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

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URL: https://www.frixlaw.com/law-library/documents/fr%3A94-8780

## Record

- **Collection:** Federal Register
- **Document type:** Uncategorized Document
- **Published:** April 13, 1994

## Text

UNITED STATES NUCLEAR REGULATORY COMMISSION

Biweekly Notice

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

I. Background

Pursuant to Public Law 97-415, the U.S. Nuclear Regulatory
Commission (the Commission or NRC staff) is publishing this regular
biweekly notice. Public Law 97-415 revised section 189 of the Atomic
Energy Act of 1954, as amended (the Act), to require the Commission to
publish notice of any amendments issued, or proposed to be issued,
under a new provision of section 189 of the Act. This provision grants
the Commission the authority to issue and make immediately effective
any amendment to an operating license upon a determination by the
Commission that such amendment involves no significant hazards
consideration, notwithstanding the pendency before the Commission of a
request for a hearing from any person.
This biweekly notice includes all notices of amendments issued, or
proposed to be issued from March 21, 1994, through April 1, 1994. The
last biweekly notice was published on March 30, 1994 (59 FR 14884).

Notice of Consideration of Issuance of Amendments to Facility Operating
Licenses, Proposed No Significant Hazards Consideration Determination,
and Opportunity for a Hearing

The Commission has made a proposed determination that the following
amendment requests involve no significant hazards consideration. Under
the Commission's regulations in 10 CFR 50.92, this means that operation
of the facility in accordance with the proposed amendment would not (1)
involve a significant increase in the probability or consequences of an
accident previously evaluated; or (2) create the possibility of a new
or different kind of accident from any accident previously evaluated;
or (3) involve a significant reduction in a margin of safety. The basis
for this proposed determination for each amendment request is shown
below.
The Commission is seeking public comments on this proposed
determination. Any comments received within 30 days after the date of
publication of this notice will be considered in making any final
determination.
Normally, the Commission will not issue the amendment until the
expiration of the 30-day notice period. However, should circumstances
change during the notice period such that failure to act in a timely
way would result, for example, in derating or shutdown of the facility,
the Commission may issue the license amendment before the expiration of
the 30-day notice period, provided that its final determination is that
the amendment involves no significant hazards consideration. The final
determination will consider all public and State comments received
before action is taken. Should the Commission take this action, it will
publish in the Federal Register a notice of issuance and provide for
opportunity for a hearing after issuance. The Commission expects that
the need to take this action will occur very infrequently.
Written comments may be submitted by mail to the Rules Review and
Directives Branch, Division of Freedom of Information and Publications
Services, Office of Administration, U.S. Nuclear Regulatory Commission,
Washington, DC 20555, and should cite the publication date and page
number of this Federal Register notice. Written comments may also be
delivered to Room P-223, Phillips Building, 7920 Norfolk Avenue,
Bethesda, Maryland from 7:30 a.m. to 4:15 p.m. Federal workdays. Copies
of written comments received may be examined at the NRC Public Document
Room, the Gelman Building, 2120 L Street, NW., Washington, DC 20555.
The filing of requests for a hearing and petitions for leave to
intervene is discussed below.
By May 13, 1994, the licensee may file a request for a hearing with
respect to issuance of the amendment to the subject facility operating
license and any person whose interest may be affected by this
proceeding and who wishes to participate as a party in the proceeding
must file a written request for a hearing and a petition for leave to
intervene. Requests for a hearing and a petition for leave to intervene
shall be filed in accordance with the Commission's ``Rules of Practice
for Domestic Licensing Proceedings'' in 10 CFR Part 2. Interested
persons should consult a current copy of 10 CFR 2.714 which is
available at the Commission's Public Document Room, the Gelman
Building, 2120 L Street, NW., Washington, DC 20555 and at the local
public document room for the particular facility involved. If a request
for a hearing or petition for leave to intervene is filed by the above
date, the Commission or an Atomic Safety and Licensing Board,
designated by the Commission or by the Chairman of the Atomic Safety
and Licensing Board Panel, will rule on the request and/or petition;
and the Secretary or the designated Atomic Safety and Licensing Board
will issue a notice of a hearing or an appropriate order.
As required by 10 CFR 2.714, a petition for leave to intervene
shall set forth with particularity the interest of the petitioner in
the proceeding, and how that interest may be affected by the results of
the proceeding. The petition should specifically explain the reasons
why intervention should be permitted with particular reference to the
following factors: (1) the nature of the petitioner's right under the
Act to be made a party to the proceeding; (2) the nature and extent of
the petitioner's property, financial, or other interest in the
proceeding; and (3) the possible effect of any order which may be
entered in the proceeding on the petitioner's interest. The petition
should also identify the specific aspect(s) of the subject matter of
the proceeding as to which petitioner wishes to intervene. Any person
who has filed a petition for leave to intervene or who has been
admitted as a party may amend the petition without requesting leave of
the Board up to 15 days prior to the first prehearing conference
scheduled in the proceeding, but such an amended petition must satisfy
the specificity requirements described above.
Not later than 15 days prior to the first prehearing conference
scheduled in the proceeding, a petitioner shall file a supplement to
the petition to intervene which must include a list of the contentions
which are sought to be litigated in the matter. Each contention must
consist of a specific statement of the issue of law or fact to be
raised or controverted. In addition, the petitioner shall provide a
brief explanation of the bases of the contention and a concise
statement of the alleged facts or expert opinion which support the
contention and on which the petitioner intends to rely in proving the
contention at the hearing. The petitioner must also provide references
to those specific sources and documents of which the petitioner is
aware and on which the petitioner intends to rely to establish those
facts or expert opinion. Petitioner must provide sufficient information
to show that a genuine dispute exists with the applicant on a material
issue of law or fact. Contentions shall be limited to matters within
the scope
of the amendment under consideration. The contention must be one which,
if proven, would entitle the petitioner to relief. A petitioner who
fails to file such a supplement which satisfies these requirements with
respect to at least one contention will not be permitted to participate
as a party.
Those permitted to intervene become parties to the proceeding,
subject to any limitations in the order granting leave to intervene,
and have the opportunity to participate fully in the conduct of the
hearing, including the opportunity to present evidence and cross-
examine witnesses.
If a hearing is requested, the Commission will make a final
determination on the issue of no significant hazards consideration. The
final determination will serve to decide when the hearing is held.
If the final determination is that the amendment request involves
no significant hazards consideration, the Commission may issue the
amendment and make it immediately effective, notwithstanding the
request for a hearing. Any hearing held would take place after issuance
of the amendment.
If the final determination is that the amendment request involves a
significant hazards consideration, any hearing held would take place
before the issuance of any amendment.
A request for a hearing or a petition for leave to intervene must
be filed with the Secretary of the Commission, U.S. Nuclear Regulatory
Commission, Washington, DC 20555, Attention: Docketing and Services
Branch, or may be delivered to the Commission's Public Document Room,
the Gelman Building, 2120 L Street, NW., Washington, DC 20555, by the
above date. Where petitions are filed during the last 10 days of the
notice period, it is requested that the petitioner promptly so inform
the Commission by a toll-free telephone call to Western Union at 1-
(800) 248-5100 (in Missouri 1-(800) 342-6700). The Western Union
operator should be given Datagram Identification Number N1023 and the
following message addressed to (Project Director): petitioner's name
and telephone number, date petition was mailed, plant name, and
publication date and page number of this Federal Register notice. A
copy of the petition should also be sent to the Office of the General
Counsel, U.S. Nuclear Regulatory Commission, Washington, DC 20555, and
to the attorney for the licensee.
Nontimely filings of petitions for leave to intervene, amended
petitions, supplemental petitions and/or requests for a hearing will
not be entertained absent a determination by the Commission, the
presiding officer or the Atomic Safety and Licensing Board that the
petition and/or request should be granted based upon a balancing of
factors specified in 10 CFR 2.714(a)(1)(i)-(v) and 2.714(d).
For further details with respect to this action, see the
application for amendment which is available for public inspection at
the Commission's Public Document Room, the Gelman Building, 2120 L
Street, NW., Washington, DC 20555, and at the local public document
room for the particular facility involved.

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

Date of amendment requests: February 18, 1994
Description of amendment requests: The proposed amendment would
modify Technical Specifications (TS) 5.3.1, Fuel Assemblies, and TS
5.6.1, Criticality. In addition, the proposed Amendment would add a new
Technical Specification 3/4.9.13, Boron Concentration-Storage Pool, and
its associated BASES. This proposed amendment is requested to allow
credit to be taken for burnup of spent fuel assemblies in establishing
storage locations within the PVNGS spent fuel pools.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensees have
provided their analysis about the issue of no significant hazards
consideration, which is presented below:
Standard 1--Involve a significant increase in the probability or
consequences of an accident previously evaluated.
This amendment does not involve a significant increase in the
probability or consequences of an accident previously evaluated.
Radiological consequences of the fuel handling accident are not
impacted by the formation of new storage regions since the fuel
assembly design is unchanged. However, even though the probability of
occurrence of a fuel misplacement error has increased slightly, the
consequences are markedly reduced by the crediting of 2150 ppm of
soluble boron in the spent fuel storage pool. The increase is also not
significant because of the types of administrative controls being put
into place in Regions 2 and 3. Furthermore, a fuel assembly
misplacement error is not considered an accident, as defined in the
UFSAR.
Standard 2--Create the possibility of a new or different kind of
accident from any accident previously evaluated.
This amendment does not create the possibility of a new or
different kind of accident from any accident previously evaluated. No
changes are being made to the fuel assemblies or the storage racks, and
controls will be employed to control the placement of assemblies in
Regions 2 and 3. As such, there is no possibility of a new or different
kind of accident being created. The existing design basis covers all
possible accident scenarios in the spent fuel storage pool.
Standard 3--Involve a significant reduction in a margin of safety.
This amendment request will not involve a significant reduction in
a margin of safety. There is no reduction in the margin of safety since
a keff less than or equal to 0.95 is met under all analyzed
conditions using conservative assumptions which do not credit the
soluble boron in the spent fuel storage pool except under some accident
conditions, as allowed by NRC guidelines. The original mechanical
analyses are unchanged for thermal and seismic/structural
considerations, as these analyses were originally performed for a fully
loaded spent fuel storage pool.
The NRC staff has reviewed the licensees' analysis and, based on
that 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: Phoenix Public Library, 12
East McDowell Road, Phoenix, Arizona 85004
Attorney for licensees: Nancy C. Loftin, Esq., Corporate Secretary
and Counsel, Arizona Public Service Company, P.O. Box 53999, Mail
Station 9068, Phoenix, Arizona 85072-3999
NRC Project Director: Theodore R. Quay

Commonwealth 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: March 11, 1994
Description of amendment request: The proposed amendments would
revise Technical Specification 3/4.7.D, ``Primary Containment Isolation
Valves.'' The proposed amendments will add check valves installed in
the reference leg instrumentation line. The valves have been installed
as part of the modifications required to meet NRC Bulletin (IEB) 93-03.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:

(1) Involve a significant increase in the probability or
consequences of an accident previously evaluated because:
The proposed license amendment adds the reference leg backfill
check valves to the Technical Specifications. As such, the proposed
amendment does not change the probability nor does it change the
consequences of any previously evaluated accident for Dresden and
Quad Cities Stations.
The proposed modifications (and proposed Technical Specification
amendments) add reference leg backfill instrument lines and check
valves to the reactor vessel level instrumentation. The proposed
modifications will eliminate the phenomenon described in IEB 93-03
(dissolved gases in the [Reactor Vessel Instrumentation System]
RVLIS piping may produce uncertainties in the level instrumentation
during RPV depressurization) by providing degassed Control Rod Drive
(CRD) water to the RVLIS reference leg piping. The proposed design
ensures that a continuous column of water, free of non-condensible
gases is maintained in the RVLIS reference leg piping. As such, the
proposed modifications do not affect any accident precursors or
initiators. Therefore the proposed modifications for the reference
leg backfill instrument lines do not increase the probability of any
previously evaluated accidents for Dresden Station and Quad Cities
Station.
The proposed plant modifications for the reference leg backfill
check valves will not increase the radiological consequences of any
previously evaluated accident. The radiological impact from a
reference leg backfill instrument line break is bounded by Dresden's
and Quad Cities' Instrument Line Break analysis (UFSAR Section
15.6.2). Therefore, the proposed plant changes will not increase the
consequences of any previously evaluated accident.
(2) Create the possibility of a new or different kind of
accident from any accident previously evaluated because:
The proposed modification connects the non-safety-related CRD
system to each safety-related division of RPV instrumentation and
Feedwater Level Control System. The backfill check valves will
eliminate the potential for reference leg leakage if CRD piping
integrity is lost. These check valves are classified as safety-
related and will be maintained and controlled such that overall
plant safety is maintained. The addition of the reference leg
backfill check valves to the Technical Specifications does not
create the possibility of a new or different kind of accident for
Dresden Station or Quad Cities Station.
(3) Involve a significant reduction in the margin of safety
because:
Primary containment integrity is not compromised by the addition
of a pair of check valves that provide isolation for the reference
leg backfill lines. These valves have been demonstrated to meet the
intent of the criteria specified in General Design Criterion (GDC)
55. The maintenance and control applied toward all the reference leg
backfill check valves ensures that overall plant safety is
maintained. Therefore, the addition of the reference leg backfill
valves to Technical Specification 3.7.D.1 and 3.7.D.2 does not
reduce the margin of safety for Dresden Station or Quad Cities
Station.

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: for Dresden, Morris Public
Library, 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; Sidley and
Austin, One First National Plaza, Chicago, Illinois 60690
NRC Project Director: James E. Dyer

Commonwealth 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: March 26, 1993
Description of amendment request: The proposed amendments would
revise Technical Specification 3/4.6 for Dresden and Quad Cities
Stations to allow Single Loop Operation (SLO) with the recirculation
loop suction and discharge valves open. The amendments would also
delete outdated and unnecessary portions of Technical Specification
3.6.H for Dresden, Units 2 and 3, and provide more consistency to the
BWR Standard Technical Specifications (NUREG-0213, Revision 4).
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:

Commonwealth Edison has evaluated this proposed amendment and
determined that it involves no significant hazards considerations.
According to 10 CFR 50.92(c), a proposed amendment to an operating
license involves no significant hazards considerations if operation
of the facility in accordance with the proposed amendment would not:
1. Involve a significant increase in the probability or
consequences of an accident previously evaluated; or
2. Create the possibility of a new or different kind of accident
from any accident previously evaluated; or
3. Involve a significant reduction in a margin of safety.
The proposed changes do not involve a significant increase in
the probability or consequences of an accident previously evaluated
because:
The proposed change to delete the requirement to close the
suction valve of the idle loop during SLO potentially affects two
transient or accident analysis previously evaluated. The first is
the Loss-of-Coolant Accident (LOCA) which has been analyzed for the
full range of break sizes, from a small rupture, where the makeup
flow is greater than the coolant loss rate, to the largest, a highly
improbable circumferential recirculation line break. The design
basis LOCA at Dresden and Quad Cities is the double-ended guillotine
break in a recirculation line. LPCI is one of the Emergency Core
Cooling Systems that would be initiated to flood the core following
the Design Basis LOCA accident.
The LOCA analysis for Quad Cities, Units 1 and 2, takes no
credit for closure of the recirculation suction valve to properly
direct LPCI flow into the lower plenum of the reactor. Instead, the
LPCI loop selection logic is relied upon to automatically close the
recirculation discharge valve of the selected intact loop. For
Dresden, Units 2 and 3, LPCI is not credited to inject because the
limiting failure is the LPCI injection valve. The LOCA ECCS analyses
previously performed for SLO remain applicable and the severity of a
postulated LOCA event has not increased. The proposed changes do not
physically change the plant in any manner that would increase the
probability of a LOCA.
The second transient considered is the inadvertent startup of an
idle recirculation pump in an unisolated loop. This event is
precluded, however, when the loop is unisolated because the
discharge valve must be closed for the pump to start. To further
decrease the probability of the occurrence of this transient,
Section 3.6.H.3.e is added to require the pump to be electrically
prohibited from starting. In addition, leaving the loop unisolated
results in an increase in the temperature of the water in the loop,
and a correspondingly lower reactivity insertion should the
transient occur. For these reasons, neither the probability nor the
consequences of an inadvertent idle pump start have increased.
The additional requirements introduced in Section 3.6.H.5.a-b
and 4.6.H.5 to monitor temperatures between the two loops and the
reactor coolant do not cause an increase in the probability or
consequences of an accident because they limit stresses in the
vessel and primary piping system to acceptable levels.
The procedures that are currently followed at Dresden, Units 2
and 3, regarding inadvertent entrance into a region of instability,
defined as Region A, B, and C in Reference (e), are more
conservative than those recommended by the NRC Bulletin, and do not
allow operation in the stability regions defined in the Dresden
Technical Specifications. Removing Sections 3.6.H.3.b-c and Section
4.6.H.3 only removes outdated material from the Dresden Technical
Specifications and does not increase the probability or consequences
of an accident previously evaluated.
The removal of Section 3.6.H.4 allowing operation without forced
circulation below 25% of rated power at Dresden, Units 2 and 3, will
not increase the probability or consequences of an accident
previously evaluated. This change is conservative because it will
prohibit operation in a condition susceptible to instabilities. This
section also is not included in the Standard Technical
Specifications. In the same manner, Section 2.1.A.4 is removed from
the Quad Cities Technical Specifications.
The removal of Section 3.6.H.3.a from the Dresden Technical
Specifications will not increase the probability or consequences of
an accident, because a one-pump run-up transient is bounded by the
two-pump run-up transient.
The change in initiation time for SLO requirements for Quad
Cities from 12 hours to 24 hours does not represent a significant
change, and still allows adequate time to implement the
requirements. Therefore, no increase in the probability or
consequences of an accident will be caused by this change.
For the reasons stated above, no increase in the probability or
consequences of an accident previously evaluated is introduced by
the proposed changes.
The proposed changes do not create the possibility of a new or
different kind of accident from any accident previously evaluated
because:
The proposed change to eliminate the requirement to close the
recirculation suction valve of the idle recirculation loop during
SLO only removes unnecessary conservatism which is not required to
ensure proper LPCI injection into the vessel during a LOCA. The LPCI
loop selection logic already ensures that the intact loop's
recirculation pump discharge valves will close when selected for
LPCI injection. Since all ECCS functions will continue to perform as
designed, no new accident scenarios are created. Also, by requiring
the idle loop to be electrically prohibited from starting, the
possibility of a new event is not created.
Section 3.6.H.5.a-b and 4.6.H.5 provide for temperature
monitoring prior to starting an idle pump, and monitoring to
maintain acceptable primary system stress levels. Since the changes
[do] not adversely affect the performance of any safety related
systems, no new accident scenarios are created.
The change eliminating the requirements for actions when a
region of instability is entered during SLO will not create a new or
different type of accident because procedures are already in place
that are consistent with NRC guidance in this area. These procedures
are more conservative than the current Technical Specifications.
The removal of Section 3.6.H.4 (Dresden) and Section 2.1.A.4
(Quad Cities) does not create the possibility of a new or different
kind of accident because the units will not be allowed to operate
without forced circulation with these sections removed, and Section
3.6.H.4 added. The possibility of accidents occurring from operating
in this mode has been eliminated, and no new types of accidents are
created.
There is no possibility of a new type of accident being created
from the removal of Section 3.6.H.3.a from the Dresden Technical
Specifications. The analysis behind the reduced flow MCPR curves
provides more thermal margin during SLO than two-loop operation,
because the one-pump run-up transient is less severe than the two-
pump run-up transient.
The speed requirement change (Section 3.6.H.3.d for Dresden) for
the operating recirculation pump prior to idle loop startup is in a
conservative direction, and no new types of accidents are created.
The increase in allowed time to initiate SLO requirements for
Quad Cities (Section 2.1.A.4) does not represent a significant
change, and still allows adequate time to implement necessary
requirements. No new types of accidents are created by this change.
The proposed changes do not involve a significant reduction in a
margin of safety because:
The change to eliminate the requirements to close the
recirculation suction valves of the idle loop during SLO maintains
the assumptions of the LOCA analyses. The LPCI loop selection logic
will automatically close the recirculation pump discharge valve of
the unbroken loop to ensure proper LPCI injection. Therefore, the
current MAPLHGR limits at Dresden and Quad Cities will continue to
ensure that Appendix K criteria are satisfied.
During normal dual loop operation, LPCI loop selection logic is
relied upon to close the discharge valve of the unbroken loop
following a LOCA. This function is performed during SLO, provided
the discharge valve and the logic that automatically closes this
valve upon the occurrence of a LOCA signal remain operable. Since
the assumptions of the accident analysis are preserved by the
proposed change, there is no reduction in any safety margin.
The safeguards in place preventing the inadvertent start of an
idle recirculation pump are more than adequate protection against
this transient. Three concurrent failures are required for this
transient to occur. The transient would also be less severe due to
the warmer water in the loop. Therefore, no reduction in a margin of
safety will occur with this change.
The addition of Sections 3.6.H.5.a-b and 4.6.H.5 will not
decrease margin to safety, since the temperature monitoring
requirements will maintain acceptable stresses in the primary system
during idle pump starts.
For Dresden, the current procedures for entrance into a region
of stability provide more margin to safety than the current
Technical Specifications require, because operation in a stability
region is not allowed.
The elimination of Section 3.6.H.4 in the Dresden Technical
Specifications and Section 2.1.A.4 in the Quad Cities Technical
Specifications will not decrease a margin of safety because it
prohibits operations in a potentially unstable region. This change
is in a conservative direction.
The elimination of Section 3.6.H.3.a in the Dresden Technical
Specifications does not cause a decrease in margin to safety,
because there is more thermal margin to SLO than two-loop operation.
Changing the active loop speed requirement from 65% to 43% prior
to idle loop startup for Dresden is in the conservative direction;
therefore, margin to safety is increased.
The increase in allowed time to initiate SLO requirements for
Quad Cities still provides adequate time to implement these
requirements, and is not a significant change. Margin to safety is
not decreased by this change.
Margin of safety does not, therefore, decrease due to the
proposed Technical Specification amendment.

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: for Dresden, Morris Public
Library, 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; Sidley and
Austin, One First National Plaza, Chicago, Illinois 60690
NRC Project Director: James E. Dyer

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

Date of application for amendments: March 11, 1994
Description of amendment requests: The proposed amendments would
provide surveillance requirements for new hydraulic snubbers, which
will be installed on the Main Steam Lines (MSLs) during the current
Unit 1 refueling outage. This outage began on March 13, 1994, and it is
scheduled to end on July 3, 1994. These snubbers will also be installed
on Unit 2 during the Unit 2 refuel outage (Q2R13) currently scheduled
for the first quarter of 1995.
The amendment request would also change the Snubber Visual
Inspection Intervals and Corrective Actions in Technical Specifications
Sections 3.6.1 and 4.6.1 to the format and content of the BWR
Standardized Technical Specifications (STS), as revised by the
provisions of Generic Letter (GL) 84-13 ``Technical Specification for
Snubbers'', dated May 3, 1984 and GL 90-09 ``Alternative Requirements
for Snubber Visual Inspection Intervals and Corrective Actions'', dated
December 11, 1990.
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:

Commonwealth Edison Company has evaluated the proposed Technical
Specification Amendment and determined that it does not represent a
significant hazards consideration. Based on the criteria for
defining a significant hazards consideration established in 10 CFR
50.92, operation of Quad Cities Station Units 1 and 2 (Quad Cities)
in accordance with the proposed amendment will not:
(1) Involve a significant increase in the probability or
consequences of an accident previously evaluated because:
The proposed changes adopt the format and content of the BWR-
STS, as modified by the provisions of GL 84-13 and GL 90-09. As
such, these proposed changes are administrative in nature and have
no effect on the accident analyses or system operation.
The proposed schedule for snubber visual inspection intervals
described in GL 90-09 will maintain the same level of confidence as
the existing schedule as documented in Generic Letter 90-09,
Alternative Requirements for Snubber Visual Inspection Intervals and
Corrective Actions, dated December 11, 1990. Also, the surveillance
requirement and schedule for snubber functional testing remains the
same providing a 95 percent confidence level that 90 to 100 percent
of the snubbers operate within the specified limits. The proposed
visual inspection schedule is separate from functional testing and
adds to the confidence level that the installed snubbers will serve
their design function and are being maintained operable. Accident
analyses assume that snubbers are initially operable. Compliance
with the Technical Specification Surveillance Requirements for
functional testing in conjunction with the revised visual inspection
schedule assures continued operability of the snubbers. Therefore,
no initial assumptions are being changed and thus neither the
probability nor consequences of any accidents previously evaluated
are significantly increased.
(2) Create the possibility of a new or different kind of
accident from any accident previously evaluated because:
The proposed changes adopt the format and content of the BWR-
STS, as modified by the provisions of GL 84-13 and GL 90-09. As
such, these proposed changes are administrative in nature and have
no effect on the accident analyses or system operation.
The proposed schedule for snubber visual inspection intervals
will maintain the same level of confidence as the existing schedule
as documented in Generic Letter 90-09, Alternative Requirements for
Snubber Visual Inspection Intervals and Corrective Actions, dated
December 11, 1990. Also, the surveillance requirement and schedule
for snubber function testing remains the same providing a 95 percent
confidence level that 90 to 100 percent of the snubbers operate
within the specified limits. The proposed visual inspection schedule
is separate from functional testing and adds to the confidence level
that the installed snubbers will serve their design function and are
being maintained operable. As a result, the supported piping,
components, etc. will be maintained operable, so that supported
safety systems will perform as designed. Therefore, the possibility
of a new or different kind of accident is not created.
(3) Involve a significant reduction in the margin of safety
because:
The proposed changes adopt the format and content of the BWR-
STS, as modified by the provisions of GL 84-13 and G[L] 90-09. As
such, these proposed changes are administrative in nature and have
no effect on the accident analyses or system operation. In addition,
the proposed amendment maintains the same level of confidence as the
current technical specification that snubbers are operable through
the current snubber functional testing and the revised snubber
visual inspection
schedule and the associated corrective action requirements. Therefore
the proposed changes do not impact 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: Dixon Public Library, 221
Hennepin Avenue, Dixon, Illinois 61021
Attorney for licensee: Michael I. Miller, Esquire; Sidley and
Austin, One First National Plaza, Chicago, Illinois 60690
NRC Project Director: James E. Dyer

Connecticut Yankee Atomic Power Company, Docket No. 50-213, Haddam Neck
Plant, Middlesex County, Connecticut

Date of amendment request: February 25, 1994
Description of amendment request: The proposed amendment will add a
new Technical Specification 3/4.7.12, ``Ultimate Heat Sink.''
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:

The proposed changes do not involve an SHC [significant hazards
consideration] because the changes would not:
1. Involve a significant increase in the probability or
consequences of an accident previously evaluated.
The ultimate heat sink (Connecticut River) provides the cooling
water necessary to ensure the removal of the normal heat loads and
normal cooldown loads of the plant and to mitigate the effects of
accidents at the plant within acceptable limits. By placing a
technical specification limit on the maximum temperature of the
ultimate heat sink for plant operation, CYAPCO will assure that
sufficient heat removal capacity is available. The intake structure
draws water from the ultimate heat sink for circulation by the
service water system and circulating water system. By adding this
new requirement to the technical specifications, CYAPCO will ensure
that the design basis, as stated in the Final Safety Analysis
Report, for the ultimate heat sink is not violated.
The change does not affect any initiating event. Thus, the
change does not affect the probability of occurrence of any design
basis accidents previously evaluated.
There are no adverse impacts on the design basis accidents due
to the addition of the ultimate heat sink temperature limitation.
This administrative change has no effect on the consequences of the
previously evaluated accidents.
2. Create the possibility of a new or different kind of accident
from any previously analyzed.
Currently, the Haddam Neck Plant controls the ultimate heat sink
temperature limit to less than 90 deg.F for Mode 1, 2, 3, and 4 via
a plant procedure. The proposed change institutes a technical
specification in place of this administrative control.
As such, the administrative change is consistent with the
current plant practice and has no effect on plant operation. Since
there are no changes in the way the plant is operated, there is no
possibility of an accident of a different type than previously
evaluated due to the change.
3. Involve a reduction in a margin of safety.
The proposed change does not impact the physical protective
boundaries, nor does it affect the performance of safety systems.
There is no degradation in operability and surveillance requirements
for the ultimate heat sink. Therefore, there will be no adverse
impact on the margin of safety as defined in the basis for any
technical specifications.

The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: Russell Library, 123 Broad
Street, Middletown, Connecticut 06457.
Attorney for licensee: Gerald Garfield, Esquire, Day, Berry &
Howard, Counselors at Law, City Place, Hartford, Connecticut 06103-
3499.
NRC Project Director: John F. Stolz

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

Date of amendment request: January 28, 1994
Description of amendment request: This amendment request is an
additional followup to the amendment request of May 29, 1992, published
in the Federal Register on July 8, 1992, (57 FR 30242) which changed
the Technical Specifications Section 1.0, Definitions, to accommodate a
24-month fuel cycle and which proposed the extension of the test
intervals for specific surveillance tests. This amendment proposes
extending the surveillance intervals to 24 months for leak testing
containment isolation valves. The changes requested by the licensee are
in accordance with Generic Letter 91-04, ``Changes in Technical
Specification Intervals to Accommodate a 24-Month Fuel Cycle.'' In
addition, the request corrects an administrative error.
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. A significant increase in the probability or consequences of
an accident previously evaluated will not occur.
It is proposed that the interval between leakage tests of the
containment isolation valves listed in the Technical Specifications
be revised from 24 months to 24 months (+25%), consistent with an
exemption request to 10 CFR [Part] 50 Appendix J for type C tests
which requests an identical extension in the interval between tests.
The proposed changes do not involve a significant increase in
the probability or consequences of a previously analyzed accident.
These changes propose extending the surveillance intervals for
containment leakage testing. The changes do not involve any physical
changes to the plant or alter the way equipment functions. Other
system testing (e.g., on-line tests) provides assurance of system
operability. An evaluation of past equipment performance provides
additional assurance that the longer surveillance intervals will not
degrade system performance. The 25% increase in the surveillance
interval for type C leak rate testing is compensated for by a
proportionate increase in the margin between specified leakage limit
and the allowable leakage limit. Valves that are sealed with fluid
are exempted from the 10 CFR [Part] 50, Appendix J leakage
requirements. The Technical Specifications establish separate
acceptance criteria for such cases base[d] on system design
considerations. Additionally, in most cases, containment isolation
valve redundancy (two valves in series) provides additional
assurance that actual leakage would be lower than the test results
would indicate.
2. The possibility of a new or different kind of accident from
any accident previously evaluated has not been created.
The proposed license amendment does not create the possibility
of a new or different kind of accident. These changes propose
extending the surveillance intervals for containment leakage
testing. The changes do not involve any physical changes to the
plant or alter the way equipment functions. Other system testing
(e.g., on-line tests) provides assurance of system operability. An
evaluation of past equipment performance provides additional
assurance that the longer surveillance intervals will not degrade
system performance. The 25% increase in the surveillance interval
for type C leak rate testing is compensated for by a proportionate
increase in the margin between specified leakage limit and the
allowable leakage limit. Also, containment isolation valve
redundancy (two valves in series) provides additional assurance, in
most cases, that leakage would be lower than the test results would
indicate.
3. Does the proposed amendment involve a significant reduction
in the margin of safety?
The proposed change[s] does[do] not involve a significant
reduction in the margin of safety. These changes propose extending
the surveillance intervals for containment leakage testing. Other
system testing (e.g., on-line tests) provides assurance of system
operability. An evaluation of past equipment performance provides
additional assurance that the longer surveillance intervals will not
degrade system performance. The 25% increase in the surveillance
interval for type C leak rate testing is compensated for by a
proportionate increase in the margin between specified leakage limit
and the allowable leakage limit. Also, containment isolation valve
redundancy (two valves in series), in most cases, provides
additional assurance that leakage would be lower than the test
results would indicate.

The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: White Plains Public Library,
100 Martine Avenue, White Plains, New York 10610.
Attorney for licensee: Brent L. Brandenburg, Esq., 4 Irving Place,
New York, New York 10003.
NRC Project Director: Robert A. Capra

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

Date of amendment request: November 11, 1993
Description of amendment request: The amendments would change the
Technical Specification (TS) surveillance requirements for the
emergency core cooling system (ECCS) subsystems. Specifically, the
changes would revise the minimum developed head requirement for the
centrifugal charging pumps (CCPs), the safety injection pumps (SIPs),
and the residual heat removal pumps (RHRPs); revise the sum of the
minimum injection flowrates for the CCPs, SIPs, and the RHRPs; and
revise the total maximum pump flowrate (runout limit) for the CCPs and
the SIPs.
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:

(Amendment would not) involve a significant increase in the
probability or consequences of an accident previously evaluated.
The TS [technical specification] changes proposed by this
amendment request are not considered to be initiators of any Design
Basis Accidents (DBA). During normal operation the SIPs and the
RHRPs are in standby, they are not operating. In the event of an
accident resulting in an Engineered Safeguard (ES) actuation, the
pumps would start to provide flow to the reactor vessel. The minor
changes proposed for these pumps (SIPs and RHRPs) would not cause
any accidents or events that have been previously evaluated.
During normal operation, a CCP is operating. The proposed minor
changes provided by this submittal only impact the performance of
these pumps in response to an ES actuation. The proposed changes do
not affect, in any way, how these pumps are operated during normal
operation. As such, the minor changes proposed for the CCPs would
not cause any accidents or events that have been previously
evaluated. Accordingly, the proposed TS changes would not increase
the probability of an accident that has been previously evaluated.
The purpose of the ECCS subsystem is to ensure sufficient flow
is provided to the core in the event of a LOCA [loss of coolant
accident], that is to mitigate the consequences of a LOCA. A LOCA
analysis was performed to determine the impact of the proposed TS
changes. The analysis was performed in accordance with the NRC
approved LOCA methodology for McGuire Nuclear Station. The results
of the analysis demonstrate that the acceptance criteria of 10 CFR
50.46 are still satisfied. Further, the purpose of the proposed TS
changes are to prevent runout of the ECCS subsystem pumps during the
injection and recirculation phases of a LOCA. Accordingly, the
proposed TS changes would not increase the consequences of an
accident that has been previously evaluated.
(Amendment would not) create the possibility of a new or
different kind of accident from any kind of accident previously
evaluated.
The proposed TS changes would not require any modifications to
any structures, systems or components at McGuire Nuclear Station.
Some minor changes to certain testing procedures for the ECCS
subsystem pumps would be necessary. These minor changes would only
involve specific values identified within the procedure and would
not result in any changes on how the test would be performed. No
other changes to procedures on how the station is operated or
maintained would occur. Accordingly, the proposed TS change would
not create a new or different kind of accident than what has been
previously evaluated.
(Amendment would not) involve a significant reduction in a
margin of safety.
The results of the analysis that was performed to determine the
impact of the proposed TS changes would have in mitigating a LOCA
indicate that the acceptance criteria of 10 CFR 50.46 are still
satisfied. The analysis that was performed demonstrate that the Peak
Clad Temperature (PCT) would remain below 2200 deg.F. The proposed
changes ensure that the ECCS subsystem pumps will be operated within
the limits specified by the manufacturer. Accordingly, the proposed
TS changes would not significantly reduce any margins 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: Atkins Library, University of
North Carolina, Charlotte (UNCC Station), North Carolina 28223
Attorney for licensee: Mr. Albert Carr, Duke Power Company, 422
South Church Street, Charlotte, North Carolina 28242
NRC Project Director: David B. Matthews, Director

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

Date of amendment request: February 25, 1994
Description of amendment request: The amendments would add four
instruments to the Technical Specification (TS) Tables 3.3-10 and 4.3-7
as part of the accident monitoring instrumentation, and delete five
instruments from the TS Tables that are not part of the accident
monitoring instrumentation.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:

1. Will operation of the facility in accordance with this
proposed change involve a significant increase in the probability or
consequences of an accident previously evaluated?
RESPONSE: No
This proposed change does not involve any significant increase
in the probability or consequences of any accident previously
evaluated because no changes in the types, categories, hardwares and
setpoints of the instruments involved were made; only the
designation of which instruments should be listed in the T/S
[technical specification] Tables and labeled as PAM [Post-Accident
Monitoring] in the control room is changed through this proposed
change.
2. Will operation of the facility in accordance with this
proposed change create the possibility of a new or different kind of
accident from any accident previously evaluated?
RESPONSE: No
This proposed change will not create the possibility of a new or
different kind of accident from any accident previously evaluated
because no changes in the types, categories, hardwares and setpoints
of the instruments involved were made; only the designation of which
instruments should be listed in the T/S Tables and labeled as PAM in
the control room is changed through this proposed change.
3. Will operation of the facility in accordance with this
proposed change involve a significant reduction in a margin of
safety?
RESPONSE: No
This proposed change does not involve a significant reduction in
a margin of safety because no changes in the types, categories,
hardwares and setpoints of the instruments involved were made; only
the designation of which instruments should be listed in the T/S
Tables and labeled as PAM in the control room is changed through
this proposed change.

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: Atkins Library, University of
North Carolina, Charlotte (UNCC Station), North Carolina 28223
Attorney for licensee: Mr. Albert Carr, Duke Power Company, 422
South Church Street, Charlotte, North Carolina 28242
NRC Project Director: David B. Matthews, Director

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

Date of amendment request: March 3, 1994
Description of amendment request: This amendment removes
restrictions from the Arkansas Nuclear One, Unit No. 1 (ANO-1)
technical specifications (TSs) that prohibit use of the auxiliary
building crane to move spent fuel shipping casks.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:

Criterion 1--Does Not Involve a Significant Increase in the
Probability or Consequences of an Accident Previously Evaluated.
The ANO procedures, load paths, crane equipment certification,
operator training and other related heavy load handling topics were
evaluated as part of the control of heavy loads issue and
found acceptable. Spent fuel cask handling is discussed in Section
9.6.2.6 of the ANO-1 SAR [safety analysis report], which shows that the
cask will never travel over spent fuel. ANO-1 SAR Section 9.6.2.6
further evaluates the unlikely event of a cask drop accident and shows
that the consequences are acceptable. Deletion of TS 3.8.15 to allow
handling of a spent fuel shipping cask by the auxiliary building crane
will have no actual impact on the cask drop or any other previously
analyzed accident and therefore, does not involve a significant
increase in the probability or consequences of any accident previously
evaluated.
Criterion 2--Does Not Create the Possibility of a New or
Different Kind of Accident from any Previously Evaluated.
The proposed amendment will allow handling of a spent fuel
shipping cask by the auxiliary building crane where [sic] previously
prohibited pending NRC evaluation of the spent fuel cask drop
accident and crane design. The cask handling methods and cask drop
accident are discussed and evaluated in ANO-1 SAR Section 9.6.2.6.
Additionally, the NRC performed an independent evaluation of the
radiological consequences of a cask drop accident, as documented in
the ANO-1 SER [safety evaluation report] dated June 6, 1973. The
evaluation of the unlikely event of a cask drop accident included
assessment of equipment failures and has shown the consequences to
be within acceptable bounds. Since no new accident scenarios can be
identified related to the proposed amendment request, this change is
bounded by the analysis described in the SAR and does not create the
possibility of a new or different kind of accident from any
previously evaluated.
Criterion 3--Does Not Involve a Significant Reduction in the
Margin of Safety.
Although allowing use of the Auxiliary Building crane where
[sic] previously prohibited by the TS could increase the possibility
of a cask drop accident, the margin of safety is preserved in that
the acceptable consequences of the cask drop accident evaluation in
SAR Section 9.6.2.6 are not affected by this change. The proposed
amendment request will not adversely affect the adequacy and
conservatism of the cask drip accident evaluation. Therefore, the
cask handling issue at ANO-1 continues to exhibit an acceptable
margin of safety and does not involve a significant reduction in the
margin of safety.

The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: Tomlinson Library, Arkansas
Tech University, Russellville, Arkansas 72801
Attorney for licensee: Nicholas S. Reynolds, Esquire, Winston and
Strawn, 1400 L Street, N.W., Washington, D.C. 20005-3502
NRC Project Director: William D. Beckner

Florida Power and Light Company, et al., Docket Nos. 50-335 and 50-389,
St. Lucie Plant, Unit Nos. 1 and 2, St. Lucie County, Florida

Date of amendment request: February 22, 1994
Description of amendment request: The proposed amendments will
relocate the instrument response time limits for the Reactor Protective
System and Engineered Safety Features Actuation System from the
Technical Specifications (TS) to the Updated Safety Analysis Report for
both units. The proposed changes are line-item TS improvements and
conform to the guidance given in Enclosures 1 and 2 of NRC Generic
Letter 93-08.
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:

Pursuant to 10 CFR 50.92, a determination may be made that a
proposed license amendment involves no significant hazards
consideration if operation of the facility in accordance with the
proposed amendment would not: (1) involve a significant increase in
the probability or consequences of an accident previously evaluated;
or (2) create the possibility of a new or different kind of accident
from any accident previously evaluated; or (3) involve a significant
reduction in a margin of safety. Each standard is discussed as
follows:
1. Operation of the facility in accordance with the proposed
amendment would not involve a significant increase in the
probability or consequences of an accident previously evaluated.

The proposed amendments for St. Lucie Units 1 and 2 simply
relocate tables of response time limits for instrumentation of the
Reactor Protective System (RPS) and Engineered Safety Features
Actuation System (ESFAS) from the Technical Specifications (TS) to
the Updated Safety Analysis Report (UFSAR). The proposed amendments
conform to the guidance given in Enclosures 1 and 2 of USNRC Generic
Letter 93-08 (GL 93-08). Neither the response time limits nor the
surveillance requirements for performing response time testing will
be altered by this submittal. The overall RPS and ESFAS system
functional capabilities will not be changed and assurance that
actions of the protective and engineered safety features systems are
completed within the time limits assumed in the accident analyses is
unaffected by the proposed TS changes. Therefore, operation of the
facility in accordance with the proposed amendment will not involve
a significant increase in the probability or consequences of an
accident previously evaluated.

2. Operation of the facility in accordance with the proposed
amendment would not create the possibility of a new or different kind
of accident from any accident previously evaluated.

The proposed amendment will not change the physical plant or the
modes of plant operation defined in the Facility License. The change
does not involve the addition or modification of equipment nor does
it alter the design or operation of plant systems. Therefore,
operation of the facility in accordance with the proposed amendment
would not create the possibility of a new or different kind of
accident from any accident previously evaluated.

3. Operation of the facility in accordance with the proposed
amendment would not involve a significant reduction in a margin of
safety.

The measurement of instrumentation response times at the
frequencies specified in the TS provides assurance that actions
associated with the protective and engineered safety features
systems are accomplished within the time limits assumed in the St.
Lucie Units 1 and 2 accident analyses. The response time limits, and
the measurement frequencies remain unchanged by the proposed
amendments. The proposed changes do not alter the basis for any
other Technical Specification that is related to the establishment
of or maintenance of a nuclear safety margin. Therefore, operation
of the facility in accordance with the proposed amendment would not
involve a significant reduction in a margin of safety.

Based on the discussion presented above and on the supporting
Evaluation of Proposed TS Changes, FPL has concluded that this proposed
license amendment involves no significant hazards consideration.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: Indian River Junior College
Library, 3209 Virginia Avenue, Fort Pierce, Florida 34954-9003.
Attorney for licensee: Harold F. Reis, Esquire, Newman and
Holtzinger, 1615 L Street, NW., Washington, DC 20036.
NRC Project Director: Herbert N. Berkow.

Florida Power and Light Company, et al., Docket Nos. 50-335 and 50-
389, St. Lucie Plant, Unit Nos. 1 and 2, St. Lucie County, Florida

Date of amendment request: February 28, 1994.
Description of amendment request: The proposed amendments will
delete the minimum frequency criteria prescribed for quality assurance
audits from Administrative Controls sections 6.5.2.8 and 6.8.4 of the
Technical Specifications (TS). Audit periodicity will thereby be
controlled by the program described in the Florida Power and Light
Company (FPL) Topical Quality Assurance Report.
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:
Pursuant to 10 CFR 50.92, a determination may be made that a
proposed license amendment involves no significant hazards
consideration if operation of the facility in accordance with the
proposed amendment would not: (1) involve a significant increase in the
probability or consequences of an accident previously evaluated; or (2)
create the possibility of a new or different kind of accident from any
accident previously evaluated; or (3) involve a significant reduction
in a margin of safety. Each standard is discussed as follows:
1. Operation of the facility in accordance with the proposed
amendment would not involve a significant increase in the probability
or consequences of an accident previously evaluated.
The proposed amendment relocates administrative control criteria
for minimum audit frequencies from the facility Technical
Specifications to the FPL Quality Assurance (QA) Program. The QA
Program is described in the FPL Topical Quality Assurance Report
pursuant to 10CFR50, Appendix B. The change does not alter the bases
upon which assurance is provided that safety-related activities are
performed correctly nor does it involve the conditions and assumptions
utilized in the analyses of plant transients and accidents. Therefore,
operation of the facility in accordance with the proposed amendment
will not involve a significant increase in the probability or
consequences of an accident previously evaluated.
2. Operation of the facility in accordance with the proposed
amendment would not create the possibility of a new or different kind
of accident from any accident previously evaluated.
The proposed amendment will not change the physical plant or the
modes of plant operation defined in the Facility License. The change
does not involve the addition or modification of equipment nor does it
alter the design or operation of plant systems. Therefore, operation of
the facility in accordance with the proposed amendment would not create
the possibility of a new or different kind of accident from any
accident previously evaluated.
3. Operation of the facility in accordance with the proposed
amendment would not involve a significant reduction in a margin of
safety.
The proposed amendment does not alter the bases for assurance that
safety-related activities are performed correctly or that compliance
with the required Limiting Conditions for Operation will be achieved.
The change does not alter the basis for any Technical Specification
that is related to the establishment of or maintenance of a nuclear
safety margin. Therefore, operation of the facility in accordance with
the proposed amendment would not involve a significant reduction in a
margin of safety.
Based on the discussion presented above and on the supporting
Evaluation of Proposed TS Changes, FPL has concluded that this proposed
license amendment involves no significant hazards consideration.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: Indian River Junior College
Library, 3209 Virginia Avenue, Fort Pierce, Florida 34954-9003.
Attorney for licensee: Harold F. Reis, Esquire, Newman and
Holtzinger, 1615 L Street, NW., Washington, DC 20036.
NRC Project Director: Herbert N. Berkow.

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

Date of amendment request: March 1, 1994.
Description of amendment request: The proposed amendments would
revise Technical Specification (TS) 3.2.4, ``Quadrant Power Tilt
Ratio,'' to add an exception to the requirements of TS 3.0.4.
Specifically, to add ACTION statement ``d. The provisions of
Specification 3.0.4 are not applicable.''
Basis for proposed no significant hazards consideration
determination: In 1992, the Vogtle TSs were amended in accordance with
the recommendations of Generic Letter 87-09 to revise the wording of TS
3.0.4 and to delete from TS 3.2.4 the statement that the provisions of
TS 3.0.4 are not applicable. With the revised wording of TS 3.0.4, the
statement of the non-applicability of the provisions of TS 3.0.4 was
redundant for many individual specifications, and its deletion caused
no change in ACTION requirements. However, in the case of TS 3.2.4, the
deletion had the unintended effect of prohibiting power escalation
above 50% rated thermal power (RTP) whenever the quadrant power tilt
ratio (QPTR) exceeds 1.0.2. This unnecessarily delays power escalation.
The proposed amendment would correct this error and restore the
originally intended meaning of TS 3.2.4. The intent of TS 3.2.4 is to
permit the escalation of reactor power above 50% RTP for limited times
and
under specified conditions when the QPTR is greater than 1.02.
With the original requirements of TS 3.2.4 restored, plant
operation and power escalation during startup would be the same as
previously approved. Therefore, the proposed change (1) does not
involve a significant increase in the probability or consequences or an
accident previously evaluated, (2) does not create the possibility of a
new or different kind of accident than previously evaluated, and (3)
does not involve a significant reduction in the margin of safety.
As required by 10 CFR 50.91(a), the licensee has provided its
analysis of the issue of no significant hazards consideration, which is
presented below:

1. The proposed change to the Technical Specifications does not
involve a significant increase in the probability or consequences of
an accident previously evaluated because it does not allow operation
in a condition that is not already allowed by the Technical
Specifications.
2. The proposed change to the Technical Specifications does not
create the possibility of a new or different kind of accident from
any accident previously evaluated because it will not allow
operation under conditions different from those already allowed by
the technical specifications.
3. The proposed addition to the Technical Specifications does
not involve a significant reduction in the margin of safety because
the action requirements will continue to be met in the same manner
as currently required by the Technical Specifications.

Accordingly, the NRC staff proposes to determine that the amendment
request involves no significant hazards consideration.
Local Public Document Room location: Burke County Public Library,
412 Fourth Street, Waynesboro, Georgia 30830.
Attorney for licensee: Mr. Arthur H. Domby, Troutman Sanders,
NationsBank Plaza, Suite 5200, 600 Peachtree Street, NE., Atlanta,
Georgia 30308.
NRC Project Director: David B. Matthews, Director.

GPU Nuclear Corporation, et al., Docket No. 50-289, Three Mile Island
Nuclear Station, Unit No. 1, Dauphin County, Pennsylvania

Date of amendment request: February 7, 1994.
Description of amendment request: The purpose of the request is to
require the TMI-1 annual radioactive effluent release report for the
previous calendar year be required to be submitted prior to May 1 of
each year. Changing the TMI-1 due date to prior to May 1 can enable the
licensee to combine the reports for TMI-1 and TMI-2 into a single
report with a common due date.
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. Submission of the annual radioactive effluent release report
on or before April 30 each year in accordance with the change, as
compared to the current requirement of March 1, does not involve a
significant increase in the probability of occurrence or the
consequences of an accident previously evaluated. The date on which
the report is due has no impact on plant operations or effluents. It
does not change the control of plant activities or the monitoring of
plant effluents.
2. Operation of TMI-1 in accordance with the proposed amendment
will not create the possibility of a new or different kind of
accident from any accident previously evaluated. The proposed
amendment has no impact on plant operations, plant effluents or the
control of plant operations or effluents. Therefore, there is no
potential to create a new or different kind of accident.
3. Operation of TMI-1 in accordance with the proposed amendment
does not involve a significant reduction in a margin of safety.
There is no specified margin of safety in regards to the due date
for the annual radioactive effluent release report.

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: Government Publications
Section, State Library of Pennsylvania, Walnut Street and Commonwealth
Avenue, Box 1601, Harrisburg, Pennsylvania 17105.
Attorney for licensee: Ernest L. Blake, Jr., Esquire, Shaw,
Pittman, Potts & Trowbridge, 2300 N Street, NW., Washington, DC 20037.
NRC Project Director: John F. Stolz.

GPU Nuclear Corporation, et al., Docket No. 50-289, Three Mile Island
Nuclear Station, Unit No. 1, Dauphin County, Pennsylvania

Date of amendment request: March 2, 1994.
Description of amendment request: The purposes of the request are
to change the plant Technical Specifications (TS) to modify Operational
Safety Instrumentation requirements to specify completion times which
allows for performance of maintenance or surveillance within a
reasonable time and to be consistent with the allowable outage time for
other safety-related equipment when only one train is affected.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:

1. Operation of the facility in accordance with the proposed
amendment would not involve a significant increase in the
probability of occurrence or the consequence of an accident
previously evaluated.
The proposed amendment permits time to restore instrumentation
channels to operable status which is consistent with existing times
allowed for outage of other safety-related equipment affecting one
train. With regard to the 1 hour timeclock, this time is sufficient
to perform the required action necessary to restore minimum required
conditions. Allowing 6 hours to reduce reactor power in an orderly
manner without challenging plant systems is reasonable, based on
operating experience. Thus, the proposed amendment maintains an
adequate degree of equipment availability without requiring
unnecessary initiation of a plant shutdown for partial equipment
outages.
Therefore, it can be concluded that the proposed changes do not
involve a significant increase in the probability or consequences of
an accident previously evaluated.
2. Operation of the facility in accordance with the proposed
amendment would not create the possibility of a new or different
kind of accident from any accident previously evaluated.
The proposed amendment affects the [Reactor Protection System]
RPS and the [Engineered Safeguards Actuation System] ESAS by
providing timeclocks to perform corrective actions. During this
timeclock period the safety function of the RPS can still be
completed by the remaining minimum required channels. If an accident
occurred while one ESAS train was inoperable due to faulty pressure
switches or a faulty manual actuation channel, the redundant train
would complete the safety function. The proposed time allowed for
the pressure switches in one train or the faulty manual actuation
channel to be out of service is bounded by the allowable time for
other safety-related equipment such that only one train is affected.
The proposed 8 hour timeclock associated with the [Reactor Building]
RB purge radiation monitor, RMA-9, provides adequate time to confirm
a problem exists and perform minor troubleshooting. The containment
isolation function for the RB purge valves would still be maintained
by a redundant 4 psig ESAS signal and a redundant reactor trip
containment isolation signal.
Therefore, the proposed amendment does not create the
possibility of a new or different accident.
3. Operation of the facility in accordance with the proposed
amendment would not involve a significant reduction in a margin of
safety.
As noted in the Bases for Specification 3.5.1, every reasonable
effort will be made to maintain all safety instrumentation in
operation. If RPS or ESAS instrumentation is found to be inoperable
or require maintenance to assure reliability, the proposed amendment
will allow the performance of maintenance and surveillance in a
reasonable time period. The change does not result in a significant
reduction in a margin of safety for the RPS because the automatic
functions and various alternative manual trip methods are still
available. Also, this change does not result in a significant
reduction in a margin of safety for the ESAS because at least one
train of safety features is required for continued operation within
the specified timeclocks with automatic and manual trip functions.
Thus, operation of the facility in accordance with the proposed
amendment does not involve a significant reduction in a margin of
safety.

The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 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: Government Publications
Section, State Library of Pennsylvania, Walnut Street and Commonwealth
Avenue, Box 1601, Harrisburg, Pennsylvania 17105.
Attorney for licensee: Ernest L. Blake, Jr., Esquire, Shaw,
Pittman, Potts & Trowbridge, 2300 N Street, NW., Washington, DC 20037.
NRC Project Director: John F. Stolz.

GPU Nuclear Corporation, et al., Docket No. 50-289, Three Mile Island
Nuclear Station, Unit No. 1, Dauphin County, Pennsylvania

Date of amendment request: March 11, 1994.
Description of amendment request: The purposes of the request are
to (1) change the plant Technical Specifications (TS) to specify an
allowable outage time for the Emergency Feedwater (EFW) Pumps during
surveillance activities and (2) change the requirement to test
redundant components for operability to a requirement to ensure
operability based on verification of completion of appropriate
surveillance activities. Basis for proposed no significant hazards
consideration determination: As required by 10 CFR 50.91(a), the
licensee has provided its analysis of the issue of no significant
hazards consideration, which is presented below:

1. Operation of the facility in accordance with the proposed
amendment would not involve a significant increase in the
probability of occurrence or the consequences of an accident
previously evaluated.
The proposed amendment specifies an allowable outage time for
testing of the EFW pumps. Also, this proposed change reflects the
current NRC staff position regarding the need for additional testing
to assure OPERABILITY. The allowable outage time this change
provides for EFW pump testing is acceptable because the operator
action required to make the motor driven EFW pump OPERABLE is
minimal and can be performed in a very short time by the Control
Room Operator who is continuously present during the time by the
motor-driven EFW pump is in the Pull-To-Lock position.
The changes affecting OPERABILITY determinations of redundant
train/components for reactor building isolation valves and the
control room air treatment systems reflect the current NRC staff
position. Verifying that the required periodic surveillance testing
is current and there are no known reasons to suggest the redundant
train/component is inoperable, provides adequate assurance of system
OPERABILITY.
Therefore, it can be concluded that the proposed changes do not
involve a significant increase in the probability or consequences of
an accident previously evaluated.
2. Operation of the facility in accordance with the proposed
amendment would not create the possibility of a new or different
kind of accident from any accident previously evaluated.
The proposed amendment specifies an allowable outage time and
deletes unnecessary redundant equipment testing. These changes do
not change system operational requirements or response to system
transients. Therefore, the proposed amendment does not create the
possibility of a new or different kind of accident.
3. Operation of the facility in accordance with the proposed
amendment would not involve a significant reduction in a margin of
safety.
The proposed amendment specifies an allowable outage time and
replaces redundant equipment testing with verification that
surveillance is current as an adequate means to ensure OPERABILITY.
These changes do not involve any activities associated with the
margin of safety envelope. Thus, operation of the facility in
accordance with the proposed amendment does not involve a
significant reduction in a margin of safety.

The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 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: Government Publications
Section, State Library of Pennsylvania, Walnut Street and Commonwealth
Avenue, Box 1601, Harrisburg, Pennsylvania 17105.
Attorney for licensee: Ernest L. Blake, Jr., Esquire, Shaw,
Pittman, Potts & Trowbridge, 2300 N Street, NW., Washington, DC 20037.
NRC Project Director: John F. Stolz.

Houston Lighting & Power Company, City Public Service Board of San
Antonio, Central Power and Light Company, City of Austin, Texas, Docket
Nos. 50-498 and 50-499, South Texas Project, Units 1 and 2, Matagorda
County, Texas

Date of amendment request: March 16, 1994.
Description of amendment request: The licensee proposes to amend
the South Texas Project technical specifications (TS) by modifying TS
3.4.9.3, ``Reactor Coolant System--Overpressure Protection Systems,''
Figure 3.4-4, ``Nominal Maximum Allowable PORV Setpoint for the Cold
Overpressure System,'' for the cold overpressure mitigation system
(COMS) with a revised setpoint curve. The proposed amendment would
account for the pressure losses of the reactor coolant flow through the
reactor core with either two or four reactor coolant pumps (RCP)
operating. It was determined that the original COMS setpoint curve
neglected reactor coolant pressure losses due to flow through the
reactor core with the RCPs operating. The resulting pressure at the
reactor vessel downcomer at the elevation equivalent to the core mid-
plane was higher than the pressure at the sensing point located in the
residual heat removal system suction line connected to the reactor
coolant system (RCS) hot leg. The proposed amendment would lower the
power operated relief valve (PORV) setpoint limit by a quantity equal
to the pressure difference between the pressure at the reactor vessel
downcomer at the elevation of the core mid-plane and the pressure at
the location of the residual heat removal system pressure transmitters.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:

(1) The proposed change does not involve a significant increase
in the probability or consequences of a previously evaluated
accident.
The presently existing pressurizer Power Operated Relief Valves
(PORVs) setpoints, provided by the Cold Overpressure Mitigation
System Curve (Figure 3.4-4) of Technical Specification 3.4.9.3, are
nonconservative in that they do not account for reactor coolant
pressure losses due to flow through the reactor core with Reactor
Coolant Pump (RCPs) in operation. When RCP operation is considered,
the pressure at the reactor vessel downcomer, at an elevation
equivalent to core midplane, is higher than the pressure sensing
point located in the Residual Heat Removal System suction line
connection to the Reactor Coolant System (RCS) hot leg. Houston
Lighting & Power Company became aware of this condition and re-
analyzed the Cold Overpressurization Event for the South Texas
Project. The re-analysis has resulted in modifications to Figure
3.4-4 of Technical Specification 3.4.9.3. The re-analysis reduced
the PORVs setpoint to account for the pressure losses and provides
for a setpoint for two Reactor Coolant Pump operation and for four
Reactor Coolant Pump operation.
The proposed decrease in the PORVs setpoint reduces the pressure
versus temperature limit for the RCS under start-up and shut-down
operations. The decreased PORV setpoints for post-overpressure
incidents will ensure that RCS pressure will be maintained within
acceptable limits during low temperature water solid operation for
both two and four pump operation.
The proposed change is based on a re-analysis which accounts for
reactor coolant pressure losses through the reactor core. Reflecting
actual reactor coolant pressure losses and adjusting the PORV
setpoint as necessary has no adverse effect on the probability or
consequences of an accident previously evaluated. Therefore, the
proposed changes not only do not involve a significant increase in
the probability or consequences of an accident previously evaluated,
but actually maintain the original design basis.
(2) The proposed change does not create the possibility of a new
or different kind of accident from any previously evaluated.
The proposed decrease in the PORV setpoints ensures that
staggered operation of the two PORVs are maintained, thus minimizing
the potential for large pressure undershoots resulting from multiple
valve operation which may compromise the Reactor Coolant Pump No. 1
Seal integrity. It also restricts the total number of discharge
ports at any given moment to that absolutely necessary for pressure
control. In addition, operation of either PORV provides the required
design basis relief capacity and the required redundancy necessary
to meet single failure criteria.
The proposed change is the result of a re-analysis of a
previously evaluated accident. Therefore, the proposed changes do
not create the possibility of a new or different kind of accident
from any accident previously evaluated.
(3) The proposed change does not involve a significant reduction
in the margin of safety.
The proposed change corrects an error present in the original
analysis by accounting for reactor coolant pressure losses through
the reactor core. The revised COMS curves are the result of a re-
analysis of the original COMS analysis. The new analysis preserves
the originally intended margin of safety. Therefore, the proposed
changes do not involve a significant reduction in a margin of
safety.

The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
request for amendments involves no significant hazards consideration.
Local Public Document location: Wharton County Junior College, J.M.
Hodges Learning Center, 911 Boling Highway, Wharton, Texas 77488.
Attorney for licensee: Jack R. Newman, Esq., Newman & Holtzinger,
P.C., 1615 L Street, NW, Washington, DC 20036.
NRC Project Director: Suzanne C. Black.

Houston Lighting & Power Company, City Public Service Board of San
Antonio, Central Power and Light Company, City of Austin, Texas, Docket
Nos. 50-498 and 50-499, South Texas Project, Units 1 and 2, Matagorda
County, Texas.

Date of amendment request: March 21, 1994.
Description of amendment request: The licensee proposes to revise
Technical Specifications 3.1.2.3 ``Reactivity Control Systems Charging
Pumps--Shutdown'' and 3.1.2.1 ``Boration Systems Flow Paths--
Shutdown.'' The amendment would allow energizing of an inoperable
centrifugal charging pump in preparation for switching of the
centrifugal charging pumps, provided the pump discharge is isolated
from the reactor coolant system.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration. The NRC staff has reviewed the licensee's analysis
against the standards of 10 CFR 50.92(c). The NRC staff's review is
presented below:
(1) The proposed change does not involve a significant increase in
the probability or consequences of a previously evaluated accident.
The proposed change is to modify the note which permits energizing
of an inoperable centrifugal charging pump for testing purposes,
provided the pump discharge is isolated from the reactor coolant
system, to include pump energization for switching purposes.
The proposed changes can potentially impact two events during Modes
4, 5, and 6: (1) Cold overpressurization of the reactor coolant system,
and (2) boron dilution resulting in a return to criticality. The
requirements of Specification 3.1.2.3 with regard to the cold
overpressure mitigation system analysis would remain valid because the
inoperable centrifugal charging pump would be isolated from the reactor
coolant system. A return to criticality would be prevented because the
action statement of Specification 3.1.2.1 would be entered if the boron
injection flow path could not be restored following centrifugal
charging pump switching. Therefore, allowing energization of an
inoperable pump for switching would have an insignificant effect on the
probability of an overpressurization and boron dilution accident.
Energization of an inoperable pump is currently permitted for
testing purposes provided the pump discharge is isolated from the
reactor coolant system. It is operationally desirable to maintain flow
to the reactor coolant pump seals during the centrifugal charging pump
switching process. This proposed change will not only protect the
reactor coolant system from overpressurization at low temperatures, but
will also provide the capability of maintaining reactor coolant pump
seal injection flow during the switching process.
Therefore, there is no increase in the probability or consequences
of a previously evaluated accident.
(2) The proposed change does not create the possibility of a new or
different kind of accident from any previously evaluated.
Required boron injection flow paths would be maintained in Modes 4,
5, and 6 except during centrifugal charging pump switching. In the
event the boron injection flow path could not be restored after
centrifugal charging pump switching, the action statement of
Specification 3.1.2.1 would be entered. The proposed changes would not
affect the operability of safety-related equipment and reactor coolant
pump seal injection flow could be maintained. The plant operators are
knowledgeable of the potential situation being created by energizing
two centrifugal charging pumps and will follow direct administrative
controls to isolate the pumps from the reactor coolant system.
Therefore, the possibility of a new or different kind of accident is
not created.
(3) The proposed change does not involve a significant reduction in
the margin of safety.
Cold overpressure mitigating system requirements in Specification
3.1.2.3 would continue to be maintained as a result of the proposed
change. Thus, 10 CFR 50 Appendix G limits will not be affected.
Although the boron injection flow path required by Specification
3.1.2.1 may briefly be compromised, there is no significant reduction
in a margin of safety because core alterations would be halted and
positive reactivity changes would not be made if the boron injection
path could not be maintained after centrifugal charging pump switching.
This action, coupled with the short time period required for
centrifugal charging pump switching, would preclude a return to
criticality event. Therefore, there is no significant reduction in a
margin of safety.
Based on this review, it appears that the standards of 10 CFR
50.92(c) are satisfied. Therefore, the NRC staff proposes to determine
that the request for amendments involves no significant hazards
consideration.
Local Public Document Room Location: Wharton County Junior College,
J.M. Hodges, Learning Center, 911 Boling Highway, Wharton, Texas 77488.
Attorney for licensee: Jack R. Newman, Esq., Newman & Holtzinger,
P.C., 1615 L Street, NW, Washington, DC 20036.
NRC Project Director: Suzanne C. Black.

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: February 16, 1994 (Reference LAR 94-
05).
Description of amendment request: The proposed amendments would
revise the combined Technical Specifications (TS) for the Diablo Canyon
Power Plant Unit Nos. 1 and 2 to revise TS 3/4.7.2.1, ``Steam Generator
Pressure/Temperature Limitation,'' 3/4.7.7.1, ``Snubbers,'' 3/4.7.8.1,
``Sealed Source Contamination,'' 3/4.7.11, ``Area Temperature
Monitoring,'' and 3/4.7.13, ``Flood Protection,'' in accordance with
the Commission's Final Policy Statement on TS Improvements for Nuclear
Power Reactors. These TS would be relocated to plant administrative
controls and the final safety analysis report by reference.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:

a. Do the changes involve a significant increase in the
probability or consequences of an accident previously evaluated?
These proposed changes simplify the TS, meet regulatory
requirements for relocated TS, and implement the recommendations of
the Commission's Final Policy Statement on TS Improvements. Future
changes to these requirements will be controlled by 10 CFR 50.59.
The proposed changes are administrative in nature and do not involve
any modifications to any plant equipment or affect plant operation.
Therefore, the proposed changes do not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
b. Do the changes create the possibility of a new or different
kind of accident from any accident previously evaluated?
The proposed changes are administrative in nature, do not
involve any physical alterations to any plant equipment, and cause
no change in the method by which any safety-related system performs
its function.
Therefore, the proposed changes do not create the possibility of
a new or different kind of accident from any accident previously
evaluated.
c. Do the changes involve a significant reduction in a margin of
safety?
The proposed changes do not alter the basic regulatory
requirements and do not affect any safety analyses. Therefore, the
proposed changes do not involve a significant reduction in a margin
of safety.

The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment 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: Theodore R. Quay.

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: February 16, 1994 (Reference LAR 94-
04).
Description of amendment request: The proposed amendments would
revise the combined Technical Specifications (TS) for the Diablo Canyon
Power Plant Unit Nos. 1 and 2 to revise TS 4.2.2, ``Heat Flux Hot
Channel Factor--FQ(z),'' and 6.9.1.8, ``Core Operating Limits
Report,'' to implement the revised methodology for calculating the
penalty to FQ(z). The specific TS changes proposed are as follows:
(1) The 2 percent FQ(z) penalty listed in TS 4.2.2.2.e.1)
would be deleted and the statement revised to indicate the use of an
appropriate factor to be specified in the Core Operating Limits Report
(COLR).
(2) TS 6.9.1.8.b.1. would be changed to reference Revision 1 of
WCAP 10216-P-A, ``Relaxation of Constant Axial Offset Control
FQ(z) Surveillance Technical Specification,'' dated February 1994.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:

a. Does the change involve a significant increase in the
probability or consequences of an accident previously evaluated?
The Heat Flux Hot Channel Factor, FQ(z), is not involved in
the initiation of any accident. Verifying FQ(z) is below its
limit ensures initial conditions for accident analyses are met. The
proposed changes have been previously approved by the NRC and
provide for application of a more conservative FQ(z) penalty
which will ensure that possible FQ(z) margin decreases are
adequately accounted for. Therefore, if the FQ(z) does exceed
its limit, the appropriate actions in TS 3.2.1 and TS 3.2.2. will be
taken and are adequate to ensure design basis accidents analyses
assumptions are met.
Therefore, the proposed changes do not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
b. Does the change create the possibility of a new or different
kind of accident from any accident previously evaluated?
FQ(z) is not involved in the initiation of any accident.
The FQ(z) surveillance provides assurance that the initial
conditions for accident assumptions are met. FQ(z) is a
measurement of a physical property and is not involved in the
initiation of any accident.
Therefore, the proposed changes do not create the possibility of
a new or different kind of accident from any accident previously
evaluated.
c. Does the change involve a significant reduction in a margin
of safety?
The FQ(z) surveillance ensures that certain core parameters
are maintained consistent with supporting assumptions regarding the
core for postulated accidents. The methodology used in Revision 1 to
WCAP-10216-P-A adequately accounts for FQ(z) increases between
monthly flux maps. Using the methodology of Revision 1 to WCAP-
10216-P-A results in a FQ(z) penalty which is more conservative
than the current TS FQ(z) penalty of 2 percent. If the
FQ(z) increases above the TS limit, appropriate actions in TS
3.2.1 and TS 3.2.2. are adequate to ensure design basis accidents
analyses assumptions are met.
Therefore, the proposed changes do not involve a significant
reduction in a margin of safety.

The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment 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: Theodore R. Quay

Portland General Electric Company, et al., Docket No. 50-344, Trojan
Nuclear Plant, Columbia County, Oregon

Date of amendment request: January 27, 1993, revised March 8, 1994.
Description of amendment request: The proposed amendment, as
revised, by Portland General Electric Company, PGE or the licensee,
would change the Trojan Nuclear Plant (Trojan) Appendix A Technical
Specifications to reflect the permanently defueled status of the
facility. The permanent cessation of power generation at Trojan and the
May 5, 1993 amendment to the license which granted the licensee a
Possession Only License for the facility has rendered many of the
existing provisions of the current Appendix A Technical Specifications
inappropriate. PGE has developed Permanently Defueled Technical
Specifications (PDTS) for Trojan using NUREG-1431, ``Standard Technical
Specifications, Westinghouse Plants,'' as a basis for the PDTS scope
and format.
Based on a series of discussions between the PGE staff and the NRC
on February 9, 1994 and February 28, 1994 the licensee has revised
several requirements contained in the original June 27, 1993 amendment
request. These revisions were forwarded to the NRC staff by letter
dated March 8, 1994. The March 8, 1994 revision updates the June 27,
1993 submittal deleting reference to sections that had been relocated
out of the Technical Specifications by amendments granted since June
1993. It also provided supplemental information concerning the deletion
and/or relocation of certain existing Trojan Technical Specifications
requested by the NRC staff. The March 8, 1994 submittal also clarified
the long term organization at the site, established a line of
succession for the operational command and control function, modified
the review and audit functions performed by the Independent Review and
Audit Committee assuring their independence of review, required
independent review of certain programs and manuals, limited annual and
quarterly doses and operability and usage of the effluent treatment
systems to conform to Appendix I to 10 CFR Part 50, continued the
existing requirement for surveillance of former structural
modifications to the facility, and required the submission of an annual
radioactive effluent release report in accordance with 10 CFR 50.36a.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.92(a), the licensee has
provided an analysis of the issue of no significant hazards
consideration. In accordance with the requirements of 10 CFR 50.92,
Issuance of Amendment, this license amendment request, as revised, is
judged to involve no significant hazards consideration based upon the
following:

1. Operation of the facility in accordance with the proposed
amendment would not involve a significant increase in the
probability or consequences of an accident previously evaluated.
The licensee analysis shows that the worst case design basis
accident for this plant, in its permanently shutdown defueled state,
is a fire in the radioactive waste annex building. The licensee has
also identified a second design basis accident scenario, a fuel
handling accident in the vicinity of the Trojan spent fuel pool.
Other Trojan Final Safety Analysis Report (FSAR) accident scenarios
addressed in Chapter 15 are no longer applicable to Trojan in the
permanently defueled mode. The proposed amendment, as revised, does
not lessen any of the requirements associated with either the
radioactive waste annex building or the spent fuel pool therefore
the probability of either accident occurring is unchanged. The
proposed amendment, as revised, does not change the consequences of
the accident since it does not affect the magnitude, detection, or
mitigation of either accident scenario. Additionally, the ability of
the radioactive waste annex building and the spent fuel pool to
withstand other applicable FSAR events, natural phenomena, and fires
is either unchanged from the existing licensing basis or is improved
during the permanently defueled condition.
2. Operation of the facility in accordance with the proposed
amendment would not create the possibility of a new or different
kind of accident from any accident previously evaluated.
Maintaining the permanently defueled facility in accordance with
the PDTS, as revised by the March 8, 1994 letter, does not create
the possibility of a new or different kind of accident from any
previously considered. Most of the existing plant systems and
functions will not be operational in the permanently defueled
condition since power operations are prohibited and all of the fuel
at Trojan is stored in the spent fuel pool. However,
all structures, systems and components that are necessary for safe fuel
handling and storage activities will be maintained operable during the
permanently defueled condition. The proposed PDTS, as revised, provide
operation and surveillance requirements and administrative controls
which are sufficient to ensure that the required structures, systems
and components will be maintained operable in the permanently defueled
condition.
3. Operation of the facility in accordance with the proposed
amendment does not involve a significant reduction in a margin of
safety.
The proposed PDTS, as revised by the March 8, 1994 letter, are
sufficient to ensure no reduction in a margin of safety, in part,
because of the reduced range of design basis accidents against which
the facility must be protected now that the facility is prohibited
from power operations and is permanently defueled. Only a fire in
the radioactive waste storage facility or a fuel handling accident
are relevant during the permanently defueled condition. The margins
of safety for both of these accidents will remain the same or
improve by maintaining the facility in accordance with the proposed
PDTS, as revised. None of the other Chapter 15 FSAR accidents are
applicable since power operations are prohibited and the facility is
permanently defueled. Additionally, the margins of safety for other
applicable FSAR events, natural phenomena, and fires are either
unchanged from the existing licensing basis or is improved during
the permanently defueled condition.

The NRC staff has reviewed the licensee's analysis, and based on
this review, it appears that the three standards of 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request, as revised by the March 8, 1994 supplement, involves
no significant hazards consideration.
Local Public Document Room location: Branford Price Millar Library,
Portland State University, 934 S.W. Harrison Street, P.O. Box 1151,
Portland, Oregon 97207.
Attorney for licensee: Leonard A. Girard, Esq., Portland General
Electric Company, 121 S.W. Salmon Street, Portland, Oregon 97204.
NRC Project Director: Seymour H. Weiss

Southern Nuclear Operating Company, Inc., Docket Nos. 50-348 and 50-
364, Joseph M. Farley Nuclear Plant, Units 1 and 2, Houston County,
Alabama

Date of amendments request: October 14, 1993
Description of amendments request: The proposed changes would
revise surveillance test intervals and allowed outage times for reactor
trip system (RTS) and engineered safety feature actuation system
(ESFAS) instrumentation. The proposed changes would also revise certain
RTS/ESFAS functions, minimum channels operable, channel calibration,
and channel functional test requirements to ensure they are in concert
with the Westinghouse Standard Technical Specifications and WCAP-10271,
``Evaluation of Surveillance Frequencies and Out-of-Service Times for
Reactor Protection Instrumentation Systems.''
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:

(1) The proposed changes to the RTS/ESFAS STIs [surveillance
test intervals] and AOTs [allowed outage times], and Minimum
Channels Operable, Channel Calibration and Channel Functional Test
requirements will not involve a significant increase in the
probability or consequences of an accident previously evaluated.
The determination that the result of the proposed changes
associated with STI and AOT are within all acceptable criteria
[that] has been established in the SERs [Safety Evaluations]
prepared for WCAP-10271; WCAP-10271 Supplement 1; WCAP-10271
Supplement 2; and WCAP-10271 Supplement 2, Revision 1.
Implementation of the proposed changes results in a slight increase
in the Reactor Trip System yearly unavailability. This slight
increase, which is primarily due to less frequent surveillance,
results in a slight increase in Core Damage Frequency (CDF) and
public health risk. The values determined by the WOG and presented
in the WCAP for the increase in CDR were verified by Brookhaven
National Laboratory (BNL) as part of an audit and sensitivity
analyses for the NRC staff. Based on the small value of the increase
compared to the range of uncertainty in the CDF, the increase is
considered acceptable. Increasing STIs and AOTs is not expected to
affect the probability or consequences of previously evaluated
accidents.
The change associated with the Minimum Channel Operable
requirement for the RTS Turbine Trip by Turbine Throttle Valve
Closure provides additional operating flexibility based on the
Westinghouse Standard Technical Specifications, Revision 5. The new
action statement ensures that any inoperable channel is placed in
trip, and the remaining operable channels fulfill the necessary
reactor trip diversity function. The change associated with the
ESFAS Minimum Channel Operable requirement for Containment
Pressure--High-High assures that the Technical Specifications
reflect the correct as-built design actuation logic while ensuring
the function continues to meet the single failure criteria. The
proposed change to the turbine trip reactor trip function Channel
Calibration reflects the assumptions in WCAP-10271 and current
Farley calibration practices. The proposed change to the safety
injection ESF [engineered safety feature] input for the reactor trip
Channel Functional Test is consistent with the assumptions in WCAP-
10271 and current Farley surveillance testing practices. The
proposed changes to the ESF permissive interlocks Channel
Calibration and Channel Functional Test requirements are in concert
with the NRC SER for WCAP-10271 and Farley surveillance practices.
The proposed change to the ESF manual initiation functions Channel
Functional Test reflects the proper surveillance requirements for a
Westinghouse Solid State Protection System (SSPS). The proposed
changes to these RTS/ESFAS Channel Calibration and Channel
Functional Test requirements are also consistent with Westinghouse
Standard Technical Specifications.
(2) The proposed changes will not create the possibility of a
new or different kind of accident from any accident previously
evaluated. The proposed changes to not involve hardware changes and
do not result in a change in the manner in which the RTS/ESFAS
provides plant protection or the manner in which surveillance
testing is perform to demonstrate operability. Therefore, a new or
different kind of accident will not occur as a result of these
changes.
(3) The proposed changes do not involve a significant reduction
in a margin of safety. The proposed changes do not alter the manner
in which safety limits, limiting safety system setpoints, or
limiting conditions for operation are determined. The impact of
reduced testing, other than as addressed above, is to allow a longer
time interval over which instrument uncertainties (e.g., drift) may
act. Evaluations have been performed to assure that the plant
setpoints properly account for these instrument uncertainties over
the longer time interval. RTS diversity is still provided by the
Turbine Throttle Valve closure logic circuits. Steam Line Isolation
diversity continues to be provided by the ESFAS Containment
Pressure--High-High. Changes to certain RTS/ESFAS Channel
Calibration and Channel Functional Test surveillances clarify what
tests are required and when the tests are performed. Implementation
of the proposed changes is expected to result in an overall
improvement in safety as noted below.
a. Less frequent testing will potentially result in fewer
inadvertent reactor trips and ESF component actuation.
b. Longer allowed outage times provide for better assessments of
problems and easier repairs, ultimately resulting in better
equipment performance.
c. Less frequent distraction of the plant operator and shift
supervisor to attend to and support instrumentation testing will
improve the effectiveness of the operating staff in monitoring and
controlling plant operation.

The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: Houston-Love Memorial Library,
212 W. Burdeshaw Street, Post Office Box 1369, Dothan, Alabama 36302
Attorney for licensee: James H. Miller, III, Esq., Balch and
Bingham, Post Office Box 306, 1710 Sixth Avenue North, Birmingham,
Alabama 35201
NRC Project Director: S. Singh Bajwa

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 No. 1, Lake County, Ohio

Date of amendment request: June 29, 1992, as supplemented on
February 22, 1994
Description of amendment request: The proposed amendment would
modify the following Technical Specifications and their associated
bases to permit longer allowable outage times (AOT) and increase
surveillance testing intervals from monthly to quarterly: 3/4.3.1
``Reactor Protection System Instrumentation;'' 3/4.3.2 ``Isolation
Actuation Instrumentation;'' 3/4.3.3 ``Emergency Core Cooling System
Actuation Instrumentation;'' 3/4.3.4. ``Recirculation
Pump Trip Actuation Instrumentation ATWS Recirculation Pump Trip System
Instrumentation;'' 3/4.3.5 ``Reactor Core Isolation Cooling System
Actuation Instrumentation;'' 3/4.3.6 ``Control Rod Block
Instrumentation;'' 3/4.3.9 ``Plant Systems Actuation Instrumentation;''
and 3/4.4.2 ``Safety Valves.''
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration. The NRC staff has reviewed the licensee's analysis
against the standards of 10 CFR 50.92(c). The staff's review is
presented below.
The proposed changes do not involve a significant increase in the
probability or consequences of an accident previously evaluated because
the proposed changes do not involve a change to the plant design or
operation. The changes simply involve the frequency at which testing of
the instrumentation is performed and the AOT for instruments. There may
be a small increase in average instrument failure frequency as a result
of increasing the surveillance interval. However, the proposed changes
will require that a check be made for the majority of instruments to
assure that making them inoperable for surveillance testing or repair
does not result in a loss of function. This added check assures that a
loss of function has not occurred, or if it has, that the appropriate
ACTION statement be entered promptly. Therefore, these proposed changes
do not result in a significant increase in either the probability or
consequences of any accident previously analyzed.
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 result in any change in the plant
design or operation. The changes are for increased AOT and decreased
frequency of instrumentation testing. The instrumentation involved are
those instruments which sense plant problems and/or accidents, and then
initiate systems or alarms to respond to the plant problem/accident.
The proposed changes do not modify any of the instruments, or the
initiation logic formed by the instruments. Therefore, no new or
different type of an accident has been created.
The proposed changes do not involve a significant reduction in a
margin of safety because the small increase in average instrument
failure frequency is offset by safety benefits such as a reduction in
the number of inadvertent test-induced scrams, a reduction in wear due
to excessive equipment test cycling, and better optimization of plant
personnel resources. In addition, the proposed changes will require
that a check be made for the majority of instruments to assure that
making them inoperable for surveillance testing or repair does not
result in a loss of function. This added check assures that a loss of
function has not occurred, or if it has, that the appropriate ACTION
statement be entered promptly. Therefore, these proposed changes do not
result in a significant reduction in a margin of safety.
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, Ohio 44081.
Attorney for licensee: Jay Silberg, Esq., Shaw, Pittman, Potts &
Trowbridge, 2300 N Street, NW., Washington, D.C. 20037.
NRC Project Director: John N. Hannon.

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: November 15, 1993
Brief description of amendments: The proposed amendment would
revise the Comanche Peak Steam Electric Station (CPSES) Units 1 and 2
technical specifications to increase the maximum permitted power at
which the post-refueling power ascension reactor coolant system (RCS)
flow verification can be performed.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:

1. The proposed changes do not involve a significant increase in
the probability or consequences of a previously evaluated accident.
The proposed change increases the power at which the
verification of the Reactor Coolant System flow rate with a
precision heat balance can be performed. The only potentially
relevant concern for this change is the possibility of having
insufficient flow to support the accident analyses at the higher
(85%) power level. Power level and RCS flow are important parameters
in determining the severity of an event but have no impact on the
initiation of an event or accident. Thus, the change does not
involve a significant increase in the probability of any previously
analyzed accident.
Although accidents tend to be more severe at higher initial
power levels, the acceptance criteria of the applicable safety
analyses continue to be met [even when the test is conducted at the
higher power level]. Thus, the proposed change would not involve an
increase in the consequences of any previously analyzed accident.
2. The proposed changes do not create the possibility of a new
or different kind of accident from any accident previously
evaluated.
The proposed change merely increases the power at which the
initial post-refueling startup verification of RCS flow with a
precision heat balance may be performed. Since the new power level
is within the normal operating range of the reactor, it does not
create the possibility of a new or different kind of accident from
any accident previously evaluated.
3. The proposed changes do not involve a significant reduction
in the margin of safety.
The proposed change increases the power which can be attained
prior to verification of the Reactor Coolant System flow with a
precision heat balance. The power value is increased from 75% RTP
[rated thermal power] to 85% RTP. An analysis was performed which
demonstrates that the safety analysis DNB [departure from nucleate
boiling] limit will not be exceeded if the initial power is 85% or
less, even with a significant reduction in flow. A flow reduction
significantly different from the expected flow is highly unlikely,
however, since RCS flow is verified by measurement of elbow tap
differential pressure prior to operation in Mode 1. This flow
measurement, although less accurate than the precision heat balance,
is sufficient to assure adequate flow at 85% power. Adequate
limitations on power level, F delta-H verification, and the power
range neutron flux-high setpoint are imposed during post-refueling
power ascension to ensure that if the RCS flow is not verified with
a precision heat balance until 85% RTP, the results of the accident
analyses would remain valid.
The evaluation of the DNB-limited events, initiated from a power
level of 85% RTP, considered the limitations imposed during the
power ascension. Based on this evaluation, it is concluded that,
even though 85% RTP is a more severe initial condition, the
applicable event acceptance criteria would continue to be met;
therefore, the proposed change does not involve a significant
reduction in the margin of safety.

The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: University of Texas at
Arlington Library, Government Publications/Maps, 701 South Cooper, P.O.
Box 19497, Arlington, Texas 76019
Attorney for licensee: George L. Edgar, Esq., Newman and
Holtzinger, 1615 L Street, N.W., Suite 1000, Washington, D.C. 20036
NRC Project Director: Suzanne C. Black

TU Electric Company, Docket Nos. 50-445 and 50-446, Comanche Peak Steam
Electric Sta

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Source: Frix Law Library, https://www.frixlaw.com/law-library/documents/fr%3A94-8780. Public record. Not legal advice.