[Federal Register Volume 62, Number 185 (Wednesday, September 24, 1997)]
[Notices]
[Pages 50000-50018]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: X97-30924]
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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 August 29, 1997, through September 12, 1997.
The last biweekly notice was published on September 10, 1997 (62 FR
47696).
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 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
[[Page 50001]]
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 October 24, 1997, the licensee may file a request for a hearing
with respect to issuance of the amendment to the subject facility
operating license and any person whose interest may be affected by this
proceeding and who wishes to participate as a party in the proceeding
must file a written request for a hearing and a petition for leave to
intervene. Requests for a hearing and a petition for leave to intervene
shall be filed in accordance with the Commission's ``Rules of Practice
for Domestic Licensing Proceedings'' in 10 CFR Part 2. Interested
persons should consult a current copy of 10 CFR 2.714 which is
available at the Commission's Public Document Room, the Gelman
Building, 2120 L Street, NW., Washington, DC and at the local public
document room for the particular facility involved. If a request for a
hearing or petition for leave to intervene is filed by the above date,
the Commission or an Atomic Safety and Licensing Board, designated by
the Commission or by the Chairman of the Atomic Safety and Licensing
Board Panel, will rule on the request and/or petition; and the
Secretary or the designated Atomic Safety and Licensing Board will
issue a notice of a hearing or an appropriate order.
As required by 10 CFR 2.714, a petition for leave to intervene
shall set forth with particularity the interest of the petitioner in
the proceeding, and how that interest may be affected by the results of
the proceeding. The petition should specifically explain the reasons
why intervention should be permitted with particular reference to the
following factors: (1) the nature of the petitioner's right under the
Act to be made a party to the proceeding; (2) the nature and extent of
the petitioner's property, financial, or other interest in the
proceeding; and (3) the possible effect of any order which may be
entered in the proceeding on the petitioner's interest. The petition
should also identify the specific aspect(s) of the subject matter of
the proceeding as to which petitioner wishes to intervene. Any person
who has filed a petition for leave to intervene or who has been
admitted as a party may amend the petition without requesting leave of
the Board up to 15 days prior to the first prehearing conference
scheduled in the proceeding, but such an amended petition must satisfy
the specificity requirements described above.
Not later than 15 days prior to the first prehearing conference
scheduled in the proceeding, a petitioner shall file a supplement to
the petition to intervene which must include a list of the contentions
which are sought to be litigated in the matter. Each contention must
consist of a specific statement of the issue of law or fact to be
raised or controverted. In addition, the petitioner shall provide a
brief explanation of the bases of the contention and a concise
statement of the alleged facts or expert opinion which support the
contention and on which the petitioner intends to rely in proving the
contention at the hearing. The petitioner must also provide references
to those specific sources and documents of which the petitioner is
aware and on which the petitioner intends to rely to establish those
facts or expert opinion. Petitioner must provide sufficient information
to show that a genuine dispute exists with the applicant on a material
issue of law or fact. Contentions shall be limited to matters within
the scope of the amendment under consideration. The contention must be
one which, if proven, would entitle the petitioner to relief. A
petitioner who fails to file such a supplement which satisfies these
requirements with respect to at least one contention will not be
permitted to participate as a party.
Those permitted to intervene become parties to the proceeding,
subject to any limitations in the order granting leave to intervene,
and have the opportunity to participate fully in the conduct of the
hearing, including the opportunity to present evidence and cross-
examine witnesses.
If a hearing is requested, the Commission will make a final
determination on the issue of no significant hazards consideration. The
final determination will serve to decide when the hearing is held.
If the final determination is that the amendment request involves
no significant hazards consideration, the Commission may issue the
amendment and make it immediately effective, notwithstanding the
request for a hearing. Any hearing held would take place after issuance
of the amendment.
If the final determination is that the amendment request involves a
significant hazards consideration, any hearing held would take place
before the issuance of any amendment.
A request for a hearing or a petition for leave to intervene must
be filed with the Secretary of the Commission, U.S. Nuclear Regulatory
Commission, Washington, DC 20555-0001, Attention: Rulemakings and
Adjudications Staff, or may be delivered to the Commission's Public
Document Room, the Gelman Building, 2120 L Street, NW., Washington DC,
by the above date. A copy of the petition should also be sent to the
Office of the General Counsel, U.S. Nuclear Regulatory Commission,
Washington, DC 20555-0001, and to the attorney for the licensee.
Nontimely filings of petitions for leave to intervene, amended
petitions, supplemental petitions and/or requests for a hearing will
not be entertained absent a determination by the Commission, the
presiding officer or the Atomic Safety and Licensing Board that the
petition and/or request should be granted based upon a balancing of
factors specified in 10 CFR 2.714(a)(1)(i)-(v) and 2.714(d).
For further details with respect to this action, see the
application for amendment which is available for public inspection at
the Commission's Public Document Room, the Gelman Building, 2120 L
Street, NW., Washington, DC, and at the local public document room for
the particular facility involved.
Duke Energy Corporation, Docket Nos. 50-269, 50-270, and 50-287,
Oconee Nuclear Station, Units 1, 2, and 3, Oconee County, South
Carolina
Date of amendment request: July 15, 1997
Description of amendment request: The proposed amendments would:
(1) add Technical Specification (TS) 3.5.7, ``Main Steam Line Break
Detection and Feedwater Isolation,'' to identify operability
requirements and Bases for the main steamline break (MSLB) detection
isolation circuitry, the feedwater isolation circuitry, the main
feedwater main control valves, and the main feedwater startup control
valves; (2) revise TS 3.5.1, ``Operation Safety Instrumentation'' to
add a reference to TS 3.5.7; (3) revise Table 3.5.1-1, ``Instruments
Operating Conditions,'' to reflect operability requirements for the
main steam header pressure and MSLB detection channels, the feedwater
isolation channels, and the feedwater isolation channels manual
pushbuttons; and (4) revise Table 4.1-1, ``Instrument Surveillance
Requirements,'' and Table 4.1-2, ``Minimum Equipment Test Frequency,''
to include surveillance requirements for the subject circuitry and
components.
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:
[[Page 50002]]
A. Involve a significant increase in the probability or
consequences of an accident previously evaluated?
NO
This proposed Technical Specification amendment does not create
any conditions or events which lead to accidents (events) previously
evaluated in the UFSAR [Updated Final Safety Analysis Report], other
than a loss of Main Feedwater (FDW). The new MSLB detection and
feedwater isolation circuitry addressed by this change is designed
so that a credible single failure will not cause a loss of FDW to
the steam generator unless [an] MSLB is detected. Single failures
are not assumed if entry into a Technical Specification action
statement occurs.
During [an] MSLB, the circuitry is intentionally stopping and
isolating FDW. Operators are currently instructed to isolate FDW on
indication of [an] MSLB. The new circuitry will automatically stop
FDW to eliminate the need for this operator action. Thus the
probability of the stopping (loss) of FDW is not increased. The NRC
has also stated that the stopping of FDW to mitigate [an] MSLB is an
acceptable response to address the concerns of Inspection and
Enforcement Bulletin 80-04.
The Emergency Feedwater (EFW) System is an accident mitigation
system. The MSLB modification and associated Technical Specification
to keep the turbine driven emergency feedwater pump (TDEFW) pump
from starting following [an] MSLB will not initiate any accidents.
The potential for containment overpressurization currently
exists without the installed modification and associated Technical
Specification. The new MSLB detection and feedwater isolation
circuitry will assist in reducing the potential for the
overpressurization of containment. The EFW circuitry is designed so
that the TDEFWP will still auto start for any event other than [an]
MSLB. The TDEFWP can still be manually started during [an] MSLB or
FDW line break accident as needed. This action is similar to other
manual actions to align EFW for the MSLB scenarios that are already
described in the ONS [Oconee Nuclear Station] UFSAR. This new
circuitry and associated Technical Specification creates no new
credible single failures that could prevent the TDEFWP from auto
starting (except for the MSLB). The motor driven EFW pumps and EFW
flow control valves are not adversely affected by this change and
will provide EFW flow for scenarios other than Station Blackout.
Both FDW and EFW will still provide their design functions of
supplying feedwater to the steam generators, as evaluated in the
UFSAR. The ability to shut down following a 10CFR50 Appendix R fire
is not adversely affected. This Technical Specification change does
not adversely affect containment integrity and radiological release
pathways.
B. Create the possibility of a new or different kind of accident
from the accident previously evaluated?
NO
No accidents different than already evaluated in the UFSAR are
postulated. The FDW System will still perform its design function of
supplying feedwater to the steam generators as evaluated in the
UFSAR. The EFW System will still provide its function of supplying
feedwater to the steam generators, as evaluated in the UFSAR for
events resulting in the loss of the FDW System.
C. Involve a significant reduction in a margin of safety?
NO
The design pressure of containment is specified to be 59 psig in
the bases to several Technical Specifications. With the potential
for unrestricted FDW and EFW flow during [an] MSLB inside
containment, the design pressure of the containment could be
exceeded. The proposed Technical Specifications address equipment
which will function to isolate FDW in the unlikely event of [an]
MSLB accident. Therefore, the proposed Technical Specifications do
not increase the potential for the containment to be pressurized or
increase the expected pressure of containment following [an] MSLB.
No plant safety limits, set points, or design parameters are
adversely affected. The fuel, fuel cladding, and Reactor Coolant
System are not impacted.
Duke [Duke Energy Corporation] has concluded based on the above
that there are no significant hazards considerations involved in
this amendment request.
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: Oconee County Library, 501
West South Broad Street, Walhalla, South Carolina 29691
Attorney for licensee: J. Michael McGarry, III, Winston and Strawn,
1200 17th Street, NW., Washington, DC 20036
NRC Project Director: Herbert N. Berkow
Duke Power Company, Docket Nos. 50-269, 270 and 50-287, Oconee
Nuclear Station, Units 1, 2 and 3, Oconee County, South Carolina
Date of amendment request: August 28, 1997 (TSC 96-09)
Description of amendment request: The proposed changes would add
new limiting conditions for operation and new surveillance requirements
for the Emergency Condenser Circulating Water System, the Essential
Siphon Vacuum System, and the Siphon Seal Water System to reflect
design changes and modifications to these systems.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
[1. Will the change] involve a significant increase in the
probability or consequences of an accident previously evaluated?
NO.
This Technical Specification change does not create any
conditions or events which lead to accidents previously evaluated in
the UFSAR [Updated Final Safety Analysis Report]. The new ECCW
[Emergency Condenser Circulating Water] System Technical
Specification 3.19, along with the new ECCW Surveillance
requirements specified in Technical Specification Table 4.1-2, are
conservative in nature. No existing Technical Specification
requirements are being deleted with this revision. Surveillance and
operability requirements are being added for the upgraded ECCW
System.
The ECCW System is only required following the occurrence of
loss of offsite power (LOOP) events. The most limiting of these LOOP
events is the loss of coolant accident concurrent with the LOOP
(LOCA/LOOP). Therefore, the ECCW System is not considered to be an
accident initiator. As a result, the proposed new ECCW Technical
Specification requirements will not result in any increase in the
probability of any design basis accidents or events evaluated in the
UFSAR.
The credit for restarting a CCW [Condenser Circulating Water]
pump within 1.5 hours following a LOOP, to ensure suction to LPSW
[Low Pressure Service Water] is maintained, is being replaced by
credit for maintaining the ECCW siphon using the new siphon support
systems (ESV [Essential Siphon Vacuum] System and SSW [Siphon Seal
Water] System) in conjunction with the upgraded ECCW System.
Therefore, obsolete requirements specified in Selected Licensee
Commitments (SLCs) 16.9.7 and 16.9.8 will be revised or deleted
accordingly. Replacement of the CCW pump restart during a LOOP with
the ability to maintain ECCW siphon flow will not create any
conditions or events which lead to accidents previously evaluated in
the UFSAR.
The modifications to upgrade the ECCW System were performed to
improve the reliability of the ECCW System. The proposed new ECCW
Technical Specification provides additional surveillance and
operability requirements to ensure that the upgraded ECCW System
will function reliably during the design basis events which require
its operation. Therefore, these proposed new Technical Specification
requirements will not increase the consequences of any accidents
previously evaluated in the UFSAR.
[2. Will the change] create the possibility of a new or
different kind of accident from the accident previously evaluated?
NO.
No accidents different than those already evaluated in the UFSAR
are postulated. The upgraded ECCW System will more reliably perform
its design function of supplying water to the suction of the Low
Pressure Service Water (LPSW) System as evaluated in the UFSAR. The
new Technical Specification requirements will increase the
reliability of the upgraded ECCW System. In addition, the ECCW
System is not an accident initiator since it is used following
certain design basis events such as a LOCA/LOOP.
[[Page 50003]]
[3. Will the change] involve a significant reduction in a margin
of safety?
NO.
The proposed Technical Specifications address equipment which
will function in certain design basis events, such as a LOCA/LOOP,
to ensure a reliable water supply to the LPSW System. The LPSW
System must function to remove decay heat from primary systems and
the reactor building during a LOCA/LOOP. The proposed Technical
Specifications addressing the upgraded ECCW System will further
enhance the reliability of the ECCW System and will result in
greater assurance that the LPSW System can perform its safety
functions. No plant safety limits, setpoints, or design parameters
are adversely affected. The fuel, fuel cladding, and Reactor Coolant
System are not impacted. The proposed Technical Specifications
provide additional, conservative, operational requirements beyond
the current Technical Specifications which address the ECCW System.
Duke [Duke Energy Corporation] has concluded based on this
information that there are no significant hazards considerations
involved in this amendment request.
The NRC has reviewed the licensee's analysis and, based on
thisreview, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: Oconee County Library, 501
West South Broad Street, Walhalla, South Carolina 29691
Attorney for licensee: J. Michael McGarry, III, Winston and Strawn,
1200 17th Street, NW., Washington, DC 20036
NRC Project Director: Herbert N. Berkow
Duke Energy Corporation, Docket Nos. 50-269, 50-270, and 50-287,
Oconee Nuclear Station, Units 1, 2, and 3, Oconee County, South
Carolina
Date of amendment request: September 4, 1997
Description of amendment request: The proposed changes would
incorporate changes to the Oconee Final Safety Analysis Report and
Technical Specification Bases to address a potential unreviewed safety
question associated with implementation of revised small break loss-of-
coolant accident analysis. The proposed changes would address operation
of the facility and single failure criteria related to the high
pressure injection 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) Involve a significant increase in the probability or
consequences of an accident previously evaluated:
No. None of the proposed changes [have] any impact upon the
probability of any accident which has been evaluated in the UFSAR
[Updated Final Safety Analysis Report].
None of these changes have any impact upon the ability of the
HPI [high-pressure injection] System to mitigate the consequences of
a small break LOCA [loss-of-coolant accident], which is addressed
below. The small break LOCA is the limiting design basis accident
with respect to the HPI System operability requirements.
The proposed changes to the Bases of Specification 3.3.1 and
Chapter 15 of the Oconee UFSAR include operator actions that have
not previously been reviewed and approved by the [NRC] staff for
licensing basis small break LOCA analyses. However, these operator
actions have been included in the Emergency Operating Procedure for
over 10 years and crediting these actions in the safety analyses
does not result in any change to the operator's response to a small
break LOCA. These actions are simply changes to the assumptions
contained in the licensing basis small break LOCA analyses. The
operability requirements for the HPI System contained in
Specification 3.3.1 are supported by a spectrum of small break LOCA
analyses based on the approved Evaluation Model described in FTI
[Framatome Technologies, Inc.] topical report BAW-10192P. These
small break LOCA analyses demonstrate that the acceptance criteria
of 10CFR 50.46 are satisfied.
The operability requirements in Technical Specification 3.3.1.c
assure that the HPI System can withstand the worst single failure
and still result in two HPI pumps injecting through two trains. The
full power small break LOCA analyses supporting this proposed
license amendment have been performed in accordance with the
approved Evaluation Model described in FTI topical report BAW-
10192P.
When at or below 75% FP [full power], one HPI train provides
sufficient flow to mitigate a small break LOCA. The 60% power level
currently in Specification 3.3.1 is justified by analyses using the
Evaluation Model described in FTI topical report BAW-10192P,
considering the worst case break location and size described in LER
[Licensee Event Report] 269/90-15 and Attachment 2 to this
submittal. The proposed changes to the Bases of Technical
Specification 3.3.1 describe the operator actions credited to
justify the adequacy of the current specification and eliminate the
need for the administrative restrictions imposed by LER 269/90-15.
These requirements ensure that, following the worst single failure,
one train of HPI would remain available to mitigate a small break
LOCA.
In summary, the technical analyses described in this license
amendment justify the adequacy of this specification and assure that
operability of the HPI System is maintained in a manner consistent
with the requirements of the design basis accidents. Therefore, it
is concluded that this amendment request will not significantly
increase the probability or consequences of an accident previously
evaluated.
(2) Create the possibility of a new or different kind of
accident from any kind of accident previously evaluated:
No. The proposed changes to the Bases of Technical Specification
3.3.1 and Chapter 15 of the Oconee UFSAR do not result in any new
operator actions or changes in plant operation. The proposed changes
involve crediting operator actions in the licensing basis small
break LOCA analyses that have been included in the Emergency
Operating Procedure for years. No new initiating events or
potentially unanalyzed conditions have been created. Therefore, this
proposed amendment will not create the possibility of any new or
different kind of accident.
(3) Involve a significant reduction in a margin of safety.
No. The HPI System requirements associated with the proposed
UFSAR and Technical Specification Bases changes are supported by
analyses which demonstrate that the acceptance criteria of 10 CFR
50.46 are not violated for any small break LOCA. These analyses were
performed in accordance with the Evaluation Model described in FTI
topical report BAW-10192P. Therefore, it is concluded that the
proposed amendment request will not result in a significant decrease
in the margin of safety.
Duke [Duke Energy Corporation] has concluded, based on the
above, that there are no significant hazards considerations involved
in this amendment request.
The NRC 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: Oconee County Library, 501
West South Broad Street, Walhalla, South Carolina 29691
Attorney for licensee: J. Michael McGarry, III, Winston and Strawn,
1200 17th Street, NW., Washington, DC 20036
NRC Project Director: Herbert N. Berkow
Entergy Operations, Inc., et al., Docket No. 50-416, Grand Gulf
Nuclear Station, Unit 1, Claiborne County, Mississippi
Date of amendment request: August 6, 1997
Description of amendment request: The proposed amendment would
eliminate the provisions in Technical Specification 3.8.1, ``AC Sources
- Operating,'' for accelerated testing of the emergency diesel
generators (DG). The proposed changes are the following: (1) the
frequency of verifying DG starts and operation in Surveillance
Requirements 3.8.1.2 and 3.8.1.3, respectively, would be changed to 31
days, from the present reference to Table 3.8.1-1, and (2) Table 3.8.1-
1, ``Diesel Generator Test
[[Page 50004]]
Schedule,'' would be deleted. The emergency DG provide emergency AC
power to the site with the loss of offsite AC power.
Basis for proposed no significant hazardsconsideration
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. This request does not involve a significant increase in the
probability or consequences of an accident previously evaluated.
[These] change[s] will provide flexibility to structure the
standby diesel generator maintenance program based on the risk
significance of the structures, systems, and components [(SSCs)]
that are within the scope of the Maintenance Rule [(10 CFR 50.65,
``Requirements for Monitoring the Effectiveness of Maintenance at
Nuclear Power Plants)]. The removal of the diesel generator
accelerated testing is acceptable as the maintenance rule applies
site and system specific performance criteria to monitor diesel
generator performance. This criteria includes a running availability
and reliability goal as well as specific goals to monitor
maintenance preventable functional failures. The performance
criteria for the diesel generator reliability and availability
established by the maintenance rule and the causal determinations
and corrective actions required for maintenance preventable
functional failures are considered to be an acceptable method for
monitoring diesel generator performance.
The proposed change[s] [have] no effect on the probability of
the initiation of an accident, because the emergency diesel
generators do not serve as the initiator of any event. Additionally,
as diesel generator performance will continue to be [ensured] by the
maintenance rule, the proposed changes do not affect the ability to
mitigate the consequences of an accident previously evaluated. The
changes do not impact the diesel [generator]'s design sources,
operating characteristics, system functions, or system
interrelationships. The failure mechanisms for the accident
previously evaluated are not affected and no additional failure
modes are created that could cause an accident that has been
previously evaluated. Since the diesel generator performance and
reliability will continue to be [ensured] by the maintenance rule,
the proposed changes do not involve a significant increase in the
probability or consequences of an accident previously evaluated.
2. This request does not create the possibility of a new or
different kind of accident from any accident previously evaluated.
[These] proposed change[s] [do] not involve a change to the
plant design or operation. As a result, the proposed change[s] [do]
not affect any of the parameters or conditions that could contribute
to the initiation of any accidents. The proposed changes only affect
the methods used to monitor and [ensure] diesel generator
performance. The performance criteria for both the diesel generator
reliability and unavailability established by the maintenance rule,
and the causal determinations and corrective actions required for
maintenance preventable functional failures, [are] considered by
[the Nuclear Regulatory Commission (NRC) in] GL [(Generic Letter)]
94-01[, ``Removal of Accelerated Testing and Special Reporting
Requirements for Emergency Diesel Generators,'' issued May 31,
1994,] to be an acceptable method for monitoring diesel generator
performance.
No SSC, method of operation, or system interface is altered by
[these] change[s]. The changes do not impact the diesel
[generator]'s design sources, operating characteristics, system
functions, or system interrelationships. The failure mechanisms for
the accidents are not affected, and no additional failure modes are
created. Because the diesel generator performance and reliability
will continue to be [ensured] by the maintenance rule, the proposed
changes do not create the possibility of a new or different kind of
accident from any accident previously evaluated.
3. This request does not involve a significant reduction in a
margin [of] safety.
The proposed changes only affect the methods used to monitor and
[ensure] diesel generator performance and reliability. The
performance criteria for the diesel generator reliability and
availability established by the maintenance rule, and the causal
determinations and corrective actions required for maintenance
preventable functional failures, [are] considered by [NRC in] GL 94-
01 to be an acceptable method for monitoring diesel generator
performance. No margin [of] safety as defined in the bases for any
technical specification is impacted by these changes. [These]
change[s] [do] not impact any uncertainty in the design,
construction, or operation of any SSC. Diesel generator response to
accident initiators is unchanged. No SSC, method of operating, or
system interface is altered by [these] change[s]. The changes do not
impact the diesel [generator]'s design sources, operating
characteristics, system functions, or system interrelationships.
Because the diesel generator performance and reliability will
continue to be [ensured] by the maintenance rule, the proposed
changes do not involve a significant reduction in the margin [of]
safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: Judge George W. Armstrong
Library, 220 S. Commerce Street, Natchez, MS 39120
Attorney for licensee: Nicholas S. Reynolds, Esquire, Winston and
Strawn, 1400 L Street, N.W., 12th Floor, Washington, DC 20005-3502
NRC Project Director: James W. Clifford, Acting
Florida Power Corporation, et al., Docket No. 50-302, Crystal River
Nuclear Generating Plant, Unit No. 3, Citrus County, Florida
Date of amendment request: August 26, 1997
Description of amendment request: The proposed amendment would
revise the Crystal River Unit 3 (CR3) Technical Specifications Bases
(TSB) to change the design basis of the Emergency Diesel Generator
(EDG) Air Handling System. Specifically, TSB Sections B 3.8.1 and B
3.8.2 would be revised to indicate that a single or dual fan operation
depending upon fan supply air temperature, would maintain the
temperature of the EDG engine and control rooms within the EDG
manufacturer's limits.
Basis for proposed no significant hazardsconsideration
determination:
The EDG Air Handling System provides continuous ventilation, and
dissipates internal heat gains in the EDG engine and control rooms when
the diesel is operating. Presently, the CR3 plant documentation
requires operation of only one cooling fan per room to maintain the EDG
room temperature within the manufacturer's limit and is inconsistent
with the Final Safety Analysis Report (FSAR) which requires operation
of two fans.
As part of its EDG upgrade to increase their service ratings and
associated cooling analysis, the licensee has determined that operation
of either a single or dual cooling fans depending upon fan supply air
temperature, would achieve the required room cooling limits. The
licensee has determined that reliance on the operation of two cooling
fans instead of one involves an unreviewed safety question and requires
a license amendment.
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 not involve a significant increase in the probability or
consequences of an accident previously evaluated.
The proposed change does not involve a significant increase in
the probability of an accident previously evaluated. The EDG room
cooling fans support operation of the EDGs which are used to
mitigate design basis accidents. Although EDG availability is a
contributor to the risk of station blackout, the CR-3 licensing
basis assumes a station blackout without regard to EDG reliability.
Therefore, the probability of previously evaluated accidents is not
significantly increased.
For design basis accidents, the proposed change does not involve
a significant increase in the consequences of an accident previously
evaluated. The proposed change to operate both cooling fans for each
EDG to
[[Page 50005]]
provide adequate ventilation potentially increases the probability
of malfunction of equipment important to safety. However, the
proposed changes do not affect the independence of the EDGs or the
independence of the EDG Air Handling System and, based on single
failure criteria, one EDG will be fully operable and capable of
meeting its mission at all times as required by the CR-3 Technical
Specifications. Therefore, no significant increase in the
consequences of an accident previously evaluated, including the
offsite radiological dose exists.
Based on the above, the probability of an accident previously
evaluated has not been significantly increased, and this change does
not involve a significant increase in the consequences of an
accident previously evaluated.
2. Does not create the possibility of a new or different kind of
accident from any accident previously evaluated
The proposed change does not create the possibility of a new or
different kind of accident from any accident previously evaluated.
Neither the fans nor the EDGs are initiators of any new accidents.
The EDG room cooling fans support operation of the EDGs, which are
used to mitigate design basis accidents. Reliance on two fans rather
than one has reduced the redundancy of the EDG Air Handling System
and increased the probability of a malfunction of an EDG. However,
the proposed changes do not affect the independence of the EDGs or
the independence of the EDG Air Handling System and, based on single
failure criteria, one EDG will be fully operable and capable of
meeting its mission at all times as required by the CR-3 Technical
Specifications. Results of analyses to evaluate the failure of an
EDG to operate following a design basis accident are documented in
the FSAR. Therefore, this change does not create the possibility of
a new or different kind of accident.
3. Does not involve a significant reduction in the margin of
safety
The proposed change does not involve a significant reduction in
the margin of safety. The EDG room cooling fans support operation of
the EDGs. Following this change, two fans will be required to
maintain the EDG engine room and EDG control room temperatures
within the design basis limit when the fan supply air temperature is
greater than or equal to 85 deg.F. Reliance on two fans rather than
one has reduced the redundancy of the EDG Air Handling System and
slightly increased the probability of malfunction of an EDG, but
only after it has run for some period of time. However, the proposed
changes do not affect the independence of the EDGs or the
independence of the EDG Air Handling System and, based on single
failure criteria, one EDG will be fully operable and capable of
meeting its mission at all times as required by the CR-3 Technical
Specifications. Therefore, this change does not result in a
significant reduction to the margin 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: Coastal Region Library, 8619
W. Crystal Street, Crystal River, Florida 34428
Attorney for licensee: R. Alexander Glenn, General Counsel, Florida
Power Corporation, MAC - A5A, P. O. Box 14042, St. Petersburg, Florida
33733-4042
NRC Project Director: Frederick J. Hebdon
Florida Power Corporation, et al., Docket No. 50-302, Crystal River
Nuclear Generating Plant, Unit No. 3, Citrus County, Florida
Date of amendment request: September 9, 1997
Description of amendment request: The proposed amendment would
revise the Crystal River 3 (CR3) Final Safety Analysis Report (FSAR) to
reflect the revised analysis for the hypothetical Makeup System Letdown
Line Failure Accident. In the original analysis, the event was modeled
as being terminated by an automatic isolation of the failed letdown
line on low reactor coolant system pressure. The revised analysis has
modeled the event as being terminated by manual operator action to
isolate the line. The licensee has determined that reliance on a manual
operator action in place of the automatic action involves an unreviewed
safety question (USQ) and requires prior Nuclear Regulatory Commission
(NRC) approval. Other FSAR changes are being proposed to clarify that
this accident is a hypothetical event that is presented only to
demonstrate that the dose consequences are below 10 CFR Part 100
limits. The licensee submitted its proposed FSAR changes which, upon
NRC approval, will be incorporated in the next revision to the FSAR.
Basis for proposed no significant hazardsconsideration
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 not involve a significant increase in the probability or
consequences of an accident previously evaluated.
This change involves a revision to the analysis for the Makeup
System Letdown Line Failure Accident. The revised analysis assesses
the resultant change in consequences of this event based on the
actions specified in EOP-3 [Emergency Operating Procedure - 3] to
manually isolate the letdown line failure. No changes have been made
to any precursors to this event. Therefore, the probability of an
accident previously evaluated has not been increased.
This change has resulted in an increase in the calculated doses
due to the greater release of reactor coolant prior to termination
of the leak. Although the doses have increased, they remain
significantly less than the limits of 10 CFR 100. These doses also
remain lower than the resultant doses for the design basis LOCA
[loss-of-coolant-accident].
The revised analysis evaluates the consequences of this accident
based on the replacement of the automatic isolation of the letdown
line with a manual operator action to isolate the letdown line. This
action was added to EOP-3 when it was identified that the manual
initiation of the HPI [high pressure injection] system directed by
the EOP would interfere with the automatic isolation signal assumed
to terminate this event. Manual initiation of the HPI system for a
LSCM [loss of subcooling margin] event is consistent with the
symptomatic philosophy of the EOPs. This philosophy is utilized in
order to manage a wide range of event/leaks that would be indicated
by a LSCM. Early initiation of the HPI system is intended to ensure
adequate core cooling as the primary concern during a LSCM event.
Prior to the addition of the EOP step to manually isolate the
letdown line, the EOP directed actions towards locating and
isolating the source of the leak resulting in the LSCM. However, due
to the potential significance of the letdown line failure which can
result in RCS [reactor coolant system] leakage outside the reactor
building, the manual action was added early in EOP-3 to isolate the
letdown line. This action is proactive in ensuring early isolation
of the potential leakage path and is consistent with the concept of
a ``simple'' operator action (Reference 9) [NRC to Florida Power
Corporation letter, Long-term modifications regarding emergency core
cooling system Small Break Analysis problem, dated September 26,
1978].
Crediting a manual operator action instead of the automatic
isolation introduces the possibility of a malfunction of a different
type (i.e., operator error). The revised analysis assumes that
operator action to isolate the letdown line occurs 10 minutes
following a LSCM. Although the probability of operator error during
this action may be greater than the probability of the failure of
the automatic function, the consequences of this error would be
small. Several indications would be available to the operator to
identify the continued loss of coolant through this line. As
discussed above, the radiological dose calculated by this event
remains a small fraction of the limits of 10 CFR Part 100.
Therefore, adequate time would exist for the identification of an
operator error and correction of this error before any significant
increase in the consequences of this event would occur.
Additionally, the probability for operator error in this event
is considered to be small due to the extensive training plant
operators receive regarding the EOPs and the simple nature of the
action. Validation of the required actions in the EOPs, including
isolation of the letdown line, is performed on the plant simulator
to ensure the validity of the EOPs as well as to ensure that these
actions can be performed as required.
[[Page 50006]]
The clarification added to FSAR Section 5.4.4.2 and 14.2.2.6.1
reflects the previously approved evaluation for pipe rupture
criteria outside the reactor building for CR-3. A break in the high
energy portion of the letdown line outside containment is not
considered a credible event. This accident is presented only to
demonstrate that the dose consequences from a postulated break in
the letdown line outside containment remain below the 10 CFR Part
100 limits.
Based on the above, this change does not involve a significant
increase in the consequences of an accident previously evaluated.
2. Does not create the possibility of a new or different kind of
accident from any accident previously evaluated.
This change does not involve any modification to the plant nor a
change in the operation of the plant prior to the postulated failure
of the letdown line. This change only evaluates the radiological
dose consequences of the actions taken following the line failure.
The addition of the action to manually isolate the letdown line for
a LSCM event is consistent with the need to isolate potential RCS
leakage paths and replaces the automatic isolation that was
previously assumed to occur. Therefore, this change does not create
the possibility of a new or different kind of accident.
3. Does not involve a significant reduction in the margin of
safety.
This change does not result in a reduction to the margin of
safety as defined in the Bases for any Technical Specifications. As
discussed above, the radiological doses for the revised analysis
have increased but remain a small fraction of the 10 CFR Part 100
limits.
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: Coastal Region Library, 8619
W. Crystal Street, Crystal River, Florida 34428
Attorney for licensee: R. Alexander Glenn, General Counsel, Florida
Power Corporation, MAC - A5A, P. O. Box 14042, St. Petersburg, Florida
33733-4042
NRC Project Director: Frederick J. Hebdon
Florida Power and Light Company, et al., Docket Nos. 50-335 and 50-
389, St. Lucie Plant, Unit Nos. 1 and 2, St. Lucie County, Florida
Date of amendment request: August 22, 1997
Description of amendment request: The proposed amendment revises
Technical Specification (TS) 4.0.5, Surveillance Requirements for
Inservice Inspection and Testing of ASME Code Class 1, 2, and 3
components, to relocate the Inservice Testing Program requirements from
TS 4.0.5 to the Administrative Controls Section 6.8, Procedures and
Programs. The proposed amendment also provides conforming changes to
several Surveillance Requirements to change the reference from TS 4.0.5
to the Inservice Testing Program.
Basis for proposed no significant hazardsconsideration
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.
There are no changes to the testing and evaluation related to pumps
and valves in the Inservice Testing Program. The only substantive
change allows the implementation of alternate testing provisions
where Code-requirements are impractical and the NRC has not formally
provided written approval. Since impractical testing would not be
performed in any event, the actual testing program is unaffected.
(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 cannot 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
implementation of this administrative change 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 remain unchanged by the proposed
amendments, therefore, these changes do not involve a significant
reduction in the margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: Indian River Community College
Library, 3209 Virginia Avenue, Fort Pierce, Florida 34981-5596
Attorney for licensee: M.S. Ross, Attorney, Florida Power & Light,
P.O. Box 14000, Juno Beach, Florida 33408-0420
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: August 27, 1997
Description of amendment request: The licensee proposed modifying
the Turkey Point Units 3 and 4 Technical Specifications (TS) to delete
a sentence from section 6.2.2.f and add clarification to section
6.2.2.f of the Administrative section of TS to allow the use of up to
12 hour shifts without routine heavy use of overtime.
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 change does not involve a physical or procedural
change to any structure, system or component that significantly
effects the probability or consequences of any accident or
malfunction of equipment important to safety. The proposed changes
will allow the use of 12 hour shifts for a nominal 40 hours per
week.
This change is only administrative in nature and has no
significant impact on the probabilities or consequences of any
evaluated accident or malfunction of equipment important to safety.
(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 amendment will not change the physical plant or
modes of plant operation defined in the Turkey Point Units 3 and 4
operating license. The proposed amendment will not involve addition
or modification of permanent equipment for any systems structures or
components at Turkey Point.
The change does modify the controls on working shift hours for
operating personnel without significantly changing the hours worked
per week and retains the current limitations on excessive overtime.
The changes are administrative in nature.
Consequently, operation of either unit in accordance with the
proposed amendment would not create the possibility of a new or
different kind of accident from any accident previously evaluated.
[[Page 50007]]
(3) Operation of the facility in accordance with the proposed
amendments would not involve a significant reduction in a margin of
safety.
The proposed amendment will allow the use of 12 hour shifts by
virtue of the administrative change. This will result in fewer
turnovers per day and will allow more contiguous days off between
work shifts. The sum of these 12 hour work shift features will be
more rested crews with better communications between shifts. The
proposed change will not alter the basis for any Technical
Specification that is related to the establishment of, or
maintenance of, a nuclear safety margin.
Consequently, operation of Turkey Point Units 3 and 4 in
accordance with this proposed amendment would not involve a
significant reduction in 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
Houston Lighting & Power Company, City Public Service Board of San
Antonio, Central Power and Light Company, City of Austin, Texas,
Docket Nos. 50-498 and 50-499, South Texas Project, Units 1 and 2,
Matagorda County, Texas
Date of amendment request: August 18, 1997
Description of amendment request: The proposed amendment would
revise Technical Specification 3.7.1.6, Atmospheric Steam Relief
Valves, to ensure the automatic feature of the steam generator power
operated relief valve remains operable during Modes 1 and 2. In
addition, the proposed change adds a surveillance requiring that a
channel calibration on the steam generator power operated relief valve
be performed every 18 months.
Basis for proposed no significant hazardsconsideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
The South Texas Project proposed to revise Technical
Specification 3.7.1.6 to ensure the automatic feature of the Steam
Generator Power Operated Relief Valve remains operable during Modes
1 and 2. The South Texas Project has evaluated this proposed
amendment and determined that it involves no significant hazards
considerations based on the following:
A. The proposed change does not involve a significant increase
in the probability or consequences of an accident previously
evaluated.
The methodologies used in the accident analyses remain
unchanged. The automatic actuation of the Steam Generator Power
Operated Relief Valves is not a new design feature. The effects of
the inadvertent opening of a Steam Generator Power Operated Relief
Valve are currently analyzed as described in Section 15.1.4 of the
Updated Final Safety Analysis Report. The radiological consequences
for the SBLOCA [small-break loss-of-coolant accident] event
presented in the Updated Final Safety Analysis Report remain
unchanged. The calculated Peak Clad Temperature remains
substantially below the 2200 deg.F acceptance limit of
10[]CFR[]50.46.
B. The proposed change does not create the possibility of a new
or different kind of accident from any accident previously
evaluated.
The automatic actuation of the steam generator power operated
relief valves is not an accident initiator for the SBLOCA event. The
automatic actuation of the steam generator power operated relief
valves currently exists at the South Texas Project and is not a new
design feature. The description of the Steam Generator Power
Operated Relief Valves currently exists in the Updated Final Safety
Analysis Report. This change does not represent a change to the
facility and does not affect the safety functions and reliability of
systems, structures, or components in any new manner. Operating
procedures have a temporary administrative control to ensure the
automatic actuation of the Steam Generator Power Operated Relief
Valves remains operable in Modes 1 and 2. This condition will become
permanent with the approval of the Technical Specification Amendment
proposal.
C. The proposed change does not involve a significant reduction
in a margin of safety.
The proposed change results in the calculated Peak Clad
Temperature remaining well below the acceptance limit of
10[]CFR[]50.46 and comparable to the results currently described in
the Updated Final Safety Analysis Report.
Therefore, the South Texas Project has concluded that the
proposed change does not involve a significant hazards
considerations.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
request for amendments involves no significant hazards consideration.
Local Public Document Room location: Wharton County Junior College,
J. M. Hodges Learning Center, 911 Boling Highway, Wharton, TX 77488
Attorney for licensee: Jack R. Newman, Esq., Morgan, Lewis &
Bockius, 1800 M Street, N.W., Washington, DC 20036-5869
NRC Project Director: James W. Clifford, Acting
Northeast Nuclear Energy Company, et al., Docket No. 50-336,
Millstone Nuclear Power Station, Unit No. 2, New London,
Connecticut
Date of amendment request: September 2, 1997
Description of amendment request: The proposed changes to the
Technical Specifications (TSs) would modify the maximum allowed
containment pressure specified in TS 3.6.1.4, ``Containment Systems
Internal Pressure,'' from 2.1 psig to 1.0 psig. The TS Bases, Section
3/4.6.1.4, would also be revised to reflect the new maximum allowed
containment pressure.
Basis for proposed no significant hazardsconsideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
The proposed change does not involve an SHC [significant hazards
consideration] because the changes would not:
1. Involve a significant increase in the probability or
consequences of an accident previously evaluated.
The proposed change will reduce the maximum allowed value for
containment pressure specified in Technical Specification 3.6.1.4,
``Containment Systems Internal Pressure.'' This change will improve
the margin between the peak containment pressure following a main
steam line break (most limiting accident for peak containment
pressure at Millstone Unit No. 2) and the containment design
pressure limit of 54 psig. Reducing the initial containment pressure
will result in a reduction in peak containment pressure.
To ensure the assumption of a lower initial containment pressure
is maintained, a change to Technical Specification 3.6.1.4 is
necessary.
The proposed change to Technical Specification 3.6.1.4 will
allow one of the initial assumptions used in the analysis for peak
containment pressure following a main steam line break to be
changed. However, this change will not affect how any of the plant
systems function to mitigate design basis accidents and will not
require any changes to mitigation procedures. The acceptance
criteria of a peak containment pressure less than the design limit
of 54 psig remains the same. Therefore, this change does not
significantly increase the probability or consequences of an
accident previously evaluated.
2. Create the possibility of a new or different kind of accident
from any accident previously evaluated.
The proposed change does not alter the way any structure,
system, or component functions and does not alter the manner in
which the plant is operated. It does not
[[Page 50008]]
introduce any new failure modes and conservatively alters an
assumption made in the main steam line break safety analysis.
Therefore, the change will not create the possibility of a new
or different kind of accident from any accident previously
evaluated.
3. Involve a significant reduction in a margin of safety.
This proposed change will reduce the maximum allowed value for
containment pressure specified in Technical Specification 3.6.1.4,
``Containment Systems Internal Pressure.'' This change will improve
the margin between the peak containment pressure following a main
steam line break (most limiting accident for peak containment
pressure at Millstone Unit No. 2) and the containment design
pressure limit 54 psig. Starting at a lower initial containment
pressure will result in a lower peak containment pressure. To ensure
the assumption of a lower initial containment pressure is
maintained, a change to Technical Specification 3.6.1.4 is
necessary.
This more restrictive change in the maximum allowed containment
pressure will result in the use of a lower initial containment
pressure in the analysis of a main steam line break accident.
However, the analysis acceptance criteria of a peak accident
containment pressure less than 54 psig, will remain the same.
Therefore, there is no significant reduction in a margin of safety
as defined in the Bases of Technical Specification 3.6.1.4.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: Learning Resources Center,
Three Rivers Community-Technical College, 574 New London Turnpike,
Norwich, CT 06360, and the Waterford Library, ATTN: Vince Juliano, 49
Rope Ferry Road, Waterford, CT 06385
Attorney for licensee: Lillian M. Cuoco, Esq., Senior Nuclear
Counsel, Northeast Utilities Service Company, P.O. Box 270, Hartford,
CT 06141-0270 NRC Deputy Director: Phillip F. McKee
Northeast Nuclear Energy Company, et al., Docket No. 50-336,
Millstone Nuclear Power Station, Unit No. 2, New London County,
Connecticut
Date of amendment request: September 2, 1997
Description of amendment request: The proposed amendment would
change the Technical Specifications (TSs) to: (1) Combine TS 3.6.2.1,
``Containment Spray System,'' and TS 3.6.2.2, ``Containment Air
Recirculation System,'' into one specification which would reduce the
allowed outage time for one inoperable containment spray (CS) train or
one inoperable containment air recirculation (CAR) cooler from 30 days
to 7 days; increase the allowed outage time for two inoperable CAR
coolers from 48 hours to 7 days; add an allowed outage time of 48 hours
(instead of entering TS 3.0.3) for one inoperable CS train and two
inoperable CAR coolers or three or four inoperable CAR coolers; provide
specific guidance on when to enter TS 3.0.3; and expand the applicable
TS Bases to reflect these changes; (2) Modify the definition of
containment integrity and TS 3.6.1.1, ``Containment Integrity,'' to
indicate that the operability of the automatic isolation valve system
is satisfied by the use of the containment isolation trip push buttons
in Mode 4, and expand the TS Bases to reflect these change; (3) Add an
exception to the reactor coolant flow rate surveillance requirement, TS
4.1.1.3, whenever there is a reduction in reactor coolant system
boration while in Modes 2 and 3 because the reactor coolant pumps are
required to be in operation; (4) Delete the reactor coolant system
leakage surveillance requirements, TS 4.4.6.2.a and TS 4.4.6.2.b, which
require monitoring the containment atmosphere particulate radioactivity
and containment sump inventory, respectively; (5) Modify emergency core
cooling system surveillance requirement, TS 4.5.2.e, to allow the use
of alternative methods to verify that the throttle valves in Table 4.5-
1 are in the correct position and expand the TS bases to address the
alternative methods; (6) Modify TS 5.5.1,'' Emergency Core Cooling
Systems,'' by deleting the word ``original'' since the design has been
modified; and (7) Make editorial changes to terminology and item
numbering.
Basis for proposed no significant hazardsconsideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
The proposed changes do not involve an SHC [significant hazards
consideration] because the changes would not:
1.Involve a significant increase in the probability or
consequences of an accident previously evaluated.
The proposed change to combine Technical Specifications 3.6.2.1
and 3.6.2.2 into one specification reduces the allowed outage time
for one inoperable containment spray (CS) train or one inoperable
containment air recirculation (CAR) cooler from 30 days to 7 days;
increases the allowed outage time for two inoperable CAR coolers
from 48 hours to 7 days; adds an allowed outage time of 48 hours
(instead of entering Technical Specification 3.0.3) for one
inoperable CS train and two inoperable CAR coolers, or three or four
inoperable CAR coolers; and provides specific guidance when it is
necessary to enter Technical Specification 3.0.3 will not affect how
these systems function to mitigate design basis accidents.
Therefore, this change does not significantly increase the
probability or consequences of an accident previously evaluated.
The proposed changes to modify the definition of containment
integrity, modify the Technical Specification 3.6.1.1, ``Containment
Integrity,'' and expand the Bases to explain why automatic
containment isolation valves are operable in Mode 4 have no affect
on any containment isolation valve or Engineered Safety Feature
Actuation System (ESFAS) component. These components will still
function as designed to mitigate design basis accidents. Therefore,
this change does not significantly increase the probability or
consequences of an accident previously evaluated.
The proposed change to provide an exception to Surveillance
Requirement 4.1.1.3 when the plant is in Modes 1 and 2 will not
result in any new approach to plant operation, it simply removes the
requirement to perform an unnecessary surveillance. The minimum
coolant flow through the core during a reduction in Reactor Coolant
System (RCS) boron concentration will still be met. Therefore, this
change does not significantly increase the probability or
consequences of an accident previously evaluated.
The proposed change to delete Surveillance Requirements (SRs)
4.4.6.2.a and 4.4.6.2.b does not reduce the operability requirements
for any equipment used to monitor RCS leakage. The equipment covered
by these 2 SRs, containment atmosphere particulate radioactivity
monitors and containment sump inventory monitor, provide early
indication that RCS leakage exists, but do not provide the specific
information (amount of leakage) necessary to verify operation within
the leakage limits contained in Technical Specification 3.4.6.2,
``Reactor Coolant System Leakage.'' Operability of the containment
atmosphere particulate radioactivity monitors and containment sump
inventory monitor is verified by SRs 4.4.6.1.a and 4.4.6.1.b.
Therefore, this change does not significantly increase the
probability or consequences of an accident previously evaluated.
The proposed change to Surveillance Requirement 4.5.2.e. to
allow the use of alternate methods does not reduce operability or
surveillance requirements for any of the Emergency Core Cooling
System (ECCS) throttle valves. Therefore, these ECCS throttle valves
will continue to function as designed to mitigate design basis
accidents. Therefore, this change does not significantly increase
the probability or consequences of an accident previously evaluated.
The proposed change to Technical Specification 5.5.1 has no
affect on how the ECCS operates. The ECCS will still function as
designed to mitigate design basis accidents. Therefore, this change
does not significantly increase the probability or consequences of
an accident previously evaluated.
[[Page 50009]]
The proposed changes to add information to the Bases of the
affected Technical Specifications, and make editorial changes to
terminology and item numbering will have no affect on equipment
operation. Therefore, all associated equipment will continue to
function as designed to mitigate design basis accidents. Therefore,
this change does not significantly increase the probability or
consequences of an accident previously evaluated.
Thus, this License Amendment Request does not impact the
probability of an accident previously evaluated nor does it involve
a significant increase in the consequences of an accident previously
evaluated.
2. Create the possibility of a new or different kind of accident
from any accident previously evaluated.
The proposed changes do not alter the plant configuration (no
new or different type of equipment will be installed) or require any
new or unusual operator actions. They do not alter the way any
structure, system, or component functions and do not alter the
manner in which the plant is operated. The proposed changes do not
introduce any new failure modes. They will not alter assumptions
made in the safety analysis and licensing basis. The affected
components and systems will still function as designed to mitigate
design basis accidents.
Therefore, these 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 a margin of safety.
The proposed changes will not reduce the margin of safety since
they have no impact on any safety analysis assumption. The proposed
changes do not decrease the scope of equipment currently required to
be operable or subject to surveillance testing, nor do the proposed
changes affect any instrument setpoints or equipment safety
functions. The requirement to check containment radiation and
containment sump level every 12 hours has been eliminated. However,
this equipment is still required to be operable, and the
surveillance requirements to verify operability have not been
changed. Therefore, this equipment will be available to provide
early indication of RCS leakage.
The effectiveness of Technical Specifications will be maintained
since the changes will not alter the operation of any component or
system. In addition, the changes are consistent with the new,
improved Standard Technical Specifications (STS) for Combustion
Engineering plants (NUREG-1432).
Therefore, there is not 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: 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, CT 06385
Attorney for licensee: Lillian M. Cuoco, Esq., Senior Nuclear
Counsel, Northeast Utilities Service Company, P.O. Box 270, Hartford,
CT 06141-0270
NRC Deputy Director: Phillip F. McKee
Northeast Nuclear Energy Company, et al., Docket No. 50-336,
Millstone Nuclear Power Station, Unit No. 2, New London,
Connecticut
Date of amendment request: September 3, 1997
Description of amendment request: The proposed amendment would
revise the Updated Final Safety Analysis Report (UFSAR) by changing the
length of time the emergency diesel generators (EDGs) would operate
following a loss-of-coolant accident (LOCA) based on the capacity of
the onsite diesel fuel oil supply required by the current Technical
Specifications (TSs). The UFSAR indicates that the diesel fuel oil
supply tanks contain a sufficient amount of fuel to operate one EDG for
about 7 days and the other EDG 1 hour following a LOCA based on the TS
minimum limit of 24,000 gallons of diesel fuel oil stored onsite.
Northeast Nuclear Energy Company (the licensee) has performed
calculations indicating that both EDGs can initially operate, following
a LOCA, for 24 hours and one EDG can continue to operate for an
additional 3.5 days based on the TS requirement to have a minimum of
24,000 gallons of fuel oil stored onsite. The licensee has determined
that the difference in the EDGs operating time, as a result of the new
calculations, constitutes an unreviewed safety question and requests
approval to revise the UFSAR.
Specifically, the proposed license amendment would revise the
UFSAR, Section 8.3, ``Emergency Generators,'' to reflect the operating
times for the EDGs based on the TS-required onsite fuel oil supply.
Additional requirements would also be added indicating that the
existing nonsafety-related underground fuel oil storage tank would be
required to maintain about 17,700 gallons of fuel oil when the unit is
operating in Modes 1 through 4. This requirement would be included in
the Technical Requirements Manual, which also will require that the
amount of stored fuel oil be verified by surveillance requirements
similar to the TS-required surveillances for the safety-related fuel
oil supply. This change will increase the total time that one EDG can
continue to operate following a LOCA from 3.5 to 7 days. The Emergency
Plan (EP) procedures require that an evaluation be performed within 4
hours following a LOCA or loss of normal power (offsite power) to
determine if additional fuel oil is needed from an offsite source. The
licensee has a contract with a supplier for the delivery of fuel oil to
the Millstone site. The EP procedures also require that load shedding
recommendations be made within 24 hours. The recommendations will vary
depending on the situation and are another way to extend the operating
times for the EDGs.
Basis for proposed no significant hazardsconsideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
The proposed change does not involve an SHC [significant hazards
consideration] because the changes would not:
1. Involve a significant increase in the probability or
consequences of an accident previously evaluated.
The proposed change expands FSAR Section 8.3, ``Emergency
Generators,'' to discuss the length of time the emergency diesel
generators (EDGs) will operate following a loss of coolant accident
(LOCA) and a loss of normal power (LNP), utilizing only onsite
diesel fuel oil sources. The onsite sources include the Technical
Specification required volume of 12,000 gallons in each diesel oil
supply tank and an additional approximate 17,700 gallons that will
be maintained in the underground diesel oil storage tank. This
onsite volume of diesel fuel oil is sufficient to allow two EDGs to
operate at rated load (2750 KW) for 24 hours following a design
basis LOCA and LNP. The remaining diesel fuel oil will be sufficient
for one EDG to continue operation at rated load for a total of 7
days from event initiation.
The proposed change to the FSAR has no effect on EDG operation
and reliability. The EDGs will continue to operate as designed to
supply the electrical loads assumed to mitigate the design basis
accidents. Therefore, there is no significant increase in the
probability or consequences of an accident previously evaluated.
2. Create the possibility of a new or different kind of accident
from any accident previously evaluated.
The proposed change will not alter the plant configuration (no
new or different type of equipment will be installed) or require any
new or unusual operator actions. Plant operating procedures will be
changed. However, the changes will not require the performance of
any task not currently performed by the plant operators. Emergency
Plan procedures already specify the action to provide load shedding
recommendations within 24 hours of a LOCA and LNP, and to evaluate
the need to order additional fuel from offsite sources within four
hours after the accident.
The proposed change does not alter the way any structure,
system, or component
[[Page 50010]]
functions and does not alter the manner in which the plant is
operated. It does not introduce any new failure modes and does not
alter assumptions made in the safety analysis.
Therefore, the change will not create the possibility of a new
or different kind of accident from any accident previously
evaluated.
3. Involve a significant reduction in a margin of safety.
The length of time the emergency diesel generators (EDGs) will
operate following a Loss of Coolant Accident and a Loss of Normal
Power, utilizing only the onsite diesel fuel oil sources required by
Technical Specifications has been recalculated. The new EDG run
times do not agree with the current EDG run times contained in the
Millstone Unit No. 2 Final Safety Analysis Report (FSAR), and
therefore do not agree with the current Technical Specification
Bases for 3.8.1.1, ``A.C. Sources - Operating,'' and 3.8.1.2, ``A.C.
Sources - Shutdown.''
This deviation does result in a reduction in the margin of
safety as defined in the Technical Specification Bases for 3.8.1.1,
``A.C. Sources - Operating,'' and 3.8.1.2, ``A.C. Sources -
Shutdown.'' However, this proposed change will require additional
diesel fuel oil to be maintained onsite in the non-seismic
underground diesel oil storage tank. This will ensure sufficient
diesel fuel oil will be maintained onsite to provide a 7 day supply,
assuming a seismic event does not occur. Therefore, this is not a
significant reduction in the margin of safety as defined in the
Technical Specification Bases for 3.8.1.1, ``A.C. Sources -
Operating,'' and 3.8.1.2, ``A.C. Sources - Shutdown.''
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: : Learning Resources Center,
Three Rivers Community-Technical College, 574 New London Turnpike,
Norwich, CT 06360, and the Waterford Library, ATTN: Vince Juliano, 49
Rope Ferry Road, Waterford, CT 06385
Attorney for licensee: Lillian M. Cuoco, Esq., Senior Nuclear
Counsel, Northeast Utilities Service Company, P.O. Box 270, Hartford,
CT 06141-0270
NRC Deputy Director: Phillip F. McKee
Public Service Electric & Gas Company, Docket No. 50-354, Hope
Creek Generating Station, Salem County, New Jersey
Date of amendment request: August 20, 1997
Description of amendment request: The proposed amendment would
change the Technical Specifications (TSs) to provide for: (1) the
relocation of suppression pool volume references in Limiting Condition
for Operation (LCO) 3.5.3 to the Hope Creek (HC) Updated Final Safety
Analysis Report (UFSAR) and TS Bases as appropriate; (2) the revision
of the suppression pool volume currently listed in LCO 3.5.3.b; (3) the
relocation of the suppression pool volume references in LCO 3.6.2.1.a.1
to the UFSAR and TS Bases; and (4) the revision to the suppression pool
volume reference in TS 5.2.1 to reference the TS Bases section where
this information will reside.
Basis for proposed no significant hazardsconsideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. The proposed changes do not involve a significant increase in
the probability or consequences of an accident previously evaluated.
The proposed TS revisions involve: 1) no changes to the
operation of any systems or components in normal or accident
operating conditions; and 2) no significant changes to existing
structures, systems or components. The installation of the new
strainers will be justified separately using the provisions of
10CFR50.59. The relocation of Technical Specification references to
suppression pool volume to the UFSAR and/or TS Bases will not
adversely impact the safety-related functions of the suppression
pool or its supported systems since any changes to suppression pool
volume will be subject to 10CFR50.59 provisions. The impact of the
new strainers on ECCS [emergency core cooling system] performance in
Operational Conditions 4 and 5 has been determined to be negligible,
with less than a 0.3% decrease in suppression pool water volume at
the minimum specified suppression pool water level limit. In
addition, suppression pool volume is not a parameter involved in the
initiation of any accident. Therefore these changes will not
significantly increase the probability of an accident previously
evaluated. To the extent practicable, these proposed changes were
developed consistent with the changes approved by the NRC when
developing NUREG-1433, ``Standard Technical Specifications, General
Electric Plants, BWR/4'', with the intent of having the relocated
information controlled in other plant documents subject to
10CFR50.59 provisions. Since the plant systems associated with these
proposed changes will still be capable of: 1) meeting all applicable
design basis requirements; and 2) retain the capability to mitigate
the consequences of accidents described in the HC UFSAR, the
proposed changes were determined to be justified. Therefore, these
changes will not involve a significant increase in the consequences
of an accident previously evaluated.
2. The proposed change does not create the possibility of a new
or different kind of accident from any accident previously
evaluated.
Neither the relocation of Technical Specification references to
suppression pool volume nor the revision of the suppression pool
volume references for Operational Conditions 4 and 5 (COLD SHUTDOWN
and REFUELING) will adversely impact the operation of any safety
related component or equipment. Since the proposed changes involve:
1) no changes to the operation of any systems or components; and 2)
no significant changes to existing structures, systems or
components, there can be no impact on the occurrence of any
accident. To the extent practicable, these proposed changes were
developed consistent with the changes approved by the NRC when
developing NUREG-1433, ``Standard Technical Specifications, General
Electric Plants, BWR/4'', with the intent of having the relocated
information controlled in other plant documents subject to
10CFR50.59 provisions. Furthermore, there is no change in plant
testing proposed in this change request which could initiate an
event. Therefore, these changes will not create the possibility of a
new or different kind of accident from any accident previously
evaluated.
3. The proposed change does not involve a significant reduction
in a margin of safety.
Removal and relocation of the Technical Specification references
to suppression pool volume is consistent, to the extent practicable,
with the changes approved by the NRC when developing NUREG-1433,
``Standard Technical Specifications, General Electric Plants, BWR/
4''. The information retained in the Technical Specifications for
minimum suppression pool water level and the information retained in
the UFSAR and Technical Specification Bases will ensure that the
suppression pool and supported components will remain capable of
performing their intended safety functions. Any changes to
suppression pool volume information retained in the UFSAR or
Technical Specification Bases will be subject to the provisions of
10CFR50.59 and a separate safety evaluation would be developed to
support any proposed changes that would subsequently be made. The
impact of the new strainers on ECCS performance in Operational
Conditions 4 and 5 has been determined to be negligible, with less
than a 0.3% decrease in suppression pool water volume in the minimum
specified suppression pool water level limit. By retaining the 5
inch minimum suppression pool water level limit within the TS,
adequate provisions for: 1) NPSH [net-positive suction head] for
ECCS pump suction; 2) recirculation volume; and 3) vortex prevention
are maintained. Therefore, the changes contained in this request do
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 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, NJ 08070
Attorney for licensee: Jeffrie J. Keenan, Esquire, Nuclear Business
Unit - N21,
[[Page 50011]]
P.O. Box 236, Hancocks Bridge, NJ 08038
NRC Project Director: John F. Stolz
Rochester Gas and Electric Corporation, Docket No. 50-244, R. E.
Ginna Nuclear Power Plant, Wayne County, New York
Date of amendment request: August 19, 1997
Description of amendment request: The proposed amendment would
revise the Ginna Station Improved Technical Specifications (ITSs) by
adding a note to the Containment Spray (CS) Limiting Condition for
Operation (LCO) 3.6.6 which would allow the CS pumps in MODE 4 to be
placed in pull-stop, and motor-operated valves (MOVs) 896A and 896B to
have their DC control power restored with the valves placed in the
closed position in order to perform interlock and valve testing of MOVs
857A, 857B, and 857C. A time limit of 2 hours is placed on this
configuration for each test.
Basis for proposed no significant hazardsconsideration
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 Ginna Station in accordance with the proposed
changes does not involve a significant increase in the probability
or consequences of an accident previously evaluated. The change is
to add a note to LCO 3.6.6 which allows the CS pumps to be placed in
pull-stop and MOVs 896A and 896B to have power restored and closed
in MODE 4. This does not increase the probability of any accident
previously evaluated since the CS system provides mitigation
capability only (i.e., does not initiate any accident). In addition,
there is no design basis accident previously evaluated in MODE 4
which would require the use of CS. Therefore, these changes do not
involve a significant increase in the probability or consequences of
an accident previously analyzed.
2. Operation of Ginna Station in accordance with the proposed
changes does not create the possibility of a new or different kind
of accident from any accident previously evaluated. The proposed
changes do not involve a physical alteration of the plant (i.e., no
new or different type of equipment will be installed) or changes in
the methods governing normal plant operation. Thus, this change does
not create the possibility of a new or different kind of accident
from any accident previously evaluated.
3. Operation of Ginna Station in accordance with the proposed
changes does not involve a significant reduction in a margin of
safety. The proposed changes will not reduce a margin of plant
safety because the CS function is not required for any design basis
accident in MODE 4. In addition, time restraints [are] placed on the
proposed plant configuration. As such, no question of safety is
involved, and the change does not involve a significant reduction in
a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: Rochester Public Library, 115
South Avenue, Rochester, New York 14610
Attorney for licensee: Nicholas S. Reynolds, Winston & Strawn, 1400
L Street, NW., Washington, DC 20005
NRC Project Director: Alexander W. Dromerick, Acting Director
TU Electric Company, Docket Nos. 50-445 and 50-446, Comanche Peak
Steam Electric Station, Units 1 and 2, Somervell County, Texas
Date of amendment request: August 2, 1996 (TXX-96434)
Brief description of amendments: The proposed changes would
increase the allowed outage time (AOT) for a centrifugal charging pump
from 72 hours to 7 days.
Basis for proposed no significant hazardsconsideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. Do the proposed changes involve a significant increase in the
probability or consequences of an accident previously evaluated?
There is no effect on the probability of an event; the only
potential effect is on the capability to mitigate the event. The
centrifugal charging pumps are credited in the Final Safety Analysis
Report Chapter 15 LOCA analysis for ECCS injection and for the
containment sump recirculation mode for the design-basis LOCA.
Increasing the AOT for the centrifugal charging pumps does not
affect analysis assumptions regarding functioning of required
equipment designed to mitigate the consequences of accidents.
Further, the severity of postulated accidents and resulting
radiological effluent releases will not be affected by the increased
AOT.
A reliability analysis of the charging system found the change
to have no significant impact on normal operation or on the RCP seal
cooling function. Therefore, the change would not significantly
increase in the probability of a seal LOCA.
The change potentially affects only the availability of the
charging system for accident mitigation and has no effect on the
ability of other ECCS systems to perform their functions. Through
the use of a probabilistic risk assessment, it was determined that
the proposed change would have an insignificant effect on the core
damage frequency.
Therefore, the proposed change does not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
2. Do the proposed changes create the possibility of a new or
different type of accident from any accident previously evaluated?
Unavailability of one centrifugal charging pump for a finite
period of time is currently allowed by the Technical Specifications.
Increasing the AOT from 72 hours to 7 days would not change the
method that TU Electric operates CPSES, thus would not create a new
condition. Further, the proposed change would not result in any
physical alteration to any plant system, and there would not be a
change in the method by which any safety related system performs its
function. The ECCS would still be capable of mitigating the
consequences of the design-basis accident LOCA with the one
centrifugal charging pump operable. No new unanalyzed accident would
be created.
3. Do the proposed changes involve a significant reduction in a
margin of safety?
The proposed change does not impact either the physical
protective boundaries or performance of safety systems for accident
mitigation. There is no safety analysis impact since the extension
of the centrifugal charging pump AOT interval will have no effect on
any safety limit, protection system setpoint, or limiting condition
of operation. There is no hardware change that would impact existing
safety analysis acceptance criteria, therefore there is no
significant change in the margin of safety.
In summary, the proposed change would not have a significant
impact on the margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: University of Texas at
Arlington Library, Government Publications/Maps, 702 College, P.O. Box
19497, Arlington, TX 76019
Attorney for licensee: George L. Edgar, Esq., Morgan, Lewis and
Bockius, 1800 M Street, N.W., Washington, DC 20036
NRC Project Director: James W. Clifford, Acting
Previously Published Notices Of Consideration Of Issuance Of
Amendments To Facility Operating Licenses, Proposed No Significant
Hazards Consideration Determination, And Opportunity For A Hearing
The following notices were previously published as separate
individual notices. The notice content was the same as above. They were
published as individual notices either because time did not allow the
Commission to wait for this biweekly notice or because the
[[Page 50012]]
action involved exigent circumstances. They are repeated here because
the biweekly notice lists all amendments issued or proposed to be
issued involving no significant hazards consideration.
For details, see the individual notice in the Federal Register on
the day and page cited. This notice does not extend the notice period
of the original notice.
Detroit Edison Company, Docket No. 50-341, Fermi 2, Monroe County,
Michigan
Date of amendment request: September 5, 1997 (NRC-97-0107)
Description of amendment request: The proposed amendment would add
Special Test Exception 3/4.10.7, ``Inservice Leak and Hydrostatic
Testing,'' that allows the performance of pressure testing at a reactor
coolant temperature up to 212 deg.F while remaining in Operational
Condition 4. This special test exception would also require that
certain Operational Condition 3 specifications for Secondary
Containment Isolation, Secondary Containment Integrity, Secondary
Containment Automatic Isolation Dampers, and Standby Gas Treatment
System operability be met. This change would also revise the Index,
Table 1.2, ``Operational Conditions,'' and the Bases to incorporate the
reference to the proposed special test exception. The licensee
requested that this amendment be reviewed under exigent circumstances.
Date of individual notice in the Federal Register: September 12,
1997 (62 FR 48113)
Expiration date of individual notice: October 14, 1997 NSHC
comments: September 29, 1997
Local Public Document Room location: Monroe County Library System,
3700 South Custer Road, Monroe, Michigan 48161
Attorney for licensee: John Flynn, Esq., Detroit Edison Company,
2000 Second Avenue, Detroit, Michigan 48226
NRC Project Director: John N. Hannon
Houston Lighting & Power Company, City Public Service Board of San
Antonio, Central Power and Light Company, City of Austin, Texas,
Docket Nos. 50-498 and 50-499, South Texas Project, Units 1 and 2,
Matagorda County, Texas
Date of amendment request: August 14, 1997
Brief description of amendment request: The proposed amendments
would revise the allowed tolerance of the reactor coolant system volume
provided in Technical Specification 5.4.2 to account for steam
generator tube plugging.
Date of individual notice in the Federal Register: August 26, 1997
(62 FR 45278)
Expiration date of individual notice: September 25, 1997
Local Public Document Room location: Wharton County Junior College,
J. M. Hodges Learning Center, 911 Boling Highway, Wharton, TX 77488
Notice Of Issuance Of Amendments ToFacility 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.
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 application for amendments: January 15, 1997, as
supplemented on August 22, 1997.
Brief description of amendments: The amendments revise the minimum
and maximum allowed values in Technical Specification 3.6.2.1 for
suppression chamber water volume. The amendments correct an error
identified by Carolina Power & Light Company in the previous
calculation of water volume and correct an error in the value listed in
the associated TS Bases for Unit 1 for primary system operating
pressure.
Date of issuance: August 28, 1997
Effective date: August 28, 1997
Amendment Nos.: 186 and 217
Facility Operating License Nos. DPR-71 and DPR-62: Amendments
change the Technical Specifications
Date of initial notice in Federal Register: March 26, 1997 (62 FR
14458) The Commission's related evaluation of the amendments is
contained in a Safety Evaluation dated August 28, 1997. No significant
hazards consideration comments received: No.
Local Public Document Room location: University of North Carolina
at Wilmington, William Madison Randall Library, 601 S. College Road,
Wilmington, North Carolina 28403-3297
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 application for amendments: December 4, 1996
Brief description of amendments: The amendments revise the approach
in Technical Specification 3/4.1.2 for determining a reactivity anomaly
by changing from control rod density comparison to direct comparison of
reactivity status.
Date of issuance: September 5, 1997
Effective date: September 5, 1997
Amendment Nos.: 187 and 218
Facility Operating License Nos. DPR-71 and DPR-62: Amendments
change the Technical Specifications
Date of initial notice in Federal Register: March 12, 1997 (62 FR
11484) The Commission's related evaluation of the amendments is
contained in a Safety Evaluation dated September 5, 1997. No
significant hazards consideration comments received: No.
Local Public Document Room location: University of North Carolina
at
[[Page 50013]]
Wilmington, William Madison Randall Library, 601 S. College Road,
Wilmington, North Carolina 28403-3297
Commonwealth Edison Company, Docket Nos. 50-237 and 50-249, Dresden
Nuclear Power Station, Units 2 and 3, Grundy County, Illinois
Date of application for amendments: January 24, 1997
Brief description of amendments: The amendments revise the
Technical Specification (TS) required surveillance calibration to be
performed on the reactor water level instrumentation to reflect the
modifications made to the Unit 3 instrumentation. The modifications
were made during the recent Unit 3 refueling outage to improve the
reliability of emergency core cooling system (ECCS) initiation on low
low reactor water level. The surveillance requirement for calibration
of the new level instrumentation is consistent with the ECCS low
reactor water level initiation transmitter calibration requirements of
NUREG 1433, ``Standard Technical Specifications, General Electric
Plants, BWR/4'' for similar instrumentation. The same TS change for
Unit 2 has been previously reviewed and approved by the NRC staff in
Amendment No. 145 dated June 28, 1996. In addition minor editorial
changes were made to the TS.
Date of issuance: September 10, 1997
Effective date: September 10, 1997, with full implementation within
60 days.
Amendment Nos.: 162 and 157
Facility Operating License Nos. DPR-19 and DPR-25: The amendments
revised the Technical Specifications.
Date of initial notice in Federal Register: April 18, 1997 (62 FR
19143) The Commission's related evaluation of the amendments is
contained in a Safety Evaluation dated September 10, 1997. No
significant hazards consideration comments received: No.
Local Public Document Room location: Morris Area Public Library
District, 604 Liberty Street, Morris, Illinois 60450
oit Edison Company, Docket No. 50-341, Fermi-2, Monroe County,
Michigan
application for amendment: December 15, 1994, as revised July 25,
1996, and supplemented December 13, 1996, and June 18, 1997
Brief description of amendment: The amendment revises Technical
Specification (TS) Section 6.0, Administrative Controls, by (1)
removing requirements that are adequately controlled by existing
regulations other than 10 CFR 50.36 and the TS and (2) relocating
selected requirements from TS Section 6.0 to licensee-controlled
documents or programs.
Date of issuance: September 10, 1997
Effective date: September 10, 1997, with full implementation within
90 days. Implementation of this amendment shall include the relocation
of the TS requirements to the appropriate licensee-controlled
documents, as described in the licensee's application dated December
15, 1994, as revised July 25, 1996, and supplemented December 13, 1996,
and June 18, 1997, and evaluated in the staff's safety evaluation dated
September 10, 1997.
Amendment No.: 113
Facility Operating License No. NPF-43. Amendment revises the TS.
Date of initial notice in Federal Register: June 6, 1995 (60 FR
29873) and August 14, 1996 (61 FR 42279). The December 13, 1996, and
June 18, 1997, letters provided clarifying information within the scope
of the original application and did not change the staff's initial
proposed no significant hazards considerations determination. The
Commission's related evaluation of the amendment is contained in a
Safety Evaluation dated September 10, 1997. No significant hazards
consideration comments received: No.
Local Public Document Room location: Monroe County Library System,
3700 South Custer Road, Monroe, Michigan 48161
Duquesne Light Company, et al., Docket No. 50-334, Beaver Valley
Power Station, Unit No. 1, Shippingport, Pennsylvania
Date of application for amendment: March 10, 1997
Brief description of amendment: The amendment modifies the
Technical Specifications (TSs) by reducing the reactor coolant system
specific activity limits in accordance with the NRC's guidance provided
in Generic Letter 95-05, ``Voltage-Based Repair Criteria for
Westinghouse Steam Generator Tubes by Outside Diameter Stress Corrosion
Cracking.'' The definition of DOSE EQUIVALENT I-131 is replaced with
the Improved Standard TS definition in the first sentence and an
equation is added based on dose conversion factors derived from the
International Commission on Radiation Protection (ICRP) ICRP-30. TS
3.4.8, Specific Activity, is revised by reducing the DOSE EQUIVALENT I-
131 limit from 1.0 micro Ci/gram to 0.35 micro Ci/gram for the 48-hour
limit and from 60 micro Ci/gram to 21 micro Ci/gram for the maximum
instantaneous limit. Item 4.a in TS Table 4.4-12, Primary Coolant
Specific Activity Sample and Analysis Program, TS Figure 3.4-1, and the
Bases for TS 3/4.4.8 are also modified to reflect the reduced DOSE
EQUIVALENT I-131 limit.
Date of issuance: September 10, 1997
Effective date: As of the date of issuance, to be implemented
within 60 days.
Amendment No: 205
Facility Operating License No. DPR-66. Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: May 7, 1997 (62 FR
24985) The Commission's related evaluation of the amendment is
contained in a Safety Evaluation dated September 10, 1997. No
significant hazards consideration comments received: No
Local Public Document Room location: B. F. Jones Memorial Library,
663 Franklin Avenue, Aliquippa, PA 15001
Florida Power Corporation, et al., Docket No. 50-302, Crystal River
Unit No. 3 Nuclear Generating Plant, Citrus County, Florida
Date of application for amendment: August 4, 1997, as supplemented
August 16, 1997.
Brief description of amendment: Temporary change to Technical
Specification Surveillance Requirement (SR) 3.3.8.1. The change will
allow the licensee to extend the frequency of SR 3.3.8.1 from 31 to 60
days.
Date of issuance: August 29, 1997
Effective date: August 29, 1997
Amendment No.: 157
Facility Operating License No. DPR-72. Amendment temporarily
revises Technical Specifications Surveillance Requirement 3.3.8.1.
Date of initial notice in Federal Register: August 12, 1997 (62 FR
43189) The Commission's related evaluation of the amendment is
contained in a Safety Evaluation dated August 29, 1997. No significant
hazards consideration comments received: No.
Local Public Document Room location: Coastal Region Library, 8619
W. Crystal Street, Crystal River, Florida 32629
Houston Lighting & Power Company, City Public Service Board of San
Antonio, Central Power and Light Company, City of Austin, Texas,
Docket Nos. 50-498 and 50-499, South Texas Project, Units 1 and 2,
Matagorda County, Texas
Date of amendment request: May 17, 1996, as supplemented June 14,
1996, March 17, July 29, and July 30, 1997
Brief description of amendments: The amendments modify Technical
[[Page 50014]]
Specification Section 3/4.4.5 Steam Generators, 3/4.4.6 Reactor Coolant
System Leakage, and associated Bases to allow the installation of tube
sleeves as an alternative to plugging to repair defective steam
generator tubes.
Date of issuance: September 4, 1997
Effective date: September 4, 1997
Amendment Nos.: Unit 1 - Amendment No. 90; Unit 2 - Amendment No.
77
Facility Operating License Nos. NPF-76 and NPF-80: The amendments
revised the Technical Specifications.
Date of initial notice in Federal Register: May 29, 1996 (61 FR
25938) and April 9, 1997 (62 FR 17235). The June 14, 1996, and July 29,
and July 30, 1997, submittals provided additional information that did
not affect the staff's initial no significant hazards consideration
determination. The Commission's related evaluation of the amendments is
contained in a Safety Evaluation dated September 4, 1997. No
significant hazards consideration comments received: No.
Local Public Document Room location: Wharton County Junior College,
J. M. Hodges Learning Center, 911 Boling Highway, Wharton, TX 77488
Northeast Nuclear Energy Company, et al., Docket No. 50-423,
Millstone Nuclear Power Station, Unit No. 3, New London County,
Connecticut
Date of application for amendment: June 30, 1997
Brief description of amendment: Technical Specification
Surveillance Requirements 4.7.1.5.1 and 4.7.1.5.2 require the periodic
testing of the main steam isolation valves (MSIVs) to demonstrate
operability. The amendment (1) clarifies when the MSIVs are partial
stroked or full closure tested, (2) adds a note to the Mode 4
applicability of Technical Specification 3.7.1.5 to require that the
MSIVs be closed and deactivated at less than 320 degrees F, (3) makes
editorial changes, and (4) makes changes to the associated Bases
sections.
Date of issuance: September 3, 1997
Effective date: As of the date of issuance, to be implemented
within 60 days.
Amendment No.: 148
Facility Operating License No. NPF-49: Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: July 30, 1997 ( 62 FR
40853) The Commission's related evaluation of the amendment is
contained in a Safety Evaluation dated September 3, 1997. No
significant hazards consideration comments received: No.
Local Public Document Room location: Learning Resources Center,
Three Rivers Community-Technical College, 574 New London Turnpike,
Norwich, Connecticut 06360, and the Waterford Library, ATTN: Vince
Juliano, 49 Rope Ferry Road, Waterford, Connecticut 06385
Northeast Nuclear Energy Company, et al., Docket No. 50-423,
Millstone Nuclear Power Station, Unit No. 3, New London County,
Connecticut
Date of application for amendment: May 14, 1997, as supplemented by
letter dated July 30, 1997
Brief description of amendment: Technical Specification
Surveillance Requirement 4.8.2.1.c.4 requires that each battery charger
be tested to verify that it can supply a specified current at 125
volts. The amendment increases the required test voltage.
Date of issuance: September 5, 1997
Effective date: As of the date of issuance, to be implemented
within 60 days.
Amendment No.: 149
Facility Operating License No. NPF-49: Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: June 18, 1997 (62 FR
33130) The July 30, 1997, letter provided clarifying information that
did not change the scope of the May 14, 1997, 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 September 5, 1997. No significant hazards
consideration comments received: No.
Local Public Document Room location: Learning Resources Center,
Three Rivers Community-Technical College, 574 New London Turnpike,
Norwich, Connecticut 06360, and 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: May 9, 1997, as supplemented by
letter dated July 14, 1997
Brief description of amendments: The proposed change revises the
Peach Bottom Atomic Power Station, Units 2 and 3, technical
specifications to extend the interval for replacing the primary
containment purge and exhaust valve inflatable seals.
Date of issuance: September 4, 1997
Effective date: Both units, as of date of issuance, to be
implemented within 30 days.
Amendments Nos.: 220 and 223
Facility Operating License Nos. DPR-44 and DPR-56: The amendments
revised the Technical Specifications.
Date of initial notice in Federal Register: July 2, 1997 (62 FR
35851) The supplemental letter provided clarifying information that did
not change the original no significant hazards consideration
determination. The Commission's related evaluation of the amendments is
contained in a Safety Evaluation dated September 4, 1997. No
significant hazards consideration comments received: No.
Local Public Document Room location: Government Publications
Section, State Library of Pennsylvania, (REGIONAL DEPOSITORY) Education
Building, Walnut Street and Commonwealth Avenue, Box 1601, Harrisburg,
PA 17105
Pennsylvania Power and Light Company, Docket Nos. 50-387 and 50-
388, Susquehanna Steam Electric Station, Units 1 and 2, Luzerne
County, Pennsylvania
Date of application for amendments: September 25, 1996
Brief description of amendments: These amendments (1) revise the
required number of operable gaseous radioactivity monitoring system
channels and particulate radioactivity monitoring system channels from
one in each of the monitoring systems to one in either of the
monitoring systems, (2) allow both the gaseous radioactivity monitoring
system and the particulate monitoring system to be inoperable for up to
30 days provided that grab samples are obtained and analyzed at least
once per 12 hours, and (3) add an action for the loss of all reactor
coolant system leakage detection systems (drywell floor sump level
monitoring system, gaseous radioactivity monitoring system and
particulate radioactivity monitoring system).
Date of issuance: September 3, 1997
Effective date: As of the date of issuance, to be implemented
within 30 days of issuance.
Amendment Nos.: 168 and 142
Facility Operating License Nos. NPF-14 and NPF-22: The amendments
revised the Technical Specifications.
Date of initial notice in Federal Register: November 19, 1996 (61
FR 58904) The Commission's related evaluation of the amendments is
contained in a Safety Evaluation dated
[[Page 50015]]
September 3, 1997. No significant hazards consideration comments
received: No.
Local Public Document Room location: Osterhout Free Library,
Reference Department, 71 South Franklin Street, Wilkes-Barre, PA 18701
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 application for amendments: April 25, 1997, as supplemented
June 6, 1997
Brief description of amendments: The amendments revise Technical
Specification 3.5.2 to eliminate reference to the flow path from the
residual heat removal system to the reactor coolant system hot legs.
This flow path is being eliminated to prevent excessive flow through
the residual heat removal system during all hot leg recirculation
configurations assuming worst-case single failures that could result in
excessive flow during hot leg recirculation following a loss-of-coolant
accident.
Date of issuance: September 11, 1997
Effective date: Both units, as of the date of issuance, to be
implemented within 60 days of issuance.
Amendment Nos.: 200 and 184
Facility Operating License Nos. DPR-70 and DPR-75. The amendments
revised the Technical Specifications.
Date of initial notice in Federal Register: May 14, 1997 (62 FR
26574) The June 6, 1997, supplement provided clarifying information
that did not change the initial proposed no significant hazards
consideration determination. The Commission's related evaluation of the
amendments is contained in a Safety Evaluation dated September 11,
1997. No significant hazards consideration comments received: No.
Local Public Document Room location: Salem Free Public Library, 112
West Broadway, Salem, NJ 08079Sacramento Municipal Utility District,
Docket No. 312, Rancho Seco Nuclear Generating Station, Sacramento
County, California
Date of application for amendment: December 9, 1993, as superseded
December 19, 1995, and as supplemented on January 22, 1996.
Brief description of amendment: This amendment changes the
Technical Specifications to incorporate the revised 10 CFR Part 20,
Standards for Protection Against Radiation. The amendment corrects
references from Semiannual Radioactive Effluent Release Report to
Annual Radioactive Effluent Release Report. The amendment also corrects
references from NRC Region V to NRC Region IV.
Date of issuance: August 22, 1997
Effective date: August 22, 1997
Amendment No.: 125
Facility Operating License No. NPF-1: The amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: March 2, 1994 (59 FR
10015) The information provided in the licensee's letters of December
19, 1995 and January 22, 1996 contained editorial changes and did not
involve significant changes to the original Federal Register notice.
The Commission's related evaluation of the amendment is contained in a
Safety Evaluation dated August 22, 1997. No significant hazards
consideration comments received: No.
Local Public Document Room location: : Central Library, Government
Documents, 828 I Street, Sacramento, California 95814
Southern Nuclear Operating Company, Inc., Georgia Power Company,
Oglethorpe Power Corporation, Municipal Electric Authority of
Georgia, City of Dalton, Georgia, Docket Nos. 50-321 and 50-366,
Edwin I. Hatch Nuclear Plant, Units 1 and 2, Appling County,
Georgia Date of application for amendments: January 7, 1997, as
supplemented July 2, 1997
Brief description of amendments: The amendments revise plant
Technical Specifications associated with surveillance requirements
testing that requires manually actuating every safety/relief valve
during each unit startup from a refueling outage.
Date of issuance: September 5, 1997
Effective date: As of the date of issuance to be implemented within
30 days
Amendment Nos.: 208 and 150
Facility Operating License Nos. DPR-57 and NPF-5: Amendments
revised the Technical Specifications.
Date of initial notice in Federal Register: January 29, 1997 (62 FR
4350) The Commission's related evaluation of the amendments is
contained in a Safety Evaluation dated September 5, 1997. No
significant hazards consideration comments received: No.
Local Public Document Room location: Appling County Public Library,
301 City Hall Drive, Baxley, Georgia 31513
Tennessee Valley Authority, Docket No. 50-260 Browns Ferry Nuclear
Plant, Unit 2, Limestone County, Alabama
Date of application for amendment: June 2, 1995, revised March 3,
1997, as supplemented May 13 and August 20, 1997 (TS 353)
Brief description of amendment: The amendment provides technical
specification (TS) changes for an upgrade of the power range neutron
monitor instrumentation. Changes to thermal limits specifications were
also proposed to implement average power range monitor and rod block
monitor ts improvements, and maximum extended load line limit analyses.
Date of issuance: September 11, 1997
Effective Date: September 11, 1997
Amendment No.: 249
Facility Operating License No. DPR-52: Amendment revised the TS.
Date of initial notice in Federal Register: August 16, 1995 (60 FR
42609) The March 3, 1997 revision, as supplemented May 13 and August
20, 1997, does not affect the staff's proposed finding of no
significant hazards consideration.
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated September 11, 1997. No significant hazards
consideration comments received: None.
Local Public Document Room location: Athens Public library, 405 E.
South Street, Athens, Alabama 35611
Tennessee Valley Authority, Docket No. 50-390 Watts Bar Nuclear
Plant, Unit 1, Rhea County, Tennessee
Date of application for amendment: March 27, 1997, as supplemented
May 28, June 4, and July 30, 1997.
Brief description of amendment: The amendment pertains to Cycle 2
core design changes and provides operational enhancements for reactor
trip setpoints. Part 1 addresses an increase in the containment sump
boron concentration during a large break loss-of-coolant accident and
describes changes to Technical Specification (TS) 3.5.1 and 3.5.4
regarding boron concentration. Part 2 addresses changes to TS Figure
2.1.1-1, TS Table 3.3.1-1, and TS 3.4.1 on safety limits, the trip
system and pressure, temperature and flow limits, respectively.
Date of issuance: September 11, 1997
Effective date: Sepember 11, 1997
Amendment No.: 7
Facility Operating License No. NPF-90: Amendment revises the TS.
Date of initial notice in Federal Register: July 2, 1997 (62 FR
35852) The July 30, 1997 submittal provided clarifying information
which did not affect the initial proposed no significant hazards
consideration determination.
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated September 11, 1997. No significant hazards
consideration comments received: None
[[Page 50016]]
Local Public Document Room location: Chattanooga-Hamilton County
Library, 1001 Broad Street, Chattanooga, TN 37402
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: January 31, 1997, supplemented
August 6, 1997.
Brief description of amendment: The amendment approves the use of
Option B, ``Performance-Based Requirements,'' to 10 CFR Part 50,
Appendix J, ``Primary Reactor Containment Leakage Testing for Water-
Cooled Power Reactors.''
Date of issuance: September 9, 1997
Effective date: September 9, 1997
Amendment No.: 86
Facility Operating License No. NPF-58: This amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: March 12, 1997 (62 FR
11492). The Commission's related evaluation of the amendment is
contained in a Safety Evaluation dated September 9, 1997. No
significant hazards consideration comments received: No.
Local Public Document Room location: Perry Public Library, 3753
Main Street, Perry, Ohio 44081
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: May 2, 1997
Brief description of amendment: The amendment allows the leakage
rate of one or more main steam lines to be up to 35 standard cubic feet
per hour (scfh), as long as the total leakage rate through all four
main steam lines is less than or equal to 100 scfh.
Date of issuance: September 11, 1997
Effective date: September 11, 1997
Amendment No.: 87
Facility Operating License No. NPF-58: This amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: June 18, 1997 (62 FR
33136). The Commission's related evaluation of the amendment is
contained in a Safety Evaluation dated September 11, 1997. No
significant hazards consideration comments received: No.
Local Public Document Room location: Perry Public Library, 3753
Main Street, Perry, Ohio 44081
Wisconsin Electric Power Company, Docket Nos. 50-266 and 50-301,
Point Beach Nuclear Plant, Unit Nos. 1 and 2, Town of Two Creeks,
Manitowoc County, Wisconsin
Date of application for amendments: April 14, 1997 (TSCR 198)
Brief description of amendments: These amendments revise Technical
Specification Section 15.3.1, ``Reactor Coolant System,'' to eliminate
the provisions for operation of the units at below 3.5 percent rated
power with a single reactor coolant pump.
Date of issuance: September 3, 1997
Effective date: September 3, 1997, with full implementation within
45 days
Amendment Nos.: 178 and 182
Facility Operating License Nos. DPR-24 and DPR-27: Amendments
revised the Technical Specifications.
Date of initial notice in Federal Register: May 21, 1997 (62 FR
27802) The Commission's related evaluation of the amendments is
contained in a Safety Evaluation dated September 3, 1997. No
significant hazards consideration comments received: No.
Local Public Document Room location: The Lester Public Library,
1001 Adams Street, Two Rivers, Wisconsin 54241
Wisconsin Electric Power Company, Docket Nos. 50-266 and 50-301,
Point Beach Nuclear Plant, Unit Nos. 1 and 2, Town of Two Creeks,
Manitowoc County, Wisconsin
Date of application for amendments: January 24, 1997, as
supplemented on May 15 and August 5, 1997 (TSCR 193)
Brief description of amendments: These amendments revise TS 15.5.4,
``Fuel Storage,'' to increase fuel assembly enrichment limits to 5.0
weight percent uranium-235 while maintaining Keff in the
storage pools (spent fuel pool and new fuel storage racks) less than
0.95. Date of issuance: September 4, 1997
Effective date: September 4, 1997, with full implementation within
45 days
Amendment Nos.: 179 and 183
Facility Operating License Nos. DPR-24 and DPR-27: Amendments
revised the Technical Specifications.
Date of initial notice in Federal Register: June 4, 1997 (62 FR
30647) The August 5, 1997, submittal provided clarifying information
within the scope of the original application and did not affect the
staff's initial proposed no significant hazards considerations
determination. The Commission's related evaluation of the amendments is
contained in a Safety Evaluation dated September 4, 1997. No
significant hazards consideration comments received: No.
Local Public Document Room location: The Lester Public Library,
1001 Adams Street, Two Rivers, Wisconsin 54241
Notice Of Issuance Of Amendments To Facility Operating Licenses And
Final Determination Of No Significant Hazards Consideration And
Opportunity For A Hearing (Exigent Public Announcement Or Emergency
Circumstances)
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 for the
amendment 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.
Because of exigent or emergency circumstances associated with the
date the amendment was needed, there was not time for the Commission to
publish, for public comment before issuance, its usual 30-day Notice of
Consideration of Issuance of Amendment, Proposed No Significant Hazards
Consideration Determination, and Opportunity for a Hearing.
For exigent circumstances, the Commission has either issued a
Federal Register notice providing opportunity for public comment or has
used local media to provide notice to the public in the area
surrounding a licensee's facility of the licensee's application and of
the Commission's proposed determination of no significant hazards
consideration. The Commission has provided a reasonable opportunity for
the public to comment, using its best efforts to make available to the
public means of communication for the public to respond quickly, and in
the case of telephone comments, the comments have been recorded or
transcribed as appropriate and the licensee has been informed of the
public comments.
In circumstances where failure to act in a timely way would have
resulted, for example, in derating or shutdown of a nuclear power plant
or in prevention of either resumption of operation or of increase in
power output up to the
[[Page 50017]]
plant's licensed power level, the Commission may not have had an
opportunity to provide for public comment on its no significant hazards
consideration determination. In such case, the license amendment has
been issued without opportunity for comment. If there has been some
time for public comment but less than 30 days, the Commission may
provide an opportunity for public comment. If comments have been
requested, it is so stated. In either event, the State has been
consulted by telephone whenever possible.
Under its regulations, the Commission may issue and make an
amendment immediately effective, notwithstanding the pendency before it
of a request for a hearing from any person, in advance of the holding
and completion of any required hearing, where it has determined that no
significant hazards consideration is involved.
The Commission has applied the standards of 10 CFR 50.92 and has
made a final determination that the amendment involves no significant
hazards consideration. The basis for this determination is contained in
the documents related to this action. Accordingly, the amendments have
been issued and made effective 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
application for amendment, (2) the amendment to Facility Operating
License, 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 room for the particular facility involved.
The Commission is also offering an opportunity for a hearing with
respect to the issuance of the amendment. By October 24, 1997, the
licensee may file a request for a hearing with respect to issuance of
the amendment to the subject facility operating license and any person
whose interest may be affected by this proceeding and who wishes to
participate as a party in the proceeding must file a written request
for a hearing and a petition for leave to intervene. Requests for a
hearing and a petition for leave to intervene shall be filed in
accordance with the Commission's ``Rules of Practice for Domestic
Licensing Proceedings'' in 10 CFR Part 2. Interested persons should
consult a current copy of 10 CFR 2.714 which is available at the
Commission's Public Document Room, the Gelman Building, 2120 L Street,
NW., Washington, DC and at the local public document room for the
particular facility involved. If a request for a hearing or petition
for leave to intervene is filed by the above date, the Commission or an
Atomic Safety and Licensing Board, designated by the Commission or by
the Chairman of the Atomic Safety and Licensing Board Panel, will rule
on the request and/or petition; and the Secretary or the designated
Atomic Safety and Licensing Board will issue a notice of a hearing or
an appropriate order.
As required by 10 CFR 2.714, a petition for leave to intervene
shall set forth with particularity the interest of the petitioner in
the proceeding, and how that interest may be affected by the results of
the proceeding. The petition should specifically explain the reasons
why intervention should be permitted with particular reference to the
following factors: (1) the nature of the petitioner's right under the
Act to be made a party to the proceeding; (2) the nature and extent of
the petitioner's property, financial, or other interest in the
proceeding; and (3) the possible effect of any order which may be
entered in the proceeding on the petitioner's interest. The petition
should also identify the specific aspect(s) of the subject matter of
the proceeding as to which petitioner wishes to intervene. Any person
who has filed a petition for leave to intervene or who has been
admitted as a party may amend the petition without requesting leave of
the Board up to 15 days prior to the first prehearing conference
scheduled in the proceeding, but such an amended petition must satisfy
the specificity requirements described above.
Not later than 15 days prior to the first prehearing conference
scheduled in the proceeding, a petitioner shall file a supplement to
the petition to intervene which must include a list of the contentions
which are sought to be litigated in the matter. Each contention must
consist of a specific statement of the issue of law or fact to be
raised or controverted. In addition, the petitioner shall provide a
brief explanation of the bases of the contention and a concise
statement of the alleged facts or expert opinion which support the
contention and on which the petitioner intends to rely in proving the
contention at the hearing. The petitioner must also provide references
to those specific sources and documents of which the petitioner is
aware and on which the petitioner intends to rely to establish those
facts or expert opinion. Petitioner must provide sufficient information
to show that a genuine dispute exists with the applicant on a material
issue of law or fact. Contentions shall be limited to matters within
the scope of the amendment under consideration. The contention must be
one which, if proven, would entitle the petitioner to relief. A
petitioner who fails to file such a supplement which satisfies these
requirements with respect to at least one contention will not be
permitted to participate as a party.
Those permitted to intervene become parties to the proceeding,
subject to any limitations in the order granting leave to intervene,
and have the opportunity to participate fully in the conduct of the
hearing, including the opportunity to present evidence and cross-
examine witnesses. Since the Commission has made a final determination
that the amendment involves no significant hazards consideration, if a
hearing is requested, it will not stay the effectiveness of the
amendment. Any hearing held would take place while the amendment is in
effect.
A request for a hearing or a petition for leave to intervene must
be filed with the Secretary of the Commission, U.S. Nuclear Regulatory
Commission, Washington, DC 20555-0001, Attention: Rulemakings and
Adjudications Staff or may be delivered to the Commission's Public
Document Room, the Gelman Building, 2120 L Street, NW., Washington, DC,
by the above date. A copy of the petition should also be sent to the
Office of the General Counsel, U.S. Nuclear Regulatory Commission,
Washington, DC 20555-0001, and to the attorney for the licensee.
Nontimely filings of petitions for leave to intervene, amended
petitions, supplemental petitions and/or requests for a hearing will
not be entertained absent a determination by the Commission, the
presiding officer or the Atomic Safety and Licensing Board that the
petition and/or request should be granted based upon a balancing of the
factors specified in 10 CFR 2.714(a)(1)(i)-(v) and 2.714(d).
[[Page 50018]]
Arizona Public Service Company, et al., Docket No. STN 50-529, Palo
Verde Nuclear Generating Station, Unit No. 2, Maricopa County,
Arizona
Date of application for amendment: August 28, 1997, as supplemented
by letter dated September 3, 1997.
Brief description of amendment: The amendment revises Technical
Specification Table 4.3-2 to allow for a one-time, five-day extension
of the required surveillance interval for the main steam isolation
system portion of the engineered safety feature actuation system logic.
Date of issuance: September 4, 1997
Effective date: September 4, 1997
Amendment No.: 105
Facility Operating License No. NPF-51: The amendment revised the
Technical Specifications. Press release issued requesting comments as
to proposed no significant hazards consideration: Yes. September 1,
1997. Arizona Republic Newspaper (Arizona). Comments received: No. The
Commission's related evaluation of the amendment, finding of exigent
circumstances, consultation with the State of Arizona and final
determination of no significant hazards consideration are contained in
a Safety Evaluation dated September 4, 1997.
Local Public Document Room location: Phoenix Public Library, 1221
N. Central Avenue, Phoenix, Arizona 85004
Attorney for licensee: Nancy C. Loftin, Esq., Corporate Secretary
and Counsel, Arizona Public Service Company, P.O. Box 53999, Mail
Station 9068, Phoenix, Arizona 85072-3999
NRC Project Director: William H. Bateman
Public Service Electric & Gas Company, Docket No. 50-311, Salem
Nuclear Generating Station, Unit No. 2, Salem County, New Jersey
Date of application for amendment: August 19, 1997, as supplemented
August 20, 1997.
Brief description of amendment: This amendment to the Technical
Specifications increases the allowable band for control and shutdown
rod demanded position versus indication position from plus or minus 12
steps to plus or minus 18 steps when the power level is not greater
than 85% rated thermal power.
Date of issuance: September 10, 1997
Effective date: As of date of issuance, to be implemented within 7
days.
Amendment No. 183
Facility Operating License No. DPR-75: This amendment revised the
Technical Specifications. Public comments requested as to proposed no
significant hazards consideration: Yes. The NRC published a public
notice of the proposed amendment, issued a proposed finding of no
significant hazards consideration, and requested that any comments on
the proposed no significant hazards consideration be provided to the
staff by the close of business on September 3, 1997, and stated that,
should circumstances change during the notice period, such that a
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 notice period, provided that its
final determination is that the amendment involves no significant
hazards consideration. The notice was published in the Wilmington News
Journal on August 22, 1997, and in Today's Sunbeam on August 24, 1997.
No public comments were received. The Commission's related evaluation
of the amendment, finding of exigent circumstances, consultation with
the State of New Jersey and final no significant hazards consideration
determination are contained in a Safety Evaluation dated September 10,
1997.
Local Public Document Room location: Salem Free Public Library, 112
West Broadway, Salem, NJ 08079
Attorney for licensee: Jeffrie J. Keenan, Esquire, Nuclear Business
Unit - N21, P.O. Box 236, Hancocks Bridge, NJ 08038
NRC Project Director: John F. Stolz
Dated at Rockville, Maryland, this 17th day of September 1997.
For the Nuclear Regulatory Commission
Elinor G. Adensam,
Acting Director, Division of Reactor Projects--III/IV, Office of
Nuclear Reactor Regulation
[Doc. 97-25210 Filed 9-23-97; 8:45 am]
BILLING CODE 7590-01-F