[Federal Register Volume 61, Number 129 (Wednesday, July 3, 1996)]
[Notices]
[Pages 34884-34908]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: 96-16879]
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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 June 8, 1996, through June 21, 1996. The
last biweekly notice was published on June 19, 1996 (61 FR 31171).
Notice of Consideration of Issuance of Amendments to Facility
Operating Licenses, Proposed No Significant Hazards Consideration
Determination, and Opportunity for a Hearing
The Commission has made a proposed determination that the following
amendment requests involve no significant hazards consideration. Under
the Commission's regulations in 10 CFR 50.92, this means that operation
of the facility in accordance with the proposed amendment would not (1)
involve a significant increase in the probability or consequences of an
accident previously evaluated; or (2) create the possibility of a new
or different kind of accident from any accident previously evaluated;
or (3) involve a significant reduction in a margin of safety. The basis
for this proposed determination for each amendment request is shown
below.
The Commission is seeking public comments on this proposed
determination. Any comments received within 30 days after the date of
publication of this notice will be considered in making any final
determination.
Normally, the Commission will not issue the amendment until the
expiration of the 30-day notice period. However, should circumstances
change during the notice period such that failure to act in a timely
way would result, for example, in derating or shutdown of the facility,
the Commission may issue the license amendment before the expiration of
the 30-day notice period, provided that its final determination is that
the amendment involves no significant hazards consideration. The final
determination will consider all public and State comments received
before action is taken. Should the Commission take this action, it will
publish in the Federal Register a notice of issuance and provide for
opportunity for a hearing after issuance. The Commission expects that
the need to take this action will occur very infrequently.
Written comments may be submitted by mail to the Chief, Rules
Review and Directives Branch, Division of Freedom of Information and
Publications Services, Office of Administration, U.S. Nuclear
Regulatory Commission, Washington, DC 20555-0001, and should cite the
publication date and page number of this Federal Register notice.
Written comments may also be delivered to Room 6D22, Two White Flint
North, 11545 Rockville Pike, Rockville, Maryland from 7:30 a.m. to 4:15
p.m. Federal workdays. Copies of written comments received may be
examined at the NRC Public Document Room, the Gelman Building, 2120 L
Street, NW., Washington, DC. The filing of requests for a hearing and
petitions for leave to intervene is discussed below.
By August 2, 1996, the licensee may file a request for a hearing
with respect to issuance of the amendment to the subject facility
operating license and any person whose interest may be affected by this
proceeding and who wishes to participate as a party in the proceeding
must file a written request for a hearing and a petition for leave to
intervene. Requests for a hearing and a petition for leave to intervene
shall be filed in accordance with the Commission's ``Rules of Practice
for Domestic Licensing Proceedings'' in 10 CFR Part 2. Interested
persons should consult a current copy of 10 CFR 2.714 which is
available at the Commission's Public Document Room, the Gelman
Building, 2120 L Street, NW., Washington, DC and at the local public
document room for the particular facility involved. If a request for a
hearing or petition for leave to intervene is filed by the above date,
the Commission or an Atomic Safety and Licensing Board, designated by
the Commission or by the Chairman of the Atomic Safety and Licensing
Board Panel, will rule on the request and/or petition; and the
Secretary or the designated Atomic Safety and Licensing Board will
issue a notice of a hearing or an appropriate order.
As required by 10 CFR 2.714, a petition for leave to intervene
shall set forth with particularity the interest of the petitioner in
the proceeding, and how that interest may be affected by the results of
the proceeding. The petition should specifically explain the reasons
why intervention should be permitted with particular reference to the
following factors: (1) the nature of the petitioner's right under the
Act to be made a party to the proceeding; (2) the nature and extent of
the petitioner's property, financial, or other interest in the
proceeding; and (3) the possible effect of any order which may be
entered in the proceeding on the petitioner's interest. The petition
should also identify the specific aspect(s) of the subject matter of
the proceeding as to which petitioner wishes to intervene. Any person
who has filed a petition for leave to intervene or who has been
admitted as a party may amend the petition without requesting leave of
the Board up to 15 days prior to the first prehearing conference
scheduled in the proceeding, but such an amended petition must satisfy
the specificity requirements described above.
[[Page 34885]]
Not later than 15 days prior to the first prehearing conference
scheduled in the proceeding, a petitioner shall file a supplement to
the petition to intervene which must include a list of the contentions
which are sought to be litigated in the matter. Each contention must
consist of a specific statement of the issue of law or fact to be
raised or controverted. In addition, the petitioner shall provide a
brief explanation of the bases of the contention and a concise
statement of the alleged facts or expert opinion which support the
contention and on which the petitioner intends to rely in proving the
contention at the hearing. The petitioner must also provide references
to those specific sources and documents of which the petitioner is
aware and on which the petitioner intends to rely to establish those
facts or expert opinion. Petitioner must provide sufficient information
to show that a genuine dispute exists with the applicant on a material
issue of law or fact. Contentions shall be limited to matters within
the scope of the amendment under consideration. The contention must be
one which, if proven, would entitle the petitioner to relief. A
petitioner who fails to file such a supplement which satisfies these
requirements with respect to at least one contention will not be
permitted to participate as a party.
Those permitted to intervene become parties to the proceeding,
subject to any limitations in the order granting leave to intervene,
and have the opportunity to participate fully in the conduct of the
hearing, including the opportunity to present evidence and cross-
examine witnesses.
If a hearing is requested, the Commission will make a final
determination on the issue of no significant hazards consideration. The
final determination will serve to decide when the hearing is held.
If the final determination is that the amendment request involves
no significant hazards consideration, the Commission may issue the
amendment and make it immediately effective, notwithstanding the
request for a hearing. Any hearing held would take place after issuance
of the amendment.
If the final determination is that the amendment request involves a
significant hazards consideration, any hearing held would take place
before the issuance of any amendment.
A request for a hearing or a petition for leave to intervene must
be filed with the Secretary of the Commission, U.S. Nuclear Regulatory
Commission, Washington, DC 20555-0001, Attention: Docketing and
Services Branch, or may be delivered to the Commission's Public
Document Room, the Gelman Building, 2120 L Street, NW., Washington DC,
by the above date. Where petitions are filed during the last 10 days of
the notice period, it is requested that the petitioner promptly so
inform the Commission by a toll-free telephone call to Western Union at
1-(800) 248-5100 (in Missouri 1-(800) 342-6700). The Western Union
operator should be given Datagram Identification Number N1023 and the
following message addressed to (Project Director): petitioner's name
and telephone number, date petition was mailed, plant name, and
publication date and page number of this Federal Register notice. A
copy of the petition should also be sent to the Office of the General
Counsel, U.S. Nuclear Regulatory Commission, Washington, DC 20555-0001,
and to the attorney for the licensee.
Nontimely filings of petitions for leave to intervene, amended
petitions, supplemental petitions and/or requests for a hearing will
not be entertained absent a determination by the Commission, the
presiding officer or the Atomic Safety and Licensing Board that the
petition and/or request should be granted based upon a balancing of
factors specified in 10 CFR 2.714(a)(1)(i)-(v) and 2.714(d).
For further details with respect to this action, see the
application for amendment which is available for public inspection at
the Commission's Public Document Room, the Gelman Building, 2120 L
Street, NW., Washington, DC, and at the local public document room for
the particular facility involved.
Boston Edison Company, Docket No. 50-293, Pilgrim Nuclear Power
Station, Plymouth County, Massachusetts
Date of amendment request: May 1, 1996
Description of amendment request: The proposed amendment would
modify Table 3.1.1, ``Reactor Protection System (SCRAM) Instrumentation
Requirement,'' Table 3.2.C.1, ``Instrumentation that Initiates Rod
Blocks,'' and Technical Specification 3/4.4, ``Standby Liquid
Control.''
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. Does the change involve a significant increase in the
probability or consequences of an accident previously evaluated?
Note 7 to Table 3.1.1 and Note 6 to Table 3.2.C.1
The changes to Note 7 to Table 3.1.1 and the addition of Note 6
to Table 3.2.C.1 are proposed to clarify their requirements, the
appropriate action to take, and their relationship to plant modes.
This revised scram and rod block applicability is acceptable because
control rods withdrawn from a core cell containing no fuel
assemblies have a negligible impact on the reactivity of the core,
and, therefore, these features are not required to be operable (i.e.
provide the capability to scram). Provided all rods otherwise remain
inserted, the RPS [Reactor Protection System] functions serve no
purpose and are not required. In this condition, the required
shutdown margin (Specification 3.3.A.1) and the required one-rod-out
interlock (Specification 3.10.A) ensure that no event requiring the
RPS or Rod Block will occur.
The Actions of Table 3.1.1 for inoperable equipment were
previously revised in Amendment 147 to be consistent with
the improved STS [Standard Technical Specifications]. Action (A)
requires fully inserting all insertable control rods in core cells
containing one or more fuel assemblies. Since Specification 3.10.A
requires all control rods to be fully inserted during fuel movement,
the proposed applicable conditions cannot be entered while moving
fuel. In addition, Specification 3.10.D used for controlling
multiple control rod removal, requires all control rods in a 3X3
array centered on the CRDs [Control Rod Drive] being removed to be
fully inserted and electrically disarmed and all other control rods
fully inserted. The only possible action is control rod withdrawal,
which is addressed by Action A.
Hence operating Pilgrim in accordance with the proposed changes
will not involve a significant increase in the probability or
consequences of an accident previously evaluated.
Section 3/4.4
The proposed change involves reformatting, renumbering, and
rewording of the existing Technical Specifications and Bases along
with other changes to the Technical Specifications discussed above.
The reformatting, renumbering, and rewording along with the other
changes listed involves no technical changes to existing Technical
Specifications, and does not impact initiators of analyzed events.
It also does not impact the assumed mitigation of accidents or
transient events. Therefore, the change does not involve a
significant increase in the probability or consequences of an
accident previously evaluated.
The proposed change relocates requirements to other sections of
the Technical Specifications, to plant procedures, or to the
Technical Specifications BASES. The procedure change and BASES
change processes require any changes that reflect plant design as
described in the FSAR [Final Safety Analysis Report] be evaluated in
accordance with 10 CFR 50.59. Since any changes will be evaluated
per 10 CFR 50.59, no increase (significant or insignificant) in the
probability or consequences of an accident previously evaluated will
be allowed. Therefore, this change will not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
[[Page 34886]]
The proposed change provides more stringent requirements than
previously existed in the Technical Specifications. The more
stringent requirements will not result in operation that will
increase the probability of initiating an analyzed event. If
anything the new requirements may decrease the probability or
consequences of an analyzed event by incorporating the more
restrictive changes discussed above. The change will not alter
assumptions relative to mitigation of an accident or transient
event. The more restrictive requirements will not alter the
operation of process variables, structures, systems, or components
as described in the safety analyses. Therefore, the change will not
involve a significant increase in the probability or consequences of
an accident previously evaluated.
The proposed change deletes the requirements for Standby Liquid
Control (SLC) System operability during Hot Shutdown, Cold Shutdown,
and Refueling. The SLC System is not assumed in the initiation of
any previously evaluated events and therefore the proposed change
will not increase the probability or consequence of a previously
analyzed accident. The SLC System is not assumed to operate in the
mitigation of any previously analyzed accidents which are assumed to
occur during Hot Shutdown, Cold Shutdown or Refueling. This change
will not result in operation that will increase the probability of
initiating an analyzed event. This change will not alter assumptions
relative to mitigation of an accident or alter the operation of
process variables, structures, systems, or components as described
in the safety analyses. Therefore, this change will not involve a
significant increase in the probability or consequences of an
accident previously evaluated.
The proposed change adds an action for both SLC subsystems
inoperable that delays the requirement to initiate plant shutdown
immediately and allows time to recover at least one subsystem before
subjecting the plant to a potentially unnecessary transient.
Allowing a short period of time to recover one subsystem is
acceptable because of the large number of independent control rods
available to shut down the reactor and the diversity of means
available to cause control rod insertion. This change will not alter
assumptions relative to mitigation of an accident or alter the
operation of process variables, structures, systems, or components
as described in the safety analyses. Therefore, this change will not
involve a significant increase in the probability or consequences of
an accident previously evaluated.
The proposed change deletes requirements for demonstrating
operability of the redundant subsystems which eliminates excessive
and unnecessary testing of safety significant equipment. This is
consistent with guidance 10.1 of Generic Letter 93-05, ``Line-Item
Technical Specifications Improvements to Reduce Surveillance
Requirement for Testing During Power Operations''. The change does
not affect the ability of the SLC system to perform on demand, and
by actually lowering the number of demands to demonstrate
operability, reduces the probability of equipment failure. Since the
change will not alter assumptions relative to mitigation of an
accident or alter the operation of process variables, structures,
systems, or components as described in the safety analyses, the
change will not involve a significant increase in the probability or
consequences of an accident previously evaluated.
The proposed change replaces the requirement to verify B-10
enrichment concentration by test anytime boron is added to the
solution and each refueling outage with verifying the enrichment
prior to addition. Since enrichment of the solution in the tank
cannot change by any other means but chemical addition, ensuring
that only properly enriched material is available for addition is
adequate to maintain enrichment at the required level. This change
will not alter assumptions relative to mitigation of an accident or
alter the operation of process variables, structures, systems, or
components as described in the safety analyses. Therefore, this
change will not involve a significant increase in the probability or
consequences of an accident previously evaluated.
2. The operation of Pilgrim Station in accordance with the
proposed amendment will not create the possibility of a new or
different kind of accident from any accident previously evaluated.
Note 7 to Table 3.1.1 and Note 6 to Table 3.2.C.1
The changes to Note 7 to Table 3.1.1, and the addition of Note 6
to Table 3.2.C.1 are proposed to clarify their requirements, the
appropriate action to take, and their relationship to plant modes.
This revised scram and rod block applicability is acceptable because
control rods withdrawn from a core cell containing no fuel
assemblies have a negligible impact on the reactivity of the core,
and, therefore, are not required to be operable. Provided all rods
otherwise remain inserted, the RPS functions serve no purpose and
are not required. In this condition, the required shutdown margin
(Specification 3.3.A.1) and the required one-rod-out interlock
(Specification 3.10.A) ensure that no event requiring the RPS or Rod
Block will occur.
The Actions of Table 3.1.1 for inoperable equipment were
previously revised in Amendment 147 to be consistent with
the improved STS. Action (A) requires fully inserting all insertable
control rods in core cells containing one or more fuel assemblies.
Since Specification 3.1O.A requires all control rods to be fully
inserted during fuel movement, the proposed applicable conditions
cannot be entered while moving fuel. In addition, Specification
3.10.D, used for controlling multiple control rod removal, requires
all control rods in a 3X3 array centered on the CRDs being removed
to be fully inserted and electrically disarmed and all other control
rods fully inserted. The only possible action is control rod
withdrawal, which is addressed by Action A. Hence, operating Pilgrim
in accordance with the proposed changes will not create the
possibility of a new or different kind of accident from any accident
previously evaluated.
Section 3/4.4
The proposed change involves reformatting, renumbering, and
rewording of the existing Technical Specifications and Bases along
with other changes to the Technical Specifications discussed above.
The reformatting, renumbering, and rewording along with the other
changes listed involves no technical changes to existing Technical
Specifications. These changes are administrative and do not impact
the assumed mitigation of accidents or transient events. Therefore,
these changes do not involve a significant increase in the
probability or consequences of an accident previously evaluated.
The proposed change relocates requirements to other Technical
Specification sections, to plant procedures, or to the Technical
Specification BASES. Relocating requirements will not alter the
plant configuration (no new or different type of equipment will be
installed) or changes in methods governing normal plant operation.
Relocating requirements will not impose different requirements and
adequate control of information will be maintained. Relocating
requirements will not alter assumptions made in the safety analysis
and licensing basis. Therefore, these changes will not create the
possibility of a new or different kind of accident from any accident
previously evaluated.
The proposed changes make some existing requirements more
restrictive and add additional requirements to the Technical
Specifications but will not alter the plant configuration (no new or
different type of equipment will be installed) or change methods
governing normal plant operation. These changes do impose different
requirements, however, they are consistent with assumptions made in
the safety analyses. Therefore, these changes will not create the
possibility of a new or different kind of accident from any accident
previously evaluated.
The proposed change relaxes the modes of applicability for the
SLC. Relaxing the applicability will not involve a physical
alteration of the plant (no new or different type of equipment will
be installed) or changes in methods governing normal plant
operation. Therefore, this change will not create the possibility of
a new or different kind of accident from any accident previously
evaluated.
3. The operation of Pilgrim Station in accordance with the
proposed amendment will not involve a significant reduction in a
margin of safety.
Note 7 to Table 3.1.1 and Note 6 to Table 3.2.C.1
This revised scram and rod block applicability is acceptable
because control rods withdrawn from a core cell containing no fuel
assemblies have a negligible impact on the reactivity of the core,
and, therefore, are not required to be operable (provide a scram).
Provided all rods otherwise remain inserted, the RPS functions serve
no purpose and are not required. In this condition, the required
shutdown margin (Specification 3.3.A.1) and the required one-rod-out
interlock (Specification 3.10.A) ensure that no event requiring the
RPS or Rod Block will occur.
The Actions of Table 3.1.1 for inoperable equipment were
previously revised in
[[Page 34887]]
Amendment 147 to be consistent with the improved STS.
Action (A) requires fully inserting all insertable control rods in
core cells containing one or more fuel assemblies. Since
Specification 3.10.A requires all control rods to be fully inserted
during fuel movement, the proposed applicable conditions cannot be
entered while moving fuel. In addition, Specification 3.10.D, used
for controlling multiple control rod removal, requires all control
rods in a 3X3 array centered on the CRDs being removed to be fully
inserted and electrically disarmed and all other control rods fully
inserted. The only possible action is control rod withdrawal, which
is adequately addressed by Action A.
Therefore, operating Pilgrim in accordance with the proposed
changes will not involve a significant reduction in a margin of
safety.
Section 3/4.4
The administrative changes involve no technical changes. These
proposed changes will not reduce a margin of safety because there is
no impact on any safety analysis assumptions. Also, because the
change is administrative in nature, no question of safety is
involved. Therefore, these changes do not involve a significant
reduction in a margin of safety. The change relocates requirements
to other Technical Specification sections, to plant procedures, or
to the Technical Specification BASES. These changes will not reduce
a margin of safety since there is no impact on any safety analysis
assumptions. In addition, the requirements to be transposed are the
same as the existing Technical Specifications. Since any changes to
plant procedures and Technical Specification BASES are required to
be evaluated per 10 CFR 50.59, no reduction (significant or
insignificant) in a margin of safety will be allowed. Therefore,
these changes will not involve a significant reduction in a margin
of safety.
The addition of new requirements and making existing ones more
restrictive either increases or does not affect the margin of
safety. These changes do not impact any safety analysis assumptions.
As such, no question of safety is involved. Therefore, these changes
will not involve a significant reduction in a margin of safety.
The proposed change would remove a backup (in the Hot Shutdown,
Cold Shutdown, and Refueling Modes) to the available systems for
reactivity control; however, this backup is not considered in the
margin of safety when determining the required reactivity for
shutdown and refueling events. This change will have no impact on
any safety analysis assumptions. As such, no question of safety is
involved. Therefore, this change does not involve a significant
reduction in a margin of safety.
The SLC system is not assumed to function in any DBA or
transient and is not the primary success path of a safety sequence
analysis. It is a backup to the CRD scram function, therefore,
allowing a short period of time to recover one subsystem will have
no impact on any safety analysis assumptions. As such, no question
of safety is involved. Therefore, this change does not involve a
significant reduction in a margin of safety.
The change does not alter the requirements for enrichment/
concentration of the boron solution necessary to satisfy 10 CFR
50.62. Since enrichment of the solution in the tank cannot change by
any other means but chemical addition, this change does not involve
a significant reduction in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 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: Plymouth Public Library, 11
North Street, Plymouth, Massachusetts 02360
Attorney for licensee: W. S. Stowe, Esquire, Boston Edison Company,
800 Boylston Street, 36th Floor, Boston, Massachusetts 02199
NRC Project Director: Jocelyn A. Mitchell, Acting Director
Carolina Power & Light Company, et al., Docket Nos. 50-325 and 50-
324, Brunswick Steam Electric Plant, Units 1 and 2, Brunswick
County, North Carolina
Date of amendments request: November 15, 1995
Description of amendments request: The proposed amendments would
revise the Technical Specifications (TS) to alter the wording of TS
4.8.2.5.a in accordance with the guidance of Generic Letter (GL) 91-09,
``Modification of Surveillance Interval For The Electrical Protection
Assemblies In Power Supplies For The Reactor Protection 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, which is presented below:
1. The proposed amendments do not involve a significant increase
in the probability or consequences of an accident previously
evaluated because the proposed change does not alter the design,
function, or operation of the EPAs [Electrical Protective
Assemblies]. The proposed amendments modify the surveillance
requirement for an electrical protective device on the Reactor
Protection System [RPS]. The RPS-EPA units are designed to protect
RPS equipment from abnormal operating voltage or frequency. The
proposed change will preclude the need to test the RPS-EPA units
during power operation. This will eliminate the potential for
reactor scrams and Group isolations during performance of the
surveillance, thus, preventing unwarranted challenges to safety
systems. The proposed change does not affect any accident precursor
or initiator. Therefore, the probability of an accident is not
affected by the proposed change. The proposed amendments do not
affect the operability of the RPS-EPA units. The proposed change
does not affect the ability of the Reactor Protection System to
maintain the integrity of the fuel cladding, protect the reactor
coolant pressure boundary, or limit the amount of energy released to
primary containment. Therefore, the consequences of an accident is
not affected by the proposed change.
2. The proposed amendments do not create the possibility of a
new or different kind of accident from any accident previously
evaluated. As stated above, these proposed amendments do not alter
the design, functions, or operation of the EPAs. The RPS relay trip
logic remains protected from power supplies operating with abnormal
voltage or frequency. Additionally, the redundancy of this
protection is not changed.
Thus, the proposed amendments do not create the possibility of a
new or different kind of accident.
3. The proposed amendments do not involve a significant
reduction in a margin of safety because the benefit to safety by
reducing the frequency of testing during power operation and
attendant possible challenges to safety systems more than offsets
any risk to safety from relaxing the surveillance requirement to
test the EPAs during power operation. The testing of each EPA
channel involves a dead-bus transfer and the momentary interruption
of power results in a half scram and half isolation. Generic Letter
91-09 notes that many plants have encountered problems with the
reset of the half trip resulting in inadvertent scrams and group
isolations that challenge safety systems during power operation.
Eliminating EPA testing at power operation increases the margin of
safety by eliminating the potential for trips due to testing that
challenge safety systems. An insignificant reduction in the margin
of safety is introduced by increasing the test interval up to a
maximum of a refuel cycle which will produce a small increase in
risk that an inoperable EPA would not be detected. The elimination
of potential challenges to safety systems provides a safety benefit
that offsets the increased risks of component failure.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: University of North Carolina
at Wilmington, William Madison Randall Library, 601 S. College Road,
Wilmington, North Carolina 28403-3297
Attorney for licensee: William D. Johnson, Vice President and
Senior Counsel, Carolina Power & Light Company, Post Office Box 1551,
Raleigh, North Carolina 27602
NRC Project Director: Eugene V. Imbro
[[Page 34888]]
Carolina Power & Light Company, Docket No. 50-261, H. B. Robinson
Steam Electric Plant, Unit No. 2, Darlington County, South Carolina
Date of amendment request: June 6, 1996
Description of amendment request: The proposed change would revise
technical specifications (TS) Section 4.2.3 to allow the licensee to
defer the ultrasonic inspection of the reactor coolant pump flywheel
for one operating cycle.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. The proposed change does not involve a significant increase
in the probability or consequences of an accident previously
evaluated.
The safety function of the Reactor Coolant Pump (RCP) flywheel
is to provide a coastdown period during which the RCPs would
continue to provide reactor coolant flow to the core after a loss of
power to the RCPs. The maximum loading on the RCP motor flywheel
results from overspeed following a large break Loss of Coolant
Accident (LOCA). The estimated maximum obtainable speed in the event
of a Reactor Coolant System (RCS) piping break was established
conservatively, and the proposed one-time change does not affect
that analysis.
The RCP flywheels have been carefully designed and manufactured
from high quality steel. Twenty-two inspections have been performed
at HBRSEP, Unit No. 2 over the past 25 years and no indications have
been discovered that would affect the integrity of the flywheel. The
Westinghouse Owners Group (WOG) has performed an extensive study
documented in WCAP-14535, ``Topical Report on Reactor Coolant Pump
Flywheel Inspection Elimination,'' that includes an evaluation of
industry experience, a stress and fracture evaluation, and a risk
assessment, and has concluded that RCP flywheel inspections may be
safely eliminated.
Reduced coastdown times due to a single failed flywheel would
not place the plant in an unanalyzed condition since a locked rotor
(i.e., an instantaneous coastdown) is analyzed in the Updated Final
Safety Analysis Report (UFSAR). The proposed change also does not
increase the amount of radioactive material available for release or
modify any systems used for mitigation of releases during an
accident. Therefore, the proposed change does not involve an
increase in the probability of consequences of an accident
previously evaluated.
2. The proposed change does not create the possibility of a new
or different kind of accident from any accident previously
evaluated.
The proposed change will not change the design, configuration,
or method of operation of the plant. Therefore, the proposed change
will not create the possibility of a new kind of accident from any
previously evaluated.
3. The proposed change does not involve a significant reduction
in the margin of safety.
The RCP flywheels have been carefully designed and manufactured
from high quality steel. Twenty-two inspections have been performed
at HBRSEP, Unit No. 2 over the past 25 years and no indications have
been discovered that would affect the integrity of the flywheel. The
Westinghouse Owners Group (WOG) has performed an extensive study
documented in WCAP-14535, ``Topical Report on Reactor Coolant Pump
Flywheel Inspection Elimination,'' that includes an evaluation of
industry experience, a stress and fracture evaluation, and a risk
assessment, and has concluded that RCP flywheel inspections may be
safety eliminated. The proposed change would only result in a one-
time deferral of the scheduled inspection for one operating cycle.
In consideration of the historical integrity of the HBRSEP, Unit No.
2 RCP flywheels, the industry experience, the results of the WOG
study, and the deferral of the risk of RCP flywheel damage during
disassembly and inspection, we conclude that a one operating cycle
deferral of the scheduled RCP flywheel inspection will not result in
a reduction in the margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: Hartsville Memorial Library,
147 West College Avenue, Hartsville, South Carolina 29550
Attorney for licensee: William D. Johnson, Vice President and
Senior Counsel, Carolina Power & Light Company, Post Office Box 1551,
Raleigh, North Carolina 27602
NRC Project Director: Eugene V. Imbro
Carolina Power & Light Company, et al., Docket No. 50-400, Shearon
Harris Nuclear Power Plant, Unit 1, Wake and Chatham Counties,
North Carolina
Date of amendment request: May 31, 1996
Description of amendment request: The proposed amendment would
change the plant Technical Specifications (TS) Table 3.3-7, Seismic
Monitoring Instrumentation, and TS Table 4.3-4, Seismic Monitoring
Instrumentation Surveillance Requirements, to correct the location
described for one of the three Triaxial Peak Accelerograph Recorders.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. The proposed amendment does not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
These recorders are passive components which serve only a
recording function. They can neither initiate an accident nor serve
to mitigate accident consequences. The proposed change serves only
to correct the location, commensurate with design documents, for one
of the three recorders described in the Technical Specifications.
Accordingly, this change is administrative in nature. Therefore,
there would be no increase in the probability or consequences of an
accident previously evaluated.
2. The proposed amendment does not create the possibility of a
new or different kind of accident from any accident previously
evaluated.
The proposed correction is an administrative change to correct
the location of a recorder currently described in the Technical
Specifications. No physical alterations to plant equipment are being
made, and there will be no changes that alter how 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.
3. The proposed amendment does not involve a significant
reduction in the margin of safety.
Technical Specification Bases 3/4.3.3.3 specify the acceptance
level for seismic instrumentation as ``consistency'' with the
recommendations of Regulatory Guide 1.12. Since the regulatory guide
states only that one recorder should be provided at a ``selected
location on the reactor piping,'' it is not material whether it is
installed on Loop 1 versus Loop 2. Therefore, the proposed change
does not affect a margin of safety as defined in the Bases to the
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: Cameron Village Regional
Library, 1930 Clark Avenue, Raleigh, North Carolina 27605
Attorney for licensee: William D. Johnson, Vice President and
Senior Counsel, Carolina Power & Light Company, Post Office Box 1551,
Raleigh, North Carolina 27602
NRC Project Director: Eugene V. Imbro
[[Page 34889]]
Florida Power and Light Company, Docket Nos. 50-250 and 50-251,
Turkey Point Plant Units 3 and 4, Dade County, Florida
Dates of amendment request: December 18, 1995, May 3 and June 11,
1996
Description of amendment request: The licensee proposed to change
the Turkey Point Units 3 and 4 Technical Specifications (TS) to uprate
the core thermal output of Turkey Point Units 3 and 4 from 2200 MWt to
2300 MWt. The proposed TS changes were divided into eight groups. The
submittal included a ``No Significant Hazards'' evaluation for each of
the eight groups. The groupings are as follows:
TS changes associated with the uprated power level, the revised
core safety limits, revised DNB [departure from nucleate boiling]
parameters, Engineered Safety Features Actuation System (ESFAS) and
reactor trip setpoint changes, and Reactor Coolant Pump (RCP) Breaker
Position Trip, were evaluated together. The safety of these proposed
changes were verified by the accident analyses that were completed in
support of the uprated power.
TS changes associated with reducing the SI [safety injection] pump
discharge head requirement and increasing usable volume requirements
for the Demineralized Water Storage Tank (DWST) and the Condensate
Storage Tank (CST) were addressed together.
TS changes associated with pressurizer and main steam safety valve
(MSSV) setpoint tolerance increases were assessed together.
TS changes associated with operation at reduced power with
inoperable MSSVs were assessed separately.
TS changes associated with the service period for heatup and
cooldown pressure-temperature limit curves were assessed together.
The Surveillance Requirement change for the emergency containment
cooling [ECC] unit operability was handled separately since this was a
design change that required extensive evaluations.
TS change associated with the methyl iodide removal efficiency in
the Control Room Emergency Ventilation System was assessed separately.
All LOCA [loss-of-coolant accident] related changes dealing with
the peaking factor increase, COLR [core operating limit report]
changes, Evaluation Model references, and relocation of peaking factors
from the TS and subsequent inclusion in the COLR were included in one
``No Significant Hazards'' evaluation. All of the items are closely
related since the LOCA analysis is performed to ensure peaking factor
acceptability.
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.
LICENSE CONDITION, RATED THERMAL POWER, CORE SAFETY LIMITS,
REACTOR TRIP SYSTEM INSTRUMENTATION TRIP SETPOINTS, ESFAS
INSTRUMENTATION TRIP SETPOINTS, DNB PARAMETERS AND RCP BREAKER
POSITION TRIP
(1) Operation of the facility in accordance with the proposed
amendments would not involve a significant increase in the
probability or consequences of an accident previously evaluated.
The proposed changes do not involve an increase in the
probability or consequences of an accident previously evaluated
because operation with these revised values will not cause any
design or analysis acceptance criteria to be exceeded. The
structural and functional integrity of all plant systems are
unaffected. The overtemperature Delta T and overpower Delta T
reactor trip functions as well as ESFAS functions are part of the
accident mitigation response and are not accident initiators. All
proposed changes have been assessed and no design and analysis
acceptance criteria have been exceeded. Therefore the probability of
occurrence previously evaluated is not affected.
The proposed changes do not affect the integrity of the fission
product barriers utilized for mitigation of dose consequences as a
result of an accident. Dose consequences were reviewed and
reanalyzed (as needed) and found acceptable. Therefore, the
probability or consequences of an accident previously evaluated are
not significantly increased.
(2) Operation of the facility in accordance with the proposed
amendments would not create the possibility of a new or different
kind of accident from any accident previously evaluated.
The proposed changes do not create the possibility of a new or
different kind of accident from any accident previously evaluated
because their effects do not affect accident initiation sequences.
All new operating configurations have been evaluated and no new
limiting single failures have been identified. In addition, no new
failure modes have been identified. Therefore, it is concluded that
no new or different kind of accident from any accident previously
evaluated has been created as a result of these revisions.
(3) Operation of the facility in accordance with the proposed
amendments would not involve a significant reduction in a margin of
safety.
The proposed changes do not involve a reduction in a margin of
safety because the margin of safety associated with these parameters
as verified by the results of the accident analyses, are within
acceptable limits. All transients impacted have been analyzed and
have met the applicable accident analyses acceptance criteria (e.g.,
DNBR [departure from nucleate boiling ratio], RCS [reactor coolant
system] pressure, secondary side pressure, etc.). The margin of
safety required for each affected safety analysis is maintained. The
adequacy of the revised Technical Specifications values has been
confirmed such that there is no reduction in the margin of safety.
Therefore, the proposed changes do not involve a significant
reduction in the margin of safety.
AVAILABLE VOLUME CHANGE FOR CONDENSATE STORAGE TANK (CST) AND
DEMINERALIZED WATER STORAGE TANK (DWST), AND REDUCED SAFETY
INJECTION (SI) PUMP DISCHARGE HEAD REQUIREMENT.
(1) Operation of the facility in accordance with the proposed
amendments would not involve a significant increase in the
probability or consequences of an accident previously evaluated.
The revised tank volumes and SI head requirements have been
evaluated with respect to system performance and analysis impacts.
All accident analysis acceptance criteria continue to be met. The
design function of all affected systems have been reviewed and all
system design criteria continue to be met. The structural and
functional integrity of the affected systems are unaffected. These
changes are not initiators for any accident and therefore the
probability of occurrence of an accident previously evaluated has
not increased.
The proposed changes do not affect the integrity of the fission
product barriers for mitigation of dose consequences. All dose
consequences remain well within the 10 CFR 100 limits. Therefore
there is no increase in the probability or consequences of an
accident previously evaluated.
(2) Operation of the facility in accordance with the proposed
amendments would not create the possibility of a new or different
kind of accident from any accident previously evaluated.
The revised tank volumes and SI head requirements do not create
the possibility of a new or different kind of accident from any
accident previously evaluated because these modifications do not
affect accident initiation sequences. No new operating configuration
is being imposed by the adjustments that would create a new failure
scenario. In addition, no new failure modes or limiting single
failures have been identified. Therefore, it is concluded that no
new or different kind of accident from any accident previously
evaluated have been created as a result of these revisions.
(3) Operation of the facility in accordance with the proposed
amendments would not involve a significant reduction in a margin of
safety.
The proposed changes do not involve a reduction in a margin of
safety because the margin of safety associated with these
parameters, as verified by the results of the accident analyses and
system evaluations, are within acceptance limits. The margin of
safety required for each affected safety analysis is maintained.
Therefore, the proposed changes do not involve a significant
reduction in the margin of safety.
PRESSURIZER AND MAIN STEAM SAFETY VALVE SETPOINT TOLERANCES
[[Page 34890]]
(1) Operation of the facility in accordance with the proposed
amendments would not involve a significant increase in the
probability or consequences of an accident previously evaluated.
The revised tolerances for main steam safety valves and
pressurizer safety valves do not involve an increase in the
probability or consequences of an accident previously evaluated
because operation with these revised values will not cause any
design or analytical acceptance criteria, such as those applicable
to primary and secondary side pressures to be exceeded. The
structural and functional integrity of the valves are unaffected by
this proposed change. The tolerance changes do not initiate or cause
initiation of any transient. Therefore, the probability of
occurrence previously evaluated is not affected.
The changes do not affect the integrity of the fission product
barriers utilized for dose consequence mitigation. Therefore, the
probability or consequences of an accident previously evaluated is
not increased.
(2) Operation of the facility in accordance with the proposed
amendments would not create the possibility of a new or different
kind of accident from any accident previously evaluated.
The revised valve tolerances do not create the possibility of a
new or different kind of accident from any accident previously
evaluated because the tolerances do not affect accident initiation
sequences. No new operating configuration is being imposed by the
tolerances that would create a new failure scenario. In addition, no
new failure modes or limiting single failures have been identified.
Therefore, it is concluded that no new or different kind of accident
from any accident previously evaluated have been created as a result
of these revisions.
(3) Operation of the facility in accordance with the proposed
amendments would not involve a significant reduction in a margin of
safety.
The changes to valve tolerances do not involve a reduction in a
margin of safety because the margin of safety associated with the
MSSVs and the pressurizer safety valves, as verified by the results
of the accident analyses and valve evaluations, are within
acceptable limits. Transients impacted by this change have been
analyzed and have met the applicable accident analyses acceptance
criteria, such as those applicable to primary and secondary side
pressure. The margin of safety required for each affected safety
analysis is maintained. This conclusion is not changed by the valve
tolerances for the main steam safety valves and the pressurizer
safety valves. Therefore, the changes do not involve a significant
reduction in the margin of safety.
OPERATION AT REDUCED POWER WITH INOPERABLE MAIN STEAM SAFETY
VALVES
(1) Operation of the facility in accordance with the proposed
amendments would not involve a significant increase in the
probability or consequences of an accident previously evaluated.
The proposed changes do not involve a significant increase in
the probability or consequences of an accident previously evaluated.
The proposed maximum allowable power level values will ensure that
the secondary side steam pressure will not exceed 110 percent of the
design pressure following a Loss of Load/Turbine Trip event, when
one or more main steam safety valves (MSSVs) are declared
inoperable. The proposed change will not impact the classification
of the Loss of Load/Turbine Trip event as a Condition II probability
event (faults of moderate frequency) per ANSI - N18.2, 1973.
Accordingly, since the proposed maximum allowable power level will
maintain the capability of the MSSVs to perform their pressure
relief function associated with a Loss of Load/Turbine Trip event,
there will be no effect on the probability or consequences of an
accident previously evaluated.
(2) Operation of the facility in accordance with the proposed
amendments would not create the possibility of a new or different
kind of accident from any accident previously evaluated.
The proposed changes do not create the possibility of a new or
different kind of accident from any accident previously evaluated.
The proposed changes do not involve any change to the configuration
of any plant equipment, and no new failure modes have been defined
for any plant system or component. The proposed maximum allowable
power level as specified in TS Table 3.7-1 will improve the
capability of the MSSVs to perform their pressure relief function to
ensure the secondary side steam pressure does not exceed 110 percent
of design pressure following a Loss of Load/Turbine Trip event.
Therefore, since the function of the MSSVs is improved by the
proposed changes, the possibility of a new or different kind of
accident from any accident previously evaluated is not created.
(3) Operation of the facility in accordance with the proposed
amendments would not involve a significant reduction in a margin of
safety.
The proposed changes to the Technical Specifications do not
involve a significant reduction in a margin of safety. The algorithm
methodology used to calculate the maximum allowable power level is
conservative and bounding since it is based on a number of
inoperable MSSVs per loop; i.e., if only one MSSV in one loop is out
of service, the required action to reduce power to the maximum
allowable power level would be the same as if one MSSV in each loop
were out of service. Another conservatism with the algorithm
methodology is with the assumed minimum total steam flow rate
capability of the operable MSSVs. The assumption is that if one or
more MSSVs are inoperable per loop, the inoperable MSSVs are the
largest capacity MSSVs, regardless of which capacity MSSVs are
actually inoperable.
Therefore, since the maximum allowable power level calculated
for the proposed changes using the algorithm methodology are more
conservative and ensure that 110 percent of secondary side steam
pressure is not exceeded following a Loss of Load/Turbine Trip
event, this proposed license amendment will not involve a
significant reduction in a margin of safety.
SERVICE PERIOD FOR HEATUP AND COOLDOWN PRESSURE-TEMPERATURE
LIMIT CURVES
(1) Operation of the facility in accordance with the proposed
amendments would not involve a significant increase in the
probability or consequences of an accident previously evaluated.
Calculation of the service period for the heatup and cooldown
curves does not involve an increase in the probability or
consequences of an accident previously evaluated because the
calculations were completed to verify the adequacy of the existing
curves and to determine an appropriate service period. The use of
approved methods and the acceptable results have shown that no
design or analysis criteria are changed. The structural and
functional integrity of the reactor vessel has been verified.
No fission product barriers or inputs to dose analyses are
adversely affected by these calculations and reverification of the
existing heatup/cooldown curves. Therefore, the probability or
consequences of an accident previously evaluated are not increased.
(2) Operation of the facility in accordance with the proposed
amendments would not create the possibility of a new or different
kind of accident from any accident previously evaluated.
The revised service period does not create the possibility of a
new or different kind of accident from any accident previously
evaluated because the recalculation of an acceptable service period
does not affect accident initiation sequences. No new operating
configuration is being imposed by the calculations that would create
a new failure scenario. In addition, no new failure modes or
limiting single failures have been identified. Therefore, the types
of accidents defined in the UFSAR continue to represent the credible
spectrum of events to be analyzed which determine safe plant
operation. Therefore, it is concluded that no new or different kind
of accident from any accident previously evaluated have been created
as a result of these revisions.
(3) Operation of the facility in accordance with the proposed
license amendments would not involve a significant reduction in a
margin of safety.
Calculations were performed to determine the service period
appropriate for the existing curves. The changes to service period
do not involve a reduction in a margin of safety because the margin
of safety associated with the heatup/cooldown curves, as verified by
the results of the analyses, are unchanged. Therefore, the proposed
change to the service period does not involve a significant
reduction in the margin of safety.
MODIFICATION TO SURVEILLANCE REQUIREMENT FOR EMERGENCY
CONTAINMENT COOLING SYSTEM
(1) Operation of the facility in accordance with the proposed
amendments would not involve a significant increase in the
probability or consequences of an accident previously evaluated.
The purpose of the ECC units is to help mitigate the
consequences of an accident (i.e., to help maintain the containment
pressure and temperature within their design
[[Page 34891]]
values following a design basis accident). The ECC units do not
operate during normal operation of the plant. Failure of the ECC
units would not initiate a plant transient or accident. Therefore,
the proposed change involving the ECC units would not affect the
probability of occurrence of an accident previously evaluated.
Evaluations demonstrate that, with two ECC units operating
during a LOCA or MSLB [main steamline break], the containment
pressure and temperature will be maintained within their design
values. These evaluations also demonstrate that, with two ECC units
operating during a LOCA or MSLB, the temperature of the CCWS
[component cooling water system] will be maintained within its
design temperature. Therefore, the proposed change involving the ECC
units would not affect the consequences of an accident previously
evaluated.
(2) Operation of the facility in accordance with the proposed
amendments would not create the possibility of a new or different
kind of accident from any accident previously evaluated.
The purpose of the ECC units is to mitigate design basis
accidents, and failure of the ECC units would not cause a plant
transient or accident. Furthermore, a single failure of an ECC unit
during a LOCA or MSLB would not lead to a new or different kind of
accident. Although the revised Technical Specifications require two
ECC units to start automatically on a LOCA signal, they would also
require that all three ECC units be operable. On a single failure of
an operating ECC unit, there would be sufficient time to start the
standby ECC unit to accomplish the design function of the ECC
system. Therefore, 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
amendments would not involve a significant reduction in a margin of
safety.
The proposed change in the actuation logic of the ECC units
would not cause either the containment pressure and temperature or
the CCWS temperature to exceed their design values. While the energy
released into containment and subsequently transferred to the CCWS
will increase as a result of the thermal uprate, this increase is
insignificant and will not result in either the containment or CCWS
exceeding a design limit. Therefore, the proposed change would not
affect the margin of safety.
CONTROL ROOM EMERGENCY VENTILATION SYSTEM
(1) Operation of the facility in accordance with the proposed
amendments would not involve a significant increase in the
probability or consequences of an accident previously evaluated.
The proposed change does not affect the integrity of the fission
product barriers utilized for mitigation of dose consequences as a
result of an accident. Only the iodide removal efficiency of the
control room emergency ventilation system is increased, and this
change is in the conservative direction.
To assure consistency between testing efficiency and analysis
assumptions for post-accident control room doses, the methyl iodide
removal efficiency required to be demonstrated by laboratory test,
is being increased from 90% to 99%. This increase in testing
efficiency is consistent with the recommendations set by the NRC
staff in Regulatory Guide 1.52 to support analysis efficiencies for
elemental iodine and methyl iodide removal of 95%, respectively.
Testing performed to verify methyl iodide removal efficiency will be
performed under conditions representative of the control room
environment.
Since this change in removal efficiency is in the conservative
direction, plant safety will not be adversely impacted.
(2) Operation of the facility in accordance with the proposed
amendments would not create the possibility of a new or different
kind of accident from any accident previously evaluated.
The proposed change to the control room emergency ventilation
system iodide removal efficiency does not create the possibility of
a new or different kind of accident from any accident previously
evaluated because operation of the control room emergency
ventilation system is not identified in any accident initiation
sequence. The system is provided to minimize operator exposure to
airborne radioactivity released as a result of an accident. The new
operating configuration has been evaluated and no new limiting
single failures have been identified as a result of the proposed
modification. Therefore, it is concluded that no new or different
kind of accidents from any accident previously evaluated have been
created as a result of these revisions.
(3) Operation of the facility in accordance with the proposed
amendments would not involve a significant reduction in a margin of
safety.
The proposed changes do not involve a reduction in the margin of
safety because the margin of safety associated with this change is
in the conservative direction. Thus, plant safety will not be
adversely impacted and the margin of safety required for the
affected safety analysis is maintained. The adequacy of the revised
Technical Specification values to maintain the plant in a safe
operating condition has been confirmed, since the testing will be
done to a more conservative criteria (i.e., 99% efficiency).
Therefore, the proposed changes do not involve a significant
reduction in the margin of safety.
RELOCATION OF FQ(Z) [HEAT FLUX HOT CHANNEL FACTOR] AND F
Delta H [NUCLEAR ENTHALPY RISE HOT CHANNEL FACTOR] LIMITS FROM
TECHNICAL SPECIFICATIONS TO CORE OPERATING LIMITS REPORT AND
EDITORIAL CORRECTIONS
(1) Operation of the facility in accordance with the proposed
amendments would not involve a significant increase in the
probability or consequences of an accident previously evaluated.
The relocation of the values for FQ and F Delta H from the
Technical Specifications to the Core Operating Limits Report is
administrative in nature and has no impact on the probability or
consequences of any Design Bases Event (DBE) occurrence which was
previously evaluated. The determination of the FQ and F Delta H
limits will be performed using methodology approved by the NRC and
poses no significant increase in the probability or consequences of
any accident previously evaluated.
The changes being proposed as editorial in nature do not affect
assumptions contained in the safety analyses, the physical design
and/or operation of the plant, nor do they affect Technical
Specifications that preserve safety analysis assumptions. Therefore,
these proposed changes do not affect the probability or consequences
of accidents previously analyzed.
(2) Operation of the facility in accordance with the proposed
amendments would not create the possibility of a new or different
kind of accident from any accident previously evaluated.
The relocation of the FQ and F Delta H limits from the
Technical Specifications to the Core Operating Limits Report is
administrative in nature and has no impact, nor does it contribute
in any way to the possibility of a new or different kind of accident
from any accident previously evaluated.
The determination of the FQ and F Delta H limits will be
performed using NRC-approved methodology and are submitted to the
NRC as a revision to the COLR to allow the NRC staff to trend
peaking factors. The Technical Specifications will continue to
require operation within the required core operating limits and
appropriate actions will be taken if the FQ and F Delta H
limits are exceeded. Therefore, the proposed amendments does not in
any way create the possibility of a new or different kind of
accident from any accident previously evaluated.
The editorial changes proposed are administrative in nature and
do not affect assumptions contained in plant safety analyses, the
physical design and/or operation of the facility, nor do they affect
Technical Specifications that preserve safety analysis assumptions.
Therefore, these changes do not create the possibility of a new or
different kind of accident.
(3) Operation of the facility in accordance with the proposed
amendments would not involve a significant reduction in a margin of
safety.
The relocation of the FQ and F Delta H limits from the
Technical Specifications to the Core Operating Limits Report is
administrative in nature and has no impact on the margin of safety.
The determination of the FQ and F Delta H limits will be
performed using methodology approved by the NRC and does not
constitute a significant reduction in the margin of safety.
The supporting Technical Specification values are defined by the
accident analyses which are performed to conservatively bound the
operating conditions defined by the Technical Specifications.
Performance of analysis and evaluation have confirmed that the
operating envelope defined by the Technical Specifications continues
to be bounded by the analytical basis, which in no case exceeds the
acceptance limits. Therefore, the margin of safety provided in the
analyses in accordance with the acceptance limits is maintained and
not significantly reduced.
[[Page 34892]]
The changes being proposed as editorial in nature do not relate
to or modify the safety margins defined in, and maintained by the
Technical Specifications. Therefore, the proposed changes which
correct administrative errors and clarify existing Technical
Specification requirements do not involve any reduction in a margin
of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: Florida International
University, University Park, Miami, Florida 33199
Attorney for licensee: J. R. Newman, Esquire, Morgan, Lewis &
Bockius, 1800 M Street, NW., Washington, DC 20036
NRC Project Director: Frederick J. Hebdon
Florida Power and Light Company, Docket Nos. 50-250 and 50-251,
Turkey Point Plant Units 3 and 4, Dade County, Florida
Dates of amendment request: April 19, 1996, May 10, 1996, and May
28, 1996
Description of amendment request: The licensee proposed to change
the Turkey Point Units 3 and 4 Technical Specifications (TS) to address
frequency extension for actions required on a periodic basis, delete
the separate notification requirement for an inoperable startup
transformer, and allow the operating RHR loop to be removed from
operation during refueling operations under certain conditions.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below.
(1) Operation of the facility in accordance with the proposed
amendments would not involve a significant increase in the
probability or consequences of an accident previously evaluated.
The proposed amendments do not involve a significant increase in
the probability or consequences of an accident previously evaluated
because the proposed amendments are purely administrative in nature.
These amendments will not involve a significant increase in the
probability or consequences of an accident previously evaluated
because they do not affect assumptions contained in plant safety
analyses, the physical design and/or operation of the plant, nor do
they affect Technical Specifications that preserve safety analysis
assumptions. Therefore, the proposed changes do not affect the
probability or consequences of accidents previously analyzed.
(2) Operation of the facility in accordance with the proposed
amendments would not create the possibility of a new or different
kind of accident from any accident previously evaluated.
The use of the modified specifications can not create the
possibility of a new or different kind of accident from any
previously evaluated since the proposed amendments will not change
the physical plant or the modes of plant operation defined in the
facility operating license. No new failure mode is introduced due to
the administrative changes and clarifications, since the proposed
changes do not involve the addition or modification of equipment nor
do they alter the design or operation of affected plant systems,
structures, or components.
(3) Operation of the facility in accordance with the proposed
amendments would not involve a significant reduction in a margin of
safety.
The operating limits and functional capabilities of the affected
systems, structures, and components are unchanged by the proposed
amendments. The modified specifications which correct administrative
errors and clarify existing Technical Specification requirements do
not significantly reduce any of the margins of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: Florida International
University, University Park, Miami, Florida 33199
Attorney for licensee: J. R. Newman, Esquire, Morgan, Lewis &
Bockius, 1800 M Street, NW., Washington, DC 20036
NRC Project Director: Frederick J. Hebdon
Gulf States Utilities Company, Cajun Electric Power Cooperative,
and Entergy Operations, Inc., Docket No. 50-458, River Bend
Station, Unit 1, West Feliciana Parish, Louisiana
Date of amendment request: May 30, 1996
Description of amendment request: The proposed amendment would
revise the technical specifications surveillance requirement (SR)
3.8.3.4 to specify a 5-start pressure for the air receivers associated
with the Division III, High Pressure Core Spray emergency diesel
generator.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. Does the change involve a significant increase in the
probability or consequence of an accident previously evaluated?
The purpose of the proposed Technical Specification change is to
establish consistency between the basis for the air start pressure
required for the Division I and II diesels and the value required
for the Division III diesel. The value of 160 psig currently
specified in SR 3.8.3.4 is representative of a 5-start value for the
Division I and II diesels, however, this value is not representative
of a 5-start for the Division III diesel. While the 160 psig value
does serve to satisfy the requirements of 10 CFR 50.36 with regard
to maintaining the lowest functional level required for the Division
III diesel to perform its design safety function, the current value
does not serve to maintain the design margin utilized when sizing
the air receivers for the purpose of satisfying the Standard Review
Plan guidance contained in section 9.5.6 (NUREG-0800 Revision 2).
The proposed value fully complies with the guidance provided in
NUREG-0800 and is more conservative than the value currently
included in the Technical Specifications. The proposed value is well
within the capability of the air system's design and will not
subject the air system to excessive pressures or undue cycling of
the system's compressors. The proposed change has no effect on the
probability of an accident as diesel generators have no bearing on
the initiation of any analyzed event. In addition, the capability of
the Division III diesel to perform its design basis function (i.e.,
starting, accelerating to rated speed and voltage, and connecting to
its respective bus within 13 seconds) is not affected by this
change. The ability of the diesel to support the mitigation of
analyzed accidents is not affected and hence the consequences of any
analyzed event are not affected. Therefore, the proposed change does
not increase the probability or the consequences of previously
analyzed accidents.
2. Does the change create the possibility of a new or different
kind of accident from any accident previously evaluated?
The proposed change does not introduce any new failure modes.
All of the affected components remain within their applicable design
limits. In addition, the environmental qualification of any plant
equipment is not adversely affected by the proposed change. Since
the performance of this system is not adversely affected by this
change and the design margins of this system are not challenged in a
manner differently than previously analyzed, the proposed change
does not create the possibility of a new or different kind of
accident from any accident previously evaluated.
3. Does this change involve a significant reduction in a margin
of safety?
The proposed change raises the required starting air pressure
for the Division III above that currently required by the Technical
Specifications to establish consistency between the basis of the
Division III value with the value used for the Division I and II
diesels. Issuance of the proposed change will establish a 5 start
air receiver pressure for each of the three safety-related diesels
at
[[Page 34893]]
River Bend. While the proposed value is slightly less than the 5
start value discussed in River Bend's SER, the proposed value is
supported by the River Bend site-specific test data and does not
adversely affect existing analyses or system performance. Therefore,
the proposed change does not result in a reduction in a margin of
safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: Government Documents
Department, Louisiana State University, Baton Rouge, LA 70803
Attorney for licensee: Mark Wetterhahn, Esq., Winston & Strawn,
1400 L Street, NW., Washington, DC 20005
NRC Project Director: William D. Beckner
Nebraska Public Power District, Docket No. 50-298, Cooper Nuclear
Station, Nemaha County, Nebraska
Date of amendment request: June 6, 1996, as supplemented by letters
dated June 7 and 9, 1996
Description of amendment request: The proposed amendment would
revise the technical specification Limited Safety System Setting for
the MINIMUM CRITICAL POWER RATIO (MCPR) for dual recirculation loop
operation and for single recirculation loop operation.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. Does the proposed license amendment involve a significant
increase in the probability or consequences of an accident
previously evaluated?
The purpose of the Safety Limit Minimum Critical Power Ratio
(SLMCPR) is to provide statistical confidence that less than 0.1% of
the fuel rods in a core would experience transition boiling during
the most limiting analyzed Anticipated Operational Occurrence
(transient). While transition boiling in a BWR does not in and of
itself signal the onset of fuel cladding failure, this criterion has
been selected as a conservative and convenient parameter for the
evaluation of fuel designs. Therefore, while this safety limit does
not provide any control over either the probability or consequences
of any accident previously evaluated, it does ensure that evaluated
transients remain within NRC-approved criteria. Revision of the
SLMCPR will establish in the CNS Technical Specifications a valid
limit, based on the NRC approved GESTAR II methodology using cycle-
specific inputs. This change will result in the input of more
restrictive core operating limits into the plant process computer,
ensuring that CNS will be operated within the constraints of the new
SLMCPR limits of 1.07 for dual recirculation loop operation, and
1.08 for single recirculation loop operation. No plant hardware
modifications are associated with this change. Therefore, since this
proposed change will not change the physical configuration of the
plant, nor result in operational changes which invalidate
assumptions used in any CNS accident analysis, this change does not
involve an increase in the probability or consequences of any
accident previously evaluated.
2. Does the proposed License Amendment create the possibility of
a new or different kind of accident from any accident previously
evaluated?
This change revises the SLMCPR values in the CNS Technical
Specifications in accordance with a cycle specific analysis
performed for the remainder of the current cycle. The SLMCPR ensures
that less than 0.1% of the fuel rods in a core would experience
transition boiling during the most limiting Anticipated Operational
Occurrence. Increasing the SLMCPR from 1.06 to 1.07 for dual
recirculation loop operation and from 1.07 to 1.08 for single
recirculation loop operation will ensure that the specified
statistical confidence will be met for all analyzed transients. This
change does not involve any plant hardware changes. The only
operational changes will be the institution of appropriate thermal
restrictions on reactor core operation in accordance with the SLMCPR
changes. Therefore, this proposed change will not create the
possibility of a new or different kind of accident from any
previously evaluated.
3. Does the proposed change create a significant reduction in
the margin of safety?
This change will establish in the CNS Technical Specifications,
SLMCPR values that ensure the margin of safety to the NRC approved
Anticipated Operational Occurrence evaluation acceptance criteria
will be met. Increasing the SLMCPR institutes more restrictive
thermal limitations on core operation. The change of the SLMCPR from
1.06 to 1.07 for dual recirculation loop operation, and from 1.07 to
1.08 for single loop operation will ensure that the acceptance
criteria for evaluated transients will continue to be met, and that
the appropriate limit is reflected in the CNS Technical
Specifications. Therefore, this proposed change does not create a
reduction in the margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: Auburn Memorial Library, 1810
Courthouse Avenue, Auburn, NE 68305
Attorney for licensee: Mr. John R. McPhail, Nebraska Public Power
District, Post Office Box 499, Columbus, NE 68602-0499
NRC Project Director: William D. Beckner
Niagara Mohawk Power Corporation, Docket No. 50-410, Nine Mile
Point Nuclear Station, Unit 2, Oswego County, New York
Date of amendment request: May 15, 1996
Description of amendment request: The proposed amendment would
revise Technical Specification (TS) 3/4.3.2, ``Isolation Actuation
Instrumentation,'' to establish a range of allowable and trip setpoints
for high temperature (varying as a function of ambient temperature) in
the Main Steam Line Tunnel Lead Enclosure Area. Specifically, a new TS
Figure 3.3.2-1 would be added to provide a curve of allowable
temperature values and a curve of trip temperature setpoints, both
plotted over a range of ambient temperatures. The new Figure would be
referenced by Table 3.3.2-2 at item 1.d.3 (High Temperature Main Steam
Line Tunnel Lead Enclosure Trip Function) by a new footnote stating:
The trip setpoint and allowable value for a channel may be
established based on Figure 3.3.2-1, if:
a. The actual ambient temperature readings for all operable
channels in the Lead Enclosure Area are equal to or greater than the
ambient temperature used as the basis for the setpoint, and
b. The absence of steam leaks in the Main Steam Line Tunnel Lead
Enclosure Area is verified by visual inspection prior to increasing
a channel setpoint, and
c. A surveillance is implemented in accordance with Note (d) of
Table 4.3.2.1-1.
Similarly, TS Surveillance Table 4.3.2.1-1 would be supplemented at
item 1.d.3 (High Temperature Main Steam Line Tunnel Lead Enclosure)
with a new footnote stating:
(d) In addition to the normal shift channel check, if a channel
setpoint has been established using Figure 3.3.2-1, then once per
shift, the actual ambient temperature reading for all operable
channels in the Lead Enclosure Area shall be verified to be equal to
or greater than the ambient temperature used as the basis for the
setpoint.
Basis for proposed no significant hazards consideration
determination: The main steam tunnel high temperature isolation
actuation instrumentation is part of the Leak Detection System (LDS).
It is used to detect leakage early at 25 gallons per minute (gpm) and
initiate signals to automatically close the Main Steam Isolation Valves
before a pipe break could occur. The existing temperature setpoints for
the tunnel lead enclosure are based upon transient analyses for steam
leaks in the steam tunnel utilizing
[[Page 34894]]
winter temperatures as an initial condition. The licensee finds that a
change is needed because actual temperatures in the tunnel, especially
during the summer, are approaching the setpoints when steam leakage is
not occurring. Under the present conditions, a minor disturbance in the
turbine building ventilation system could cause an unwarranted
isolation actuation at full power with resulting Main Steam Isolation
Valve closure and reactor scram.
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 operation of NMP2 [Nine Mile Point Unit 2] in accordance
with the proposed amendment will not involve a significant increase
in the probability or consequences of an accident previously
evaluated.
The LDS instrumentation in the main steam line tunnel isolates
the Main Steam Isolation Valves upon sensing a steam leak of 25 gpm.
For an elevated ambient temperature in the Lead Enclosure area, a
setpoint established using the proposed Figure 3.3.2-1 ensures that
the Main Steam Isolation Valves continue to receive an isolation
signal upon sensing a steam leak of 25 gpm. Verifying the absence of
any steam leak in the area prior to raising any temperature
instrument setpoint ensures that the ability to sense a 25 gpm leak
is not compromised by an increased ambient temperature resulting
from a smaller steam leak. The periodic surveillance to verify the
actual ambient temperature ensures the continued validity of the
ambient temperature used for the setpoint basis, and provides
sufficient advance indication to take appropriate compensatory
action. Accordingly, this change will not involve a significant
increase in the consequences of any accident previously evaluated.
Furthermore, the LDS function provides a mitigation action for a
postulated main steam line pipe leak which could lead to a pipe
break. This function does not affect any accident precursors, and
the proposed change does not affect the function of the LDS system.
Accordingly, this change will not involve a significant increase in
the probability of any accident previously evaluated.
2. The operation of NMP2 in accordance with the proposed
amendment will not create the possibility of a new or different kind
of accident from any previously evaluated.
The qualification of safety-related equipment in the main steam
lead enclosure is evaluated using actual temperatures and component
qualified life is adjusted accordingly. The temperature elements are
the only safety-related equipment affected by this change,
therefore, the instrumentation response to previously evaluated
accidents will not be adversely affected. This change will not
affect the performance of safety related structures. Accordingly,
the design capabilities of those structures, systems and components
affected by the proposed change are not challenged in a manner not
previously evaluated so as to create the possibility of a new or
different kind of accident from any previously evaluated.
3. The operation of NMP2 in accordance with the proposed
amendment will not involve a significant reduction in a margin of
safety.
The proposed change provides a range of setpoints and allowable
values for the Main Steam Line Tunnel Lead Enclosure temperatures.
The calculation of the allowable values and trip setpoints was
performed using the same methodologies as previously employed. For
an elevated ambient temperature in the Lead Enclosure area, a
setpoint established using the proposed Figure 3.3.2-1 ensures that
the Main Steam Isolation Valves receive an isolation signal upon
sensing a steam leak of 25 gpm, resulting in a main steam line
isolation prior to a pipe break. Therefore, the proposed change
provides the same level of protection against a main steam line
break as the existing setpoint values. The proposed setpoints will
provide increased scram avoidance, and thereby reduce unnecessary
challenges to the plant shutdown systems. Accordingly, the proposed
change does not result in a significant reduction in a margin of
safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: Reference and Documents
Department, Penfield Library, State University of New York, Oswego, New
York 13126
Attorney for licensee: Mark J. Wetterhahn, Esquire, Winston &
Strawn, 1400 L Street, NW., Washington, DC 20005-3502
NRC Project Director: Jocelyn A. Mitchell, Acting Director
PECO Energy Company, Public Service Electric and Gas Company,
Delmarva Power and Light Company, and Atlantic City Electric
Company, Dockets Nos. 50-277 and 50-278, Peach Bottom Atomic Power
Station, Units Nos. 2 and 3, York County, Pennsylvania
Date of application for amendments: March 25, 1996
Description of amendment request: These amendments revise the
safety limit minimum critcal power ratios (SLMCPRs) to support use of
GE-13 fuel at Peach Bottom Atomic Power Station.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1)The proposed TS [technical specification] changes do not
involve a significant increase in the probability or consequences of
an accident previously evaluated.
The derivation of the revised GE13 SLMCPRs for incorporation
into the TS, and its use to determine cycle-specific thermal limits,
have been performed using USNRC [U.S. Nuclear Regulatory
Commission]-approved methods within the existing fuel licensing
criteria as discussed in NEDE-32198P, ``GE13 Compliance With
Amendment 22 of NEDE-24011-P-A (GESTAR II),'' and cannot increase
the probability or severity of an accident.
The basis of the SLMCPRs calculation is to ensure that greater
than 99.9% of all fuel rods in the core avoid boiling transition if
the limit is not violated. The new SLMCPRs preserve the existing
margin to transition boiling and fuel damage in the event of a
postulated accident. The fuel licensing acceptance criteria for the
SLMCPRs calculation apply to the GE13 fuel in the same manner that
they have applied to previous fuel designs. The probability of fuel
damage is not increased. Therefore, the proposed TS changes do not
involve an increase in the probability or consequences of an
accident previously evaluated.
2) The proposed TS changes do not create the possibility of a
new or different kind of accident from any accident previously
evaluated.
The SLMCPR for the GE13 fuel design is a Technical Specification
numerical value, designed to ensure that transition boiling does not
occur in 99.9% of all fuel rods in the core during the limiting
postulated accident. It cannot create the possibility of any new
type of accident. The new SLMCPRs are calculated using USNRC-
approved methods and have the same calculational basis as the SLMCPR
for other GE fuel designs previously used at PBAPS, Units 2 and 3.
Therefore, the proposed TS changes do not create the possibility of
a new or different kind of accident, from any accident previously
evaluated.
3) The proposed TS changes do not involve a significant
reduction in a margin of safety.
The margin of safety as defined in the TS Bases will remain the
same. The new SLMCPRs are calculated using USNRC-approved methods
which are in accordance with the current fuel licensing criteria.
The SLMCPRs for the GE13 fuel remain high enough to ensure that
greater than 99.9% of all fuel rods in the core will avoid boiling
transition if the limit is not violated, thereby preserving the fuel
cladding integrity. Therefore, the proposed TS changes do not
involve a reduction in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: Government Publications
[[Page 34895]]
Section, State Library of Pennsylvania, (REGIONAL DEPOSITORY) Education
Building, Walnut Street and Commonwealth Avenue, Box 1601, Harrisburg,
Pennsylvania 17105
Attorney for licensee: J. W. Durham, Sr., Esquire, Sr. V. P. and
General Counsel, PECO Energy Company, 2301 Market Street, Philadelphia,
Pennsylvania 19101
NRC Project Director: John F. Stolz
PECO Energy Company, Public Service Electric and Gas Company,
Delmarva Power and Light Company, and Atlantic City Electric
Company, Docket No. 50-277, Peach Bottom Atomic Power Station, Unit
No. 2, York County, Pennsylvania
Date of application for amendment: June 13, 1996
Description of amendment request: The proposed amendment to the
Technical Specifications (TS) will permit a one time performance of
Surveillance Requirement 3.3.1.1.12, for the Average Power Range
Monitor Flow Biased High Scram function, with a delayed entry into its
associated TS Conditions and Required Actions for up to 6 hours
provided core flow is maintained at or above 82 percent. This change
would be in effect until the end of refueling outage 2R11, currently
scheduled for early October 1996.
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:
i) The proposed change does not involve a significant increase
in the probability or consequences of an accident previously
evaluated.
The APRM system provides monitoring and accident mitigation
functions to limit peak flux in the core during startup and run
modes. This proposed TS change for delaying entry into Conditions
and Required Actions associated with SR 3.3.1.1.12 for the APRM flow
bias function will have no impact on the APRM system or any system
that interfaces with it. No pressure boundary interfaces or process
control parameters will be challenged.
This change does not affect the operation of any equipment.
Delaying entry into Conditions and Required Actions associated with
SR 3.3.1.1.12 does not affect either the initiator of any accident
previously evaluated or any equipment required to mitigate the
consequences of an accident, or the isotopic inventory in the fuel.
Thus, the change does not increase either the probability or the
consequences of accidents previously evaluated.
ii) The proposed change does not create the possibility of a new
or different kind of accident from any previously evaluated.
Because there is no direct pressure boundary interface or
process control function associated with the APRM system or its
interfacing electronics, the possibility of a new or different type
of accident than any previously evaluated will not be created.
Although the flow bias instrument loop does employ flow transmitters
to measure recirculation drive flow, delaying entry into Conditions
and Required Actions associated with SR 3.3.1.1.12 will have no
impact on their pressure boundary function. Also, failure of the
sensing line associated with these transmitters has already been
accounted for in the initial plant design by including excess flow
check valves for sensing line break isolation.
The proposed change does not introduce a new mode of plant
operation and does not involve the installation of any new equipment
or modifications to the plant. Therefore, it does not create the
possibility of a new or different kind of accident from any accident
previously evaluated.
iii)The proposed change does not involve a significant reduction
in a margin of safety.
The APRM flow biased high scram function is not specifically
credited in the safety analysis. However, it is intended to provide
an additional margin of protection from transient induced fuel
damage during operation where recirculation flow is reduced to below
the minimum required for rated power operation.
The margin of safety associated with this change refers to the
margin inherent in the accident analyses that takes credit for the
clamped high flux scram only (i.e., margin between scramming at 120%
peak flux and the peak flux necessary for fuel damage). The current
reactor operating state (end of cycle coast down extended core flow)
dictates that only the 120% flux trip be enforced. This trip remains
functional during the APRM flow biased high scram calibration.
Currently, the Conditions and Required Actions associated with
SR 3.3.1.1.12 permit a one hour delay prior to entry because it
minimizes risk while allowing time for restoration or tripping of
channels by operations personnel. Because the APRM flow biased
function is not enforced during end of cycle, coast down, extended
core flow conditions, extending entry in associated Conditions and
Required Actions from one to six hours has no impact on the margin
associated with the clamped high flux scram. In the event core flow
drops below 82%, the flow point below which APRM setpoints
automatically become flow biased, the associated Conditions and
Required Actions will be entered.
Therefore, extending entry into associated Conditions and
Required Actions associated with SR 3.3.1.1.12, provided core flow
remains at or above 82%, from one to six hours does not reduce any
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, (REGIONAL DEPOSITORY) Education
Building, Walnut Street and Commonwealth Avenue, Box 1601, Harrisburg,
PA 17105
Attorney for licensee: J. W. Durham, Sr., Esquire, Sr. V. P. and
General Counsel, PECO Energy Company, 2301 Market Street, Philadelphia,
PA 19101
NRC Project Director: John F. Stolz
Power Authority of the State of New York, Docket No. 50-333, James
A. FitzPatrick Nuclear Power Plant, Oswego County, New York
Date of amendment request: May 16, 1996
Description of amendment request: The proposed amendment to the
James A. FitzPatrick Technical Specifications (TSs) proposes to delete
the requirement for the Plant Operating Review Committee (PORC) to
review the fire protection program and implementing procedures. This
proposal will reduce the administrative burden on the committee while
making PORC's responsibilities more consistent with the other
responsibilities described in Section 6.1.5.6 of the TS.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
Operation of the FitzPatrick plant in accordance with the
proposed amendment would not involve a significant hazards
consideration as defined in 10 CFR 50.92, since it would not:
1. Involve a significant increase in the probability or
consequences of an accident previously evaluated.
The proposed changes delete the Plant Operating Review Committee
(PORC) review of the fire protection program and implementing
procedures, and deleted fire protection inspection and audit
requirements that are redundant to those performed under the
cognizance of the Safety Review Committee (SRC). The changes do not
introduce any new modes of plant operation, make any physical
changes, or alter any operational setpoints. Therefore, the changes
do not degrade the performance of any safety system assumed to
function in the accident analysis. Consequently, there is no effect
on the probability or consequences of an accident.
2. Create the possibility of a new or different kind of accident
from those previously evaluated.
No physical changes to the plant or changes to equipment
operating procedures are proposed. The changes are administrative
and will not have any direct affect on equipment important to
safety. Therefore the changes cannot create the possibility of a new
or different kind of accident.
[[Page 34896]]
3. Involve a significant reduction in the margin of safety.
Adequacy of the fire protection program and implementing
procedures is assured by the fire protection license condition, the
procedure review and approval process implemented by Amendment 222,
the provisions of 10 CFR 50.59, and inspections and audits performed
under the cognizance of the SRC. Consequently, deleting PORC's
responsibility for review of the fire protection program and
implementing procedures, and deleting the inspection and audit
requirements contained in Specification 6.14.A and 6.14.B will not
degrade the fire protection program. Therefore, the proposed changes
do not involve a significant reduction in the margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: Reference and Documents
Department, Penfield Library, State University of New York, Oswego, New
York 13126
Attorney for licensee: Mr. Charles M. Pratt, 1633 Broadway, New
York, New York 10019
NRC Project Director: Jocelyn A. Mitchell, Acting Director
Power Authority of the State of New York, Docket No. 50-333, James
A. FitzPatrick Nuclear Power Plant, Oswego County, New York
Date of amendment request: May 30, 1996
Description of amendment request: The proposed amendment would
revise Minimum Critical Power Ratio Safety Limit and associated basis.
The changes are required to support introduction of General Electric
Company supplied, GE12, 10x10 fuel into the Cycle 13 reactor core.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
Operation of the FitzPatrick plant in accordance with the
proposed Amendment would not involve a significant hazards
consideration as defined in 10 CFR 50.92, since it would not:
1. Involve a significant increase in the probability or
consequences of an accident previously evaluated because:
A change in the SLMCPR [Safety Limit Minimum Critical Power
Ratio] does not affect initiation of any accident. Operation in
accordance with the revised SLMCPR ensures the consequences of
previously analyzed accidents are not changed.
2. Create the possibility of a new or different kind of accident
from any accident previously evaluated because:
The SLMCPR establishes a performance limit for the fuel.
Therefore changing the limit will not initiate any accident.
3. Involve a significant margin of safety because:
The analyses performed to determine the revised SLMCPR assure
maintenance of the same margin of safety as presently exists for the
prevention of onset of transition boiling.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: Reference and Documents
Department, Penfield Library, State University of New York, Oswego, New
York 13126
Attorney for licensee: Mr. Charles M. Pratt, 1633 Broadway, New
York, New York 10019
NRC Project Director: Jocelyn A. Mitchell, Acting Director
Power Authority of the State of New York, Docket No. 50-333, James
A. FitzPatrick Nuclear Power Plant, Oswego County, New York
Date of amendment request: May 30, 1996
Description of amendment request: The proposed amendment would
revise Anticipated Transient Without Scram (ATWS) Recirculation Pump
Trip Reactor Pressure - High setpoint when either zero or one Safety
Relief Valves are out-of service.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
Operation of the FitzPatrick plant in accordance with the
proposed Amendment would not involve a significant hazards
consideration as defined in 10 CFR 50.92, since it would not:
1. Involve a significant increase in the probability or
consequences of an accident previously evaluated because:
A change in the ATWS Recirculation Pump Trip Reactor Pressure -
High setpoint does not affect initiation of any accident. Operation
in accordance with the revised setpoints ensures the consequences of
previously analyzed accidents are not changed.
2. Create the possibility of a new or different kind of accident
from any accident previously evaluated because:
RPV [reactor pressure vessel] pressure following an ATWS with
MSIV [main steam isolation valve] closure event (worst case
transient for RPV pressurization) remains within acceptable limits
with the revised setpoint. Therefore changing the setpoint will not
lead to a new type of accident.
3. Involve a significant reduction in a margin of safety
because:
The analyses performed to determine the revised ATWS
Recirculation Pump Trip Reactor Pressure - High setpoint assure
maintenance of the same margin of safety as presently exists for
limiting RPV pressure following an ATWS with MSIV closure (limiting
transient).
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: Reference and Documents
Department, Penfield Library, State University of New York, Oswego, New
York 13126
Attorney for licensee: Mr. Charles M. Pratt, 1633 Broadway, New
York, New York 10019
NRC Project Director: Jocelyn A. Mitchell, Acting Director
Power Authority of the State of New York, Docket No. 50-333, James
A. FitzPatrick Nuclear Power Plant, Oswego County, New York
Date of amendment request: May 30, 1996
Description of amendment request: The proposed amendment would
eliminate selected response time testing requirements. The affected
Technical Specifications (TS) are TS 4.1.A, ``Surveillance
Requirements, Reactor Protection System,'' and TS 4.2.A, ``Surveillance
Requirements, Instrumentation, Primary Containment Isolation
Functions.''
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
Operation of the FitzPatrick plant in accordance with the
proposed Amendment would not involve a significant hazards
consideration as defined in 10 CFR 50.92, since it would not:
1. Involve a significant increase in the probability or
consequences of an accident previously evaluated because:
The purpose of the proposed TS change is to eliminate response
time testing requirements for selected sensors in the RPS [reactor
protection system] and Primary Containment Isolation System. The
BWROG [Boiling Water Reactor Owners Group] has completed an
evaluation which demonstrates that response time testing is
redundant to the other TS required testing. These other tests
[[Page 34897]]
in conjunction with actions taken in response to NRC Bulletin 90-01,
``Loss of Fill-Oil in Transmitters
Manufactured by Rosemount,'' and Supplement 1 to Bulletin 90-01,
are sufficient to identify failure modes or degradation in
instrument response time and ensure operation of the associated
systems within acceptable limits. Furthermore, failure modes
detected by response time testing are detectable by other TS
required testing. This evaluation was documented in Reference 1 [See
application dated May 30, 1996]. NYPA [New York Power Authority] has
confirmed the applicability of this evaluation to the FitzPatrick
Plant. In addition, NYPA will complete the actions identified in the
NRC staff's safety evaluation of NEDO-32291-A.
Because of the continued application of other existing TS
required tests such as channel calibrations, channel checks, channel
functional tests, and logic system functional tests, the response
time of these systems will be maintained within the acceptance
limits assumed in plant safety analyses and required for successful
mitigation of an initiating event. The proposed changes do not
affect the capability of the associated systems to perform their
intended function within their required response time, nor do the
proposed changes themselves affect the operation of any equipment.
Therefore, the proposed change does not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
2. Create the possibility of a new or different kind of accident
from those previously evaluated because:
The proposed changes do not affect the ability of the systems to
perform their intended function within the acceptance limits assumed
in plant safety analyses and required for successful mitigation of
an initiating event. No new failure modes are introduced by the
changes. Therefore, the proposed changes do not create the
possibility of a new or different kind of accident from any accident
previously evaluated.
3. Involve a significant reduction in the margin of safety.
The current TS required response time test limits are based on
the maximum allowable values assumed in the plant safety analyses.
These analyses conservatively establish the margin of safety. As
described above, the proposed changes do not affect the capability
of the associated systems to perform their intended function within
the allowed response time used as the basis for the plant safety
analysis. Plant and system response to an initiating event will
remain in compliance within the assumptions of the safety analyses,
and therefore the margin of safety is not affected.
Further, although not explicitly evaluated, the proposed changes
will provide an improvement to plant safety and operation by
reducing the time safety systems are unavailable, reducing safety
systems actuations, reducing plant shutdown risk, limiting radiation
exposure to plant personnel, and eliminating the diversion of key
personnel to conduct unnecessary testing. Therefore, the overall
effect of the changes should increase the margin the safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: Reference and Documents
Department, Penfield Library, State University of New York, Oswego, New
York 13126
Attorney for licensee: Mr. Charles M. Pratt, 1633 Broadway, New
York, New York 10019
NRC Project Director: Jocelyn A. Mitchell, Acting Director
Public Service Electric & Gas Company, Docket No. 50-354, Hope
Creek Generating Station, Salem County, New Jersey
Date of amendment request: March 6, 1996, as supplemented by letter
dated May 30, 1996
Description of amendment request: The proposed change to Hope Creek
Technical Specification (TS) 3.8.1, ``A.C. Sources - Operating'', would
decrease the minimum fuel oil storage capacity of the Emergency Diesel
Generator Fuel Oil Storage Tanks, from 48,800 to 44,800 gallons. In
addition, footnote ** is deleted from TS 3.8.1.1.b.2. The proposed
change would also add an Action Statement to address remedial action
when a fuel oil transfer pump becomes inoperable.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. Will not involve a significant increase in the probability or
consequences of an accident previously evaluated.
TANK LEVEL
Amendment 59 provides an allowance for transferring fuel oil
from a pair of storage tanks associated with an inoperable
[Emergency Diesel Generator] EDG to another pair of storage tanks in
order to demonstrate compliance with PSE&G's commitment to
Regulatory Guide 1.137. The proposed change is consistent with that
transfer strategy and extends this allowance to include using fuel
oil in operable EDG storage tanks in order to reduce the amount of
stored fuel oil. Transfer from operable EDG storage tanks is,
actually, less complex than transferring from an inoperable EDG
storage tank since power to the transfer pumps would be available.
The low level alarm setpoint is the only physical change to be
made. No change is being made to the EDGs, to the fuel oil storage
tanks, or to the fuel oil transfer system and since EDG fuel oil
supply is associated with mitigating the consequences of an
accident, there is no change in the probability of any accident
analyzed in the [Updated Final Safety Analysis Report] UFSAR.
Since the proposed change still ensures the minimum fuel oil
storage capacity meets the existing licensing basis and since off-
site replacement oil is expected to be available within 60 hours
there is no change in the consequences of an accident previously
evaluated.
TRANSFER PUMP ACTION STATEMENT
Since no change is being made to the EDGs, to the fuel oil
storage tanks or to the fuel oil transfer system, and since EDG fuel
oil supply is associated with mitigating the consequences of an
accident, there is no change in the probability of any accident
analyzed in the UFSAR.
The proposed change provides compensatory action in the event a
single fuel oil transfer pump is inoperable without having to
immediately declare the EDG inoperable. The change ensures the
affected EDG remains fully capable of functioning as assumed in the
safety analyses, therefore, there is no significant impact on the
consequences of an accident previously evaluated.
Therefore, the proposed changes will not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
2. Will not create the possibility of a new or different kind of
accident from any previously evaluated.
TANK LEVEL AND TRANSFER PUMP ACTION STATEMENT
The proposed changes will result in a setpoint change to the low
level alarm. No other physical changes to the EDGs, to the fuel oil
storage tanks, or to the fuel oil transfer system will result from
the proposed changes. Operation including the proposed changes will
not impair the diesel generators from performing as provided in the
design basis. In addition, EDG fuel oil supply is associated with
mitigating accident consequences, not accident prevention.
Therefore, the proposed change will not create the possibility of a
new or different kind of accident from any previously evaluated.
3. Will not involve significant reduction in a margin of safety.
TANK LEVEL
The margin of safety is provided by the on-site storage of an
adequate supply of diesel fuel oil to ensure uninterrupted EDG
operation for seven days. Although the proposed change may result in
a reduction of stored fuel oil, the new minimum continues to provide
for an on-site seven day supply of diesel fuel oil.
TRANSFER PUMP ACTION STATEMENT
The margin of safety is provided by the ability of the fuel oil
transfer pumps to supply an adequate flow of the stored fuel to each
EDG day tank. The proposed change continues to provide 100% capacity
to the EDG day tank for a minimum of three days with no operator
action. With the proposed action, adequate transfer capability is
[[Page 34898]]
provided for a minimum of seven days fuel oil supply at which time
refilling of the tanks would provide an indefinite supply. With both
transfer pumps on a single EDG inoperable, the remaining three EDGs
would provide adequate power for safe shutdown. Transfer of fuel oil
from the storage tanks with inoperable transfer pumps can still be
effected using temporary hoses.
Since the proposed changes do not involve the addition of plant
equipment, are consistent with the intent of the existing Technical
Specifications, are consistent with allowances for fuel oil
transfers approved in Amendment 59, meets the intent of Regulatory
Guide 1.137, and are consistent with the design basis of the diesel
generators and the accident analysis, no action proposed by this
request will occur that will 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: Pennsville Public Library, 190
S. Broadway, Pennsville, New Jersey 08070
Attorney for licensee: M. J. Wetterhahn, Esquire, Winston and
Strawn, 1400 L Street, NW., Washington, DC 20005-3502
NRC Project Director: John F. Stolz
Public Service Electric & Gas Company, Docket Nos. 50-272 and 50-
311, Salem Nuclear Generating Station, Unit Nos. 1 and 2, Salem
County, New Jersey
Date of amendment request: May 10, 1996
Description of amendment request: The proposed amendments would
change Technical Specification Sections, 1.0, 2.0, 3/4 1.0, 3/4 2.0,
5.0 and 6.0. These changes support the Margin Recovery Program (MRP)
and support increased steam generator tube plugging, improved fuel
reliability, reduced fuel costs, longer fuel cycles, reduced spent fuel
storage, and enhanced reactor safety. These changes incorporate the
results of the revised safety analyses (margin recovery) and the
establishment of a Core Operating Limits 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:
1. The proposed change does not involve a significant increase
in the probability or consequences of an accident previously
evaluated.
The accidents potentially affected by the parameters and
assumptions associated with the MRP have been evaluated/ analyzed
and all design standards and applicable safety criteria are met. The
consideration of these changes does not result in a situation where
the design, material, or construction standards that were applicable
prior to the change have been altered. Therefore, the changes
occurring with the MRP will not result in any additional challenges
to plant equipment that could increase the probability of any
previously evaluated accident.
The changes associated with the MRP do not affect plant systems
such that their function in the control of radiological consequences
is adversely affected. The safety evaluation documents that the
design standards and applicable safety criteria limits continue to
be met and therefore fission barrier integrity is not challenged.
The MRP changes have been shown not to adversely affect the response
of the plant to postulated accident scenarios. In all cases, the
calculated doses are within the regulatory criteria and therefore do
not constitute an increase in consequences. These changes will,
therefore, not affect the mitigation of the radiological
consequences of any accident described in the Updated Final Safety
Analysis Report (UFSAR).
Based on the above, it is concluded that the probability or
consequences of an accident previously evaluated is not
significantly increased by the proposed changes.
2. The proposed change does not create the possibility of a new
or different kind of accident from any accident previously
evaluated.
The possibility for a new or difference[t] type of accident from
any accident previously evaluated is not created since the changes
associated with the MRP do not result in a change to the design
basis of any plant component or system. The evaluation of the
effects of the MRP changes shows that all design standards and
applicable safety criteria limits are met. These changes therefore
do not cause the initiation of a new accident nor create any new
failure mechanisms. Component integrity is not challenged. The
changes do not result in any event previously deemed incredible
being made credible. The MRP changes will not result in more adverse
conditions and will not result in any increase in the challenges to
safety systems.
Therefore, the consideration of the MRP as described in the
safety evaluation does not create the possibility of a new or
different type of accident from any accident previously evaluated.
3. The proposed change does not involve a significant reduction
in a margin of safety.
The margin of safety is maintained by assuring compliance with
acceptance limits reviewed and approved by the NRC. Since all of the
appropriate acceptance criteria for the various analyses and
evaluations have been met, by definition there has not been a
reduction in any margin of safety.
Therefore, the margin of safety as defined in the Bases to the
Salem Unit 1 and 2 Technical Specifications has not been
significantly reduced.
Based on the above, PSE&G has determined that the proposed
changes do not involve a significant hazards consideration.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: Salem Free Public library, 112
West Broadway, Salem, New Jersey 08079
Attorney for licensee: Mark J. Wetterhahn, Esquire, Winston and
Strawn, 1400 L Street, NW, Washington, DC 20005-3502
NRC Project Director: John F. Stolz
South Carolina Electric & Gas Company (SCE&G), South Carolina
Public Service Authority, Docket No. 50-395, Virgil C. Summer
Nuclear Station, Unit No. 1, Fairfield County, South Carolina
Date of amendment request: April 16, 1996
Description of amendment request: The proposed amendment would
revise the Virgil C. Summer Nuclear Station, Unit 1 (VCSNS), Technical
Specifications (TS) to implement the amended regulation to 10 CFR Part
50, Appendix J, Option B (new rule), to provide a performance-based
option for leakage-rate testing of containment. The proposed amendment
will revise the VCSNS TS 3/4.6 ``Containment Systems,'' TS Bases 3/4.6,
and TS 6.8 ``Administrative Controls - Programs and Procedures,'' to
adopt the implementation requirements of 10 CFR Part 50, Appendix J,
Option B. The proposed amendment utilizes the guidelines (guidelines)
provided in ``Option B'' of Regulatory Guide (RG) 1.163 ``Performance-
Based Containment Leak-Test Program, September 1995,'' and NEI 94-01,
``Industry Guideline for Implementing Performance-Based Option of 10
CFR 50, Appendix J, July 26, 1995.'' The licensee has stated that the
proposed amendment is within these prescribed guidelines and does not
propose any deviations to the established methods which would impact
already approved analyses/justifications and established review
process.
The proposed change will remove the prescriptive TS requirements
for the performance of containment leakage testing and allow leakage
testing to be conducted as determined appropriate through the
performance-based or risk-based alternatives described in the VCSNS
Containment Leakage Rate Testing Program developed in accordance with
RG 1.163 and NEI 94-01. Since the requirements of Appendix J to 10 CFR
Part 50 will continue to
[[Page 34899]]
apply, the type of testing will not change. The proposed request does
not modify any plant equipment or systems.
The requirements of Appendix J will continue to govern the type of
test, testing methodology, and acceptance criteria for Type A, B, and C
testing. The performance-based testing of Option B eliminates or
modifies prescriptive regulatory requirements for which the burden is
marginal to safety for which the reviews and analyses have been
presented in NUREG-1493, ``Performance-Based Containment Leak-Test
Program, Final Report, September 1995.''
Earlier leakage testing performed at VCSNS has demonstrated low
overall containment leakage and supports the implementation of Option
B.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. The probability or consequences of an accident previously
evaluated is not significantly increased.
There is no increase in the probability of an accident since
there is no work that would affect containment integrity. The
testing of containment isolation valves (CIVs) and other containment
penetration sealing devices is not postulated as an accident
precursor or initiating event.
Type A testing is capable of determining the total leakage from
both local leakage paths and gross containment leakage paths. Our
Type B and C testing has consistently provided accurate leakage
rates for valves and penetrations.
Administrative controls govern maintenance and testing such that
there is very low probability that unacceptable maintenance or
alignments can occur. Prior to and following maintenance on CIVs and
penetrations, a local leak rate test (LLRT) is required to be
performed. As a result, Type A testing is not required to accurately
quantify the leakage through containment penetrations.
Any specific exemptions to the requirements of Appendix J will
require approval by the NRC before implementation.
Therefore, this proposed change does not involve a significant
increase in the possibility or consequences of an accident
previously evaluated.
2. The possibility of an accident or a malfunction of a
different type than any previously evaluated is not created.
The proposed request does not involve any physical changes to
the plant, affect the operation of the plant, or change testing
methods or acceptance criteria. The history of containment testing
verifies that containment integrity has been maintained.
The frequency changes allowed by implementation of Option B will
not significantly decrease the level of confidence in the ability of
the reactor building to limit offsite doses to allowable values. No
accident or malfunction can be the result of the allowed changes to
test schedule or frequency.
Since the proposed request will not directly impact equipment,
procedures or operations, the changes will not create the
possibility of any new or different kind of accident from any
previously evaluated.
3. The margin of safety has not been significantly reduced.
The reason for performing containment leakage rate testing is to
assure that the leakage paths are identified, and that any accident
release will be restricted to those paths assumed in the safety
analysis. The purpose for the schedule is to assure that containment
integrity is verified on a periodic basis.
Implementation of Option B to provide flexibility in the
scheduled requirements does not mean that containment integrity will
be compromised. The historical leakage rate test results for VCSNS
and for the nuclear industry support extension of testing
frequencies and demonstrate that structural integrity has been
maintained.
Therefore, the margin of safety has not been significantly
reduced.
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: Fairfield County Library, 300
Washington Street, Winnsboro, SC 29180
Attorney for licensee: Randolph R. Mahan, South Carolina Electric &
Gas Company, Post Office Box 764, Columbia, SC 29218
NRC Project Director: Eugene V. Imbro
Southern Nuclear Operating Company, Inc., Docket No. 50-364, Joseph
M. Farley Nuclear Plant, Unit 2, Houston County, Alabama
Date of amendment request: April 22, 1996
Description of amendment request: The amendment would revise the
Technical Specifications to implement the L* Tubesheet Region Plugging
Criterion, which would allow a steam generator tube to remain in
service with bands of axial degradation in the tubesheet region
provided sufficient non-degraded tubing remains to satisfy regulatory
guidance concerning structural and leakage integrity.
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 Farley Nuclear Plant Unit 2 steam generators
in accordance with the proposed license amendment does not involve a
significant increase in the probability or consequences of an
accident previously evaluated.
The supporting technical evaluations of the subject criteria
demonstrate that the presence of the tubesheet enhances the tube
integrity in the region of the hardroll by precluding tube
deformation beyond its initial expanded outside diameter. The
resistance to both tube rupture and tube collapse is strengthened by
the presence of the tubesheet in that region. The result of the
hardroll of the tube into the tubesheet is an interference fit
between the tube and the tubesheet. Tube rupture [cannot] occur
because the contact between the tube and tubesheet does not permit
sufficient movement of tube material. In a similar manner, the
tubesheet does not permit sufficient movement of tube material to
permit buckling collapse of the tube during postulated LOCA [loss-
of-coolant accident] loadings.
The type of degradation for which the L* criterion has been
developed (cracking with an axial or near axial orientation) has
been found not to significantly reduce the axial strength of a tube.
An evaluation including analysis and testing has been done to
determine the strength reduction for axial loads with simulated
axial and near axial cracks. This evaluation provides the basis for
the acceptance criteria for tube degradation subject to the L*
criterion.
The SRE [sound roll expansion] L* length is sufficient to
preclude significant leakage from tube degradation located below the
L* length. The existing Technical Specification leak rate
requirements and accident analysis assumptions remain unchanged in
the unlikely event that significant leakage from this region does
occur. Any leakage from the tube within the tubesheet at any
elevation in the tubesheet is fully bounded by the existing steam
generator tube rupture analysis included in the Farley Nuclear Plant
Final Safety Analysis Report. A conservative leakage allowance for
each L* tube is provided to determine the impact of L* criterion
upon offsite doses in the event of a postulated double ended
guillotine break of the main steam line outside of containment, but
upstream of the main steam line isolation valves. Since Farley Unit
2 has implemented the Interim Plugging Criteria (IPC) for ODSCC at
the tube support plates, projected steam line break (SLB) leakage at
the end of the next successive operating cycle must be evaluated.
Per Generic Letter 95-05, plants implementing the IPC can utilize
SLB leakage limits higher than the originally assumed 1.0 gpm
primary to secondary leakage value provided an analysis of offsite
doses consistent with the Standard Review Plan methodology is
performed. This analysis performed for the Farley Unit 2 plant
indicates that primary to secondary leakage of 11.2 gpm in the
faulted loop (0.1 gpm in the intact loops) will result in offsite
doses at the site boundary of less than 10% of the 10 CFR [Part] 100
guidelines. The total projected SLB leakage from all leakage sources
must remain below this value. Per attachment 4 addressing the L*
methodology,
[[Page 34900]]
the number of tube ends to which L* criterion can be applied is
limited to 600 per steam generator. Using a bounding SLB leakage
allowance per L* tube, the SLB leakage component from 600 L* tube
ends will be less than 0.33 gpm in the faulted loop. The proposed
alternate plugging criterion does not adversely impact any other
previously evaluated design basis accident. As the current Unit 2
IPC SLB leakage has been calculated to be less than 2 gpm in the
faulted loop, [an] SLB leakage margin of over 9 gpm is provided for
this cycle.
As noted above, tube rupture and pullout is not expected for
tubes using the L* criterion. In addition to the L* length, a
minimum length of SRE below the identified degradation must be
established. The aggregate L* distance of SRE provides the
structural integrity to prevent tube pullout. Conservatively, it is
assumed that the degraded band length does not provide any support
in resisting tube pullout.
Therefore SNC [Southern Nuclear Operating Company, Inc.]
concludes that Operation of the Farley Nuclear Plant Unit 2 steam
generators in accordance with the proposed license amendment does
not involve a significant increase in the probability or
consequences of an accident previously evaluated.
2. The proposed license amendment does not create the
possibility of a new or different kind of accident from any accident
previously evaluated.
Implementation of the proposed L* criterion does not introduce
any significant changes to the plant design basis. Use of the
criterion does not provide a mechanism to result in an accident
initiated outside of the region of the tubesheet expansion. The
structural integrity of L* tubes will be maintained during all plant
conditions. Any hypothetical accident as a result of any tube
degradation in the expanded portion of the tube would be bounded by
the existing tube rupture accident analysis. If it is postulated
that a circumferential separation of an L* tube were to occur below
the PLRL [pullout load reaction length], tube structural and leakage
integrity will be maintained during all plant conditions.
Verification of the L* distance of non-degraded tube roll expansion
prevents the postulated separated tube from lifting out of the
tubesheet during all plant conditions. Verification of the L*
criterion prevents tube displacement of any magnitude, and
therefore, postulated axial cracks existing a minimum of 0.5 inch
from either the bottom of the roll transition or top of tubesheet,
whichever is lower, from migrating out of the tubesheet.
Therefore, SNC concludes that the proposed license amendment
does not create the possibility of a new or different kind of
accident from any accident previously evaluated.
3. The proposed license amendment does not involve a significant
reduction in a margin of safety.
The use of the L* criterion has been concluded to maintain the
integrity of the tube bundle commensurate with the requirements of
draft Regulatory Guide 1.121 under normal and postulated accident
conditions. The safety factors used in the verification of the
strength of the degraded tube are consistent with the safety factors
in the ASME [American Society of Mechanical Engineers] Boiler and
Pressure Vessel Code used in steam generator design. The L* length
has been verified by testing to be greater than the length of roll
expansion required to preclude significant leakage during normal and
postulated accident conditions. The leak testing acceptance criteria
are based on the primary to secondary leakage limit in the Technical
Specifications and the leakage assumptions used in the FSAR accident
analyses. The L* distance provides for structural integrity during
all plant conditions.
Implementation of the L* criterion will decrease the number of
tubes which must be taken out of service with tube plugs or repaired
with sleeves. Both plugs and sleeves reduce the RCS [reactor coolant
system] flow margin, thus implementation of the L* criterion will
maintain the margin of flow that would otherwise be reduced in the
event of increased plugging or sleeving.
Therefore, SNC, concludes based on the above, it is concluded
that the proposed change does not result in a significant reduction
in a loss of margin with respect to plant safety as defined in the
Final Safety Analysis Report [FSAR] or the bases of the FNP [Farley
Nuclear Plant] 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: Houston-Love Memorial Library,
212 W. Burdeshaw Street, Post Office Box 1369, Dothan, Alabama 36302
Attorney for licensee: M. Stanford Blanton, Esq., Balch and
Bingham, Post Office Box 306, 1710 Sixth Avenue North, Birmingham,
Alabama 35201
NRC Project Director: Herbert N. Berkow
Union Electric Company, Docket No. 50-483, Callaway Plant, Unit 1,
Callaway County, Missouri
Date of application request: May 29, 1996
Description of amendment request: The application requests staff
review and approval of a modification to the facility, as described in
the safety analysis report, that involves an unreviewed safety
question. The modification will reduce the single failure trip
potential for the main feedwater control and bypass 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, which is presented below:
1. The proposed change does not involve a significant increase
in the probability or consequences of an accident previously
evaluated.
The Callaway safety analysis assumes the MFC&BVs [main feedwater
control and bypass valves] close during certain events in order to
terminate fluid inventory addition to faulted steam generators and
thereby preclude the diversion of auxiliary feedwater to the main
feedwater system. This feature is necessary because each feedwater
line at Callaway is equipped with only one MFIV [main feedwater
isolation valve]. It should be noted that the safety analysis simply
requires the valves to close and does not prescribe a mechanism for
accomplishing that action.
The following are accidents that credit feedwater isolation or
AFW [auxiliary feedwater] addition. There is no impact by the
proposed modification on the consequences of each accident.
Feedwater System Malfunctions That Result In An Increase
In Feedwater Flow
Inadvertent Opening Of A Steam Generator Relief or
Safety Valve
Steam System Piping Failure
Loss of Nonemergency AC Power to the Station Auxiliaries
Loss of Normal Feedwater Flow
Feedwater System Pipe Break
Decrease in Reactor Coolant Inventory
The modification will not change the radiological consequences
of FSAR [final safety analysis report] Chapter 15 accidents because
the feedwater isolation function (and NSSS [nuclear steam supply
system] break response) has not changed. Therefore, there will be no
increase in the consequences of an accident evaluated previously in
the FSAR.
An analysis was performed to quantify the impact of the proposed
modification on the probability of MFCV [main feedwater control
valve] failure (closure) during normal plant operation. Comparison
of this failure probability for the existing design (1.20E-1 per
year) versus the proposed design (6.99E-2 per year) indicates that
the percentage reduction in the system failure probability at power
is 41.75%. Thus, the proposed design results in a reduction in the
probability of inadvertent MFCV failures at power and hence, a
reduction in the probability of a reactor trip and subsequent
challenges to other safety systems.
While this modification reduces the probability of a reactor
trip, it slightly increases the unavailability of the feedwater
isolation function. This is because the original design required
actuation of only one FWIS [feedwater isolation system] train to
close the MFC&BVs, whereas the new design requires actuation of both
trains. The impact of the modification on the probability of
incurring a feedwater isolation failure was therefore quantified,
utilizing PRA [probabilistic risk assessment] techniques. Fault
trees were developed for both the new and existing designs. Failure
probabilities for each event were then obtained from the IPE
[individual plant examination] and utilized to calculate failure
probabilities for the feedwater isolation safety function. This
calculation considered hardware failures
[[Page 34901]]
only, i.e., failure of an MFIV to close after receiving an actuation
signal. The failure probability of feedwater isolation, based on the
proposed design, was determined to be 6.1E-5 per demand (1 event
every 16,400 demands). The existing design was found to have a
failure probability of 2.8E-5 per demand (1 event every 35,700
demands). Therefore, this modification will not significantly
increase the probability or consequences of an accident evaluated
previously in the FSAR.
2. The proposed change does not create the possibility of a new
or different kind of accident from any accident previously
evaluated.
The modification maintains the present de-energize-to-actuate
configuration of the MFC&BV trip solenoid valves.
Thus, the proposed modification does not create the possibility
of an accident of a different type than any previously evaluated.
3. The proposed change does not involve a significant reduction
in a margin of safety.
Credit is taken in the accident analyses for the MFIVs to close
on demand for feedwater isolation. Because of this, the MFIVs have
been incorporated into the Callaway Technical Specifications. Action
Statements and surveillance requirements have been developed to
assure the availability of the valves when needed.
The MFC&BVs are not addressed by any of the Callaway Technical
Specifications or their bases. Therefore, this modification will not
involve a significant reduction in the margin of safety as defined
in the basis for any technical specification.
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: Callaway County Public
Library, 710 Court Street, Fulton, Missouri 65251
Attorney for licensee: Gerald Charnoff, Esq., Shaw, Pittman, Potts
& Trowbridge, 2300 N Street, NW., Washington, DC 20037
NRC Project Director: William H. Bateman
Wisconsin Electric Power Company, Docket Nos. 50-266 and 50-301,
Point Beach Nuclear Power Plant, Unit Nos. 1 and 2, Town of Two
Creeks, Manitowoc County, Wisconsin
Date of amendment request: May 29, 1996
Description of amendment request: The proposed amendment would
revise Technical Specification (TS) Section 15.4.4, ``Containment
Tests,'' to incorporate the provisions of 10 CFR Part 50, Appendix J,
``Primary Reactor Containment Leakage Testing for Water-Cooled Power
Reactors,'' Option B. Revisions would also be made to TS Sections 15.1,
``Definitions,'' 15.3.6, ``Containment System,'' and 15.6,
``Administrative Controls,'' to support the proposed changes to Section
15.4.4.
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:
1. Operation of this facility under the proposed Technical
Specifications will not create a significant increase in the
probability or consequences of an accident previously evaluated.
The proposed change does not involve a change to structures,
systems, or components which would affect the probability or
consequences of an accident previously evaluated in the PBNP [Point
Beach Nuclear Plant] Final Safety Analyses Report (FSAR).
Furthermore, containment leakage rate testing is not an initiator of
any accident. The proposed change simply provides a mechanism within
the Technical Specifications for implementing a performance-based
method of determining the frequency for leakage rate testing which
has been approved by the NRC. The proposed change does not affect
reactor operations or accident analysis and has no significant
radiological consequences. Therefore, this change will not create a
significant increase in the probability or consequences of an
accident previously evaluated.
2. Operation of this facility under the proposed Technical
Specifications change will not create the possibility of a new or
different kind of accident from any accident previously evaluated.
The proposed change does not involve a change to the plant
design or operation. As a result, the proposed change does not
affect any of the parameters or conditions that contribute to
initiation of any accidents. This change involves a potential
reduction of Type A, B, and C test frequency. Except for the method
of defining the test frequency, the methods for performing the
actual tests are not changed. No new accident modes are created by
extending the testing intervals. No safety-related equipment or
safety functions are altered as a result of this change. Extending
the test frequency has no influence on, nor does it contribute to,
the possibility of a new or different kind of accident or
malfunction from those previously analyzed. Therefore, the proposed
change will not create the possibility of a new or different kind of
accident from any accident previously evaluated.
3. Operation of this facility under the proposed Technical
Specifications change will not create a significant reduction in a
margin of safety.
The proposed change potentially affects only the frequency of
Type A, B, and C testing. Except for the method of defining test
frequency, the methods for performing the actual tests are not
changed. The proposed change is based on NRC accepted provisions and
maintains necessary levels of system and component reliability
affecting containment integrity. Evaluation of the performance-based
approach to leakage rate testing, as documented in NUREG-1493,
concludes that the impact on public health and safety due to revised
testing intervals is negligible. Furthermore, the proposed change
will not reduce the availability of systems associated with
containment integrity when they are required to mitigate accident
conditions. Therefore, the proposed change will not create 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: Joseph P. Mann Library, 1516
Sixteenth Street, Two Rivers, Wisconsin 54241
Attorney for licensee: Gerald Charnoff, Esq., Shaw, Pittman, Potts,
and Trowbridge, 2300 N Street, NW., Washington, DC 20037
NRC Project Director: Gail H. Marcus
Wisconsin Public Service Corporation, Docket No. 50-305, Kewaunee
Nuclear Power Plant, Kewaunee County, Wisconsin
Date of amendment request: June 4, 1996
Description of amendment request: The proposed amendment would
revise the Kewaunee Nuclear Power Plant Technical Specifications (TS)
by reducing the surveillance test frequencies for the radiation
monitoring system (Table TS 4.1-1) and the control rods (Table TS 4.1-
3) in accordance with the guidance of Generic Letter 93-05, ``Line-Item
Technical Specifications Improvements to Reduce Surveillance
Requirements for Testing During Power Operation,'' dated September 27,
1993.
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:
Table TS 4.1-1, ``Minimum Frequencies for Checks, Calibrations
and Test of Instrument Channels,'' Item 19
The proposed changes were reviewed in accordance with the
provisions of 10 CFR 50.92 to determine that no significant hazards
exist. The proposed changes will not:
1. Involve a significant increase in the probability or
consequences of an accident previously evaluated.
The radiation monitors are not accident initiators; therefore,
they cannot increase the probability of an accident occurring. The
reliability of the radiation monitors is not expected to decrease
due to the decreased surveillance frequency; therefore, this change
does not increase the consequences of an accident.
[[Page 34902]]
The addition of comment (a) to the Check, Calibrate, and Test
columns is merely a clarification of the existing information in the
table and does not change the intent of the Technical
Specifications.
2. Create the possibility of a new or different kind of accident
from any accident previously evaluated.
The proposed change revises only the testing frequency and does
not revise the test method or operational performance of the
radiation monitors. The radiation monitors are not accident
initiators; therefore, they cannot create a new or different kind of
accident.
3. Involve a significant reduction in the margin of safety.
Quarterly testing of the radiation monitoring system channels
will continue to verify operability of the monitors. Decreasing the
test surveillance frequency is not expected to decrease the
reliability of the radiation monitors. This change is acceptable in
accordance with Generic Letter 93-05 and NUREG-1366, ``Improvements
to Technical Specifications Surveillance Requirements.''
Table TS 4.1-3, ``Minimum Frequencies for Equipment Tests,''
Item 1
The proposed change in test frequency for control rod exercising
was reviewed in accordance with the provisions of 10 CFR 50.92 to
determine that no significant hazards exist. It has been determined
that the proposed change will not:
1. Involve a significant increase in the probability or
consequences of an accident previously evaluated.
The proposed change revises only the testing frequency for
control rod exercising. The control rod exercise surveillance
procedure will continue to be conducted, on a quarterly basis, to
ensure that the equipment remains operable. The reduced frequency of
control rod exercising reduces the probability of an inadvertent
reactor trip occurring during testing due to a dropped control rod.
Surveillance procedure SP 49-075 is conducted to verify rod
movement. In accordance with NUREG-1366, the frequency of a stuck
control rod occurring is very low. This condition is most often
discovered during reactor startup or during low power physics
testing. The reduction in control rod exercising is, therefore,
considered acceptable and is not expected to affect the probability
of a stuck control rod occurring.
2. Create the possibility of a new or different kind of accident
from any accident previously evaluated.
The proposed change revises only the testing frequency and does
not revise the test method or the design of the control rod system.
Therefore, a new or different kind of accident will not be created
by this change.
3. Involve a significant reduction in the margin of safety.
Quarterly control rod exercising will continue to verify
movement of the control rods. No adverse consequences are expected
to occur due to decreasing the test frequency. This change is
acceptable in accordance with Generic Letter 93-05 and NUREG-1366.
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 Wisconsin,
Cofrin Library, 2420 Nicolet Drive, Green Bay, Wisconsin 54311-7001
Attorney for licensee: Bradley D. Jackson, Esq., Foley and Lardner,
P. O. Box 1497, Madison, Wisconsin 53701-1497
NRC Project Director: Gail H. Marcus
Wisconsin Public Service Corporation, Docket No. 50-305, Kewaunee
Nuclear Power Plant, Kewaunee County, Wisconsin
Date of amendment request: June 10, 1996
Description of amendment request: The proposed amendment would
revise Technical Specification 4.2.b, ``Steam Generator Tubes,'' and
its associated basis, by allowing the use of Westinghouse laser-welded
sleeves to repair defective steam generator tubes. A description of the
sleeving repair process and supporting technical justification are
contained in WCAP-13088, Revision 3, ``Westinghouse Series 44 and 51
Steam Generator Generic Sleeving Report.'' WCAP-13088, and a non-
proprietary version (WCAP-13089), were submitted to the Nuclear
Regulatory Commission on April 13, 1995, to support a similar TS
amendment request for the DC Cook Nuclear Power Plant, Unit 1.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. Operation of the KNPP [Kewaunee Nuclear Power Plant] in
accordance with the proposed license amendment does not involve a
significant increase in the probability or consequences of an
accident previously evaluated.
The LWS [laser-welded sleeve] configuration has been designed
and analyzed in accordance with the requirements of the ASME
[American Society of Mechanical Engineers] Code. Fatigue and stress
analyses of the sleeved tube assemblies produced acceptable results;
i.e., the applied stresses and fatigue usage for the sleeve and weld
are bounded by the limits established in the ASME Code. ASME Code
minimum material property values are used for the structural and
plugging limit analysis. Ultrasonic inspection is used to verify
that minimum weld fusion zone thicknesses are produced. Mechanical
testing of 7/8'' tubesheet sleeves installed in roll expanded tubes
has shown that the individual joint structural strength of Alloy 690
LWSs provides margin to acceptance limits. These acceptance limits
bound the most limiting loadings (3 times normal operating pressure
differential) recommended by RG [Regulatory Guide] 1.121. Therefore,
each individual joint provides for structural integrity exceeding RG
recommendations. A hypothetical loss of integrity of one of the
joints will not result in a loss of structural integrity for the
sleeve. Leakage testing for 3/4'' and 7/8'' full length tubesheet
sleeves has demonstrated that unacceptable levels of primary-to-
secondary leakage are not expected during all plant conditions for
non-welded tubesheet sleeve lower joints. The welded joint produces
a hermetic seal, and therefore will not leak under any plant
conditions. Laser welded sleeves will not contribute to the current
SLB [steam-line break] primary-to-secondary leakage limit of 34 gpm
in the faulted loop. The 34 gpm leakage limit was calculated in
accordance with the standard review plan methodology to support
implementation of the voltage-based repair criteria for tube support
plate intersections.
The sleeve minimum acceptable wall thickness (used for
developing the depth based plugging limit for the sleeve) is
determined using the guidance of RG 1.121 and the pressure stress
equation of Section III of the ASME Code. With respect to the design
of the sleeve for KNPP, the limiting requirement of the RG which
applies to part throughwall degradation is that the minimum
acceptable wall must maintain a factor of safety consistent with the
analysis conditions as defined by the ASME Code. A bounding set of
design and transient loading input conditions was used for the
minimum wall thickness evaluation in the generic evaluation.
Evaluation of the minimum acceptable wall thickness for normal,
upset and postulated accident condition loading per the ASME Code
indicates the limiting condition is established for the normal
operating conditions, and the minimum acceptable wall thickness for
this case bounds the upset and faulted condition values.
According to RG recommendations, an allowance for non-
destructive evaluation (NDE) uncertainty and operational growth of
existing tube wall degradation indications within the sleeve must be
accounted for when determining the sleeve plugging limit. A
conservative tube wall degradation growth rate per cycle and an NDE
uncertainty has been assumed for determining the sleeve TS plugging
limit. The sleeve wall degradation extent determined by NDE, which
would require plugging sleeved tubes, is developed using the
guidance of RG 1.121 and is defined in WCAP-13088 [non-proprietary
WCAP-13089] to be 25% throughwall (plugging limit = 100% -
structural limit + NDE uncertainty + growth) for KNPP.
The hypothetical consequences of failure of the sleeve joint
would be bounded by the current SG [steam generator] tube rupture
analysis included in the KNPP Updated Safety Analysis Report. Due to
the slight reduction in diameter caused by the sleeve wall
thickness, primary coolant release rates would be slightly less than
assumed for the SG tube rupture analysis (depending on break
location), and therefore, would result
[[Page 34903]]
in lower total primary fluid mass release to the secondary system.
The proposed TS change to use Alloy 690 LWSs does not adversely
impact any other previously evaluated design basis accidents or the
results of LOCA [loss of coolant accident] and non-LOCA accident
analyses for the current TS minimum reactor coolant system flow
rate. The results of the analyses and testing, as well as plant
operating experience, demonstrates that the sleeve assembly is an
acceptable means of maintaining tube integrity. Plugging limit
criteria are established using the guidance of RG 1.121.
Furthermore, per RG 1.83 recommendations, the sleeved tube will be
monitored through periodic inspections with present NDE techniques.
These measures demonstrate that installation of sleeves spanning
degraded areas of the tube will restore the tube to a condition
consistent with its original design basis.
Corrosion testing of free span LWS joint has indicated that the
corrosion resistance (relative to roll transitions) can be increased
by greater than a factor of ten with the application of a PWHT [post
weld heat treatment] step. Estimations of joint susceptibility based
on expected far field stresses after heat treatment using the
expected original tube-to-tubesheet hydraulic expansion transition
residual stresses and actual time to crack in these transitions at
KNPP indicate that LWS joint lifetime should exceed the current
plant license. Consistent with other license amendments addressing
LWS, all free span laser welds will receive a PWHT; therefore, rapid
corrosion degradation of the free span joint is not expected.
Recently performed corrosion testing of LWS joints in locked tube
conditions indicates that with PWHT the stress corrosion cracking
resistance and initiation potential in the parent tube weld region
is greatly enhanced. Similar test results and conclusions would be
expected for KNPP. The Model 51 SG tube span between the top of the
tubesheet and the first support plate is such that even lower PWHT
residual stresses would be expected. Also, the weld placement within
the hydraulically expanded area and sleeve installation sequence
have been optimized to provide for some level of heat treatment at
the upper transition above the weld and lower far field residual
stress levels. While no parent tube degradation has been detected at
this elevation, or any other elevation in a laser welded sleeve
assembly, the relocation of the weld serves to provide further
resistance to PWSCC [primary water stress corrosion cracking] at
this elevation. The suggested target PWHT temperature has also been
optimized in that this temperature provides for adequate PWHT while
maintaining the parent tube far field stresses.
Approximately 19,500 LWSs have been installed in the U.S. Of
this number, over 300 which have up to 3 cycles of operation were
inspected in 1995 using the CECCO-5 probe. No degradation of the
sleeves or the parent tube was detected. Operating experience in
Europe has shown good performance of the LWS joint for up to 5
cycles of operation. In 1994, approximately 11,200 LWSs were
installed in the Doel-4 Plant. After one year of operation, all in-
service sleeves were inspected using the +point probe. No service
induced corrosion was detected. In 1995, approximately 18,600 LWSs
were installed in two different U.S. plants. Due to their limited
operational time, these sleeves have not been inspected.
Conformance of sleeve design with the applicable sections of the
ASME Code and results of the leakage and mechanical tests support
the conclusion that installation of LWSs will not increase the
probability or consequences of an accident previously evaluated.
2. The proposed license amendment request does not create the
possibility of a new or different kind of accident from an accident
previously evaluated.
Installation of LWSs will not introduce significant or adverse
changes to the plant design basis and does not represent a potential
to affect any other plant component. Stress and fatigue analysis of
the repair has shown that the ASME Code and RG 1.121 criteria are
not exceeded. Installation of LWSs maintains overall tube bundle
structural and leakage integrity at a level consistent to that of
the originally supplied tubing during all plant conditions; stresses
are bounded by the Code and the tubing is leaktight. Sleeving of
tubes does not provide a mechanism resulting in an accident outside
of the area affected by the sleeves. Any hypothetical accident as a
result of potential tube or sleeve degradation in the repaired
portion of the tube is bounded by the existing tube rupture accident
analysis. Since the sleeve design does not affect any component or
location of the tube outside of the immediate area repaired, in
addition to the fact that the installation of sleeves and the impact
on current plugging level analyses is accounted for, the possibility
that laser welded sleeving creates a new or different type of
accident is not supported.
Installation of LWSs will reduce the potential for primary-to-
secondary leakage during postulated steam line break while not
significantly impacting primary coolant flow area in the event of a
LOCA. By effectively isolating degraded areas of the tube through
repair, the potential for steam line break leakage is reduced.
3. The proposed license amendment does not involve a significant
reduction in the margin of safety.
The LWS repair of degraded SG tubes as identified in WCAP-13088
[non-proprietary WCAP-13089] has been shown by analysis to
restore the integrity of the tube bundle consistent with its
original design basis conditions; i.e., tube/sleeve operational and
faulted conditions stresses and cumulative fatigue usage are bounded
by the ASME Code requirements and the repaired tubes are leaktight.
The safety factors used in the design of sleeves for the repair of
degraded tubes are consistent with the safety factors in the ASME
Code used in SG design. The design of the LWS lower joint for 7/8''
tube sleeves has been verified by testing to sufficiently preclude
leakage during normal and postulated accident conditions. The
portions of the installed sleeve assembly which represents the
reactor coolant pressure boundary will be monitored for the
initiation and progression of sleeve/tube wall degradation, thus
satisfying the requirements of RG 1.83. The portion of the tube
bridged by the sleeve joints is effectively removed from the
pressure boundary, and the sleeve then forms the new pressure
boundary. The areas of the sleeved tube assembly which require
inspection are defined in WCAP-13088 [non-proprietary WCAP-13089].
Since the installed sleeves represent a portion of the pressure
boundary, a baseline inspection of these areas is required prior to
operation with sleeves installed.
The effect of sleeving on the design transients and accident
analyses has been reviewed based on the installation of sleeves up
to the level of SG tube plugging coincident with the minimum reactor
coolant flow rate. The installation of sleeves is evaluated as the
equivalent of some level of SG tube plugging. This is based on
determining the minimum reactor coolant flow for the LOCA
evaluation. Information provided in WCAP-13088 [non-proprietary
WCAP-13089] describes the method to determine the flow equivalent
for all combinations of tubesheet and tube support plate sleeves.
Therefore, installation of LWSs will not result in 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 Wisconsin,
Cofrin Library, 2420 Nicolet Drive, Green Bay, Wisconsin 54311-7001
Attorney for licensee: Bradley D. Jackson, Esq., Foley and Lardner,
P. O. Box 1497, Madison, Wisconsin 53701-1497
NRC Project Director: Gail H. Marcus
Wisconsin Electric Power Company, Docket Nos. 50-266 and 50-301,
Point Beach Nuclear Power Plant, Unit Nos. 1 and 2, Town of Two
Creeks, Manitowoc County, Wisconsin
Date of amendment request: June 4, 1996 (VPNPD-96-035)
Description of amendment request: The proposed amendment would
revise Technical Specification (TS) Section 15.2.3, ``Limiting Safety
System Settings and Protective Instrumentation,'' and Section 15.5.3,
``Design Features - Reactor,'' to incorporate changes associated with
the operation of Point Beach Nuclear Plant (PBNP), Unit 2, with
replacement steam generators.
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 this facility under the proposed Technical
Specifications will not
[[Page 34904]]
create a significant increase in the probability or consequences of
an accident previously evaluated.
The proposed changes do not involve a change to structures,
systems, or components which would affect the probability or
consequences of an accident previously evaluated in the PBNP Final
Safety Analyses Report (FSAR). The proposed setpoints maintain the
margin to safe operation of Unit 2 with the replacement steam
generators. In order to maintain one set of safety analyses for both
units, the analyses for operation of Unit 2 with the replacement
steam generators were performed to encompass the operation of Unit
1. Therefore, the proposed changes apply to the operation of both
units and maintain the margin of safety for each. The proposed
change to the description of nominal RCS [reactor coolant system]
volume is an administrative change and has no effect on plant
operation. Therefore, the probability or consequences of a
previously evaluated accident are not significantly increased as a
result of these changes.
2. Operation of this facility under the proposed Technical
Specifications change will not create the possibility of a new or
different kind of accident from any accident previously evaluated.
The proposed changes do not involve a change to the plant
design. The proposed setpoints maintain the margin to safe operation
of Unit 2 with the replacement steam generators. In order to
maintain one set of safety analyses for both units, the analyses for
operation of Unit 2 with the replacement steam generators were
performed to encompass the operation of Unit 1. Therefore, the
proposed changes apply to the operation of both units and maintain
the margin of safety for each. These changes do not affect any of
the parameters or conditions that contribute to initiation of any
accidents. The proposed change to the description of nominal RCS
volume is an administrative change and has no effect on plant
operation or initiation of any accidents. Therefore, the proposed
changes will not create the possibility of a new or different kind
of accident from any accident previously evaluated.
3. Operation of this facility under the proposed Technical
Specifications change will not create a significant reduction in a
margin of safety.
The proposed setpoints maintain the margin to safe operation of
Unit 2 with the replacement steam generators. In order to maintain
one set of safety analyses for both units, the analyses for
operation of Unit 2 with replacement steam generators were performed
to encompass the operation of Unit 1. Therefore, the proposed
changes apply to the operation of both units and maintain the margin
of safety for each. The proposed change to the description of
nominal RCS volume is an administrative change and has no effect on
plant operation. Therefore, the proposed changes will not create 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: Joseph P. Mann Library, 1516
Sixteenth Street, Two Rivers, Wisconsin 54241
Attorney for licensee: Gerald Charnoff, Esq., Shaw, Pittman, Potts,
and Trowbridge, 2300 N Street, NW., Washington, DC 20037
NRC Project Director: Gail H. Marcus
Wisconsin Electric Power Company, Docket Nos. 50-266 and 50-301,
Point Beach Nuclear Power Plant, Unit Nos. 1 and 2, Town of Two
Creeks, Manitowoc County, Wisconsin
Date of amendment request: June 4, 1996 (VPNPD-96-036)
Description of amendment request: The proposed amendment would
revise Technical Specification (TS) Section 15.2.1, ``Safety Limit,
Reactor Core,'' 15.2.3, ``Limiting Safety System Settings, Protective
Instrumentation,'' and Section 15.3.1.G, ``Operational Limitations,''
to maintain safety margin for Unit 2 with replacement steam generators.
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 this facility under the proposed Technical
Specifications will not create a significant increase in the
probability or consequences of an accident previously evaluated.
The proposed change does not involve a change to structures,
systems, or components which would affect the probability or
consequences of an accident previously evaluated in the PBNP [Point
Beach Nuclear Plant] Final Safety Analyses Report (FSAR). The
proposed changes maintain the margin to safe operation for Unit 2
with the replacement steam generators. In order to maintain one set
of safety analyses for both units, the analyses for operation of
Unit 2 with the replacement steam generators were performed to
encompass the operation of Unit 1. Therefore, the proposed changes
apply to the operation of both units and maintain the margin of
safety for each. The proposed changes do not change, degrade, or
preclude the prevention or mitigation of the consequences of any
accident described in the FSAR. Therefore, the probability or
consequences of a previously evaluated accident are not
significantly increased as a result of these changes.
2. Operation of this facility under the proposed Technical
Specifications change will not create the possibility of a new or
different kind of accident from any accident previously evaluated.
The proposed changes do not involve a change to the plant
design. The proposed changes maintain the margin to safe operation
for Unit 2 with the replacement steam generators. In order to
maintain one set of safety analyses for both units, the analyses for
operation of Unit 2 with the replacement steam generators were
performed to encompass the operation of Unit 1. Therefore, the
proposed changes apply to the operation of both units and maintain
the margin of safety for each. These changes do not affect any of
the parameters or conditions that contribute to initiation of any
accidents. In addition, the safety functions of safety-related
systems and components, which are related to accident mitigation,
have not been altered. Therefore, the proposed changes will not
create the possibility of a new or different kind of accident from
any accident previously evaluated.
3. Operation of this facility under the proposed Technical
Specifications change will not create a significant reduction in a
margin of safety.
The proposed changes maintain the margin to safe operation for
Unit 2 with the replacement steam generators. In order to maintain
one set of safety analyses for both units, the analyses for
operation of Unit 2 with replacement steam generators were performed
to encompass the operation of Unit 1. Therefore, the proposed
changes apply to the operation of both units and maintain the margin
of safety for each. The proposed changes have no affect on the
availability, operability, or performance of the safety-related
systems and components described in the Technical Specifications.
Therefore, the proposed changes will not create 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: Joseph P. Mann Library, 1516
Sixteenth Street, Two Rivers, Wisconsin 54241
Attorney for licensee: Gerald Charnoff, Esq., Shaw, Pittman, Potts,
and Trowbridge, 2300 N Street, NW., Washington, DC 20037
NRC Project Director: Gail H. Marcus
Previously Published Notices of Consideration of Issuance of
Amendments to Facility Operating Licenses, Proposed No Significant
Hazards Consideration Determination, and Opportunity for a Hearing
The following notices were previously published as separate
individual notices. The notice content was the same as above. They were
published as individual notices either because time did not allow the
Commission to wait for this biweekly notice or because the action
involved exigent circumstances. They are repeated here because the
[[Page 34905]]
biweekly notice lists all amendments issued or proposed to be issued
involving no significant hazards consideration.
For details, see the individual notice in the Federal Register on
the day and page cited. This notice does not extend the notice period
of the original notice.
Florida Power and Light Company, et al., Docket No. 50-335, St.
Lucie Plant, Unit No. 1, St. Lucie County, Florida
Date of amendment request: June 1, 1996
Description of amendment request: Revise Technical Specifications
to reflect reduced reactor coolant system flows resulting from
increased percentage of plugged steam generator tubes.
Date of publication of individual notice in the Federal Register:
June 7, 1996 (61 FR 29140)
Expiration date of individual notice: June 24, 1996
Local Public Document Room location: Indian River Junior College
Library, 3209 Virginia Avenue, Fort Pierce, Florida 34954-9003
Public Service Electric & Gas Company, Docket Nos. 50-272 and 50-
311, Salem Nuclear Generating Station, Unit Nos. 1 and 2, Salem
County, New Jersey
Date of amendment request: May 31, 1996
Brief description of amendment request: The amendments (1) revise
the Reactor Vessel Level Indication System (RVLIS) Action Statements to
facilitate actions necessary for channel testing to be performed in
Mode 3, (2) revise the Channel Calibration definition to better account
for temperature detector channel calibration methodology, and (3)
delete a requirement to install a jumper in the Auxiliary Feedwater
actuation logic since a design change will result in the jumper
function being performed by a relay.
Date of publication of individual notice in Federal Register: June
17, 1996 (61 FR 30641)
Expiration date of individual notice: July 17, 1996
Local Public Document Room location: Salem Free Public Library, 112
West Broadway, Salem, New Jersey 08079
Notice of Issuance of Amendments to Facility Operating Licenses
During the period since publication of the last biweekly notice,
the Commission has issued the following amendments. The Commission has
determined for each of these amendments that the application complies
with the standards and requirements of the Atomic Energy Act of 1954,
as amended (the Act), and the Commission's rules and regulations. The
Commission has made appropriate findings as required by the Act and the
Commission's rules and regulations in 10 CFR Chapter I, which are set
forth in the license amendment.
Notice of Consideration of Issuance of Amendment to Facility
Operating License, Proposed No Significant Hazards Consideration
Determination, and Opportunity for A Hearing in connection with these
actions was published in the Federal Register as indicated.
Unless otherwise indicated, the Commission has determined that
these amendments satisfy the criteria for categorical exclusion in
accordance with 10 CFR 51.22. Therefore, pursuant to 10 CFR 51.22(b),
no environmental impact statement or environmental assessment need be
prepared for these amendments. If the Commission has prepared an
environmental assessment under the special circumstances provision in
10 CFR 51.12(b) and has made a determination based on that assessment,
it is so indicated.
For further details with respect to the action see (1) the
applications for amendment, (2) the amendment, and (3) the Commission's
related letter, Safety Evaluation and/or Environmental Assessment as
indicated. All of these items are available for public inspection at
the Commission's Public Document Room, the Gelman Building, 2120 L
Street, NW., Washington, DC, and at the local public document rooms for
the particular facilities involved.
Baltimore Gas and Electric Company, Docket Nos. 50-317 and 50-318,
Calvert Cliffs Nuclear Power Plant, Unit Nos. 1 and 2, Calvert
County, Maryland
Date of application for amendments: May 28, 1996, as supplemented
by letters dated May 31 and June 5, 1996
Brief description of amendments: These amendments authorize the
licensee to revise applicable Updated Final Safety Analysis Report
sections to reflect the installation of a variable flow controller for
the service water inlet control valves for the containment air coolers
that is not within the current licensing basis of Calvert Cliffs
Nuclear Power Plant Units No. 1 and No. 2. These amendments are being
issued pursuant to the requirements of 10 CFR 50.59(c) because the
review by Baltimore Gas and Electric Company identified the changes as
an unreviewed safety question. No changes to the Technical
Specifications are required by these amendments.
Date of issuance: June 17, 1996
Effective date: As of the date of issuance to be implemented within
30 days.
Amendment Nos.: 215 and 192
Facility Operating License Nos. DPR-53 and DPR-69: The amendments
revised the Updated Final Safety Analysis Report. Public comments
requested as to proposed no significant hazards consideration: Yes (61
FR 27371 dated May 31, 1996). The notice provided an opportunity to
submit comments on the Commission's proposed no significant hazards
consideration determination. No comments have been received. The notice
also provided for an opportunity to request a hearing by July 1, 1996,
but indicated that if the Commission makes a final no significant
hazards consideration determination, any such hearing would take place
after issuance of the amendments. The May 31 and June 5, 1996, letters
provided additional information that did not change the scope of the
May 28, 1996, application.
The Commission's related evaluation of the amendment, finding of
exigent circumstances, and a final no significant hazards consideration
determination are contained in a Safety Evaluation dated June 17, 1996.
Local Public Document Room location: Calvert County Library,
Prince Frederick, Maryland 20678
Carolina Power & Light Company, et al., Docket No. 50-400, Shearon
Harris Nuclear Power Plant, Unit 1, Wake and Chatham Counties,
North Carolina
Date of application for amendment: October 24, 1994, as
supplemented August 31, 1995 and February 8, 1996. The August 31, 1995
and February 8, 1996, letters provide clarification information. The
new information changed the scope of the October 24, 1994, letter and
was re-noticed on May 8, 1996, but did not change the initial no
significant hazards consideration determination.
Brief description of amendment: The proposed amendment would revise
the TS to allow the relocation of TS 3/4.11.2.6, Gas Storage Tanks; and
the associated Bases in the TS to licensee-controlled documents.
Date of issuance: June 12, 1996
Effective date: June 12, 1996
Amendment No.: 64
Facility Operating License No. NPF-63. Amendment revises the
Technical Specifications
Date of initial notice in Federal Register: November 23, 1994 (59
FR 60379). The Commission's related evaluation of the amendment is
contained in a Safety Evaluation dated
[[Page 34906]]
June 12, 1996. No significant hazards consideration comments received:
No
Local Public Document Room location: Cameron Village Regional
Library, 1930 Clark Avenue, Raleigh, North Carolina 27605
Commonwealth Edison Company, Docket Nos. 50-373 and 50-374, LaSalle
County Station, Units 1 and 2, LaSalle County, Illinois
Date of application for amendments: December 21, 1995
Brief description of amendments: The amendments delete the
requirement to place the reactor mode switch in the Shutdown position
if a stuck open safety/relief valve can not be closed within 2 minutes.
The operator will still be required to scram the reactor if suppression
pool average water temperature reaches 110 degrees Fahrenheit or
greater. The amendment also includes editorial changes to the index
pages.
Date of issuance: June 18, 1996
Effective date: Immediately, to be implemented within 60 days
Amendment Nos.: 113 and 98
Facility Operating License Nos. NPF-11 and NPF-18: The amendments
revised the Technical Specifications.
Date of initial notice in Federal Register: May 8, 1996 (61 FR
20844) The Commission's related evaluation of the amendments is
contained in a Safety Evaluation dated June 18, 1996. No significant
hazards consideration comments received: No
Local Public Document Room location: Jacobs Memorial Library,
Illinois Valley Community College, Oglesby, Illinois 61348
Duke Power Company, et al., Docket Nos. 50-413 and 50-414, Catawba
Nuclear Station, Units 1 and 2, York County, South Carolina
Date of application for amendments: July 18, 1994, as supplemented
by letter dated October 9, 1994
Brief description of amendments: The amendments revise the current
combined Technical Specifications (TS) for Units 1 and 2 by separating
them into individual volumes for Unit 1 and Unit 2. In addition to the
changes required by the TS split, some administrative and editorial
changes were made, such as the correction of typographical errors and
the deletion of unnecessary blank pages.
Date of issuance: June 12, 1996
Effective date: As of the date of issuance to be implemented within
30 days
Amendment Nos.: 148 and 142
Facility Operating License Nos. NPF-35 and NPF-52: Amendments
revised the Technical Specifications.
Date of initial notice in Federal Register: September 14, 1994 (59
FR 47166) The October 9, 1995 and June 6, 1996, letters provided
clarifying information that did not change the scope of the July 18,
1994, application and the initial proposed no significant hazards
consideration. The Commission's related evaluation of the amendments is
contained in a Safety Evaluation dated June 12, 1996. No significant
hazards consideration comments received: No
Local Public Document Room location: York County Library, 138 East
Black Street, Rock Hill, South Carolina 29730
Entergy Operations, Inc., System Energy Resources, Inc., South
Mississippi Electric Power Association, and Mississippi Power &
Light Company, Docket No. 50-416, Grand Gulf Nuclear Station, Unit
1, Claiborne County, Mississippi
Date of application for amendment: February 22, 1996
Brief description of amendment: The amendment increased the safety
function lift setpoint tolerances for the safety and relief valves that
are listed in Surveillance Requirement 3.4.4.1 (Page 3.4-10) of the
Technical Specifications (TSs) for the Grand Gulf Nuclear Station, Unit
1. The tolerances were increased from the current plus or minus 1
percent of the safety function (i.e., safety relief valve) lift
setpoint to plus or minus 3 percent.
Date of issuance: June 12, 1996
Effective date: June 12, 1996
Amendment No: 123
Facility Operating License No. NPF-29. Amendment revises the
Technical Specifications.
Date of initial notice in Federal Register: March 27, 1996 (61 FR
13524) The Commission's related evaluation of the amendment is
contained in a Safety Evaluation dated June 12, 1996. No significant
hazards consideration comments received: No
Local Public Document Room location: Judge George W. Armstrong
Library, 220 S. Commerce Street, Natchez, MS 39120
Illinois Power Company and Soyland Power Cooperative, Inc., Docket
No. 50-461, Clinton Power Station, Unit No. 1, DeWitt County,
Illinois
Date of application for amendment: May 1, 1996
Brief description of amendment: The amendment revises the Operating
License and Technical Specifications (TS) to implement 10 CFR Part 50,
Appendix J - Option B, by referring to Regulatory Guide 1.163,
``Performance-Based Containment Leak-Test Program.'' Specifically,
changes have been made to paragraph 2.D of the Operating License; TS
Section 1.1, ``Definitions;'' TS 3.6.1.1, ``Primary Containment;'' TS
3.6.1.1, ``Primary Containment Air Locks;'' TS 3.6.1.3, ``Primary
Containment Isolation Valves (PCIVs);'' and TS Section 5.5, ``Programs
and Manuals.''
Date of issuance: June 21, 1996
Effective date: June 21, 1996
Amendment No.: 105
Facility Operating License No. NPF-62: The amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: May 21, 1996 (61 FR
25708) The Commission's related evaluation of the amendment is
contained in a Safety Evaluation dated June 21, 1996. No significant
hazards consideration comments received: No
Local Public Document Room location: The Vespasian Warner Public
Library, 120 West Johnson Street, Clinton, Illinois 61727
Niagara Mohawk Power Corporation, Docket No. 50-410, Nine Mile
Point Nuclear Station, Unit 2, Oswego County, New York
Date of application for amendment: January 25, 1996
Brief description of amendment: The amendment revises Technical
Specification 3/4.3.3, Emergency Core Cooling System Actuation
Instrumentation, to more clearly define when, during shutdown and
refueling, the Loss of Voltage and Degraded Voltage relays for the Loss
of Power actuation trip functions are required to be operable.
Date of issuance: June 10, 1996
Effective date: As of the date of issuance to be implemented within
30 days.
Amendment No.: 72
Facility Operating License No. NPF-69: Amendment revises the
Technical Specifications.
Date of initial notice in Federal Register: May 8, 1996 (61 FR
20851) The Commission's related evaluation of the amendment is
contained in a Safety Evaluation dated June 10, 1996. No significant
hazards consideration comments received: No
Local Public Document Room location: Reference and Documents
Department, Penfield Library, State University of New York, Oswego, New
York 13126
[[Page 34907]]
Northeast Nuclear Energy Company, et al., Docket No. 50-336,
Millstone Nuclear Power Station, Unit No. 2, New London County,
Connecticut
Date of application for amendment: January 5, 1996, as supplemented
on May 31, 1996
Brief description of amendment: The amendment implements the
guidance of Generic Letter 93-08 by relocating Tables 3.3-2, ``Reactor
Protective Instrumentation Response Times'' and 3.3-5, ``Engineered
Safety Features Response Times'' from the Technical Specifications to
the Millstone Unit No. 2 Technical Requirements Manual (TRM). In
accordance with Generic Letter 93-08, the Limiting Conditions for
Operations for Technical Specifications 3.3.1.1, 3.3.2.1, and 3.7.1.6
are revised to eliminate their references to the aforementioned tables.
The amendment also revises Bases 3/4.3.1 and 3/4.3.2 to reference that
the instrument response times are located in the TRM and that these
tables in the TRM are now controlled under 10 CFR 50.59. The amendment
also removes a cycle-specific note from Tables 3.3-3 and 3.3-4.
Date of issuance: June 10, 1996
Effective date: As of the date of issuance, to be implemented
within 60 days.
Amendment No.: 198
Facility Operating License No. DPR-65: Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: February 14, 1996 (61
FR 5816) The May 31, 1996, letter provided additional information that
did not change the scope of the January 5, 1996, application and the
initial proposed no significant hazards consideration determination.
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated June 10, 1996. No significant hazards
consideration comments received: No
Local Public Document Room location: Learning Resources Center,
Three Rivers Community-Technical College, 574 New London Turnpike,
Norwich, CT 06360 and the Waterford Library, ATTN: Vince Juliano, 49
Rope Ferry Road, Waterford, Connecticut 06385
PECO Energy Company, Public Service Electric and Gas Company
Delmarva Power and Light Company, and Atlantic City Electric
Company, Docket Nos. 50-277 and 50-278, Peach Bottom Atomic Power
Station, Unit Nos. 2 and 3, York County, Pennsylvania
Date of application for amendments: February 15, 1996
Brief description of amendments: The amendments change the
Technical Specifications to implement 10 CFR Part 50, Appendix J,
Option B, by creating Technical Specification Section 5.5.12, ``Primary
Containment Leakage Rate Testing Program,'' which refers to Regulatory
Guide 1.163, ``Performance-Based Containment Leakage-Test Program.''
Date of issuance: June 18, 1996
Effective date: Both units, as of date of issuance, to be
implemented by June 28, 1996.
Amendments Nos.: 214 and 219
Facility Operating License Nos. DPR-44 and DPR-56: The amendments
revised the Technical Specifications.
Date of initial notice in Federal Register: March 27, 1996 (61 FR
13531) The Commission's related evaluation of the amendments is
contained in a Safety Evaluation dated June 18, 1996. No significant
hazards consideration comments received: No
Local Public Document Room location: Government Publications
Section, State Library of Pennsylvania, (REGIONAL DEPOSITORY) Education
Building, Walnut Street and Commonwealth Avenue, Box 1601, Harrisburg,
Pennsylvania 17105
Philadelphia Electric Company, Docket Nos. 50-352 and 50-353,
Limerick Generating Station, Units 1 and 2, Montgomery County,
Pennsylvania
Date of application for amendments: April 25, 1996
Brief description of amendments: The amendments relocate Technical
Specification Traversing In-Core Probe System Limiting Condition for
Operation 3/4.3.7.7 and its Bases 3/4.3.7.7, to the Limerick Generating
Station Technical Requirements Manual, and modify Note (f) of TS Table
4.3.1.1-1.
Date of issuance: June 11, 1996
Effective date: As of date of issuance, to be implemented within 30
days.
Amendment Nos.: 117 and 79
Facility Operating License Nos. NPF-39 and NPF-85. The amendments
revised the Technical Specifications.
Date of initial notice in Federal Register: May 8, 1996 (61 FR
20840) The Commission's related evaluation of the amendments is
contained in a Safety Evaluation dated June 11, 1996 No significant
hazards consideration comments received: No
Local Public Document Room location: Pottstown Public Library, 500
High Street, Pottstown, Pennsylvania 19464
Power Authority of The State of New York, Docket No. 50-286, Indian
Point Nuclear Generating Unit No. 3, Westchester County, New York
Date of application for amendment: March 14, 1996
Brief description of amendment: The proposed changes would allow a
one-time extension of the intervals for the steam generator tube
inspection that is due in July 1996.
Date of issuance: June 19, 1996
Effective date: As of the date of issuance to be implemented within
30 days.
Amendment No.: 166
Facility Operating License No. DPR-64: Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: May 8, 1996 (61 FR
20854) The Commission's related evaluation of the amendment is
contained in a Safety Evaluation dated June 19, 1996. No significant
hazards consideration comments received: No
Local Public Document Room location: White Plains Public Library,
100 Martine Avenue, White Plains, New York 10610
Tennessee Valley Authority, Docket Nos. 50-390 Watts Bar Nuclear
Plant, Unit 1, Rhea County, Tennessee
Date of application for amendment: February 28, as supplemented
April 15, and June 3, 1996.
Brief description of amendment: The proposed amendment would revise
the Technical Specifications (TS) to increase the surveillance
intervals for ice bed weight sampling and flow passage inspection from
9 months to 18 months. The TS would also be changed to provide an
increased ice sublimation allowance, associated with the increased
surveillance interval, by increasing the minimum total ice weight from
2,360,875 pounds to 2,403,800 pounds (1214 pounds/basket to 1236
pounds/basket).
Date of issuance: June 13, 1996
Effective date: June 13, 1996
Amendment No.: 2
Facility Operating License No. NPF-90: Amendment revises the
Technical Specifications.
Date of initial notice in Federal Register: April 10, 1996 (61 FR
15998) The Commission's related evaluation of the amendment is
contained in a Safety Evaluation dated June 13, 1996. No significant
hazards consideration comments received: No
Local Public Document Room location: Chattanooga-Hamilton County
Library, 1001 Broad Street, Chattanooga, TN 37402
[[Page 34908]]
The Cleveland Electric Illuminating Company, Centerior Service
Company, Duquesne Light Company, Ohio Edison Company, OES Nuclear,
Inc., Pennsylvania Power Company, Toledo Edison Company, Docket No.
50-440 Perry Nuclear Power Plant, Unit No. 1, Lake County, Ohio
Date of application for amendment: April 26, 1996
Brief description of amendment: The amendment corrected minor
technical and administrative errors in the Improved Technical
Specifications prior to its implementation.
Date of issuance: June 18, 1996
Effective date: June 18, 1996
Amendment No.: 85
Facility Operating License No. NPF-58: This amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: May 9, 1996 (61 FR
21213) The Commission's related evaluation of the amendment is
contained in a Safety Evaluation dated June 18, 1996. No significant
hazards consideration comments received: No
Local Public Document Room location: Perry Public Library, 3753
Main Street, Perry, Ohio 44081
TU Electric Company, Docket Nos. 50-445 and 50-446, Comanche Peak
Steam Electric Station, Unit Nos. 1 and 2, Somervell County, Texas
Date of amendment requests: April 25 (TXX-94119) and August 12,
1994 (TXX-94216), as supplemented by letters dated February 15 (TXX-
96055), March 7 (TXX-96078), and April 11, 1996 (TXX-96111).
Brief description of amendments: These amendments modified the
Administrative Controls specifications, relocate/remove requirements
that are adequately controlled by existing regulations other than 10
CFR 50.36 and the technical specifications. Guidance on the proposed
changes was developed by NRC and provided in the Standard Technical
Specifications for Westinghouse Plants, NUREG-1431. The changes also
update unit staff qualification requirements to Regulatory Guide 1.8,
Revision 2.
Date of issuance: June 12, 1996
Effective date: June 12, 1996, to be implemented witnin 60 days.
Amendment Nos.: Unit 1 - Amendment No. 50; Unit 2 - Amendment No.
36
Facility Operating License Nos. NPF-87 and NPF-89. The amendments
revised the Technical Specifications.
Date of initial notice in Federal Register: August 3, 1994 (59 FR
39599) and September 28, 1994 (59 FR 49439). The additional information
contained in the supplemental letters dated February 15, March 7, and
April 11, 1996, were clarifying in nature and thus, within the scope of
the initial notice and did not affect the staff's proposed no
significant hazards consideration determination. The Commission's
related evaluation of the amendments is contained in a Safety
Evaluation dated June 12, 1996. No significant hazards consideration
comments received: No
Local Public Document Room location: University of Texas at
Arlington Library, Government Publications/Maps, 702 College, P.O. Box
19497, Arlington, Texas 76019
TU Electric Company, Docket Nos. 50-445 and 50-446, Comanche Peak
Steam Electric Station, Unit Nos. 1 and 2, Somervell County, Texas
Date of amendment request: March 12, 1996 (TXX-96008)
Brief description of amendments: The amendments revised the
Technical Specifications to reflect the approval for the licensee to
use of the new Containment Leakage Rate Testing Program as required by
10 CFR Part 50, Appendix J, Option B for Comanche Peak Steam Electric
Station, Units 1 and 2. Implementation of the new performance based
leakage rate testing program will be based on the guidance provided by
Regulatory Guide 1.163, September 1995.
Date of issuance: June 13, 1996
Effective date: June 13, 1996, to be implemented within 60 days
Amendment Nos.: 51 and 37
Facility Operating License Nos. NPF-87 and NPF-89. The amendments
revised the Technical Specifications.
Date of initial notice in Federal Register: April 10, 1996 (61 FR
15999) The Commission's related evaluation of the amendments is
contained in a Safety Evaluation dated June 13, 1996. No significant
hazards consideration comments received: No
Local Public Document Room location: University of Texas at
Arlington Library, Government Publications/Maps, 702 College, P.O. Box
19497, Arlington, TX 76019
Union Electric Company, Docket No. 50-483, Callaway Plant, Unit 1,
Callaway County, Missouri
Date of application for amendment: September 9, 1994, as superseded
by letter dated July 25, 1995, and subsequently supplemented by letters
dated February 28, 1996, and April 9, 1996.
Brief description of amendment: The amendment would revise TS 3/
4.8.1 and its associated Bases to improve the overall emergency diesel
generator reliability and availability.
Date of issuance: June 17, 1996
Effective date: June 17, 1996, to be implemented within 30 days of
the date of issuance.
Amendment No.: 112
Facility Operating License No. NPF-30: The amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: August 30, 1995 (60 FR
45188) The February 28, 1996, and April 9, 1996 supplemental letters
provided additional clarifying information and did not change the
staff's original no significant hazards consideration determination.
The Commission's related evaluation of the amendment is contained in a
Safety Evaluation dated June 17, 1996. No significant hazards
consideration comments received: No.
Local Public Document Room location: Callaway County Public
Library, 710 Court Street, Fulton, Missouri 65251
Virginia Electric and Power Company, Docket Nos. 50-280 and 50-281,
Surry Power Station, Unit Nos. 1 and 2, Surry County, Virginia.
Date of application for amendments: April 15, 1996
Brief description of amendments: These amendments would revise the
Technical Specifications to indicate that the quadrant power tilt ratio
requirements are applicable only at power levels greater than 50% of
rated core power.
Date of issuance: June 7, 1996
Effective date: June 7, 1996
Amendment Nos.: 210 and 210
Facility Operating License Nos. DPR-32 and DPR-37: Amendments
revised the Technical Specifications.
Date of initial notice in Federal Register: May 8, 1996 (61 FR
20860) The Commission's related evaluation of the amendment is
contained in a Safety Evaluation dated June 7, 1996. No significant
hazards consideration comments received: No
Local Public Document Room location: Swem Library, College of
William and Mary, Williamsburg, Virginia 23185
Dated at Rockville, Maryland, this 26th day of June 1996.
For the Nuclear Regulatory Commission
Steven A. Varga,
Director, Division of Reactor Projects - I/II, Office of Nuclear
Reactor Regulation
[Doc. 96-16879 Filed 7-2-96; 8:45 am]
BILLING CODE 7590-01-F
