# Biweekly Notice

> Briefs, arguments, decisions, and more.

URL: https://www.frixlaw.com/law-library/documents/fr%3A97-30217

## Record

- **Collection:** Federal Register
- **Document type:** Notice
- **Published:** November 19, 1997
- **Citation:** 62 FR 61836

## Text

NUCLEAR REGULATORY COMMISSION

Biweekly Notice

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

I. Background

Pursuant to Public Law 97-415, the U.S. Nuclear Regulatory
Commission (the Commission or NRC staff) is publishing this regular
biweekly notice. Public Law 97-415 revised section 189 of the Atomic
Energy Act of 1954, as amended (the Act), to require the Commission to
publish notice of any amendments issued, or proposed to be issued,
under a new provision of section 189 of the Act. This provision grants
the Commission the authority to issue and make immediately effective
any amendment to an operating license upon a determination by the
Commission that such amendment involves no significant hazards
consideration, notwithstanding the pendency before the Commission of a
request for a hearing from any person.
This biweekly notice includes all notices of amendments issued, or
proposed to be issued from October 27, 1997, through November 6, 1997.
The last biweekly notice was published on November 5, 1997 (62 FR
59912).

Notice Of Consideration Of Issuance Of Amendments ToFacility
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

[[Page 61837]]

expiration of the 30-day notice period. However, should circumstances
change during the notice period such that failure to act in a timely
way would result, for example, in derating or shutdown of the facility,
the Commission may issue the license amendment before the expiration of
the 30-day notice period, provided that its final determination is that
the amendment involves no significant hazards consideration. The final
determination will consider all public and State comments received
before action is taken. Should the Commission take this action, it will
publish in the Federal Register a notice of issuance and provide for
opportunity for a hearing after issuance. The Commission expects that
the need to take this action will occur very infrequently.
Written comments may be submitted by mail to the Chief, Rules and
Directives Branch, Division of Freedom of Information and Publications
Services, Office of Administration, U.S. Nuclear Regulatory Commission,
Washington, DC 20555-0001, and should cite the publication date and
page number of this Federal Register notice. Written comments may also
be delivered to Room 6D22, Two White Flint North, 11545 Rockville Pike,
Rockville, Maryland from 7:30 a.m. to 4:15 p.m. Federal workdays.
Copies of written comments received may be examined at the NRC Public
Document Room, the Gelman Building, 2120 L Street, NW., Washington, DC.
The filing of requests for a hearing and petitions for leave to
intervene is discussed below.
By December 19, 1997, 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 61838]]

Baltimore Gas and Electric Company, Docket Nos. 50-317 and 50-318,
Calvert Cliffs Nuclear Power Plant, Unit Nos. 1 and 2, Calvert
County, Maryland

Date of amendments request: October 22, 1997
Description of amendments request: The proposed amendment
incorporates both steady state and transient degraded voltage setpoints
into Technical Specifications, as opposed to the current single
degraded voltage setpoint. The proposed changes ensure adequate
terminal voltage to all safety-related equipment during steady state
and transient voltage conditions. Additionally, the 4 kV voltage range
required during testing of the emergency diesel generators (EDGs) will
be decreased to ensure the new steady state degraded voltage relays are
not actuated during testing and to ensure the 4 kV motors are operated
within their voltage rating.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. Would not involve a significant increase in the probability
or consequences of an accident previously evaluated.
The proposed changes revise the current degraded voltage
setpoint and adds an additional steady state undervoltage
requirement to Unit 1 and 2 Technical Specifications. The current
degraded voltage relays will be referred to as
transient degraded voltage relays.'' The new settings
allow for calibration tolerances, potential transformer correction
factors, test equipment uncertainties, and relay drift. The nominal
settings account for the above factors, plus additional margin to
the analytical limit. The acceptable voltage range during EDG
surveillance testing is also being decreased. The setpoint and time
delay associated with the 4 kV bus loss of voltage relays is
unaffected by this amendment request.
The accident analyses credit the loading of the EDGs based on
loss of offsite power. The 4 kV emergency bus loss of voltage and
degraded voltage relays initiate starting and loading of the
emergency diesel generators (EDGs) when the preferred power source
voltage is lost or drops below a predetermined value. The relays
also initiate disconnection of the preferred power source from the 4
kV emergency busses. These actions ensure adequate terminal voltage
to all safety-related electrical equipment required to support
accident mitigation. The required voltage necessary to ensure
safety-related motors are capable of starting is 75 percent of
nominal rated equipment voltage. The required voltage necessary to
ensure these motors continue running for extended periods is 90
percent of nominal rated equipment voltage.
The degraded (transient) voltage setpoint is being changed from
3628 [plus or minus] 25 Volts to 3710 [plus or minus] 80 Volts.
Based on the most recent calculations, a minimum voltage of 3630
Volts is required to ensure at least 75 percent of the nominal
voltage is available to No. 13 Charging Pump, which is the most
limiting electrical load.
The new steady state degraded voltage relay setpoint will be
established at 3900 [plus or minus] 80 Volts. The setpoint ensures
that there is at least 90 percent of nominal voltage available to
No. 13 Charging Pump. The time delay associated with this actuation
is 101 [plus or minus] 3.5 seconds. The time delay provides adequate
time for the voltage regulator to recover bus voltage following a
voltage swing on the 500 kV system and time for the EDG voltage
regulator to stabilize. The steady state degraded voltage relays
will be tested in the same manner, and at the same frequency, as the
loss of voltage and transient degraded voltage relays.
The required voltage range during EDG surveillance testing is
being revised from 4160 [plus or minus] 420 Volts to 4160 +240, -100
Volts. The surveillance requirement verifies that the EDG voltage
regulator is maintaining an acceptable voltage. The
new value ensures the 4 kV motors are operated within their
rated voltage and prevents actuation of the steady state degraded
voltage relay during surveillance testing.
The degraded voltage relays are not initiators in any previously
evaluated accidents. Additionally, decreasing the acceptable voltage
range during EDG testing does not affect the initiation of any
previously analyzed accidents. Therefore, the proposed changes do
not involve an increase in the probability or consequences of an
accident previously analyzed.
2. Would not create the possibility of a new or different type
of accident from any accident previously evaluated.
The license amendment request revises the current degraded
voltage setpoint and adds an additional steady state degraded
voltage requirement. Additionally, the acceptable voltage range
during EDG surveillance testing is being decreased. The proposed
changes ensure adequate starting and running terminal voltage to all
safety-related electrical equipment during steady state and
transient degraded voltage conditions. The addition of the steady
state degraded voltage relays provide an extra scheme of protection
against sustained degraded voltage conditions. The facility
currently relies upon degraded voltage relays to start and load the
EDGs and to disconnect the preferred power source from the 4 kV
emergency busses. Therefore, revising the relay setpoint, adding
additional steady state degraded voltage protection, and decreasing
the acceptable voltage range during EDG testing does not create the
possibility of a new or different type of accident from any accident
previously evaluated.
3. Would not involve a significant reduction in a margin of
safety.
The safety function of the degraded voltage relays is to ensure
that the preferred power source is disconnected from the 4 kV
emergency busses during loss of voltage or degraded voltage
conditions. The relays also ensure the EDGs are started and loaded.
Ultimately, these actions ensure the minimal terminal voltage
necessary to start and run all safety-related electrical equipment
is maintained. The proposed changes revise the current degraded
voltage setpoint and adds an additional steady state undervoltage
requirement. Additionally, the acceptable voltage range during EDG
surveillance testing is being decreased to ensure actuation of the
steady state degraded voltage relays does not occur during
EDG testing, and to ensure the 4 kV motors are operated within
their rated voltage range.
Because the proposed changes ultimately ensure adequate terminal
voltage to all safety-related electrical equipment during transient
and steady state undervoltage conditions, the safety function of the
degraded voltage relays, as well as the margin of safety afforded by
these relays is unchanged.
Therefore, the changes do not involve a significant reduction in
the margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendments request involves no significant hazards consideration.
Local Public Document Room location: Calvert County Library, Prince
Frederick, Maryland 20678
Attorney for licensee: Jay E. Silbert, Esquire, Shaw, Pittman,
Potts and Trowbridge, 2300 N Street, NW., Washington, DC 20037
NRC Project Director: S. Singh Bajwa, Director

Carolina Power & Light Company, Docket No. 50-261, H. B. Robinson
Steam Electric Plant, Unit No. 2, Darlington County, South Carolina

Date of amendment request: October 2, 1997
Description of amendment request: The proposed amendment would
address an unreviewed safety question associated with the analysis of a
fuel handling accident in the Fuel Storage Building as described in
Section 15.7.4, ``Design Basis Fuel Handling Accidents,'' of the H.B.
Robinson Steam Electric Plant (HBR) Updated Final Safety Analysis
Report (UFSAR). Carolina Power & Light Company (the licensee)
determined that an assumption used in the accident analysis for depth
of water above the top of irradiated fuel in the spent fuel pit was
non-conservative. The accident analysis assumed a depth of 23 feet
instead of the correct value of 21 feet. The licensee has submitted a
revised accident analysis using the correct assumption and has proposed
that the UFSAR be

[[Page 61839]]

changed to incorporate the results of the revised analysis.
Basis for proposed no significant hazards 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 to the UFSAR is to change assumptions
associated with the evaluation of a fuel handling accident in the
Fuel Storage Building. The change in assumptions is to reduce the
decontamination factor associated with the removal of elemental
iodine from the spent fuel pool water. Because the decontamination
factor for elemental iodine is reduced, the consequences of a fuel
handling accident in the Fuel Storage Building is [sic] increased.
However, because the radiological consequences remain well within
the exposure guideline values of 10 CFR 100, paragraph 11 (i.e., 25%
or less of the values), the increase in consequences is not
significant. The change in assumptions for the fuel handling
accident in the Fuel Storage Building do [sic] not affect operation,
maintenance, or design of equipment associated with the handling of
fuel in the Fuel Storage Building, therefore, the probability of a
fuel handling accident as previously evaluated is not changed.
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 any physical alteration of
plant systems, structures or components, changes in parameters
governing normal plant operation, or methods of operation. The
proposed change does not introduce a new mode of operation or
changes in the method of normal plant operation. Therefore, 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 proposed change to the UFSAR to change the assumptions
associated with a fuel handling accident in the Fuel Storage
Building is to change the assumption for the decontamination factor
for elemental iodine to a smaller value. The new assumption for
elemental iodine decontamination factor preserves the approximate
factor of 24 margin between experimental data for elemental iodine
decontamination factor and the assumed value provided in NRC Safety
Guide 25. 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: Hartsville Memorial Library,
147 West College Avenue, Hartsville, South Carolina 29550
Attorney for licensee: R. E. Jones, General Counsel, Carolina Power
& Light Company, Post Office Box 1551, Raleigh, North Carolina 27602
NRC Project Director: James E. LyonsCommonwealth Edison Company,
Docket Nos. 50-237 and 50-249, Dresden Nuclear Power Station, Units 2
and 3, Grundy County, Illinois Docket Nos. 50-254 and 50-265, Quad
Cities Nuclear Power Station, Units 1 and 2, Rock Island County,
Illinois Date of application for amendment request: September 30, 1997
Description of amendment request: This request changes the
Technical Specifications (TS) by adding a new Section 3/4.12.C,
``Inservice Leak and Hydrostatic Testing Operation,'' to allow certain
reactor coolant pressure tests to be performed in MODE 4 when the
metallurgical characteristics of the reactor pressure vessel require
the pressure testing at or approaching temperatures
212 deg.F, which normally correspond with MODE 3.
Basis for proposed no significant hazards determination: As
required by 10 CFR 50.91(a), the licensee has provided its analysis of
the issue of no significant hazards consideration, which is presented
below: 1) Involve a significant increase in the probability or
consequences of an accident previously evaluated because of the
following:
The proposed amendment represents the addition of a Special Test
Exception to perform Pressure Testing Operations consistent with the
requirements of Section 3.10.1 of the Improved Standard Technical
Specifications (NUREG-1433). The proposed changes are consistent
with the current plant safety analyses. Implementation of these
changes will provide continued assurance that specified parameters
associated with Pressure Testing Operations will remain within their
acceptance limits, and as such, will not significantly increase the
probability or consequences of a previously evaluated accident.
The proposed changes are based on requirements specified by
Section 3.10.1 of NUREG-1433. Any such changes are consistent with
the current plant safety analyses and have been determined to
represent sufficient requirements for the assurance and reliability
of equipment assumed to operate in the safety analyses, or provide
continued assurance that specified parameters associated with
Pressure Testing Operations remain within their acceptance limits.
As such, these changes will not significantly increase the
probability or consequences of a previously evaluated accident.
The associated systems affecting Pressure Testing Operations
related to this proposed amendment are not assumed in any analyses
to initiate any accident sequence; therefore, the probability of any
accident previously evaluated is not increased by this proposed
amendment which incorporates the requirements of Section 3.10.1 of
NUREG-1433. In addition, the proposed limiting conditions for
operation and surveillance requirements for the proposed amendment
ensure a level of equipment operability sufficient to mitigate any
operational occurrences which could occur while operating under this
Special Test Exception. Furthermore, any operational occurrence
postulated during operation under this Special Test Exception is
bounded by the Design Basis Accidents. Therefore, the proposed
amendment does not increase the consequences of nay accident
previously evaluated.
There is no change to the consequences of an accident previously
evaluated because Pressure Testing Operations does not adversely
affect either the on-site or off-site does consequences resulting
from an accident. In addition, Pressure Testing Operations is not an
accident initiator. As such, there is no adverse impact on the
probability of accident initiators. Thus, there is no significant
increase in the probability of any previously analyzed accident.
2) Create the possibility of a new or different kind of accident
from any accident previously evaluated because:
The proposed amendment represents the conversion of current
Technical Specification requirements to maintain consistency with
those requirements specified in Section 3.10.1 of NUREG-1433. The
proposed changes are consistent with the current plant safety
analyses. These proposed changes do not involve revisions to the
design of the station. In addition, the proposed limiting conditions
for operation and surveillance requirements for the proposed
amendment ensure a level of equipment operability sufficient to
mitigate any operational occurrences which could occur while
operating under the Special Test Exception. Some of the changes may
involve revision in the testing of components at the station;
however, these are in accordance with the current plant safety
analyses. The proposed changes will not introduce new failure
mechanisms beyond those already considered In the current plant
safety analyses.
The associated systems that affect Pressure Testing Operations
related to the proposed amendment, are not assumed in any plant
safety analysis to initiate any accident sequence. In addition, the
proposed surveillance requirements for any such affected systems are
consistent with the requirements of Section 3.10.1 of NUREG-1433.
Therefore, the possibility of a new or different kind of accident
from any accident previously evaluated is not created.
3) Involve a significant reduction in the margin of safety
because:
ComEd proposes to revise the Technical Specifications to be
consistent with those provisions specified in Section 3.10.1 of
NUREG-1433. The proposed changes are consistent with the current
plant safety analyses. In addition, these proposed changes do not
involve revisions to the design of the station. As such, the
proposed individual changes will maintain the same level of

[[Page 61840]]

reliability of the equipment associated with Pressure Testing
Operations, assumed to operate in the plant safety analysis, or
provide continued assurance that specified parameters affecting,
will remain within their acceptance limits. Therefore, the proposed
changes provide continued assurance of Pressure Testing Operations
without adversely affecting the public health and safety and as
such, will not significantly reduce existing plant safety margins.
The proposed amendment to the Technical Specifications
implements present requirements, or the requirements in accordance
with the guidelines set forth in Section 3.10.1 of NUREG-1433. The
proposed changes have been evaluated and found to be acceptable for
use at the stations based on system design, safety analysis
requirements, and operational performance. Since the proposed
changes are based on NRC accepted provisions that are applicable at
the stations and maintain necessary levels of system or component
reliability affecting Pressure Testing Operations, the proposed
changes do not involve a significant reduction in the margin of
safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92 are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: for Dresden, Morris Area
Public Library District, 604 Liberty Street, Morris, Illinois 60450;
for Quad Cities, Dixon Public Library, 221 Hennepin Avenue, Dixon,
Illinois 61021
Attorney for licensee: Michael I. Miller, Esquire; Sidle and
Austin, One First National Plaza, Chicago, Illinois 60603
NRC Project Director: Robert A. Capra

Commonwealth Edison Company, Docket Nos. 50-373 and 50-374, LaSalle
County Station, Units 1 and 2, LaSalle County, Illinois

Date of amendment request: September 26, 1997
Description of amendment request: The proposed amendments would
revise the Technical Specifications to (1) prohibit the simultaneous
opening of the drywell and suppression chamber purge system isolation
valves, (2) upgrade the ventilation filter testing program to the
latest industry standards, and (3) specify that the auxiliary electric
equipment room is required to be habitable during design bases
accidents.
Basis for proposed no significant hazards determination: As
required by 10 CFR 50.91(a), the licensee has provided its analysis of
the issue of no significant hazards consideration, which is presented
below:
1) Involve a significant increase in the probability or
consequences of an accident previously evaluated because:
a. Drywell and Suppression Chamber Purge System
The purpose of the drywell and suppression chamber purge system
isolation valves is to mitigate the consequences of a design bases
accident. Operation of these valves will have no effect on the
probability of a design bases accident occurring.
The current TS 3.6.1.8 allows for the drywell and suppression
chamber purge system isolation valves to be open simultaneously. In
this condition, containment pressure and offsite dose during design
bases accidents would be greater than previously evaluated. The
proposed revision to TS 3.6.1.8 would prevent the simultaneous
opening of the drywell and suppression chamber purge system
isolation valves thus assuring that the consequences of design bases
accidents previously evaluated are still bounding.
b. Ventilation Filter Testing Program
The SBGTS [Standby Gas Treatment System] and Control Room and
AEER [Auxiliary Electric Equipment Room] Emergency Filtration
Systems are designed to mitigate the radiological consequences of
previously evaluated design bases accidents. Operation and testing
of these systems will have no effect on the probability of a design
bases accident occurring.
The proposed revisions associated with this change relocate the
requirements for SBGTS and Control Room and AEER Emergency
Filtration System filter testing from the current TS SRs to a new TS
administrative control program. The testing requirements are being
upgraded to the latest industry standards. Filter testing in
accordance with the proposed program will ensure that Title 10, Code
of Federal Regulations, Part 50 (10 CFR 50), Appendix A, General
Design Criteria (GDC) 19 and 10 CFR 100 limits are not exceeded.
c. Other Control Room and Auxiliary Electric Equipment Room
Emergency Filtration System Changes
The SBGTS and Control Room and AEER Emergency Filtration System
are designed to mitigate the radiological consequences of previously
evaluated design bases accidents. Operation and testing of these
systems will have no effect on the probability of a design bases
accident occurring.
The proposed revisions associated with this change acknowledge
that the AEERs are required to be habitable during design bases
accidents. This is consistent with the plants design bases.
d. Editorial Changes
The proposed revisions to TS 6.2.F.7 reformat the requirement to
establish consistency with the remainder of TS 6.2.F. There are no
technical changes being proposed.
Based upon the above, the proposed amendment will not increase
the probability or consequences of any accident previously
evaluated.
2) Create the possibility of a new or different kind of accident
from any accident previously evaluated because:
a. Drywell and Suppression Chamber Purge System
No new plant equipment is being installed, and use of currently
installed plant equipment is not affected by this proposed change.
The proposed revision to TS 3.6.1.8 provides additional limitations
on the opening of the drywell and suppression chamber purge system
isolation valves.
b. Ventilation Filter Testing Program
No new plant equipment is being installed, and use of currently
installed plant equipment is not affected by this proposed change.
These proposed revisions will demonstrate operability of the Control
Room and AEER Emergency Filtration System using the latest industry
standards.
c. Other Control Room and Auxiliary Electric Equipment Room
Emergency Filtration System Changes
No new plant equipment is being installed, and use of currently
installed plant equipment is not affected by this proposed change.
These proposed revisions will demonstrate habitability of the AEER
by imposing operability requirements on the AEER recirculation
filter units.
d. Editorial Changes
The proposed revisions to TS 6.2.F.7 reformat the requirement to
establish consistency with the remainder of TS 6.2.F. There are no
technical changes being proposed.
Based upon the above, the proposed change will not create the
possibility of a new or different kind of accident or transient
previously evaluated.
3) Involve a significant reduction in the margin of safety
because:
a. Drywell and Suppression Chamber Purge System
The current TS 3.6.1.8 requirements are non-conservative with
respect to the assumptions used when evaluating steam bypass of the
suppression chamber; specifically, a maximum allowable leakage area
of 0.03 square feet with the only credible leakage path was assumed
to be suppression chamber vacuum breaker valve seat leakage. This
proposed revision to TS 3.6.1.8 will make the TS requirements
consistent with those assumptions.
b. Ventilation Filter Testing Program
These proposed revisions will ensure operability of the Control
Room and Auxiliary Electric Equipment Room (AEER) Emergency
Filtration system using the latest industry standards. Filter
testing in accordance with the proposed program will ensure that GDC
19 and 10 CFR 100 limits are not exceeded.
c. Other Control Room and Auxiliary Electric Equipment Room
Emergency Filtration System Changes
These proposed revisions will ensure operably of the control
room and AEER Emergency Filtration System by demonstrating system
performance with the control room and AEER recirculation filter
units to ensure GDC 19 limits are not exceeded.
d. Editorial Changes
The proposed revisions to TS 6.2.F.7 reformat the requirement to
establish consistency with the remainder of TS 6.2.F. There are no
technical changes being proposed.

[[Page 61841]]

Based on the above, the proposed TS change does not involve a
significant reduction in the margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
requested amendments involve no significant hazards consideration.
Local Public Document Room location:Jacobs Memorial Library,
Illinois Valley Community College, Oglesby, Illinois 61348
Attorney for licensee: Michael I. Miller, Esquire; Sidley and
Austin, One First National Plaza, Chicago, Illinois 60603
NRC Project Director: Robert A. Capra

Commonwealth Edison Company, Docket Nos. 50-373 and 50-374, LaSalle
County Station, Units 1 and 2, LaSalle County, Illinois

Date of amendment request: October 15, 1997
Description of amendment request: The proposed amendments would
revise Technical Specification Table 4.3.7.5-1, Accident Monitoring
Instrumentation Surveillance Requirements, by deleting a footnote that
provides details concerning the calibration requirements for the
drywell hydrogen concentration analyzer and monitor.
Basis for proposed no significant hazards determination: As
required by 10 CFR 50.91(a), the licensee has provided its analysis of
the issue of no significant hazards consideration, which is presented
below:
1) Involve a significant increase in the probability or
consequences of an accident previously evaluated because:
The drywell hydrogen concentration analyzer and monitors are
required to be operable by TS 3/4.7.5, Accident Monitoring
Instrumentation. Table 4.3.7.5-1, Accident Monitoring
Instrumentation Surveillance Requirements, includes a footnote
providing unnecessary details related to the calibration of this
specific analyzer and monitors. The footnote provides information
that was determined to put the hydrogen analyzers and monitors
outside of the design basis by limiting the range of the indication
to 0% to 4% hydrogen in the drywell. The calibration method is being
corrected to provide the correct range of 0% to 10%, and requires
this note in the TS to be changed or deleted. The footnote is
proposed to be deleted from the TS, because it provides unnecessary
detail.
Deletion of the footnote will not cause an increase in the
probability of an accident, because this instrumentation is only for
accident monitoring instrumentation and thus does not affect
accident initiators or assumptions.
Deletion of the footnote will not change the consequences of an
accident previously evaluated, because this detail in the TS does
not change the requirement of performing a channel calibration at
the specified frequency. In addition, the ability to monitor
hydrogen during an accident will not be affected by deletion of the
footnote.
Therefore, this change does not involve an increase in the
probability or consequences of an accident previously evaluated.
2) Create the possibility of a new or different kind of accident
from any accident previously evaluated because:
This is monitoring instrumentation only. Deletion of the
footnote concerning specifics on how to calibrate this
instrumentation will not affect the reliability or failure modes of
the drywell hydrogen concentration analyzer and monitors. Therefore,
this change will not create the possibility of a new or different
kind of accident from any accident previously evaluated.
3) Involve a significant reduction in the margin of safety
because:
This is monitoring instrumentation only. Deletion of the
footnote concerning specifics on how to calibrate this
instrumentation will not change the requirement to perform Channel
Calibrations at the frequency specified in the TS. The details of
how to perform a Channel Calibration on the drywell hydrogen
concentration analyzer and monitors are located in plant procedures
and are in accordance with vendor recommendations. The TS
requirements for redundancy of the instrumentation and the actions
to be taken for inoperable instrumentation are also not affected by
the deletion of this footnote.
This change to the level of information regarding this
calibration is consistent with the detail for this and other
instrumentation in NUREG-1434, Revision 1, Standard Technical
Specifications, General Electric Plants, BWR/6.
Therefore, deletion of footnote * from TS Table 4.3.7.5-1 will
not involve a 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:Jacobs Memorial Library,
Illinois Valley Community College, Oglesby, Illinois 61348
Attorney for licensee: Michael I. Miller, Esquire; Sidley and
Austin, One First National Plaza, Chicago, Illinois 60603
NRC Project Director: Robert A. Capra

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

Date of amendment request: October 6, 1997
Description of amendment request: The proposed amendments would
delete all references to the steam line low pressure safety injection
function.
Basis for proposed no significant hazards 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?
Answer
Probability
Accident initiators can affect the probability of a previously
evaluated accident. The addition of a new device or piece of
equipment to the plant may introduce a new accident initiator. No
new equipment is added to the plant as a result of this change. The
proposed removal of the low steam line steam pressure will involve
removing the steam line pressure safety injection function. This
results in a reduction in the likelihood of spurious safety
injections. Spurious safety injections can result in inadvertent
ECCS [emergency core cooling system] actuations. Inadvertent ECCS
Actuation is a UFSAR [updated Final Safety Analysis Report] accident
(UFSAR 15.5.1). Therefore, this change will result in a reduction in
the probability of an accident previously evaluated.
Routine plant operating practices and conditions will not be
altered by the removal of the safety injection function. Therefore,
there is no operating practice or condition change that could
increase the probability of occurrence of a previously evaluated
accident.
There is no significant increase in the probability of an
accident previously evaluated.
Consequences
Accidents previously evaluated that could be adversely affected
are the steam line break and the feedwater line break. These
accidents will result in secondary side depressurization with
pressure reaching the current actuation setpoint. The review of
these accidents found that the consequences of the previous accident
analysis acceptance criteria remain satisfied. The specifics of the
accident analysis is discussed below.
The steam line break accident was analyzed to demonstrate short
term cooling capability. A spectrum of break sizes were evaluated to
determine the limiting break size. For smaller breaks (including the
limiting break size), the safety injection actuation on low
pressurizer pressure occurs prior to low steam line pressure safety
injection. However, for larger steam line breaks the setpoint for
low steam line pressure safety injection is reached prior to low
pressurizer pressure safety injection. The larger spectrum of breaks
were analyzed without credit for the low steam line pressure safety
injection. The results of this analysis found that there would be a
slight increase in time required for safety injection to actuate.
The low pressurizer safety injection would actuate in these
accidents due to the cooldown and depressurization of the reactor
coolant system in response to the secondary side energy removal. The
Departure from

[[Page 61842]]

Nucleate Boiling Ratios (DNBRs) were analyzed with this time delay
in safety injection. The DNBRs for these cases were found to be less
limiting than those calculated for the limiting break size.
Therefore, the removal of steam line low pressure safety injection
does not adversely affect the DNBR, fuel failure or dose
consequences of the main steam line break accident. Other acceptance
criteria would not be expected to be affected by the small change in
timing of the safety injection signal.
In addition, to the Chapter 15 accident analysis, the Chapter 6
containment response to mass and energy releases was evaluated
without credit for steam line low pressure safety injection. The
evaluation demonstrated that for steam line breaks inside of
containment, the high containment pressure safety injection set
point is reached prior to the pressure associated with steam line
low pressure safety injection. Therefore the existing containment
response evaluation is not adversely affected by the removal of the
low steam pressure safety injection. This also assures that the
existing environmental qualification envelope for McGuire is not
affected by this change. For steam line breaks outside of
containment the maximum required breaksize is 1.0 ft2, which results
in transients with safety injection caused by low pressurizer
pressure prior to low steam line pressure safety injection.
The feedwater line break accidents were analyzed to demonstrate
long term core cooling capability. During a feedwater line break,
the secondary system will depressurize if the break occurs between
the main feedwater check valve and the steam generator. However,
breaks are only required to occur at the terminal ends of feedwater
piping (i.e., at the feedwater pump or at the steam generator). For
a feedwater line break at the main feedwater pump, the main feed
check valve will prevent depressurization of the steam generator.
For a feedwater line break at the steam generator, a safety
injection on high containment pressure will occur prior to safety
injection on steam pressure. Therefore, the elimination of the steam
line low pressure safety injection does not adversely impact the
feedwater line break accident.
In summary, a review was conducted of all design basis accidents
to identify those which result in a low steam pressure safety
injection. These accidents were then evaluated to verify that the
accident analysis were within acceptance criteria. This review
revealed that all accident analysis results were within current
analysis acceptance criteria.
Therefore, there is no significant increase in the consequences
of a previously evaluated accident.
Conclusion
Elimination of the low steam line pressure safety injection
results in no 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[?]
Answer
There is no introduction of new equipment or operating practices
that could result in a new operating condition. The plant will
continue to operate in the same method with the same complement of
equipment with the exception of the actuation logic associated with
the steam line low pressure safety injection. Therefore, there is no
new operating condition that would be expected to generate a new
sequence of events which could generate a new or different accident.
There is no new equipment that could interact with other plant
structures, systems or components.
The low pressure safety injection equipment is the only plant
equipment affected by this change. There are no new equipment
failure modes which might result in a new or different accident.
Affected accidents were evaluated to validate that the accident
sequence would not deviate in a fashion which would create a new or
different accident. The analysis of the feedwater line break and
steam line break did not reveal any new or different type of
accident.
Removal of the low steam line pressure safety injection will not
create the possibility of a new or different kind of accident from
any accident previously evaluated;
(OR)
3. Involve a significant reduction in the margin of safety?
Answer
The margin of safety relevant to this change is represented by
the margin of physical protection provided by fuel cladding and the
reactor containment. Effects of this change on the safety analysis
was described under question 1 above. The results of the analysis
demonstrate that DNBR, fuel clad integrity and containment response
were not significantly affected by the removal of low steam line
pressure safety injection. Therefore, the physical protection
provide[d] by the fuel cladding and reactor containment were not
affected by this change. Accident acceptance criteria continued to
be met without credit for the safety function. The radiological
consequences of accidents was not affected by the change.
The removal of the low steam line pressure safety injection did
not significantly reduce the margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 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, North Carolina

Northeast Nuclear Energy Company (NNECO), et al., Docket No. 50-
423, Millstone Nuclear Power Station, Unit No. 3, New London
County, Connecticut

Date of amendment request: October 15, 1997
Description of amendment request: The proposed amendment would
affect nominal trip setpoints and allowable values for Reactor Trip
System (RTS) Instrumentation Trip Setpoints Table 2.2-1, and Engineered
Safety Features Actuation System (ESFAS) Instrumentation Trip Setpoints
Table 3.3-4. In addition, the proposed amendment would (1) decrease the
reactor trip setpoint for the reactor coolant pump (RCP) low shaft
speed (underspeed trip setpoint) from 95.8 percent to 92.4 percent of
rated speed, (2) make editorial changes, and (3) change the Bases to
reflect the new methodology.
Basis for proposed no significant hazards 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:
NNECO has reviewed the proposed revision in accordance with
10CFR50.92 and has concluded that the revision does not involve a
significant hazards consideration (SHC). The basis for this
conclusion is that the three criteria of 10CFR50.92(c) are not
satisfied. The proposed revision does not involve [an] SHC because
the revision would not:
1. Involve a significant increase in the probability or
consequence of an accident previously evaluated.
The proposed changes to Tables 2.2-1 and 3.3-4 involve changes
from a five column format to a two column format. The RTS trip
setpoints and ESFAS trip setpoints remain unchanged with the
exception of the RCP low shaft speed trip setpoint discussed below.
Detailed operability criteria will be moved to surveillance
procedures and analysis has demonstrated that an adequate margin for
normal trip setpoints exist and safety analysis limits are preserved
in all RTS/ESFAS functions.
Changing the RCP low shaft speed trip setpoint will not change
the probability of occurrence of the event. The existing accident
analysis (Millstone Unit No. 3 FSAR [final safety analysis report]
section 15.3.2) of the complete loss of forced reactor coolant flow
remains valid for the proposed change. Therefore, the change to the
RCP low shaft speed trip setpoint does not increase the probability
or consequences of any previously analyzed accident.
In addition, the proposed changes to Tables 2.2-1 and 3.3-4 do
not alter the intent or method by which the surveillances are
conducted. Therefore, the scope of evaluation performed gives
reasonable assurance that there will not be an adverse impact on the
consequences or the probability of any previously analyzed accident.
Therefore, the proposed revision does not involve a significant
increase in the probability or consequence of an accident previously
evaluated.
2. Create the possibility of a new or different kind of accident
from any accident previously evaluated.

[[Page 61843]]

The existing design basis adequately covers the plant response
with the proposed change to the RCP low shaft speed trip setpoint.
The change does not introduce new failure modes.
The proposed changes to Tables 2.2-1 and 3.3-4 do not modify the
design or operation of any plant system. The proposed changes do not
alter the intent or method by which the surveillances are conducted,
other than adjusting the allowable values to reflect historical
instrument performance data. Therefore, the proposed revision does
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 to Tables 2.2-1 and 3.3-4 modify the
existing five column format to a two column format to show the RTS
and ESFAS nominal trip setpoints and the process rack bistable
allowable values for individual functions. Detailed operability
criteria will be moved to the surveillance procedures. With the
exception of the low shaft speed trip discussed below, the RTS and
ESFAS setpoints remain unchanged and analysis has demonstrated that
an adequate margin for normal trip setpoints exist and safety
analysis limits are preserved in all RTS/ESFAS functions.
Since the safety limits of the design are still met, the
proposed change to the RCP low shaft speed trip setpoint does not
reduce the margin of safety.
Therefore, the proposed revision does not involve a significant
reduction in a margin of safety.
In conclusion, based on the information provided, it is
determined that the proposed revision does not involve an SHC.
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

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 requests: March 10, 1997, as supplemented by
letter dated May 20, 1997
Description of amendment requests: The proposed amendments would
revise the combined Technical Specifications (TS) for the Diablo Canyon
Power Plant Unit Nos. 1 and 2 to revise TS 3/4.4.5 and 3.4.6.2,
including associated Bases 3/4.4.5 and 3/4.4.6.2, to allow the
implementation of steam generator (SG) tube alternate repair criteria
for axial indications in the Westinghouse explosive tube expansion
(WEXTEX) region below the top of the tubesheet and below the bottom of
the WEXTEX transition that may exceed the current TS depth-based
plugging limit. The allowed primary-to-secondary operational leakage
from any one SG would be reduced from 500 gpd to 150 gpd.
Basis for proposed no significant hazards 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.
Probability
Of the various accidents previously evaluated, the proposed
changes only affect the steam generator tube rupture (SGTR) event
evaluation and the postulated steam line break (SLB) accident
evaluation. Loss-of-coolant accident (LOCA) conditions cause a
compressive axial load to act on the tube. Therefore, since the LOCA
tends to force the tube into the tubesheet rather than pull it out,
it is not a factor in this amendment request. Another faulted load
consideration is a safe shutdown earthquake (SSE); however, the
seismic analysis of Series 51 steam generators has shown that axial
loading of the tubes is negligible during an SSE.
For the SGTR event, the required structural margins of the steam
generator tubes will be maintained by the presence of the tubesheet.
Tube rupture is precluded for cracks in the Westinghouse explosive
tube expansion (WEXTEX) region due to the constraint provided by the
tubesheet. Therefore, Regulatory Guide (RG) 1.121, ``Bases for
Plugging Degraded PWR Steam Generator Tubes,'' margins against burst
are maintained for both normal and postulated accident conditions.
The W* length supplies the necessary resistive force to preclude
pullout loads under both normal operating and accident conditions.
The contact pressure results from the WEXTEX expansion process,
thermal expansion mismatch between the tube and tubesheet and from
the differential pressure between the primary and secondary side.
Therefore, the proposed change results in no significant increase in
the probability of the occurrence of an SGTR or SLB accident. 1
The proposed changes do not affect other systems, structures,
components or operational features. Therefore, based on the above
evaluation, the proposed changes do not involve a significant
increase in the probability of an accident previously evaluated.
Consequences
The consequences of an SGTR event are affected by the primary-
to-secondary leakage flow during the event. Primary-to-secondary
leakage flow through a postulated broken tube is not affected by the
proposed change since the tubesheet enhances the tube integrity in
the region of the WEXTEX expansion by precluding tube deformation
beyond its initial expanded outside diameter. The resistance to both
tube rupture and collapse is strengthened by the tubesheet in that
region. At normal operating pressures, leakage from primary water
stress corrosion cracking (PWSCC) in the W* length is limited by
both the tube-to-tubesheet crevice and the limited crack opening
permitted by the tubesheet constraint. Consequently, negligible
normal operating leakage is expected from cracks within the
tubesheet region.
SLB leakage is limited by leakage flow restrictions resulting
from the crack and tube-to-tubesheet contact pressures that provide
a restricted leakage path above the indications and also limit the
degree of crack face opening compared to free span indications. The
total leakage, that is, the combined leakage for all such tubes,
plus the combined leakage developed by any other ARC, must be below
the maximum allowable SLB leak rate limit, such that off-site doses
are maintained less than 10 CFR 100 guideline values.
Therefore, based on the above evaluation, the proposed changes
do not involve a significant increase in the consequences of an
accident previously evaluated.
2. The proposed change does not create the possibility of a new
or different kind of accident from any accident previously
evaluated.
The proposed changes do not introduce any changes or mechanisms
that create the possibility of a new or different kind of accident.
Tube bundle integrity is expected to be maintained for all plant
conditions upon implementation of the proposed steam generator
alternate tube plugging criteria.
WCAP-14797, Revision 1, ``Generic W* Tube Plugging Criteria for
51 Series Steam Generator Tubesheet Region WEXTEX Expansions,''
requires that any tubes with indications identified using the bobbin
coil probe during the bobbin sampling plan also be inspected with
the RPC coil throughout the W* length of the tubes. The use of the
RPC will: (a) identify any new or non-expected degradation mode that
may not be identified using the bobbin coil probe, and (b) confirm
and characterize the bobbin coil indication.
These changes do not introduce any new equipment or any change
to existing equipment. No new effects on existing equipment are
created nor are any new malfunctions introduced.
Therefore, based on the above evaluation, the proposed changes
do not create the possibility of a new or different kind of accident
from any accident previously evaluated.
3. The proposed change does not involve a significant reduction
in a margin of safety.
The proposed changes maintain the required structural margins of
the steam

[[Page 61844]]

generator tubes for both normal and accident conditions. RG 1.121 is
used as the basis in the development of the W* alternate tube
plugging criteria for determining that steam generator tube
integrity considerations are maintained within acceptable limits. RG
1.121 describes a method acceptable to the NRC staff for meeting
General Design Criteria 14, 15, 31, and 32 by reducing the
probability and consequences of an SGTR. RG 1.121 concludes that by
determining the limiting safe conditions of tube wall degradation
beyond which tubes with unacceptable cracking, as established by
inservice inspection, should be removed from service or repaired,
the probability and consequences of an SGTR are reduced. This RG
uses safety factors on loads for tube burst that are consistent with
the requirements of Section III of the ASME Code.
For primarily axially oriented cracking located within the
tubesheet, tube burst is precluded due to the presence of the
tubesheet. WCAP-14797 defines a length, W*, of degradation free
expanded tubing that provides the necessary resistance to tube
pullout due to the pressure induced forces (with applicable safety
factors applied). Application of the W* criteria will preclude
unacceptable primary-to-secondary leakage during all plant
conditions. The methodology for determining leakage provides for
large margins between calculated and actual leakage values in the W*
criteria.
Plugging of the steam generator tubes reduces the reactor
coolant flow margin for core cooling. Implementation of the proposed
changes are expected to result in plugging of fewer tubes than with
the current criteria. Thus, implementation of the proposed changes
will maintain the margin of flow that may have otherwise been
reduced by tube plugging.
Based on the above, it is concluded that the proposed changes do
not result in a significant reduction of margin with respect to
plant safety as defined in the FSAR Update or bases of the plant
Technical Specifications.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment requests involve no significant hazards consideration.
Local Public Document Room location: California Polytechnic State
University, Robert E. Kennedy Library, Government Documents and Maps
Department, San Luis Obispo, California 93407
Attorney for licensee: Christopher J. Warner, Esq., Pacific Gas and
Electric Company, P.O. Box 7442, San Francisco, California 94120
NRC Project Director: William H. Bateman

Philadelphia Electric Company, Docket No. 50-352, Limerick
Generating Station, Unit 1, Montgomery County, Pennsylvania

Date of amendment request: October 24, 1997
Description of amendment request: The proposed Technical
Specifications (TS) changes would revise TS Section 3/4.1.3.6 to exempt
control rod 50-27 from the coupling test for the remainder of Cycle 7.
Basis for proposed no significant hazards 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 Technical Specifications (TS) change does not
involve a significant increase in the probability or consequences of
an accident previously evaluated. The probability of occurrence of
the analyzed Control Rod Drop Accident (CRDA) is not increased by
operating with the subject control blade in a condition not known to
be coupled since the compensatory measures will assure that the
blade will remain fully inserted below 10% rated thermal power where
the CRDA is a concern. Monitoring of nuclear instrumentation
responses in the vicinity of the blade when the drive is withdrawn
above 10% power will assure the blade is tracking with the drive
with no potential to stick and then drop. Scram impact forces from
an uncoupled control rod are of insufficient energy to dislodge the
fuel support (or fuel) or to cause a threat to the pressure boundary
integrity. No reduction of system or equipment redundancy is
involved.
The CRDA analyzed in the Safety Analysis Report (SAR) remains
the limiting rod drop accident, and its consequences are unaffected
by operation of the subject blade in the proposed manner. Operation
of the control blade as described, i.e., withdrawn no further than
the 46 position and in a condition not known to be coupled, has no
adverse effect on scram performance in response to any other
postulated accident. The scram insert motion of the rod is not
affected by the potentially uncoupled condition, and since the rod
is already partially inserted at position 46, it should have a
slightly better negative reactivity insertion characteristic.
Therefore, no potential to increase onsite or offsite radiological
consequences beyond those previously analyzed in the SAR is created.
Operating the subject control blade in a condition not known to
be coupled does not result in any onsite or offsite radiological
consequences different from those previously analyzed in the SAR.
The subject control blade will be fully inserted below 10% thermal
power where the CRDA is a concern and will be monitored during drive
withdrawal above 10% thermal power to assure it is tracking with the
drive. Scram performance is not adversely affected by operation from
the near full-out position of 46. Hence, no new failure modes are
created and consequences of any postulated failures are not
increased.
Therefore, the proposed TS change does not involve an increase
in the probability or consequences of an accident previously
evaluated.
2. The proposed TS change does not create the possibility of a
new or different kind of accident from any accident previously
evaluated.
The Safety Analysis Report (SAR) analyzed Control Rod Drop
Accident (CRDA) remains the only type of accident initiated (or
contributed to) by the control rod drive/control blade interface.
The compensatory actions to be taken when operating the subject
blade in a condition not verified to be coupled assure that no new
types of accidents can occur. The subject control blade will be
fully inserted below 10% thermal power where the CRDA is a concern
and will be monitored during drive withdrawal above 10% thermal
power to assure it is tracking with the drive. Scram performance is
not adversely affected by operation from the near full-out position
of 46. Since no adverse effect on insertion or scram performance is
expected, the previously analyzed accidents encompass any potential
consequence of operating with an uncoupled control blade.
The compensatory actions to be taken when operating the subject
blade in a condition not verified to be coupled assure that no new
failure modes are created, and, therefore, no new type of equipment
malfunction is introduced by operating the subject control blade in
the proposed manner.
Therefore, the proposed TS change does not create the
possibility of a new or different kind of accident from any accident
previously evaluated.
3. The proposed TS change does not involve a significant
reduction in a margin of safety.
Operation with the subject control blade in a condition not
known to be coupled for the remainder of Cycle 7 at LGS [Limerick
Generating Station] Unit 1, but with the compensatory actions
described below, does not reduce the existing margin of safety
determined by the analysis of the Control Rod Drop Accident (CRDA).
The CRDA analyzed in the Safety Analysis Report (SAR) remains
bounding in that the subject rod will be fully inserted below 10%
rated thermal power where the CRDA is a concern. Above 10% power,
when the associated drive is withdrawn, the nuclear instrumentation
in the vicinity of the blade will be monitored to assure the blade
tracks with the drive, providing assurance that the position of the
blade can be ascertained by the drive position. If the control blade
can not be verified to have followed the drive, then the rod shall
be completely inserted and the control rod directional valves
disarmed in accordance with existing TS requirements. To minimize
any scram impact loadings, the blade will be operated at the near
full-out position of 46 except for intermediate positions
temporarily occupied during standard rod withdrawal sequences.
Operating the subject control blade in the proposed manner will have
no adverse effect on insertion or scram performance of the blade and
will preserve the margin of safety.

[[Page 61845]]

Therefore, the proposed TS change does 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
amendment request involves no significant hazards consideration.
Local Public Document Room location:Pottstown Public Library, 500
High Street, Pottstown, PA 19464
Attorney for licensee: J. W. Durham, Sr., Esquire, Sr. V. P. and
General Counsel, Philadelphia Electric Company, 2301 Market Street,
Philadelphia, PA 19101
NRC Project Director: John F. Stolz

Public Service Electric & Gas Company, Docket No. 50-272, Salem
Nuclear Generating Station, Unit No. 1, Salem County, New Jersey

Date of amendment request: October 6, 1997
Description of amendment request: The amendment to the Technical
Specifications would increase the allowable band for control and
shutdown rod demanded position versus indicated position from plus or
minus 12 steps to plus or minus 18 steps when the power level is not
greater than 85% rated thermal power. The amendment is identical to
Amendment 183 for Salem Unit 2, which was issued September 10, 1997, as
an exigent amendment.
Basis for proposed no significant hazards 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 the rod misalignment criteria of [plus or
minus] 18 steps for core powers equal to or below 85% of RATED
THERMAL POWER (RTP) does not increase the probability of previously
evaluated accidents. Increasing the magnitude of the allowed control
rod misalignment is not a contributor to the mechanistic cause of an
accident evaluated in any accident analysis. The magnitude of
control rod indicated misalignment is a parameter used to establish
the initial conditions for accident evaluation.
The proposed increase in the allowable rod misalignment from the
current [plus or minus] 12 steps for reactor powers equal to or less
than 85% RTP does not involve a significant increase in the
consequence of any previously evaluated accident. Rod misalignment
affects power distribution, shutdown margin and the ejected rod
accident. An extension of the allowable rod misalignment above and
below 85% RTP has been analyzed in Westinghouse WCAP-14672. As
provided in WCAP-14672, above 85% the allowable misalignment is
governed by the available peaking factor margins as determined by
flux maps.
[Public Service Electric & Gas] PSE&G is simplifying the
proposed change by keeping the currently allowed [plus or minus] 12
step misalignment in Technical Specifications 3.1.3.1 and 3.1.3.2.1
for reactor power greater than 85% RTP.
The PSE&G proposed change is to allow [plus or minus] 18 steps
misalignments in Technical Specifications 3.1.3.1 and 3.1.3.2.1 for
reactor power less than or equal to 85% RTP. As demonstrated in
WCAP-14672, for reactor powers less than 85% RTP, the available
peaking factor margin increases faster than any penalty associated
with a [plus or minus] 18 step misalignment.
As described in Section 4.0 of the Westinghouse WCAP, a
conservative penalty factor has been applied to the rod insertion
allowance (RIA) of the shutdown margin calculation to account for
rods misaligned an additional [plus or minus] 6 steps (for a total
of [plus or minus] 18 steps). This conservative penalty factor is
applied as part of the reload analysis in order to satisfy Technical
Specification 3.1.1.1.
In addition to the normal, or Condition 1, operational
transients, the impacts of increased rod misalignment on Condition
II, III and IV accident analysis have also been evaluated. The
proposed increase in rod misalignment does not have a significant
effect on any moderator or Doppler reactivity coefficients or
defects, boron worth or reactor kinetics parameters.
To account for the potential increase in ejected rod parameters,
conservative penalty factors have been applied to the reload safety
evaluation to cover the additional [plus or minus] 6 step
misalignment. Margin is available in the reload safety analysis to
accommodate this impact.
Therefore, the proposed amendment does not increase the
probability or consequences of any 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.
0No new accident scenarios, failure mechanisms or limiting
single failures are introduced as a result of the proposed change to
the rod misalignment criteria of [plus or minus] 18 steps below 85%
RTP. The implementation of the proposed rod misalignment criteria
will have no adverse effect on the performance of any other safety
related system. Therefore, the proposed amendment does not create
the possibility of a new or different kind of accident from any
previously evaluated.
3. The proposed change does not involve a significant reduction
in a margin of safety.
Operation of the facility in accordance with the proposed
amendment would not involve a significant reduction in the margin of
safety. The Technical Specifications allowed increase in peaking
factors as power is reduced accommodates the peaking factor penalty
associated with the additional [plus or minus] 6 step misalignment
for core powers equal to or less than 85% RTP. Therefore, there is
no change to the peaking factors assumed in the safety analysis. In
addition to peaking factors, there is no change in any other current
limit input into the safety analysis. As the input, or initial
conditions, of the safety analysis have not changed, there is no
reduction in the margin to safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: Salem Free Public Library, 112
West Broadway, Salem, NJ 08079

Public Service Electric & Gas Company, Docket No. 50-272, Salem
Nuclear Generating Station, Unit No. 1, Salem County, New Jersey

Date of amendment request: October 14, 1997
Description of amendment request: The proposed amendment will
modify the Salem Unit 1 Technical Specification (TS) 3.4.6.3, ``Primary
Coolant System Pressure Isolation Valves Limiting Condition for
Operation,'' to be consistent with Salem Unit 2 TSs.
Basis for proposed no significant hazards 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 majority of the proposed changes, as described above, are
editorial in nature. Rewording, and reformatting the Limiting
Condition for Operation, including the surveillance requirements do
not involve a significant increase to the probability or
consequences of an accident.
Those substantive changes involving the addition of (1) new
reactor coolant system pressure isolation valves, (2) providing for
a shorter test frequency upon entry into Mode 4, and (3) adding a
new surveillance test requirement, do not increase the probability
or consequences of an accident. These changes ensure that the system
and components needed to prevent and minimize the effects of inter-
system loss of coolant are properly identified in the Technical
Specifications.
Although pressure isolation valves are being added to the
Technical Specification table, these valves were already included in
the IST [inservice testing] program as pressure isolation valves and
were being tested as such. The proposed changes do not involve a
significant increase in the probability or consequences of an
accident previously evaluated.

[[Page 61846]]

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, as described above, does not physically
alter the facility or the operation of the facility. The majority of
the changes are editorial in nature and provide for improvement in
the human factors of the Technical Specifications, while properly
identifying all the pressure isolation valves in the Technical
Specifications. The addition of valves into the Technical
Specification is an administrative change that improves the quality
of the LCO [limiting condition for operation], but does not add
components to the facility.
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 margin of safety, as defined in the bases for any technical
specifications, depend upon proper identification of equipment and
performance of the proper surveillance requirements to demonstrate
equipment operability. The proposed change will ensure that the
proper valves are identified and tested in accordance with the
Technical Specification requirements.
The proposed changes do not involve a significant reduction in a
margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: Salem Free Public Library, 112
West Broadway, Salem, NJ 08079
Attorney for licensee: Jeffrie J. Keenan, Esquire, Nuclear Business
Unit - N21, P.O. Box 236, Hancocks Bridge, NJ 08038
NRC Project Director: John F. Stolz

Public Service Electric & Gas Company, Docket Nos. 50-272 and 50-
311, Salem Nuclear Generating Station, Unit Nos. 1 and 2, Salem
County, New Jersey

Date of amendment request: October 21, 1997
Description of amendment request: The proposed amendment revises
Technical Specification Tables 3.3-1 and 4.3-1 to require that
Functional Unit, 2. Power Range, Neutron Flux,'' be operable
in Mode 3, as well as in Modes 1 and 2. The change is being proposed
because the licensee has determined that the power range nuclear
instrumentation should be operable in Mode 3 whenever the reactor trip
system breakers are in the closed position and the control rods are
capable of being withdrawn.
Basis for proposed no significant hazards 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 requirement for operability of a trip and the surveillance
requirements to ensure the functionality of the trip are independent
of the probability of an accident previously evaluated. The accident
that this trip is intended to mitigate is the Rod Withdrawal from
Subcriticality event. The surveillance procedure and the requirement
for the trip to be operational when the Control Rod Drive System is
capable of rod movement mitigate the consequences of this event, and
do not increase the probability of a rod withdrawal from
subcritical.
Therefore, the probability and consequences of an accident
previously evaluated are not significantly increased.
2. The proposed change does not create the possibility of a new
or different kind of accident from any accident previously
evaluated.
The proposed changes do not involve any modifications to
existing plant equipment, do not alter the function of any plant
systems, do not introduce any new operating configurations or new
modes of plant operation, or change the safety analyses. The
proposed change is intended to ensure that the trip function is
available and will perform as designed in the event of a previously
evaluated event.
The proposed changes will 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 does not reduce the margin of safety,
because assurance of the operability of the trip function is
increased by the proposed change.
Based on the above, PSE&G [Public Service Electric & Gas
Company] has determined that the proposed changes do not involve a
significant hazards consideration.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: Salem Free Public library, 112
West Broadway, Salem, NJ
Attorney for licensee: Jeffrie J. Keenan, Esquire, Nuclear Business
Unit - N21, P.O. Box 236, Hancocks Bridge, NJ 08038
NRC Project Director: John F. Stolz

The Cleveland Electric Illuminating Company, Centerior Service
Company, Duquesne Light Company, Ohio Edison Company, Pennsylvania
Power Company, Toledo Edison Company, Docket No. 50-440, Perry
Nuclear Power Plant, Unit 1, Lake County, Ohio

Date of amendment request: August 28, 1997
Description of amendment request: The proposed change would revise
Technical Specification 3.4.11, ``Reactor Coolant System (RCS) Pressure
and Temperature (P/T) Limits,'' to incorporate the new P/T curves,
which were provided by General Electric Nuclear Energy in report number
GE-NE-B1301793-01, ``Perry Unit 1 RPV Surveillance Materials Testing
and Analysis.''
Basis for proposed no significant hazards determination: As
required by 10 CFR 50.91(a), the licensee has provided its analysis of
the issue of no significant hazards consideration which is presented
below:
1. The proposed change does not involve a significant increase
in the probability or consequences of an accident previously
evaluated.
The proposed change will provide for approved P/T limit curves
which are valid through 9 effective full-power years (EFPY) and 18
EFPY. This change will not affect any Safety Limits, Power
Distribution Limits, or Limiting Conditions for Operation. The
proposed changes incorporate operating limits which provide margin
to brittle failure of the reactor vessel based on testing of the
irradiated reactor vessel materials (base metal, weld material, and
heat affected zone material). The limits ensure that adequate safety
margins against nonductile or rapidly propagating failure exist
during normal operation, anticipated operational occurrences, and
system hydrostatic tests. The specimens have been tested and
analyzed in accordance with 10 CFR 50, Appendices G and H, using the
methods described in Generic Letter 88-11 and Regulatory Guide 1.99
Revision 2. The predicted lowest upper shelf energy at 32 EFPY was
greater than the minimum required by 10 CFR 50, Appendix G. The
adjusted reference temperature for the limiting material was lower
than the 200 degree Fahrenheit limit required by Regulatory Guide
1.99 Revision 2. As such, the integrity of the reactor pressure
coolant boundary is maintained. The changes will result in
equivalent or more conservative limits on reactor vessel pressure as
a function of temperature for all operational conditions
(hydrostatic and leak testing, non-nuclear heatup/cooldown, and core
critical operations). The methodology used to derive these values
produces limits which continue to ensure that sufficient margin is
maintained to meet the criteria of GDC 31, ``Fracture Prevention of
Reactor Coolant Pressure Boundary.'' There are no plant
modifications associated with this change and no new or revised
system interfaces. The proposed

[[Page 61847]]

changes do not increase the probability of occurrence or
consequences previously evaluated because the temperature shifts are
well within equipment operating ranges. As such, there is no
increase in the probability of occurrence or the consequences of an
accident previously evaluated.
2. The proposed change does not create the possibility of a new
or different kind of accident from any accident previously
evaluated.
The proposed changes do not create the possibility of a new or
different kind of accident from any accident previously evaluated
because the proposed changes do not involve any new modes of
operation. The only change will be operation of the plant within
operating pressure limits which are determined in a more
conservative manner. Therefore, no new failure mode or accident
sequence is introduced by this change.
The testing and analysis meets 10 CFR 50, Appendices G and H,
requirements; therefore, no new accident types, such as brittle
fracture of a reactor pressure coolant boundary component is
postulated. The adjusted reference temperature and upper shelf
energy predicted at 32 EFPY are well within the limits of 10 CFR 50,
Appendices G and H. Therefore, the possibility of an accident of a
new or different type than any previously evaluated is not created.
3. The proposed change does not involve a significant reduction
in a margin of safety.
The P/T limits are established to provide acceptable margins for
the operation of the reactor coolant system during heat up and cool
down, criticality, and hydrotest conditions. Technical Specification
3.4.11 limits the rates of change of temperature and pressure to
values consistent with the fracture toughness requirements of 10 CFR
50, Appendices G and H, and ASME Boiler and Pressure Vessel Code
Section III Appendix G. The bases section for Technical
Specification 3.4.11 refers to 10 CFR 50, Appendices G and H, and
ASME Code Section III Appendix G. Changes in these limits are
necessary because the fracture toughness properties of ferritic
materials in the reactor vessel change as a function of reactor
operating time. The specific requirements for fracture toughness and
reactor vessel material surveillance that must be considered in
developing the P/T limits are defined by 10 CFR 50, Appendices G and
H. The specific limits defined by 10 CFR 50, Appendices G and H, set
the margin of safety for the reactor pressure vessel coolant
boundary. Since the testing and analysis of the vessel specimens
meet the requirements and limits defined in 10 CFR 50, Appendices G
and H, the margin of safety as defined in the basis for Technical
Specification 3.4.11 is not reduced. The revised curves are based on
the latest NRC guidelines along with actual neutron fluence data for
Perry. The new limits conservatively account for irradiation
embrittlement effects, thereby maintaining 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: Perry Public Library, 3753
Main Street, Perry, OH 44081
Attorney for licensee: Jay Silberg, Esq., Shaw, Pittman, Potts &
Trowbridge, 2300 N Street, NW., Washington, DC 20037
NRC Project Director: Gail H. Marcus

The Cleveland Electric Illuminating Company, Centerior Service
Company, Duquesne Light Company, Ohio Edison Company, Pennsylvania
Power Company, Toledo Edison Company, Docket No. 50-440, Perry
Nuclear Power Plant, Unit 1, Lake County, Ohio

Date of amendment request: September 8, 1997
Description of amendment request: The proposed amendment would
revise Technical Specification (TS) 5.2.2.e, ``Organization - Unit
Staff,'' by removing the reference to the NRC Policy Statement on
working hours. Administrative procedures will be developed to limit the
working hours of unit staff who perform safety-related functions.
Basis for proposed no significant hazards 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 TS 5.2.2.e only alters the administrative
location of and the regulatory controls applicable to unit staff
specific overtime limits and working hours. Overtime will remain
controlled by plant administrative procedures. Changes to the
relocated overtime limits and working hours will be subject to
review and evaluation under 10 CFR 50.59, ``Changes, Tests and
Experiments.'' There is not an increase in the probability of an
accident previously evaluated because no change is being made to any
accident initiator. No previously analyzed accident scenario is
changed, and initiating conditions and assumptions remain as
previously analyzed.
There is not an increase in the radiological consequences of an
accident previously evaluated because the proposed change does not
affect accident conditions or assumptions used in evaluating the
radiological consequences of an accident. The proposed change does
not alter the source term, containment isolation, or allowable
radiological releases. Therefore, there is no increase in the
radiological consequences of an accident previously evaluated.
2. The proposed change does not create the possibility or a new
or different kind of accident from any accident previously
evaluated.
The proposed change does not create the possibility of a new or
different kind of accident from any accident previously evaluated
because the proposed change does not change the way the plant is
operated, and no new or different failure modes have been defined
for any plant system or component important to safety, nor has any
limiting single failure been identified as a result of the proposed
change. No new or different types of failures or accident initiators
are introduced by the proposed change.
The proposed change to TS 5.2.2.e only alters the administrative
location of and the regulatory controls applicable to unit staff
specific overtime limits and working hours. Therefore, there is no
possibility created for a new or different kind of accident.
3. The proposed change does not involve a significant reduction
in a margin of safety.
The proposed change does not involve a reduction in a margin of
safety because unit staff overtime is not an input in the
calculation of a safety margin with regard to Technical
Specification Safety Limits, Limiting Safety System Settings, other
Technical Specification Limiting Conditions for Operation, the
Operational Requirements Manual, or other previously defined margins
for any structure, system, or component important to safety. The
proposed change to TS 5.2.2.e only alters the administrative
location of and the regulatory controls applicable to unit staff
specific overtime limits and working hours.
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: Perry Public Library, 3753
Main Street, Perry, OH 44081
Attorney for licensee: Jay Silberg, Esq., Shaw, Pittman, Potts &
Trowbridge, 2300 N Street, NW., Washington, DC 20037
NRC Project Director: Gail H. Marcus

TU Electric Company, Docket Nos. 50-445 and 50-446, Comanche Peak
Steam Electric Station, Units 1 and 2, Somervell County, Texas

Date of amendment request: October 24, 1997
Brief description of amendments: Change to the core safety limit
curves and overtemperature N-16 reactor trip function setpoints to
support operation with Unit 1, cycle 7 core configuration.
Basis for proposed no significant hazards determination: As
required by 10 CFR 50.91(a), the licensee has provided its analysis of
the issue of no significant hazards consideration, which is presented
below:
1. Do the proposed changes involve a significant increase in the
probability or consequences of an accident previously evaluated?
A. Revision to the Unit 1 Core Safety Limits

[[Page 61848]]

Analyses of reactor core safety limits are required as part of
reload calculations for each cycle. TU Electric has performed the
analyses of the Unit 1, Cycle 7 core configuration to determine the
reactor core safety limits. The methodologies and safety analysis
values result in new operating curves which, in general, permit
plant operation over a similar range of acceptable conditions. This
change means that if a transient were to occur with the plant
operating at the limits of the new curve, a different temperature
and power level might be attained than if the plant were operating
within the bounds of the old curves.
However, since the new curves were developed using NRC approved
methodologies which are wholly consistent with and do not represent
a change in the Technical Specification BASES for safety limits, all
applicable postulated transients will continue to be properly
mitigated. As a result, there will be no significant increase in the
consequences, as determined by accident analyses, of any accident
previously evaluated.
B. Revision to Unit 1 Overtemperature N-16 Reactor Trip
Setpoints
As a result of changes discussed, the Overtemperature reactor
trip setpoint has been recalculated. These trip setpoints help
ensure that the core safety limits are protected and that all
applicable limits of the safety analysis are met.
Based on the calculations performed, no significant changes to
the safety analysis values for Overtemperature reactor trip setpoint
were required. The f(deltaI) trip reset function was revised due to
more top-skewed axial power distributions predicted for this cycle.
The analyses performed show that, using the TU Electric
methodologies, all applicable limits of the safety analysis are met.
This setpoint provides a trip function which allows the mitigation
of postulated accidents and has no impact on accident initiation.
Therefore, the changes in safety analysis values do not involve an
increase in the probability of an accident and, based on satisfying
all applicable safety analysis limits, there is no significant
increase in the consequences of any accident previously evaluated.
In addition, sufficient operating margin has been maintained in
the overtemperature setpoint such that the risk of turbine runbacks
or reactor trips due to upper plenum flow anomalies or other
operational transients will be minimized, thereby, reducing
potential challenges to the plant safety systems.
SUMMARY
The changes in the amendment request applies NRC approved
methodologies to changes in safety analysis values, new core safety
limits and new N-16 setpoint and parameter values to assure that all
applicable safety analysis limits have been met. The potential for
an operational transient to occur has not been affected and there
has been no significant impact on the consequences of any accident
previously evaluated.
2. Do the proposed changes create the possibility of a new or
different kind of accident from any accident previously evaluated?
The proposed changes involve the calculation of new reactor core
safety limits and overtemperature reactor trip setpoint resets. As
such, the changes play an important role in the analysis of
postulated accidents but none of the changes effect plant hardware
or the operation of plant systems in a way that could initiate an
accident. Therefore, the proposed changes do not create the
possibility of a new or different kind of accident from any accident
previously evaluated.
3. Do the proposed changes involve a significant reduction in a
margin of safety?
In reviewing and approving the methods used for safety analyses
and calculations, the NRC has approved the safety analysis limits
which establish the margin of safety to be maintained. While the
actual impact on safety is discussed in response to question 1, the
impact on margin of safety is discussed below:
A. Revision to the Unit 1 Reactor Core Safety Limits
The NRC-approved TU Electric reload analysis methods have been
used to determine new reactor core safety limits. All applicable
safety analysis limits have been met. The methods used are wholly
consistent with Technical Specification BASES 2.1 which is the bases
for the safety limits. In particular, the curves assure that for
Unit 1, Cycle 7, the calculated DNBR is no less than the safety
analysis limit and the average enthalpy at the vessel exit is less
than the enthalpy of saturated liquid. The acceptance criteria
remains valid and continues to be satisfied; therefore, no change in
a margin of safety occurs.
B. Revision to Unit 1 Overtemperature N-16 Reactor Trip
Setpoints
Because the reactor core safety limits for CPSES Unit 1, Cycle 7
are recalculated, the Reactor Trip System instrumentation setpoint
values for the Overtemperature N-16 reactor trip setpoint which
protect the reactor core safety limits must also be recalculated.
The Overtemperature N-16 reactor trip setpoint helps prevent the
core and Reactor Coolant System from exceeding their safety limits
during normal operation and design basis anticipated operational
occurrences. The most relevant design basis analysis in Chapter 15
of the CPSES Final Safety Analysis Report (FSAR) which is affected
by the Overtemperature reactor trip setpoint is the Uncontrolled Rod
Cluster Control Assembly Bank Withdrawal at Power (FSAR Section
15.4.2). This event has been analyzed with the new safety analysis
value for the Overtemperature reactor trip setpoint to demonstrate
compliance with event specific acceptance criteria. Because all
event acceptance criteria are satisfied, there is no degradation in
a margin of safety.
The nominal Reactor Trip System instrumentation setpoints values
for the Overtemperature N-16 reactor trip setpoint (Technical
Specification Table 2.2-1) are determined based on a statistical
combination of all of the uncertainties in the channels to arrive at
a total uncertainty. The total uncertainty plus additional margin is
applied in a conservative direction to the safety analysis trip
setpoint value to arrive at the nominal and allowable values
presented in Technical Specification Table 2.2-1. Meeting the
requirements of Technical Specification Table 2.2-1 assures that the
Overtemperature reactor trip setpoint assumed in the safety analyses
remains valid. The CPSES Unit 1, Cycle 7 Overtemperature reactor
trip setpoint is not significantly different from the previous
cycle, and thus provides operational flexibility to withstand mild
transients without initiating automatic protective actions. Although
the value of the f(deltaI) trip reset function setpoint is
different, the Reactor Trip System instrumentation setpoint values
for the Overtemperature N-16 reactor trip setpoint are consistent
with the safety analysis assumptions which have been analytically
demonstrated to be adequate to meet the applicable event acceptance
criteria. Thus, there is no reduction in a margin of safety.
Using the NRC approved TU Electric methods, the reactor core
safety limits are determined such that all applicable limits of the
safety analyses are met. Because the applicable event acceptance
criteria continue to be met, there is no significant reduction in
the margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location:University of Texas at
Arlington Library, Government Publications/Maps, 702 College, P.O. Box
19497, Arlington, TX 76019
Attorney for licensee: George L. Edgar, Esq., Morgan, Lewis and
Bockius, 1800 M Street, N.W., Washington, DC 20036
NRC Project Director: James W. Clifford, Acting

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

The following notices were previously published as separate
individual notices. The notice content was the same as above. They were
published as individual notices either because time did not allow the
Commission to wait for this biweekly notice or because the action
involved exigent circumstances. They are repeated here because the
biweekly notice lists all amendments issued or proposed to be issued
involving no significant hazards consideration.
For details, see the individual notice in the Federal Register on
the day and page cited. This notice does not extend the notice period
of the original notice.

[[Page 61849]]

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: October 13, 1997
Description of amendment request: The proposed amendments would
support replacement of the three safety-related wide range level
instruments. The engineered safety features trip setpoint for the
refueling water automatic switchover to recirculation would be revised
to account for the difference in instrument uncertainty associated with
wide range level instruments and provide additional response time
margin.
Date of publication of individual notice in Federal Register:
October 22, 1997 (62 FR 54859)
Expiration date of individual notice: November 21, 1997
Local Public Document Room location: J. Murrey Atkins Library,
University of North Carolina at Charlotte, 9201 University City
Boulevard, North Carolina

Duke Energy Corporation, Docket Nos. 50-269, 50-270, and 50-287,
Oconee Nuclear Station, Units 1, 2, and 3, Oconee County, South
Carolina

Date of amendment request: October 20, 1997
Description of amendment request: The proposed amendments would
allow use of a rerolling process as an additional repair method for
tube degradation found in the tubesheet region. The rerolling method is
designed to ensure that the area of degradation will not serve as a
pressure boundary once the repair roll is installed, thus permitting
the tube to remain in service.
Date of publication of individual notice in Federal Register:
October 28, 1997 (62 FR 55835)
Expiration date of individual notice: Comment period ends November
12, 1997; Notice period ends November 28, 1997
Local Public Document Room location: Oconee County Library, 501
West South Broad Street, Walhalla, South Carolina

Notice Of Issuance Of Amendments To Facility Operating Licenses

During the period since publication of the last biweekly notice,
the Commission has issued the following amendments. The Commission has
determined for each of these amendments that the application complies
with the standards and requirements of the Atomic Energy Act of 1954,
as amended (the Act), and the Commission's rules and regulations. The
Commission has made appropriate findings as required by the Act and the
Commission's rules and regulations in 10 CFR Chapter I, which are set
forth in the license amendment.
Notice of Consideration of Issuance of Amendment to Facility
Operating License, Proposed No Significant Hazards Consideration
Determination, and Opportunity for A Hearing in connection with these
actions was published in the Federal Register as indicated.
Unless otherwise indicated, the Commission has determined that
these amendments satisfy the criteria for categorical exclusion in
accordance with 10 CFR 51.22. Therefore, pursuant to 10 CFR 51.22(b),
no environmental impact statement or environmental assessment need be
prepared for these amendments. If the Commission has prepared an
environmental assessment under the special circumstances provision in
10 CFR 51.12(b) and has made a determination based on that assessment,
it is so indicated.
For further details with respect to the action see (1) the
applications for amendment, (2) the amendment, and (3) the Commission's
related letter, Safety Evaluation and/or Environmental Assessment as
indicated. All of these items are available for public inspection at
the Commission's Public Document Room, the Gelman Building, 2120 L
Street, NW., Washington, DC, and at the local public document rooms for
the particular facilities involved.

Duquesne Light Company, et al., Docket Nos. 50-334 and 50-412,
Beaver Valley Power Station, Unit Nos. 1 and 2, Shippingport,
Pennsylvania

Date of application for amendments: October 4, 1997
Brief description of amendments: These amendments revise the
surveillance requirements in Technical Specifications (TSs) 4.1.2.3.1,
4.1.2.4.1, 4.5.2.b, and 4.6.2.1.b and associated Bases. The subject
surveillance requirements are applicable to the charging/high-head
safety injection pumps, low-head safety injection pumps, and the
containment quench spray pumps. The proposed changes replace the
current specific test acceptance criteria contained in these
surveillance requirements with requirements to verify pump performance
in accordance with the inservice testing program, the emergency core
cooling system flow analysis, or the containment integrity safety
analysis, as applicable. The proposed changes also make minor editorial
changes in these TSs and make conforming changes in the TS Index pages.
Date of issuance: October 28, 1997
Effective date: Both units, as of date of issuance, to be
implemented within 60 days.
Amendment Nos.: 207, 86
Facility Operating License Nos. DPR-66 and NPF-73: Amendments
revised the Technical Specifications.
Date of initial notice in Federal Register: December 18, 1996 (61
FR 66706) The Commission's related evaluation of the amendments is
contained in a Safety Evaluation dated October 28, 1997. No significant
hazards consideration comments received: No.
Local Public Document Room location: B. F. Jones Memorial Library,
663 Franklin Avenue, Aliquippa, PA 15001

Entergy Operations, Inc., System Energy Resources, Inc., South
Mississippi Electric Power Association, and Mississippi Power &
Light Company, Docket No. 50-416, Grand Gulf Nuclear Station, Unit
1, Claiborne County, Mississippi

Date of application for amendment: May 27, 1997, supplemented by
October 6, 1997
Brief description of amendment: The amendment eliminated selected
response time testing (RTT) surveillance requirements (SRs) from the
Technical Specifications (TSs) for certain components of the following
systems: reactor protection system (SR 3.3.1.1.15), primary containment
and drywell isolation instrumentation (SR 3.3.6.1.8), and emergency
core cooling system (SRs 3.5.1.8 and 3.5.2.7).
Date of issuance: November 5, 1997
Effective date: November 5, 1997
Amendment No.: 133
Facility Operating License No. NPF-29: Amendment revises the
Technical Specifications.
Date of initial notice in Federal Register: June 18, 1997 (62 FR
33122) The Commission's related evaluation of the amendment is
contained in a Safety Evaluation dated November 5, 1997.No significant
hazards consideration comments received: No.
Local Public Document Room location: Judge George W. Armstrong
Library, 220 S. Commerce Street, Natchez, MS 39120

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

Date of amendment request: July 17, 1996, as supplemented October
14, 1997
Brief description of amendment: The amendment revises

[[Page 61850]]

Facility Operating License No. NPF-38 to reflect the name change
from Louisiana Power & Light Company to Entergy Louisiana, Inc.
Date of issuance: November 3, 1997
Effective date: November 3, 1997, to be implemented within 60 days.
Amendment No.: 134
Facility Operating License No. NPF-38: Amendment revised
Facility Operating License No. NPF-38.
Date of initial notice in Federal Register: April 9, 1997 (62 FR
29749) The Commission's related evaluation of the amendment is
contained in a Safety Evaluation dated November 3, 1997. The letter
dated October 14, 1997, provided clarifying information which did not
alter the initial no significant hazards determination. No significant
hazards consideration comments received: No.
Local Public Document Room location: University of New Orleans
Library, Louisiana Collection, Lakefront, New Orleans, LA 70122

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

Date of application for amendment: March 27, as supplemented April
3, May 1, and August 20, 1997.
Brief description of amendment: Change Technical Specifications
(TS) to permanently establish a primary-to-secondary leak rate of 150
gallons per day through any one steam generator and specify the steam
generator tube inservice inspection requirements for pit-like
intergranular attack degradation in the ``B'' Once-Through-Steam-
Generator.
Date of issuance: October 28, 1997
Effective date: October 28, 1997
Amendment No.: 158
Facility Operating License No. DPR-72: Amendment revised the TS.
Date of initial notice in Federal Register: June 4, 1997 (62 FR
30632) The August 20, 1997, letter provided clarifying information that
did not affect the initial no significant hazards consideration. The
Commission's related evaluation of the amendment is contained in a
Safety Evaluation dated October 28, 1997. No significant hazards
consideration comments received: No.
Local Public Document Room location: Coastal Region Library, 8619
W. Crystal Street, Crystal River, Florida 32629

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

Date of application for amendments: August 27, 1997
Brief description of amendments: The admendments change the
Administrative Section of the Technical Specifications (TS) to allow
the use of 12-hour shifts.
Date of issuance: October 27, 1997
Effective date: October 27, 1997
Amendment Nos: 194 and 188Facility Operating Licenses Nos. DPR-31
and DPR-41: Amendments revised the TS.
Date of initial notice in Federal Register: September 24, 1997 (62
FR 50006) The Commission's related evaluation of the amendments is
contained in a Safety Evaluation dated October 27, 1997. No significant
hazards consideration comments received: No.
Local Public Document Room location: Florida International
University, University Park, Miami, Florida 33199

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

Date of application for amendments: March 26, 1997.
Brief description of amendments: The amendments modify surveillance
4.7.5.1.e.2 which requires verification of the control room ventilation
system autostart function.
Date of issuance: October 28, 1997
Effective date: October 28, 1997, with full implementation within
45 days.
Amendment Nos.: 218 and 202
Facility Operating License Nos. DPR-58 and DPR-74: Amendments
revised the Technical Specifications.
Date of initial notice in Federal Register: May 21, 1997 (62 FR
27796) The Commission's related evaluation of the amendments is
contained in a Safety Evaluation dated October 28, 1997. No significant
hazards consideration comments received: No.
Local Public Document Room location: Maud Preston Palenske Memorial
Library, 500 Market Street, St. Joseph, Michigan 49085

Northeast Nuclear Energy Company, Docket No. 50-245, Millstone
Nuclear Power Station, Unit 1, New London County, Connecticut

Date of application for amendment: May 15, 1997
Brief description of amendment: The amendment revises Technical
Specification Sections 3.1 and 4.1, ``Reactor Protection System,'' and
the associated Bases to remove run mode intermediate range monitor high
flux/inoperative with the associated average power range monitor
downscale scram trip function. The amendment also makes other editorial
revisions.
Date of issuance: October 27, 1997
Effective date: As of the date of issuance, to be implemented
within 60 days.
Amendment No.: 103
Facility Operating License No. DPR-21: Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: June 18, 1997 (62 FR
33127) The Commission's related evaluation of the amendment is
contained in a Safety Evaluation dated October 27, 1997. No significant
hazards consideration comments received: No.
Local Public Document Room location: Learning Resources Center,
Three Rivers Community-Technical College, 574 New London Turnpike,
Norwich, Connecticut 06360 and at the Waterford Library, ATTN: Vince
Juliano, 49 Rope Ferry Road, Waterford, Connecticut 06385

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

Date of application for amendment: September 2, 1997
Brief description of amendment: The amendment changes the Technical
Specifications (TSs) by modifying the maximum allowed primary
containment internal pressure during normal operation from 2.1 pounds
per square inch gauge (psig) to 1.0 psig. The TS Bases, Section 3/
4.6.1.4, is also updated to reflect the new maximum allowed primary
containment internal pressure during normal operation.
Date of issuance: October 27, 1997
Effective date: As of the date of issuance, to be implemented
within 30 days.
Amendment No.: 209
Facility Operating License No. DPR-65: Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: September 24, 1997 (62
FR 50007) The Commission's related evaluation of the amendment is
contained in a Safety Evaluation dated October 27, 1997. No significant
hazards consideration comments received: No.
Local Public Document Room location: Learning Resources Center,
Three Rivers Community-Technical College, 574 New London Turnpike,
Norwich, Connecticut, and Waterford Library, ATTN: Vince Juliano, 49
Rope Ferry Road, Waterford, Connecticut

[[Page 61851]]

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

Date of application for amendment: November 25, 1996, as
supplemented December 12, 1996, April 23, May 8, July 1, August 21, and
September 29, 1997
Brief description of amendment: The amendment modifies the
Technical Specification requirements associated with the Minimum
Critical Power Ratio (MCPR) safety limits for Cycle 18 based on the
cycle-specific analysis of the current mixed core of GE11/GE10 fuel
parameters.
Date of issuance: October 29, 1997
Effective date: October 29, 1997
Amendment No.: 99
Facility Operating License No. DPR-22: Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: April 9, 1997 (62 FR
17238) The December 12, 1996, letter provided an affidavit for the
original application dated November 25, 1996. The April 23, May 8,
August 21, and September 29, 1997, letters provided clarifying
information in response to the staff's request for additional
information during a teleconference on March 18, 1997. The July 1,
1997, letter provided a nonproprietary version of the April 23, 1997,
submittal. This information was within the scope of the original
application and did not change the staff's initial proposed no
significant hazards considerations determination. Therefore, renoticing
was not warranted. The Commission's related evaluation of the amendment
is contained in a Safety Evaluation dated October 29, 1997. No
significant hazards consideration comments received: No.
Local Public Document Room location:Minneapolis Public Library,
Technology and Science Department, 300 Nicollet Mall, Minneapolis,
Minnesota 55401

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

Date of application for amendments: November 27, 1996, as
supplemented August 15, September 2, and October 3, 1997
Brief description of amendments: The amendments incorporate
Combustion Engineering steam generator tube sleeve designs and
installation and examination techniques into the plant Technical
Specifications (TS). Specifically, the amendments make changes to TS
4.12, ``Steam Generator Tube Surveillance,'' and its associated Bases
Section B.4.12, ``Steam Generator Tube Surveillance.''
Date of issuance: November 4, 1997
Effective date: November 4, 1997, with full implementation within
30 days
Amendment Nos.: 132 and 124
Facility Operating License Nos. DPR-42 and DPR-60: Amendments
revised the Technical Specifications.
Date of initial notice in Federal Register: August 13, 1997 (62 FR
43370) The August 15, September 2, and October 3, 1997, letters
provided clarifying information and updated TS pages. This information
was within the scope of the original application and did not change the
staff's initial no significant hazards considerations determination.
The Commission's related evaluation of the amendments is contained in a
Safety Evaluation dated November 4, 1997.
Local Public Document Room location: Minneapolis Public Library,
Technology and Science Department, 300 Nicollet Mall, Minneapolis,
Minnesota 55401

PECO Energy Company, Public Service Electric and Gas Company
Delmarva Power and Light Company, and Atlantic City Electric
Company, Docket Nos. 50-277 and 50-278, Peach Bottom Atomic Power
Station, Unit Nos. 2 and 3, York County, Pennsylvania

Date of application for amendments: June 4, 1997
Brief description of amendments: The proposed change revises the
Peach Bottom Atomic Power Station, Unit Nos. 2 and 3, Technical
Specifications to eliminate an inconsistency between emergency core
cooling system (ECCS) operability requirements and the auto-start and
protective trip bypass of the emergency diesel generators on an ECCS
initiation signal during certain plant configurations.
Date of issuance: October 24, 1997
Effective date: Both units, as of date of issuance, to be
implemented within 30.
Amendments Nos.: 221 and 226
Facility Operating License Nos. DPR-44 and DPR-56: The amendments
revised the Technical Specifications.
Date of initial notice in Federal Register: August 13, 1997 (62 FR
43373) The Commission's related evaluation of the amendments is
contained in a Safety Evaluation dated October 24, 1997. No significant
hazards consideration comments received: No.
Local Public Document Room location: Government Publications
Section, State Library of Pennsylvania, (REGIONAL DEPOSITORY) Education
Building, Walnut Street and Commonwealth Avenue, Box 1601, Harrisburg,
PA 17105

Public Service Electric & Gas Company, Docket No. 50-354, Hope
Creek Generating Station, Salem County, New Jersey

Date of application for amendment: May 19, 1997, as supplemented by
letter dated August 25, 1997
Brief description of amendment: This amendment changes the Hope
Creek Technical Specification (TS) 3.7.1.3, ``Ultimate Heat Sink,'' to
raise the minimum allowable ulimate heat sink (UHS) water level from 76
feet to 80 feet, lower the maximum allowable UHS temperature from
88.6 deg.F to 85 deg.F, and reflect that continued plant operation to a
UHS temperature of 87 deg.F depends upon the association of UHS
temperature and safety system availability. The associated Surveillance
Requirement, TS 4.7.1.3, is changed to decrease the river water
temperature, at which increasing temperature surveillance is required,
from 85 deg.F to 82 deg.F. The requirements of TS 3.7.1.1, ``Safety
Auxiliaries Cooling System (SACS),'' TS 3.7.1.2, ``Station Service
Water System (SSWS),'' and TS 3.8.1.1, ``Electrical Power Systems,''
are revised to reflect the revised TS 3.7.1.3. In addition, the Bases
for 3/4.7.1, ``Service Water Systems,'' are appropriately revised.
Date of issuance: October 28, 1997
Effe

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Source: Frix Law Library, https://www.frixlaw.com/law-library/documents/fr%3A97-30217. Public record. Not legal advice.