R. McPhail, Nebraska Public Power
District, Post Office Box 499, Columbus, NE 68602-0499
NRC Project Director: John N. Hannon

Nebraska Public Power District, Docket No. 50-298, Cooper Nuclear
Station, Nemaha County, Nebraska

Date of amendment request: March 27, 1997.
Description of amendment request: The proposed amendment, included
as part of the proposed conversion from the current Technical
Specifications (CTS) to the improved Technical Specifications (ITS),
would add an additional action statement to a limiting condition for
operation (LCO). The LCO is in the Improved Standard Technical
Specifications (ISTS, NUREG-1433, Revision 1) 3.6.2.3 on the residual
heat removal suppression pool cooling subsystems. The requirements in
the proposed ITS 3.6.2.3 on the subsystems do not exist in the CTS. The
Action B for ITS 3.6.2.3 would require that if the two such subsystems
were inoperable, one subsystem would have to be restored to operability
within 8 hours or the plant would be in ITS 3.0.3. ITS 3.0.3 governs
plant operation if an LCO (i.e., ISTS 3.6.2.3) and the associated
action statement are not met (i.e., Action B to ISTS 3.6.2.3).
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:

1. Does the change involve a significant increase in the
probability or consequences of an accident previously evaluated?
The proposed change provides more stringent requirements for
operation of the facility. These more stringent requirements do not
result in operation that will increase the probability of initiating
an analyzed event and do not alter assumptions relative to (the)
mitigation of an accident or transient event. The more restrictive
requirements continue to ensure * * * systems, and components
((i.e., the residual heat removal suppression pool cooling
subsystems)) are maintained consistent with the safety analyses and
licensing basis. Therefore, this (the proposed)

[[Page 25112]]

change does not involve a significant (an) increase in the
probability or consequences of any accident previously evaluated.
2. Does the change create the possibility of a new or different
kind of accident from any accident previously evaluated?
The proposed change does not involve a physical alteration of
the plant (no new or different type of equipment will be installed)
or changes in the methods governing normal plant operation. The
proposed change does impose different requirements. However, this
change is consistent with the assumptions in the safety analyses and
licensing basis. Thus, this change does not create the possibility
of a new or different kind of accident from any accident previously
evaluated is not created.
3. Does this change involve a significant reduction in a margin
of safety?
The imposition of more restrictive requirements either has no
impact on or increases the margin of plant safety. As provided in
the discussion of the change, each change in this category (i.e.,
more restrictive requirements) is, by definition, providing
additional restrictions to enhance plant safety. The change
maintains requirements (systems and components) within the safety
analyses and licensing basis. Therefore, this change does not
involve a significant reduction in a margin of safety.

The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the 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: Auburn Memorial Library, 1810
Courthouse Avenue, Auburn, NE 68305
Attorney for licensee: Mr. John R. McPhail, Nebraska Public Power
District, Post Office Box 499, Columbus, NE 68602-0499
NRC Project Director: John N. Hannon

Nebraska Public Power District, Docket No. 50-298, Cooper Nuclear
Station, Nemaha County, Nebraska

Date of amendment request: March 27, 1997.
Description of amendment request: The proposed amendment, included
as part of the proposed conversion from the current Technical
Specifications (CTS) to the improved Technical Specifications (ITS),
would add an additional test (i.e., water and sediment content within
limits) of diesel fuel oil that could be used in place of a current
test (i.e., clear and bright appearance with proper color) in the
diesel fuel oil testing program. The current tests are listed in CTS
4.9.A.2.d/e. The testing program will be in the new ITS 5.5.9. The
additional test is change number 25 to Section 5.0 of the Improved
Standard Technical Specifications (NUREG-1433, Revision 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. Does the change involve a significant increase in the
probability or consequences of an accident previously evaluated?
The proposed change provides more stringent requirements for
operation of the facility. (This) more stringent (requirement)
do(es) not result in operation that will increase the probability of
initiating an analyzed event and do(es) not alter assumptions
relative to (the) mitigation of an accident or transient event. The
more restrictive (requirement) continue(s) to ensure * * * systems
and components (i.e., the diesel generators) are maintained
consistent with the safety analyses and licensing basis. Therefore,
the proposed change does not involve an increase in the probability
or consequences of any accident previously evaluated.
2. Does the change create the possibility of a new or different
kind of accident from any accident previously evaluated?
The proposed change does not involve a physical alteration of
the plant (no new or different type of equipment will be installed)
or changes in the methods governing normal plant operation. However,
this change is consistent with the assumptions in the safety
analyses and licensing basis. Thus, this change does not create the
possibility of a new or different kind of accident from any accident
previously evaluated is not created.
3. Does this change involve a significant reduction in a margin
of safety?
The imposition of more restrictive requirements either has no
impact on or increases the margin of plant safety. As provided in
the discussion of the change, each change in this category (i.e., a
more restrictive requirement) is, by definition, providing
additional restrictions to enhance plant safety. The change
maintains (systems and components) within the safety analyses and
licensing basis. Therefore, this change does not involve a
significant reduction in a margin of safety.

The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the 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: Auburn Memorial Library, 1810
Courthouse Avenue, Auburn, NE 68305
Attorney for licensee: Mr. John R. McPhail, Nebraska Public Power
District, Post Office Box 499, Columbus, NE 68602-0499
NRC Project Director: John N. Hannon

Nebraska Public Power District, Docket No. 50-298, Cooper Nuclear
Station, Nemaha County, Nebraska

Date of amendment request: March 27, 1997.
Description of amendment request: The proposed amendment, included
as part of the proposed conversion from the current Technical
Specifications (TS) to improved TS for the Cooper Nuclear Station
(CNS), would relocate the Trip Level Settings for the Rod Block Monitor
from Table 3.2.C of the current TS to the Core Operating Limits Report.
Also, details relating to the Alternate Shutdown system design and
operation are proposed to be relocated from current TS 3.2.I and 4.2.I
to the improved TS Bases.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration. The NRC staff has reviewed the licensee's analysis
against the three criteria of 10 CFR 50.92(c), and has determined the
following:
The proposed changes relocate certain details from the Technical
Specifications to the Bases and the Core Operating Limits Report
(COLR). The Bases and the COLR containing the relocated information
will be maintained in accordance with 10 CFR 50.59. In addition, the
Bases and COLR are subject to the applicable change control provisions
of Chapter 5.0, Administrative Controls'', of the proposed improved
Technical Specifications. Since any changes to the Bases or the COLR
will be evaluated per the requirements of 10 CFR 50.59 or other
applicable change control provisions, no increase in the probability or
consequences of an accident previously evaluated will result.
Therefore, these changes do not involve a significant increase in the
probability or consequences of an accident previously evaluated.
The proposed changes do not involve any physical alterations to the
plant (no new or different type of equipment will be installed), or
changes in the methods governing normal plant operation. The proposed
changes will not impose or eliminate any requirements, and adequate
control of the information will be maintained. Thus, these changes do
not create the possibility of a new or different kind of accident from
any accident previously evaluated.
The proposed changes will not reduce a margin of safety because
they have no impact on any safety analysis assumptions. In addition,
the details to be transposed from the TS to the Bases

[[Page 25113]]

and the COLR are unchanged. Since any future changes to these details
in the Bases or the COLR will be evaluated per the requirements of 10
CFR 50.59 or other applicable change control provisions, no reduction
in a margin of safety will result. As such, these proposed changes do
not involve a significant reduction in a margin of safety.
Based on the above discussion, 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: Auburn Memorial Library, 1810
Courthouse Avenue, Auburn, NE 68305
Attorney for licensee: Mr. John R. McPhail, Nebraska Public Power
District, Post Office Box 499, Columbus, NE 68602-0499
NRC Project Director: John N. Hannon

North Atlantic Energy Service Corporation, Docket No. 50-443, Seabrook
Station, Unit No. 1, Rockingham County, New Hampshire

Date of amendment request: April 8, 1998.
Description of amendment request: The proposed change would revise
Technical Specifications (TSs) 4.4.5.3, Steam Generators--Inspection
Frequencies, and 3.4.6.2.c, Reactor Coolant System (RCS) Leakage, and
the associated bases to accommodate fuel cycles of up to 24 months with
respect to the allowed time interval between steam generator inservice
inspections.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:

1. The proposed changes do not involve a significant increase in
the probability or consequences of an accident previously evaluated.
Extending Surveillance Requirement (SR) 4.4.5.3 to accommodate a
24 month cycle for inspection of steam generator tubes structural
integrity, as well as, imposing a more restrictive Limiting
Condition for Operation (TS 3.4.6.2.c) for reactor coolant system
leakage through Category C-2 steam generators, will neither
exacerbate nor significantly increase the probability or
consequences of an accident previously evaluated in the Seabrook
Station [updated final safety analysis report] UFSAR.
The proposed changes to SR 4.4.5.3 do not alter the intent or
method by which the surveillances are conducted, do not involve
physical changes to the plant, do not alter the way structures,
systems or components (SSCs) function, and do not modify the manner
in which the plant is operated.
The proposed change to TS 3.4.6.2.c imposes more restrictive
limits on plant operations due to RCS leakage through steam
generators. The proposed change does not involve physical changes to
the plant or alter the way a SSC functions.
The proposed changes to SR 4.4.5.3 and TS 3.4.6.2.c, and their
associated Bases, will not adversely affect the ability of the steam
generators to perform their intended safety function. Furthermore,
the proposed changes do not adversely affect the physical protective
boundaries of the plant. The proposed changes do not affect accident
initiators or precursors and do not alter the design assumptions,
conditions, configuration of the facility or the manner in which the
plant is operated. The proposed changes do not alter or prevent the
ability of SSCs to perform their intended function to mitigate the
consequences of an initiating event within the acceptance limits
assumed in the Updated Final Safety Analysis Report (UFSAR). The
proposed changes are administrative in nature and do not change the
level of programmatic controls or the procedural details associated
with aforementioned surveillance requirements. While the proposed
changes will lengthen the interval between surveillances, the
increase in interval has been evaluated; and based on the reviews of
the steam generator tube eddy current test (ECT) inspections, it is
concluded that the wear growth rate of the only active degradation
mechanism (Anti-Vibration Bar (AVB) wear) identified to date at
Seabrook Station is such that sufficient margin exists between the
plugging criteria and structural limit such that no tubes are
predicted to exceed the structural limit even with the longer
surveillance interval.
Since there are no changes to previous accident analyses, the
radiological consequences associated with these analyses remain
unchanged, therefore, the proposed changes do not involve a
significant increase in the probability or consequences of an
accident previously evaluated. Therefore, the proposed changes will
not significantly increase the probability or consequences of any
previously analyzed accident.
2. The proposed changes do not create the possibility of a new
or different kind of accident from any previously analyzed.
The proposed changes to TS 3.4.6.2 and SR 4.4.5.3, and
associated Bases, do not alter the design assumptions, conditions,
configuration of the facility or the manner in which the plant is
operated. There are no changes to the source term, containment
isolation or radiological release assumptions used in evaluating the
radiological consequences in the Seabrook Station UFSAR. Existing
system and component redundancy is not being changed by the proposed
changes. The proposed changes have no impact on component or system
interactions. The proposed changes are administrative in nature and
do not change the level of programmatic controls and procedural
details associated with the aforementioned surveillance
requirements. Therefore, since there are no changes to the design
assumptions, conditions, configuration of the facility, or the
manner in which the plant is operated and surveilled, the proposed
changes do not create the possibility of a new or different kind of
accident from any previously analyzed.
3. The proposed changes do not involve a significant reduction
in a margin of safety.
The proposed change ( ) to the surveillance intervals for SR
4.4.5.3 is still consistent with the basis for the interval. The
intent or method of performing the surveillances remains unchanged.
The more restrictive limit for leakage through any one steam
generator placed in Category C-2, as well as, the requirement to do
an engineering assessment of steam generator tube integrity,
provides additional margin of ensuring safe plant operation.
In addition, there is no adverse affect on equipment design or
operation and there are no changes being made to the Technical
Specification required safety limits or safety system settings that
would adversely affect plant safety. The proposed changes are
administrative in nature and do not change the level of programmatic
controls and procedural details associated with the aforementioned
surveillance requirements. While the proposed changes will lengthen
the interval between surveillances, the increase in interval has
been evaluated; and based on the reviews of the steam generator tube
ECT inspections, it is concluded that the wear growth rate of the
only active degradation mechanism (AVB wear) identified to date at
Seabrook Station is such that sufficient margin exists between the
plugging criteria and structural limit such that no tubes are
predicted to exceed the structural limit even with the longer
surveillance interval. Therefore, extension of the current
surveillance intervals to accommodate a 24 month cycle will not
significantly degrade the ability, the availability or the
reliability of the steam generators to perform their intended safety
function, thus, it is concluded that there is no 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: Exeter Public Library, Founders
Park, Exeter, NH 03833
Attorney for licensee: Lillian M. Cuoco, Esq., Senior Nuclear Counsel,
Northeast Utilities Service Company, PO Box 270, Hartford, CT 06141-
0270
NRC Project Director: Cecil O. Thomas

Northeast Nuclear Energy Company, et al., Docket No. 50-336, Millstone
Nuclear Power Station, Unit No. 2, New London County, Connecticut

Date of amendment request: April 6, 1998.
Description of amendment request: The proposed amendment will
modify

[[Page 25114]]

the Technical Specifications (TSs) by (1) adding a surveillance
requirement to verify pressurizer heater capacity to TS 3.4.4,
``Reactor Coolant System--Pressurizer,'' (2) moving the identification
of the location of the containment air temperature detectors from the
surveillance requirements portion of TS 3.6.1.5, ``Containment
Systems--Air Temperature,'' to the TS Bases for Containment Systems,
Section 3/4.4.6.1.5, ``Air Temperature,'' and (3) modifying the action
statements and surveillance requirements of TS 3.7.1.5, ``Plant
Systems--Main Steam Isolation Valves.'' The TS Bases would also be
updated to include the list of containment air temperature detectors
and reflect the proposed changes.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:

1. Involve a significant increase in the probability or
consequences of an accident previously evaluated.
The proposed change to add a surveillance requirement (SR)
4.4.4.2 to verify pressurizer heater capacity will help ensure the
pressurizer will be able to function as designed to maintain Reactor
Coolant System pressure. There will be no effect on any design basis
accident previously evaluated or on any equipment important to
safety. Therefore, the proposed change will not result in a
significant increase in the probability or consequences of an
accident previously evaluated.
The proposed changes to modify the wording of SR 4.6.1.5 and to
relocate the list of containment air temperature detectors from SR
4.6.1.5 to the Bases will not affect the Technical Specification
limit for containment temperature or the frequency of verification
of this limit. The proposed changes do not alter the way any
structure, system, or component functions. The initial assumption
for containment temperature used in the design basis accident
analysis will remain the same. There will be no affect on any design
basis accident previously evaluated or on any equipment important to
safety. Therefore, the proposed changes will not result in a
significant increase in the probability or consequences of an
accident previously evaluated.
The proposed changes to the action statements and surveillance
requirements of Technical Specification 3.7.1.5 will not affect the
operability requirements of the main (steamline) isolation valves
(MSIVs). There will be no effect on any design basis accident
previously evaluated or on any equipment important to safety.
Therefore, the proposed changes will not result in a significant
increase in the probability or consequences of an accident
previously evaluated.
The proposed changes have no adverse effect on any of the design
basis accidents previously evaluated or on any equipment important
to safety. Therefore, the License Amendment Request does not impact
the probability of an accident previously evaluated nor does it
involve a significant increase in the consequences or an accident
previously evaluated.
2. Create the possibility of a new or different kind of accident
from any accident previously evaluated.
The proposed changes will not alter the plant configuration (no
new or different type of equipment will be installed) or require any
new or unusual operator actions. They do not alter the way any
structure, system, or component functions and do not alter the
manner in which the plant is operated. The proposed changes do not
introduce any new failure modes. Therefore, the proposed changes
will not create the possibility of a new or different kind of
accident from any accident previously evaluated.
3. Involve a significant reduction in a margin of safety.
The proposed changes will add SR 4.4.4.2 to verify pressurizer
heater capacity, relocate the list of containment temperature
detectors used to verify containment temperature from SR 4.6.1.5 to
the associated Bases, and modify the action statements and
surveillance requirements of Technical Specification 3.7.1.5.
These changes will have no adverse effect on equipment important
to safety. This equipment will continue to function as assumed in
the design basis accident analysis. Therefore, there will be no
significant reduction in the margin of safety as defined in the
Bases for the technical Specifications affected by these proposed
changes.

The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: Learning Resources Center, Three
Rivers Community-Technical College, 574 New London Turnpike, Norwich,
Connecticut, and the Waterford Library, ATTN: Vince Juliano, 49 Rope
Ferry Road, Waterford, Connecticut
Attorney for licensee: Lillian M. Cuoco, Esq., Senior Nuclear Counsel,
Northeast Utilities Service Company, P.O. Box 270, Hartford,
Connecticut
NRC Deputy Director: Phillip F. McKee

Northeast Nuclear Energy Company, et al., Docket No. 50-336, Millstone
Nuclear Power Station, Unit No. 2, New London County, Connecticut

Date of amendment request: April 13, 1998
Description of amendment request: The proposed amendment would
change the Technical Specifications (TSs) by adding a new TS 3.5.5,
``Emergency Core Cooling Systems--Trisodium Phosphate (TSP).'' Also,
the surveillance requirements in TSs 4.5.2.c.3 and 4.5.2.c.4 would be
relocated to new TS 3.5.5 as TS 4.5.5.1 and TS 4.5.5.2, respectively.
The applicable TS Index page and Bases sections will be updated to
reflect the proposed changes.
Changes to the current requirements for the TSP are also proposed.
The TSP requirements in TS 4.5.2.c.3 would become the limiting
conditions for operation in the new TS; the amount of TSP required
would increase from ``equal to or greater than 110 cubic feet'' to
``equal to or greater than 282 cubic feet'' based on the new
calculations; the applicability would be expanded to include all of
Mode 3; the action statement would allow 48 hours to restore the TSP
volume; and changes would also be made to the required tests and
specific details would be relocated to the applicable TS Bases.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:

1. Involve a significant increase in the probability or
consequences of an accident previously evaluated.
The proposed changes to relocate the current trisodium phosphate
(TSP) dodecahydrate Technical Specification requirements from the
surveillance requirements for the Emergency Core Cooling System to a
new TSP Technical Specification will not change the requirement to
store TSP inside containment. The proposed changes will require a
large quantity of TSP to be stored inside containment. This large
quantity, based on a recently revised calculation, will ensure
sufficient TSP is available for containment sump water pH control.
These proposed changes do not alter the way any structure, system,
or component functions. There will be no adverse effect on any
design basis accident previously evaluated, on any equipment
important to safety, or o n the radiological consequences of any
design basis accident. Therefore, this License Amendment Request
does not impact the probability of an accident previously evaluated
nor does it involve a significant increase in the 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 to increase the TSP volume stored inside
containment will require two of the wire mesh TSP baskets inside
containment to be replaced by two new and larger wire mesh baskets.
The design of the new baskets has been evaluated and it is
consistent with the requirements for equipment installed in
containment. The replacement of the two wire mesh baskets

[[Page 25115]]

will not result in any significant change in plant configuration and
will not require any new or unusual operator actions. It will alter
the way any structure, system, or component functions and does not
alter the manner in which the plant is operated. It will not
introduce any new failure modes. Therefore, the proposed changes
will not create the possibility of a new or different kind of
accident from any accident previously evaluated.
3. Involve a significant reduction in a margin of safety.
The proposed changes will relocate the current Technical
Specification requirements for TSP to a new Technical Specification.
The minimum required volume will be increased to reflect the results
of a new calculation performed to support the current requirement to
raise containment sump pH [equal to or greater than] 7.0. These
changes will have no adverse effect on equipment important to
safety. This equipment will continue to function as assumed in the
design basis accident analysis. Therefore, there will be no
significant reduction of the margin of safety as defined in the
Bases for the Technical Specifications affected by these proposed
changes.

The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.

Local Public Document Room location: Learning Resources Center, Three
Rivers Community-Technical College, 574 New London Turnpike, Norwich,
Connecticut, and the Waterford Library, ATTN: Vince Juliano, 49 Rope
Ferry Road, Waterford, Connecticut
Attorney for licensee: Lillian M. Cuoco, Esq., Senior Nuclear Counsel,
Northeast Utilities Service Company, P.O. Box 270, Hartford,
Connecticut
NRC Deputy Director: Phillip F. McKee

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

Date of amendment request: April 11, 1997 (supersedes July 26,
1996, application)
Description of amendment request: The proposed amendment would
modify the Monticello Technical Specifications (TS) sections 3.6.C,
Coolant Chemistry, and 3/4.17.B, Control Room Emergency Filtration
System. The changes were proposed to establish TS requirements
consistent with modified analysis inputs used for the evaluation of the
radiological consequences of the main steam line break accident.
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.
A limit is established in the plant Technical Specifications for
steady state radioiodine concentration in the reactor coolant to
ensure that in the event of a release of radioactive material to the
environment due to a postulated high energy line break up to and
including a design basis Main Steam Line Break Accident, radiation
doses are maintained within the guidelines of 10 CFR part 100. The
steady state radioiodine concentration in the reactor coolant is an
input for analysis of the radiological consequences of an accident
due to a Main Steam Line Break outside of containment and postulated
high energy line breaks. In addition, requirements are established
in the Technical Specifications for control room habitability.
During an accident, the control room emergency filtration system
provides filtered air to pressurize the Control Room to minimize the
activity, and therefore the radiological dose, inside the control
room.
A change is proposed for the steady state radioiodine
concentration. This value is conservative with respect to the value
used in the Main Steam Line Break dose consequences analysis and is
consistent with the dose consequences evaluation of a postulated
Reactor Water Cleanup (RWCU) line break. Changes are proposed to the
limiting conditions for operation and surveillance requirements for
the Control Room Emergency Filtration Train iodine removal
efficiency. These changes are consistent with the inputs used in the
analysis of the radiological consequences of the postulated RWCU
line break and the Main Steam Line Break Accident. These proposed
requirements maintain operating restrictions for analytical inputs
used in the analysis of the Main Steam Line Break Accident.
Evaluation of these events has demonstrated that the postulated
radiological consequences will remain within the licensing basis
established in the AEC [Atomic Energy Commission] Provisional
Operating License Safety Evaluation Report, dated March 18, 1970,
thus the proposed changes do not result in an increase in the
consequences of previously evaluated accidents.
The analysis of the Main Steam Line Break Accident performed
using a reactor coolant radioiodine concentration of 2
(microcuries)/gm dose equivalent Iodine-131 and a control room
ventilation filter efficiency consistent with the proposed Technical
Specifications changes demonstrated that radiological consequences
of the Main Steam Line Break are not changed significantly. The
radiological consequences of the Main Steam Line Break Accident
remain within the exposure guidelines of 10 CFR part 100 and 10 CFR
part 50 appendix A, General Design Criterion 19. The offsite dose
consequences remain bounded by the licensing basis provided in the
AEC Provisional Operating License Safety Evaluation Report, dated
March 18, 1970. The control room doses calculated for the hot
standby Main Steam Line Break Accident using the TID-14844 dose
conversion factors remain bounded by the dose consequences of the
comparable design basis loss of coolant accident.
The evaluation of the postulated RWCU line break, performed
using a reactor coolant radioiodine concentration of 0.25
(microcurie)/gm dose equivalent Iodine-131 and a control room
ventilation filter efficiency consistent with the proposed Technical
Specifications changes, demonstrated that the radiological
consequences of this event remain within the exposure guidelines of
10 CFR part 100 and 10 CFR part 50 Appendix A, General Design
Criterion 19. The offsite dose consequences remain bounded by the
Main Steam Line Break as established in the licensing basis provided
in the AEC Provisional Operating License Safety Evaluation Report,
dated March 18, 1970.
The proposed Technical Specification changes do not introduce
new equipment operating modes, nor do the proposed changes alter
existing system inter-relationships. The proposed changes do not
introduce new failure modes. The system improvements to reduce
bypass leakage during postulated accidents do not have an adverse
effect on control room habitability. Therefore, this amendment will
not cause a significant increase in the probability of an accident
previously evaluated for the Monticello plant.
2. The proposed amendment will not create the possibility of a
new or different kind of accident from any accident previously
analyzed.
The proposed Technical Specification changes do not introduce
new equipment operating modes, nor do the proposed changes alter
existing system inter-relationships. Operator action to mitigate the
consequences of the postulated RWCU line break is conservative based
on the very limited action required by the operator to close the
containment isolation valves and the availability of control room
indications to alert the operator to the postulated break. The use
of a ten (10) minute operator response time to take manual actions
in response to postulated events is consistent with Monticello's
licensing basis for similar events. The use of operator actions and
all available equipment is consistent with current regulatory
guidance for mitigating the consequences of postulated line breaks.
The proposed change to the specification for reactor coolant
dose equivalent radioiodine is conservative with respect to the re-
evaluation of the Main Steam Line Break Accident for the more
conservative hot standby initial condition for the postulated
accident. The proposed change to the specification for reactor
coolant dose equivalent radioiodine is consistent with the
postulated high energy line break of a Reactor Water Cleanup line.
The proposed changes to the limiting conditions for operation and

[[Page 25116]]

surveillance requirements for the control room emergency filtration
train iodine removal efficiency are consistent with the inputs used
in the evaluation of the radiological consequences of the postulated
RWCU line break and the Main Steam Line Break Accident. The system
improvements to reduce bypass leakage during postulated accidents do
not have an adverse effect on control room habitability. Therefore,
the proposed amendment will not create the possibility of a new or
different kind of accident.
3. The proposed amendment will not involve a significant
reduction in the margin of safety.
Surveillance data has demonstrated the proposed requirements are
within the current capability of the facility. The proposed changes
maintain margins of safety. These proposed requirements maintain
operating restrictions for analytical inputs used in the analysis of
the bounding postulated high energy line break of a Reactor Water
Cleanup line and the Main Steam Line Break Accident. The proposed
change to the specification for reactor coolant dose equivalent
radioiodine is conservative with respect to the re-evaluation of the
Main Steam Line Break Accident for the more conservative hot standby
initial condition for the postulated accident. The proposed change
to the specification for reactor coolant dose equivalent radioiodine
is consistent with the postulated high energy line break of a
Reactor Water Cleanup line. The evaluation of these postulated
events determined that the radiological consequences remain within
the exposure guidelines of 10 CFR part 100 and of 10 CFR part 50
Appendix A, General Design Criterion 19. The proposed changes to the
limiting conditions for operation and surveillance requirements for
the control room emergency filtration train iodine removal
efficiency provide assurance that the system will perform at the
filter efficiency as used in the evaluation of the radiological
consequences of the postulated events. Therefore, 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: Cynthia A. Carpenter

Pacific Gas and Electric Company, Docket Nos. 50-275 and 50-323, Diablo
Canyon Nuclear Power Plant, Unit Nos. 1 and 2, San Luis Obispo County,
California

Date of amendment request: April 10, 1998.
Description of amendment request: The proposed amendments would
revise the combined Technical Specifications (TS) for the Diablo Canyon
Power Plant Unit Nos. 1 and 2 to revise TS 6.2.2.g and 6.3 to change
the name of the Operations Manager to Operations Director and to change
the requirement for the Operations Director to hold a senior reactor
operator (SRO) license.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:

1. The proposed change does not involve a significant increase
in the probability or consequences of an accident previously
evaluated.
The proposed change to revise the title of the Operations
Manager to Operations Director is an administrative change that
clarifies the Technical Specification (TS) to reflect current
position titles.
The proposed change provides assurance that the Operations
Director will continue to have knowledge of pressurized water
reactor (PWR) operation and emergency event mitigation. The proposed
change does not detract from the Operations Director's ability to
perform his primary responsibilities. In this case, by having
previously held a senior reactor operator (SRO) license, the
Operations Director has achieved the necessary training, skills, and
experience to fully understand the operation of plant equipment and
the watch requirements for operators. In summary, the proposed
change does not affect the ability of the Operations Director to
provide the plant oversight required of his position.
Additionally, another off-shift individual that holds an SRO
license for Diablo Canyon Power Plant (DCPP) directs the licensed
activities of licensed operators (an Operations middle manager) will
have specific knowledge of operation and emergency event mitigation
at DCPP. This will assure that the change in qualification of the
Operations Director does not affect the probability of an operator
initiating an accident or increasing the consequences of an accident
due to improper direction from management. The training and
qualification programs for operators on shift will not be affected
by the proposed changes.
Therefore, the proposed changes do not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
2. The proposed change does not create the possibility of a new
or different kind of accident from any accident previously
evaluated.
The proposed change to revise the title of the Operations
Manager to Operations Director is an administrative change that
clarifies the TS to reflect current position titles.
The proposed change to TS 6.2.2g. and 6.3 do not affect the
design or function of any plant system, structure, or component, nor
does it change the way plant systems are operated. It does not
affect the performance of NRC licensed operators since the proposed
changes do not impact the training or qualification of any operator
on shift. Operation of the plant in conformance with TS and other
license requirements will continue to be supervised by personnel who
hold an SRO license. The proposed change to TS 6.2.2g and 6.3
ensures that the Operations Director will be a knowledgeable and
qualified individual by requiring the individual to have held an SRO
license at a PWR.
Therefore, the proposed changes do not create the possibility of
a new or different kind of accident from any accident previously
evaluated.
3. The proposed change does not involve a significant reduction
in a margin of safety.
The proposed change to revise the title of the Operations
Manager to Operations Director is an administrative change that
clarifies the TS to reflect current position titles.
The proposed

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Source: Frix Law Library, https://www.frixlaw.com/law-library/documents/fr%3A98-11911. Public record. Not legal advice.