[Federal Register Volume 62, Number 58 (Wednesday, March 26, 1997)]
[Notices]
[Pages 14457-14476]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: X97-10326]
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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 March 3, 1997, through March 14, 1997. The
last biweekly notice was published on March 12, 1997.
Notice Of Consideration Of Issuance Of Amendments To Facility
Operating Licenses, Proposed No Significant Hazards Consideration
Determination, And Opportunity For A Hearing
The Commission has made a proposed determination that the following
amendment requests involve no significant hazards consideration. Under
the Commission's regulations in 10 CFR 50.92, this means that operation
of the facility in accordance with the proposed amendment would not (1)
involve a significant increase in the probability or consequences of an
accident previously evaluated; or (2) create the possibility of a new
or different kind of accident from any accident previously evaluated;
or (3) involve a significant reduction in a margin of safety. The basis
for this proposed determination for each amendment request is shown
below.
The Commission is seeking public comments on this proposed
determination. Any comments received within 30 days after the date of
publication of this notice will be considered in making any final
determination.
Normally, the Commission will not issue the amendment until the
expiration of the 30-day notice period. However, should circumstances
change during the notice period such that failure to act in a timely
way would result, for example, in derating or shutdown of the facility,
the Commission may issue the license amendment before the expiration of
the 30-day notice period, provided that its final determination is that
the amendment involves no significant hazards consideration. The final
determination will consider all public and State comments received
before action is taken. Should the Commission take this action, it will
publish in the Federal Register a notice of issuance and provide for
opportunity for a hearing after issuance. The Commission expects that
the need to take this action will occur very infrequently.
Written comments may be submitted by mail to the Chief, Rules
Review and Directives Branch, Division of Freedom of Information and
Publications Services, Office of Administration, U.S. Nuclear
Regulatory Commission, Washington, DC 20555-0001, and should cite the
publication date and page number of this Federal Register notice.
Written comments may also be delivered to Room 6D22, Two White Flint
North, 11545 Rockville Pike, Rockville, Maryland from 7:30 a.m. to 4:15
p.m. Federal workdays. Copies of written comments received may be
[[Page 14458]]
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 April 25, 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: Docketing and
Services Branch, or may be delivered to the Commission's Public
Document Room, the Gelman Building, 2120 L Street, NW., Washington DC,
by the above date. Where petitions are filed during the last 10 days of
the notice period, it is requested that the petitioner promptly so
inform the Commission by a toll-free telephone call to Western Union at
1-(800) 248-5100 (in Missouri 1-(800) 342-6700). The Western Union
operator should be given Datagram Identification Number N1023 and the
following message addressed to (Project Director): petitioner's name
and telephone number, date petition was mailed, plant name, and
publication date and page number of this Federal Register notice. A
copy of the petition should also be sent to the Office of the General
Counsel, U.S. Nuclear Regulatory Commission, Washington, DC 20555-0001,
and to the attorney for the licensee.
Nontimely filings of petitions for leave to intervene, amended
petitions, supplemental petitions and/or requests for a hearing will
not be entertained absent a determination by the Commission, the
presiding officer or the Atomic Safety and Licensing Board that the
petition and/or request should be granted based upon a balancing of
factors specified in 10 CFR 2.714(a)(1)(i)-(v) and 2.714(d).
For further details with respect to this action, see the
application for amendment which is available for public inspection at
the Commission's Public Document Room, the Gelman Building, 2120 L
Street, NW., Washington, DC, and at the local public document room for
the particular facility involved.
Carolina Power & Light Company, et al., Docket Nos. 50-325 and 50-
324, Brunswick Steam Electric Plant, Units 1 and 2, Brunswick
County, North Carolina
Date of amendments request: January 15, 1997
Description of amendments request: The proposed change would revise
the values of the minimum and maximum suppression pool water volumes
corresponding to the upper and lower limits of the suppression water
levels specified in TS 3.6.2.1.a.1 such that the implementation of the
administrative controls will no longer be necessary to ensure
compliance with the Technical Specifications.
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 14459]]
1. The proposed amendments do not involve a significant increase
in the probability or consequences of an accident previously
evaluated.
The proposed change revises the values of the minimum and
maximum suppression pool water volume limits. The water inventory of
the suppression chamber is not a precursor of an accident and,
therefore, cannot increase the probability of an accident previously
evaluated. The pressure suppression chamber water pool mitigates the
consequences of loss-of-coolant accidents (LOCAs), transients, and
other events by providing a heat sink for reactor primary system
energy releases. The proposed minimum and maximum pool water volume
values will be consistent with the current suppression pool water
level limits. No changes to setpoints will be made as a result of
the proposed change. The impact of the proposed change to the
minimum and maximum suppression pool volume limits on the
suppression pool temperatures and pressures following a design basis
LOCA, an SRV [Safety Relief Valve] blowdown event, an Anticipated
Transient Without Scram (ATWS) event, an Appendix R fire event, and
a station blackout event has been evaluated and does not cause
accident parameters to exceed acceptable values. In addition, the
impact the proposed change has on the time to reach cold shutdown
when using the alternate RHR [Residual Heat Removal] shutdown
cooling function is negligible.
The potential impact the proposed change to the suppression pool
water volume limits has on SRV line loads, SRV discharge line
reflood height, wetwell pressurization, suppression pool swell
loads, vent thrust loads, and condensation oscillation and chugging
loads was also reviewed. The proposed change to the suppression pool
water volume limits has no adverse impact on any of these
parameters.
The capability of the suppression chamber water pool to perform
its mitigative functions is not affected by the proposed
change. Therefore, the proposed change does not involve a
significant increase in the consequences of an accident previously
evaluated.
2. The proposed amendment[s] would not create the possibility of
a new or different kind of accident from any accident previously
evaluated.
The proposed change revises the values of the minimum and
maximum volume of the suppression chamber water pool. The proposed
change will not alter any physical mechanism by which the
suppression chamber water pool volume is maintained between the
minimum and maximum values. The suppression pool water level will
continue to be maintained between -27 and -31 inches. As a result of
the proposed change there are no physical changes to suppression
chamber components or instrumentation. No new mode of operation is
introduced as a result of the proposed change. Analyses have been
performed which conclude that the proposed change would not affect
the operability of equipment designed to mitigate the consequences
of an accident. Therefore, the proposed change does not create the
possibility of a new or different kind of accident from any accident
previously evaluated.
3. The proposed license amendment[s do] not involve a
significant reduction in a margin of safety.
The proposed change revises the values of the minimum and
maximum suppression chamber water pool volumes. The pressure
suppression chamber water pool mitigates the consequences of several
postulated accidents and transients by providing a heat sink for the
primary coolant system. These accidents and events are the
postulated design basis LOCA, Safety Relief Valve blowdown, ATWS,
Appendix R fire and station blackout events. The consequences of the
proposed change in the suppression pool water volume limits have
been evaluated for these events.
The results of the analyses for the postulated accidents and
events indicate the temperature of the suppression pool water could
increase slightly as a consequence of the decrease in the minimum
suppression pool water volume limit. However, the containment
temperatures remain within acceptable values. The impact of the
calculated increase in containment temperature on the available Net
Positive Suction head (NPSH) for the Residual Heat Removal (RHR) and
Core Spray pumps has been evaluated for the postulated design basis
LOCA and indicate adequate NPSH is maintained throughout the event.
The potential impact of the proposed change to the suppression
pool water volume limits on SRV line loads, SRV discharge line
reflood height, wetwell pressurization, suppression pool swell
loads, vent thrust loads, and condensation oscillation and chugging
loads was evaluated with the conclusion that there are no adverse
impacts on these parameters.
In addition, a small suppression pool water temperature increase
could result due to the reduction in the minimum suppression pool
volume limit in the event reactor shutdown is conducted through a
path utilizing the suppression pool. Such a shutdown path is an
alternative to the normal RHR shutdown cooling function, and the
small potential increase in temperature results in a negligible
increase in the time required to reach cold shutdown conditions.
Cold shutdown conditions could still be reached well within the
Technical Specification requirements.
The proposed increase in the suppression pool water volume limit
does not adversely impact containment parameters as a result of
postulated accidents and events. The potential increase in
temperature of the pressure suppression pool water does not
significantly decrease the ability to maintain containment
parameters within acceptable limits. The potential increase in time
to reach cold shutdown conditions utilizing the suppression pool as
an alternative to the normal RHR shutdown cooling function is
negligible. Therefore, the proposed change to revise the minimum and
maximum suppression water pool volumes 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: University of North Carolina
at Wilmington, William Madison Randall Library, 601 S. College Road,
Wilmington, North Carolina 28403-3297.
Attorney for licensee: William D. Johnson, Vice President and
Senior Counsel, Carolina Power & Light Company, Post Office Box 1551,
Raleigh, North Carolina 27602
NRC Project Director: Mark Reinhart (Acting)
Carolina Power & Light Company, et al., Docket No. 50-400, Shearon
Harris Nuclear Power Plant, Unit 1, Wake and Chatham Counties,
North Carolina
Date of amendment request: March 14, 1997
Description of amendment request: The proposed change revises
Technical Specification 3/4.5.4, ``Refueling Water Storage Tank,'' and
its associated Bases.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. The proposed amendment does not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
The non-safety, non-seismic hydrotest pump is normally
maintained separated from the RWST [Refueling Water Storage Tank] by
a safety-related, locked closed manual operated boundary isolation
valve (1CT-22). However, performance of Technical Specification
required surveillance test OST-1506, ``Reactor Coolant System
Isolation Valve Leak Test - 18 Month Interval- Mode 3,'' requires
the short term use of the hydrotest pump during plant operating
modes. Specifically, this hydrotest pump provides a high pressure
source for leak testing the RCS [Reactor Coolant System] pressure
isolation valves in Mode 3. The test is performed prior to entry
into Mode 2, each refueling outage, whenever flow is established
through the pressure isolation valves, or whenever the plant has
been in cold shutdown for greater than 72 hours. Normally, the test
is completed in less than 8 hours. Due to the piping configuration,
a break in the non-seismic portion of the piping during these
planned evolutions could result in draining the RWST below the
minimum analyzed volume. Therefore to mitigate the consequences of a
failure in the non-seismic piping, manual actions will be needed to
isolate the break flow, (i.e., close valve 1CT-22), prior to
reducing the water volume in the RWST below the minimum analyzed
volume.
[[Page 14460]]
Based on the use of a dedicated attendant to close valve 1CT-22,
the lack of significant accessibility concerns, and the reliability
of the valve to function, it can be concluded that 30 minutes is
ample time for a valve attendant stationed at the valve to execute
the manual action. Since the RWST volume margin provides up to 103
minutes to respond to the pipe failure, it is reasonable to assume
that manual actions to isolate the postulated pipe failure can be
taken before the RWST level decreases below the minimum analyzed
volume assumed in the safety analysis.
Therefore, there would be no increase in the probability or
consequences of an accident previously evaluated.
2. The proposed amendment does not create the possibility of a
new or different kind of accident from any accident previously
evaluated.
Based on the use of a dedicated attendant to close valve 1CT-22,
the lack of significant accessibility concerns, and the reliability
of the valve to function, it can be concluded that 30 minutes is
ample time for a valve attendant stationed at the valve to execute
the manual action. Since the RWST volume margin provides up to 103
minutes to respond to the pipe failure, it is reasonable to assume
that manual actions to isolate the postulated pipe failure can be
taken before the RWST level decreases below the minimum analyzed
volume assumed in the safety analysis. As a result, the capability
of the RWST to perform its safety function is not impacted.
Therefore, the proposed change does not create the possibility
of a new or different kind of accident from any accident previously
evaluated.
3. The proposed amendment does not involve a significant
reduction in the margin of safety.
As described in the Technical Specification Bases, the
operability of the RWST ensures that a sufficient supply of borated
water is available for injection into the core by the emergency core
cooling system. This borated water is used as cooling water for the
core in the event of a LOCA [loss-of-coolant accident] and provides
negative reactivty to counteract any positive increase in reactivity
caused by reactor coolant system (RCS) cooldown. The limits on RWST
minimum volume and boron concentration assure that: (1) sufficient
water is available within containment to permit recirculation
cooling flow to the core, and (2) the reactor will remain
subcritical in the cold condition following mixing of the RWST and
the RCS water volumes with all shutdown and control rods inserted
except for the most reactive control assembly. These limits are
consistent with the assumptions of the LOCA and steam line break
analyses.
Based on the use of a dedicated attendant to close valve 1CT-22,
the lack of significant accessibility concerns, and the reliability
of the valve to function, it can be concluded that 30 minutes is
ample time for a valve attendant stationed at the valve to execute
the manual action. Since the RWST volume margin provides up to 103
minutes to respond to the pipe failure, it is reasonable to assume
that manual actions to isolate the postulated pipe failure can be
taken before the RWST level decreases below the minimum analyzed
volume assumed in the safety analysis. As a result, the capability
of the RWST to perform its safety function is not impacted.
Therefore, the proposed change does not involve a significant
reduction in the margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: Cameron Village Regional
Library, 1930 Clark Avenue, Raleigh, North Carolina 27605
Attorney for licensee: William D. Johnson, Vice President and
Senior Counsel, Carolina Power & Light Company, Post Office Box 1551,
Raleigh, North Carolina 27602
NRC Project Director: Mark Reinhart, Acting
Commonwealth Edison Company, Docket Nos. 50-254 and 50-265, Quad
Cities Nuclear Power Station, Units 1 and 2, Rock Island County,
Illinois
Date of amendment request: February 17, 1997
Description of amendment request: The proposed amendment would
change the required diesel generator load during the initial 2 hours of
a surveillance run from 2625 kW and 2750 kW to 2730 kW and 2860 kW.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1) Involve a significant increase in the probability or
consequences of an accident previously evaluated because of the
following:
The proposed changes represent a correction to the emergency
diesel generator surveillance requirement. The proposed changes are
administrative in nature and do not significantly increase the
probability or consequences of any previously evaluated accidents
for Quad Cities Station. The proposed amendment is consistent with
the current safety analyses and represents sufficient requirements
for the assurance and reliability of equipment assumed to operate in
the safety analysis. As such, these changes will not significantly
increase the probability or consequences of a previously evaluated
accident.
The associated systems related to this proposed amendment are
not assumed in any safety analysis to initiate any accident sequence
for Quad Cities Station; therefore, the probability of any accident
previously evaluated is not increased by the proposed amendment.
2) Create the possibility of a new or different kind of accident
from any accident previously evaluated because:
The proposed amendment for Quad Cities Station's Technical
Specification is required to ensure the diesel generator is tested
in accordance with the design basis requirements. The proposed
changes do not create the possibility of a new or different kind of
accident previously evaluated for Quad Cities Station. No new modes
of operation are introduced by the proposed changes. The proposed
changes are administrative in nature and maintain at least the
present level of operability. Therefore, the proposed changes do not
create the possibility of a new or different kind of accident from
any previously evaluated.
The associated systems related to this proposed amendment are
not assumed in any safety analysis to initiate any accident sequence
for Quad Cities Station; therefore, the proposed changes do not
create the possibility of a new or different kind of accident from
any previously evaluated.
3) Involve a significant reduction in the margin of safety
because:
The proposed amendment is required to ensure the diesel
generator is tested in accordance with the design basis
requirements. The proposed changes are administrative in nature and
do not adversely affect existing plant safety margins or the
reliability of the equipment assumed to operate in the safety
analysis. The proposed changes have been evaluated and found to be
acceptable for use at Quad Cities based on system design, safety
analysis requirements and operational performance. Since the
proposed changes are administrative in nature and maintain necessary
levels of system or component reliability, the proposed changes do
not involve a significant reduction in the margin of safety.
The proposed amendment for Quad Cities Station will not reduce
the availability of systems required to mitigate accident
conditions; therefore, the proposed changes do not involve a
significant reduction in the margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
requested amendments involve no significant hazards consideration.
Local Public Document Room location: Dixon Public Library, 221
Hennepin Avenue, Dixon, Illinois 61021
Attorney for licensee: Michael I. Miller, Esquire; Sidley and
Austin, One First National Plaza, Chicago, Illinois 60603
NRC Project Director: Robert A. Capra
Connecticut Yankee Atomic Power Company, Docket No. 50-213, Haddam
Neck Plant, Middlesex County, Connecticut
Date of amendment request: December 24, 1996 and January 31, 1997
[[Page 14461]]
Description of amendment request: Changes to Administrative
Controls section of the Technical Specifications needed to implement
revised management responsibilities and titles that reflect the
permanently shut down status of plant.
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:
In accordance with 10 CFR 50.92, CYAPCO [Connecticut Yankee
Atomic Power Company] and NNECO [Northeast Nuclear Energy Company]
have reviewed the attached proposed changes and have concluded that
they do not involve a Significant Hazard consideration (SHC). The
basis of this conclusion is that the three criterion of 10 CFR 50.92
are not compromised. The proposed changes do not involve an SHC
because the proposed changes will not:
1. Involve a significant increase in the probability or
consequences of an accident previously evaluated.
No design basis accidents are affected by these proposed
changes. The proposed changes are administrative in nature and are
being proposed to reflect the organizational changes which became
effective December 9, 1996.
The Haddam Neck unit changes are replacement of the Executive
Vice President, Nuclear by the Executive Vice President and Chief
Nuclear Officer along with the replacement of the Vice President,
Haddam Neck by the Unit Director.
No safety systems are adversely affected by the proposed
changes, and no failure modes are associated with the changes.
2. Create the possibility of a new or different kind of accident
from any accident previously evaluated.
There are no changes in any way that the plants are operated due
to this administrative change. The potential for an unanalyzed
accident is not created. There is no impact on plant response, and
no new failure modes are introduced. The proposed administrative and
editorial changes have no impact on safety limits or design basis
accidents, and have no potential to create a new or unanalyzed
event.
3. Involve a significant reduction in a margin of safety.
These changes do not directly affect any protective boundaries
nor do they impact the safety limits for the protective boundaries.
These proposed changes are administrative and editorial in nature.
Therefore there can be no reduction in the margin of safety.
The Commission has provided guidance concerning the application
of the standards in 10 CFR 50.92 by providing certain examples (51
FR 7751, March 4, 1986) of amendments that are considered not likely
to involve an SHC. The changes proposed herein are enveloped by
example (1), since they are purely administrative changes to the
technical specifications to reflect organizational title changes and
to achieve consistence throughout the technical specifications at
Haddam Neck.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: Russell Library, 123 Broad
Street, Middletown, CT 06457
Attorney for licensee: Ms. L. M. Cuoco, Senior Nuclear Counsel,
Northeast Utilities Service Company, Post Office Box 270, Hartford, CT
06141-0270
NRC Project Director: Seymour H. Weiss
Entergy Operations Inc., Docket No. 50-382, Waterford Steam
Electric Station, Unit 3, St. Charles Parish, Louisiana
Date of amendment request: July 17, 1996
Description of amendment request: The proposed change request
modifies Waterford Steam Electric Station, Unit 3, Technical
Specifications 3/4.7.1.3,'' CONDENSATE STORAGE POOL,'' by increasing
the minimum required contained water volume from 82 percent to 91
percent indicated level. This proposed change is required to ensure
that the minimum useable water volume in the Condensate Storage Pool
(CSP) is maintained greater than or equal to 170,000 gallons. The new
minimum level accounts for the minimum level required to prevent
Emergency Feedwater pump suction line vortexing and instrument
measurement uncertainties.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. Will operation of the facility in accordance with this
proposed change involve a significant increase in the probability or
consequences of an accident previously evaluated?
Response: No.
Increasing the minimum required CSP level will insure that the
minimum required 170,000 gallons of water is available for supply to
the Emergency Feedwater System. Maintaining the minimum required
water volume will not increase the probability of any accident
previously evaluated. Additionally, it will not affect the
consequences of any accident. Maintaining at least 170,000 gallons
of water available in the CSP will ensure that the system remains
within the bounds of the accident analysis. Therefore, the proposed
change will not involve a significant increase in the probability or
consequences of any accident previously evaluated.
2. Will operation of the facility in accordance with this
proposed change create the possibility of a new or different type of
accident from any accident previously evaluated?
Response: No.
Increasing the minimum water volume of the CSP from 82 percent
to 91 percent does not create a possibility for a new or different
kind of accident. The CSP will be operated in the same manner as
previously evaluated. Therefore, the proposed change will not create
the possibility of a new or different kind of accident from any
accident previously evaluated.
3. Will operation of the facility in accordance with this
proposed change involve a significant reduction in a margin of
safety?
Response: No.
Operation in accordance with this proposed change will ensure
that the minimum contained water volume of the CSP will remain at
least 170,000 gallons under all conditions. This will maintain the
present margin of safety. Therefore, the proposed change will 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: University of New Orleans
Library, Louisiana Collection, Lakefront, New Orleans, LA 70122
Attorney for licensee: N.S. Reynolds, Esq., Winston & Strawn 1400 L
Street N.W., Washington, D.C. 20005-3502
NRC Project Director: William D. Beckner
Entergy Operations Inc., Docket No. 50-382, Waterford Steam
Electric Station, Unit 3, St. Charles Parish, Louisiana
Date of amendment request: February 5, 1997
Description of amendment request: The proposed amendment will
change Waterford Steam Electric Station, Unit 3, Technical
Specifications 3.1.2.7, 3.1.2.8, 3.5.1, 3.5.4, 3.9.1, and Bases 3/
4.1.2. The proposed change will increase the minimum boron
concentration in the Safety Injection Tanks (SITs) and the Refueling
Water Storage Pool (RWSP) to 2050 ppm to reflect the safety analysis
for fuel Cycle 9.
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 14462]]
The Safety Injection System (SIS) is designed to provide core
cooling in the unlikely event of a loss of coolant accident (LOCA).
The cooling must be sufficient to prevent significant alteration of
core geometry, preclude fuel melting, limit the cladding metal-water
reaction, and remove the energy generated in the core for an
extended period of time following a LOCA. The SIS fluid must contain
the necessary boron concentration to maintain the core subcritical
for the duration of a LOCA.
The proposed change increases the minimum boron concentration in
the SITs and RWSP from 1720 ppm to 2050 ppm. Thus, the SIT/RWSP will
at all times contain sufficient borated water to provide adequate
shutdown margin. Sampling of the system and RWSP required by the
Technical Specifications assures that the required dissolved boron
concentration is present. In addition to its emergency core cooling
function, the SIS functions to inject borated water into the RCS to
increase shutdown margin following a rapid cooldown of the RCS as a
result of a steam line rupture.
Operation of the safety injection system is credited in the
steam line break analysis for causing a decrease in core reactivity.
The current minimum RWSP/SIT concentration to be injected is 1720
ppm. Thus an increase to 2050 ppm will have no adverse affect on
this analysis.
The Mode 5 boron dilution event identifies that with an initial
boron concentration of 1240 ppm, a Keff of 0.98, RCS partially
drained, and one charging pump operational, the minimum possible
time to criticality is greater than 90 minutes. For all other
combinations of Keff, RCS conditions, and number of charging pumps,
the time to loss of shutdown margin is greater than 55 minutes.
Thus, the proposed increase in boron concentration will not affect
the results of the Mode 5 boron dilution event.
The change to the action statement of TS 3.9.1 assures that the
more limiting reactivity condition of a Keff less than 0.95 or a
boron concentration of 2050 ppm specified in the COLR [Core
Operating Limit Report] will be adhered to during refueling
operations.
The upper limit on boron concentration has not changed;
therefore, there will be no affect on boric acid precipitation post-
LOCA.
Therefore, the proposed change will not involve a significant
increase in the probability or consequences of any accident
previously evaluated.
The proposed change does not physically alter the configuration
of the plant and, therefore, does not create the possibility of a
new or different kind of accident from any previously evaluated
accident.
Therefore, the proposed change will not create the possibility
of a new or different kind of accident from any accident previously
evaluated.
The proposed change maintains the minimum of 55 minutes to
criticality for the refueling mode boron dilution event analysis.
The proposed change continues to ensure that borated water of
sufficient concentration is injected from both the SITs and the RWSP
in the event of a LOCA or MSLB [main steam line break] and that
boric acid does not precipitate in the core during long term cooling
following a LOCA.
Therefore, the proposed change will 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: University of New Orleans
Library, Louisiana Collection, Lakefront, New Orleans, LA 70122
Attorney for licensee: N.S. Reynolds, Esq., Winston & Strawn 1400 L
Street N.W., Washington, D.C. 20005-3502
NRC Project Director: William D. Beckner
Nebraska Public Power District, Docket No. 50-298, Cooper Nuclear
Station, Nemaha County, Nebraska
Date of amendment request: February 10, 1997
Description of amendment request: The proposed amendment would
provide the requirements for avoidance and protection from thermal
hydraulic instabilities as described in NRC Generic Letter 94-02,
``Long-Term Solutions and Upgrade of Interim Operating Recommendations
for Thermal Hydraulic Instabilities in Boiling Water Reactors.''
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. The proposed amendment does not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
The proposed amendment does not involve a significant increase
in the probability or consequences of an accident previously
evaluated. In fact, it does not result in an increase in the
probability or consequences of any previously evaluated accidents.
The implementation of [Boiling Water Reactor Owners' Group] BWROG
Long-Term Stability Solution Option I-D at [Cooper Nuclear Station]
CNS does not modify the assumptions contained in the existing
accident analysis. The use of an exclusion region and the operator
actions required to avoid and minimize operation inside the region
do not increase the possibility of an accident.
Conditions of operation outside of the exclusion region are
within the analytical envelope of the existing safety analysis. The
operator action requirement to exit the exclusion region upon entry
minimizes the possibility of an oscillation occurring. The actions
to drive control rods and/or to increase recirculation flow to exit
the region are maneuvers within the envelope of normal plant
evolutions. The flow-biased scram has been analyzed and will provide
automatic fuel protection in the event of an instability. Thus, each
proposed Technical Specification requirement provides defense for
protection from an instability event within the existing assumptions
of the accident analysis.
2. The proposed amendment does not create the possibility of a
new or different kind of accident from any accident previously
evaluated.
As stated above, the proposed Technical Specification
requirements either mandate operation within the envelope of
existing plant operating conditions or force specific operating
maneuvers within those carried out in normal operation. Since
operation of the plant with all of the proposed requirements is
within the existing operating basis, an unanalyzed accident will not
be created through implementation of the proposed change.
3. The proposed amendment does not involve a significant
reduction in a margin of safety.
Each of the proposed requirements for plant thermal-hydraulic
stability provides a means for fuel protection. The combination of
avoiding possible unstable conditions and the automatic flow-biased
reactor scram provides an in-depth means for fuel protection.
Therefore, the individual or combination of means to avoid and
suppress an instability supplements the margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: Auburn Memorial Library, 1810
Courthouse Avenue, Auburn, NE 68305
Attorney for licensee: Mr. John R. McPhail, Nebraska Public Power
District, Post Office Box 499, Columbus, NE 68602-0499
NRC Project Director: William D. Beckner
Northeast Nuclear Energy Company, et al., Docket No. 50-245,
Millstone Nuclear Power Station, Unit No. 1, New London,
Connecticut
Date of amendment request: March 6, 1997
Description of amendment request: During a self assessment, the
licensee identified weaknesses in the current Technical Specifications
regarding allowed outage times for certain specific protective
instrumentation and also for reactor building access control. The
proposed amendment is designed to eliminate these weaknesses by
adopting guidance from NUREG-0123, ``Standard Technical Specifications
for General Electric Boiling Water Reactors (BWR/
[[Page 14463]]
5),'' Revision 3, and NUREG-1433, Standard Technical Specifications
General Electric Plants BWR/4,'' Revision 1.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. The operation of Millstone Nuclear Power Station, Unit No. 1,
in accordance with the proposed amendment, will not involve a
significant increase in the probability or consequences of an
accident previously evaluated.
The inherent redundancy and reliability of the protective
instrumentation trip systems ensure that the consequences of an
accident are not significantly increased. In addition, the
restrictive Allowable Outage Time (AOT) interval limits the
probability of the protective instrument channel being unavailable
and an accident requiring its function from occurring
simultaneously. The requirement that the associated trip function
maintains trip capability ensures that the protective
instrumentation response will occur such that the consequences of an
accident are not different from those previously evaluated.
Instruments addressed in the proposed TS respond to changes in
the plant. The proposed (AOTs) provide a two-hour interval where the
instrument is inoperable, yet the Technical Specification (TS)
Limiting Condition for Operation (LCO) action statement is not
immediately entered. The probability of a plant transient being
initiated by a trip of a coincident channel during surveillance
testing is reduced since the channel under test will only be tripped
for a small portion of the test interval. Therefore, AOTs provided
by the proposed TS have no effect on the probability of occurrence
of previously evaluated accidents.
The proposed TS changes provide a two-hour interval where the
instrument is inoperable, but the TS LCO action statement is not
immediately entered. If a single failure occurred on the other
channel of the trip system being tested and the channel being tested
was not in the trip condition, a valid signal might not provide the
required protective action. The probability of an event requiring
initiation of the protective function within the proposed AOT is
low. Additionally, surveillance testing is not generally performed
on multiple sensors simultaneously. So, other trip functions and
sensors remain operable and the probability of extensive
inoperabilities affecting diverse trip functions is low. A spurious
trip of a coincident channel could initiate a plant transient (for
example, a reactor scram or a main steam isolation valve closure);
however, these transients are bounded by the current analyses.
Moreover, the original TS bases submitted as part of the application
for Millstone Unit No. 1's Provisional Operating License (dated
October 7, 1970) included recognition that instruments would be
inoperable during required functional tests and calibrations. Thus,
these conditions were recognized in the original design bases and
constitute part of the licensing bases of the plant. NUREG-0123
provided specific time frames[,] ...AOTs addressed in the table
notes[,] and specific action statements. Millstone Unit No. 1 AOT
values chosen are consistent with these values and less than those
approved in NUREG-1433 which had a more detailed study performed to
lengthen the AOT value.
The existing TS definition for Instrument Functional Test would
be difficult to satisfy if the LCO condition of tripping the
inoperable channel was performed. A similar problem of complying
with the Instrument Calibration definition also exists. The TS
requirement to perform functional tests and calibrations is not
consistent with a requirement to trip the system under test. The
proposed TS changes permit more complete functional and calibration
testing. For example, the main scram contactors could be included
within the surveillance tests. Therefore, these TS clarifications do
not increase the consequences of any previously analyzed accidents.
The two-hour instrumentation AOT for the Air Ejector Off-Gas
System radiation monitors is slightly less restrictive than that
allowed by the NUREG-0123. Since this requirement was relocated from
NUREG-1433, there is no corresponding requirement for comparison.
These radiation monitors are arranged in a two-out-of-two logic;
therefore, both must trip to initiate the required action (closure
of the off-gas isolation valve to the main stack). This action,
however, is automatically delayed by 15 minutes. A high radiation
condition sensed by the monitor in service would provide sufficient
time to take corrective actions. Since a two-hour AOT is deemed
acceptable for instrumentation in system[s] such as the Reactor
Protection System and Emergency Core Cooling Systems, it is
appropriate to apply a two-hour AOT to these radiation monitors.
Additionally, the NUREG-0123 AOT of one hour does not allow
sufficient time to perform required surveillance testing without
placing undue stress on the test performer. The probability of a
plant transient (e.g., loss of condenser vacuum) resulting from a
trip of the coincident channel during surveillance testing is
reduced since the channel under test will only be tripped for a
small portion of the test interval. This transient is bounded by
existing analyses. Therefore, this proposed AOT will not
significantly increase the probability or consequences of an
accident previously evaluated.
Since no physical change is being made to the secondary
containment, or to any systems or components that interface with the
secondary containment, there is no change in the probability of any
accident analyzed in the UFSAR [Updated Final Safety Analysis
Report].
The proposed change continues to ensure the secondary
containment requirements meet the licensing basis. Also, the
proposed changes are based on Standard Technical Specifications,
NUREG-1433, ``Standard Technical Specifications General Electric
Plants, BWR/4,'' Revision 1 guidelines and implement actions to be
taken when secondary containment integrity is not met. If secondary
containment integrity is not met, existing TS 3.7.C directs the
plant to be placed in an operating condition where secondary
containment is not required, e.g., COLD SHUTDOWN. A four hour
allowable outage time is proposed which provides a period of time to
correct the problem that is commensurate with the importance of
maintaining secondary containment during RUN, STARTUP/HOT STANDBY or
HOT SHUTDOWN. The secondary containment is not an initiator for any
accident. Therefore, the proposed change will not increase the
probability of any previously analyzed accident. This short time
period ensures that the probability of an accident requiring
secondary containment integrity operability occurring during periods
when secondary containment integrity is inoperable is minimal.
The proposed surveillance requirement is based on the NUREG-1433
surveillance requiring periodic confirmation that at least one door
in each of the double-doored accesses to the secondary containment
is closed, provides additional assurance of secondary containment
system integrity. While this is a deviation from NUREG-1433 (which
requires that both doors in each access be closed except for normal
entry and exit), it is consistent with the current definition of
Secondary Containment Integrity, which requires that at least one
door in each access opening be closed. Hence, the deviation is
justifiable and represents increased passive testing which will
provide increased awareness of plant conditions. Increased awareness
of plant conditions should reduce the probability or consequences of
any accident previously evaluated.
Since the aspects of secondary containment integrity affected by
reactor building access control are being revised in this proposed
amendment to agree with the allowable outage time allowed by NUREG-
1433 upon loss of secondary containment integrity, the change does
not involve a significant increase in the probability or
consequences of an accident previously evaluated.
Since the editorial items do not alter the meaning or intent of
any requirements, they do not involve a significant increase in the
probability or consequences of an accident previously evaluated.
2. The operation of Millstone Nuclear Power Station, Unit No. 1,
in accordance with the proposed amendment, will not create the
possibility of a new or different kind of accident from any accident
previously evaluated.
The proposed change to the protective instrumentation trip
system specifications do not create the possibility of a new or
different kind of accident because they do not introduce any new
operational modes or physical modifications to the plant.
Instruments addressed in the proposed TS respond to changes in
the plant. The proposed AOTs provide a two-hour interval where the
instrument is inoperable, yet the TS LCO action statement is not
immediately entered. Given a single failure, this could impact the
response of the trip channel but not the initiation of the event.
The only
[[Page 14464]]
action resulting from the AOTs is to perform testing as required by
TS. Spurious signals during testing could initiate transients but
would be bounded by the previous transient analyses. These tests do
not subject the instruments to any conditions beyond their design
specifications and are performed in accordance with approved testing
standards. This testing ensures equipment operability by identifying
degraded conditions, initiating corrective action and properly
retesting them. Therefore, the proposed TS changes will not
introduce a new or different kind of accident than previously
evaluated.
The two-hour instrumentation AOT for the Air Ejector Off-Gas
System radiation monitors is slightly less restrictive than that
allowed by the NUREG-0123. Since this requirement was relocated from
NUREG-1433, there is no corresponding requirement for comparison.
These radiation monitors are arranged in a two-out-of-two logic;
therefore, both must trip to initiate the required action (closure
of the off-gas isolation valve to the main stack). This action,
however, is automatically delayed by 15 minutes. A high radiation
condition sensed by the monitor in service would provide sufficient
time to take corrective actions. Since a two-hour AOT is deemed
acceptable for instrumentation in system[s] such as the Reactor
Protection System and Emergency Core Cooling Systems, it is
appropriate to apply a two-hour AOT to these radiation monitors.
The proposed changes to Millstone Unit No. 1 Technical
Specifications Section 3.7/4.7 and associated bases were developed
using the guidance provided in the Standard Technical
Specifications, NUREG-1433, ``Standard Technical Specifications
General Electric Plants, BWR/4,'' Revision 1. Augmentation of the
existing surveillance requirements by incorporation of an additional
NUREG-1433 based surveillance, provides additional assurance of
secondary containment system integrity. While this is a deviation
from NUREG-1433 (which requires that both doors in each access be
closed except for normal entry and exit), it is consistent with the
current definition of Secondary Containment Integrity which requires
that at least one door in each access opening be closed. Hence, the
deviation is justifiable and represents increased passive testing
which will provide increased awareness of plant conditions.
Increased awareness of plant conditions will not create the
possibility of a new or different kind of accident from any accident
previously evaluated. Since the proposed changes do not
significantly degrade the present level of system operability and
add provisions from NUREG-1433, the proposed amendment does not
create the possibility of a new or different kind of accident from
any accident previously evaluated.
Since the editorial items do not alter plant configurations or
operating modes, they do not create the possibility of a new or
different kind of accident.
3. The operation of Millstone Nuclear Power Station, Unit No. 1,
in accordance with the proposed amendment, will not involve a
significant reduction in a margin of safety.
The protective instrumentation surveillance requirements provide
verification of the operability of the trip system instrumentation
channels. In addition, the channel that monitors the identical Trip
Function within the same trip system maintains trip capability for
the relatively short duration that the coincidence change is in
effect. This ensures that protective instrumentation reliability is
maintained. The proposed change provides for a specific time period
to perform required surveillances on instrument channels without
trips present in associated trip systems. This time allotment tends
to enhance the margin of safety by decreasing the probability of
unnecessary challenges to safety systems and inadvertent plant
transients.
The proposed TS provide a two-hour interval where the instrument
is inoperable, yet the TS LCO action statement is not immediately
entered. If a single failure occurred on the other channel of the
trip system being tested and the channel being tested was not in the
tripped condition, a valid signal might not provide the required
protective action. The probability of an event requiring initiation
of the protective function within the proposed AOT is low.
Additionally, surveillance testing is not generally performed on
multiple sensors simultaneously. So, other trip functions and
sensors remain operable and the probability of extensive
inoperabilities affecting diverse trip functions is low.
The existing TS definition for Instrument Functional Test would
be difficult to satisfy if the LCO condition of tripping the
inoperable channel was performed. A similar problem of complying
with the Instrument Calibration definition also exists. Moreover,
the original TS bases submitted as part of the application for
Millstone Unit No. 1s Provisional Operating License (dated October
7, 1970) included recognition that instruments would be inoperable
during required functional test and calibrations. Thus, these
conditions were recognized in the original design bases and
constitute part of the licensing bases of the plant. NUREG-0123
provided specific time frames[,]...AOTs addressed in the table
notes[,] and specific action statements. Millstone Unit No. 1 AOT
values chosen are consistent with these values and less than those
approved in NUREG-1433 which had a more detailed study performed to
lengthen the AOT value.
The only action resulting from the proposed TS is to perform
testing as required by TS. Spurious signals during testing could
initiate equipment or plant transients but would be bounded by the
previous transient analysis. These tests do not subject the
instruments to any conditions beyond their design specifications and
are performed in accordance with approved testing standards. This
testing ensures equipment operability by identifying degraded
conditions, initiating corrective action and properly retesting
them. Therefore, the proposed TS do not involve a significant
reduction in a margin of safety.
The two-hour instrumentation AOT for the Air Ejector Off-Gas
System radiation monitors is slightly less restrictive than that
allowed by the NUREG-0123. Since this requirement was relocated from
NUREG-1433, there is no corresponding requirement for comparison.
These radiation monitors are arranged in a two-out-of-two logic;
therefore, both must trip to initiate the required action (closure
of the off-gas isolation valve to the main stack). This action,
however, is automatically delayed by 15 minutes. A high radiation
condition sensed by the monitor in service would provide sufficient
time to take corrective actions. Since a two-hour AOT is deemed
acceptable for instrumentation in system[s] such as the Reactor
Protection System and Emergency Core Cooling Systems, it is
appropriate to apply a two-hour AOT to these radiation monitors and
does not involve a significant reduction in the margin of safety.
The addition of an allowable outage time of four hours for
Secondary Containment Integrity has negligible effect on accident
occurrence or consequences. Since the proposed change does not
involve the addition or modification of plant equipment, is
consistent with the intent of the existing Technical Specifications,
is consistent with the current industry practices as outlined in
NUREG-1433, (except for the deviation noted above), and is
consistent with the design basis of the plant and the accident
analysis, no action will occur that will involve a significant
reduction in a margin of safety.
Since the editorial items do not alter the meaning or intent of
any requirements, they do not affect 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: 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
Northern States Power Company, Docket Nos. 50-282 and 50-306,
Prairie Island Nuclear Generating Plant, Unit Nos. 1 and 2, Goodhue
County, Minnesota
Date of amendment requests: July 28, 1995, as revised February 21,
1997
Description of amendment requests: The proposed amendments would
revise the Technical Specifications (TSs) to allow use of credit for
soluble boron in spent fuel pool criticality analyses. The licensee's
February 21, 1997, submittal is a revision to its original amendment
requests dated July 28, 1995. The generic methodology for crediting
soluble boron in spent fuel rack criticality analyses was approved
[[Page 14465]]
by the NRC on October 25, 1996. However, because of changes made to the
generic methodology as a result of comments from the NRC staff, it was
necessary for NSP to revise its original amendment requests. In
addition, the licensee has revised its request by eliminating the
proposed relocation of the spent fuel pool operating limits to the Unit
1 core operating limits report and will retain these limits in the TSs.
The licensee's original application for amendments was published in
the Federal Register on September 23, 1996, (61 FR 49800).
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. The proposed amendment[s] will not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
There is no increase in the probability of a fuel assembly drop
accident in the spent fuel pool when considering the presence of
soluble boron in the spent fuel pool water for criticality control.
The handling of the fuel assemblies in the spent fuel pool has
always been performed in borated water.
The criticality analysis showed the consequences of a fuel
assembly drop accident in the spent fuel pool are not affected when
considering the presence of soluble boron.
There is no increase in the probability of the accidental
misloading of spent fuel assemblies into the spent fuel pool racks
when considering the presence of soluble boron in the pool water for
criticality control. Fuel assembly placement will continue to be
controlled pursuant to approved fuel handling procedures and will be
in accordance with the Technical Specification spent fuel rack
storage configuration limitations. The addition of the spent fuel
pool storage configuration surveillance in proposed Specification
4.20 will provide increased assurance that a spent fuel pool
inventory verification will be completed in a timely manner after
completion of a fuel handling campaign in the spent fuel pool.
There is no increase in the consequences of the accidental
misloading of spent fuel assemblies into the spent fuel pool racks
because criticality analyses demonstrate that the pool will remain
subcritical following an accidental misloading if the pool contains
an adequate boron concentration. The proposed Technical
Specifications limitations will ensure that an adequate spent fuel
pool boron concentration will be maintained.
There is no increase in the probability of the loss of normal
cooling to the spent fuel pool water when considering the presence
of soluble boron in the pool water for subcriticality control since
a high concentration of soluble boron has always been maintained in
the spent fuel pool water.
A loss of normal cooling to the spent fuel pool water causes an
increase in the temperature of the water passing through the stored
fuel assemblies. This causes a decrease in water density which would
result in a decrease in reactivity when Boraflex neutron absorber
panels are present in the racks. However, since Boraflex is not
considered to be present, and the spent fuel pool water has a high
concentration of boron, a density decrease causes a positive
reactivity addition. However, the additional negative reactivity
provided by the proposed 1800 ppm boron concentration limit, above
that provided by the concentration required to maintain Keff
less than or equal to 0.95 (750 ppm), will compensate for the
increased reactivity which could result from a loss of spent fuel
pool cooling event. Because adequate soluble boron will be
maintained in the spent fuel pool water, the consequences of a loss
of normal cooling to the spent fuel pool will not be increased.
Therefore, based on the conclusions of the above analysis, the
proposed changes will not involve a significant increase in the
probability or consequences of an accident previously evaluated.
2. The proposed amendment[s] will not create the possibility of
a new or different kind of accident from any accident previously
analyzed.
Spent fuel handling accidents are not new or different types of
accidents, they have been analyzed in Section 14.5.1 of the Updated
Safety Analysis Report.
Criticality accidents in the spent fuel pool are not new or
different types of accidents, they have been analyzed in the Updated
Safety Analysis Report and in Criticality Analysis reports
associated with specific licensing amendments for fuel enrichments
up to 5.0 weight percent U-235.
The Prairie Island Technical Specifications currently contain
limitations on the spent fuel pool boron concentration. Current
Specification 3.8.E.2, which covers the storage of restricted fuel
assemblies in an unverified condition, and Specification 3.8.B.1.c
for the loading of fuel assemblies into a cask in the spent fuel
pool, contain requirements for spent fuel pool boron concentration.
The actual boron concentration in the spent fuel pool has always
been kept at a higher value for refueling purposes. New
Specification 3.8.E.2 establishes new boron concentration
requirements for the spent fuel pool water consistent with the
results of the new criticality analysis (Exhibit E [of the February
21, 1997, submittal]).
Since soluble boron has always been maintained in the spent fuel
pool water, and is currently required by Technical Specifications
under some circumstances, the implementation of this new requirement
will have little effect on normal pool operations and maintenance.
The implementation of the proposed new limitations on the spent fuel
pool boron concentration will only result in increased sampling to
verify boron concentration. This increased sampling will not create
the possibility of a new or different kind of accident.
Because soluble boron has always been present in the spent fuel
pool and is required by current Technical Specifications as
discussed above, a dilution of the spent fuel pool soluble boron has
always been a possibility. However, it was shown in the spent fuel
pool dilution evaluation (Exhibit D [of the February 21, 1997,
submittal]) that a dilution of the Prairie Island spent fuel pool
which could reduce the rack Keff to less than 0.95 is not a
credible event. Therefore, the implementation of new limitations on
the spent fuel pool boron concentration will not result in the
possibility of a new kind of accident.
Revised Specifications 3.8.E.1, 5.6.A.1.d and 5.6.A.1.e continue
to specify the requirements for the spent fuel rack storage
configurations, the only significant changes relate to the criteria
for determining the storage configuration. Since the proposed spent
fuel pool storage configuration limitations will be similar to those
currently in the Prairie Island Technical Specifications, the new
limitations will not have any significant effect on normal spent
fuel pool operations and maintenance and will not create any
possibility of a new or different kind of accident. Verifications
will continue to be performed to ensure that the spent fuel pool
loading configuration meets specified requirements.
As discussed above, the proposed changes will not create the
possibility of a new or different kind of accident. There is no
significant change in plant configuration, equipment design or
equipment. The accident analysis in the Updated Safety Analysis
Report remains bounding.
3. The proposed amendment[s] will not involve a significant
reduction in the margin of safety.
The Technical Specification changes proposed by this License
Amendment Request and the resulting spent fuel storage operating
limits will provide adequate safety margin to ensure that the stored
fuel assembly array will always remain subcritical. Those limits are
based on a plant specific criticality analysis (Exhibit E) performed
in accordance [with] the Westinghouse spent fuel rack criticality
analysis methodology described in Reference 4 [in Exhibit A of the
February 21, 1997, submittal].
While the criticality analysis utilized credit for soluble
boron, a storage configuration has been defined using a 95/95
Keff calculation to ensure that the spent fuel rack Keff
will be less than 1.0 with no soluble boron. Soluble boron credit is
used to offset uncertainties, tolerances and off-normal conditions
and to provide subcritical margin such that the spent fuel pool
Keff is maintained less than or equal to 0.95.
The loss of substantial amounts of soluble boron from the spent
fuel pool which could lead to exceeding a Keff of 0.95 has been
evaluated (Exhibit D) and shown to be not credible.
The evaluations in Exhibit D, which show that the dilution of
the spent fuel pool boron concentration from 1800 ppm to 750 ppm is
not credible, combined with the 95/95 calculation, which shows that
the spent fuel rack Keff will remain less than 1.0 when flooded
with unborated water, provide a level of safety comparable to the
conservative criticality analysis methodology required by References
1, 2 and 3 [in Exhibit A of the February 21, 1997, submittal].
[[Page 14466]]
Therefore, the proposed changes in this license amendment will
not result in a significant reduction in the plant's margin of
safety.
Based on the evaluation above, and pursuant to 10 CFR 50,
Section 50.91, Northern States Power Company has determined that
operation of the Prairie Island Nuclear Generating Plant in
accordance with the proposed license amendment request does not
involve any significant hazards considerations as defined by NRC
regulations in 10 CFR 50, Section 50.92.
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 requests involve no significant hazards consideration.
Local Public Document Room location: Minneapolis Public Library,
Technology and Science Department, 300 Nicollet Mall, Minneapolis,
Minnesota 55401
Attorney for licensee: Jay Silberg, Esq., Shaw, Pittman, Potts, and
Trowbridge, 2300 N Street, NW, Washington, DC 20037
NRC Project Director: John N. Hannon
Pacific Gas and Electric Company, Docket Nos. 50-275 and 50-323,
Diablo Canyon Nuclear Power Plant, Unit Nos. 1 and 2, San Luis
Obispo County, California
Date of amendment requests: February 14, 1997
Description of amendment requests: The proposed amendments would
revise the combined Technical Specifications (TS) for the Diablo Canyon
Power Plant (DCPP) Unit Nos. 1 and 2 to revise the surveillance
frequencies from at least once every 18 months to at least once per
refueling interval (nominally 24 months) for 8 slave relay tests, 20
electrical system tests and 1 electrical TS Bases change, and 5
miscellaneous tests.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. The proposed change does not involve a significant increase
in the probability or consequences of an accident previously
evaluated.
The proposed TS surveillance interval increase to 24 months do
not alter the intent or method by which the inspections, tests, or
verifications are conducted; do not alter the way any structure,
system, or component functions; and do not change the manner in
which the plant is operated.
The surveillance, maintenance, and operating histories indicate
that the equipment will continue to perform satisfactorily with
longer surveillance intervals. Few surveillance and maintenance
problems were identified. No problems have recurred following
identification of root causes and implementation of corrective
actions.
There are no known mechanisms that would significantly degrade
the performance of the evaluated equipment during normal plant
operation. All potential time related degradation mechanisms have
insignificant effects in the timeframe of interest (24 months +25
percent, or 30 months). Based on the past performance of the
equipment, the probability or consequences of accidents would not be
significantly affected by the proposed surveillance interval
increases.
Deletion of the phrase ``during shutdown'' for the applicable
electrical TS will not alter the intent or method by which the
inspections, tests, or verifications are conducted; nor alter the
way any structure, system, or component functions. DCPP has
administrative programs in place which require evaluation of risk
and suitability of surveillance and maintenance activities to ensure
that performance during plant operation does not adversely affect
safety.
The administrative change for one PORV TS regarding channel
calibration only maintains the existing surveillance frequency. This
revision does not alter the intent or method by which the
inspections, tests, or verifications are conducted; nor alter the
way any structure, system, or component functions.
Therefore, the proposed changes do not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
2. The proposed change does not create the possibility of a new
or different kind of accident from any accident previously
evaluated.
For the proposed TS changes involving surveillance interval
increases to 24 months, the surveillance and maintenance histories
indicate that the equipment will continue to effectively perform its
design function over the longer operating cycles. Additionally, the
increased surveillance intervals do not result in any physical
modifications, affect safety function performance or the manner in
which the plant is operated, or alter the intent or method by which
surveillance tests are performed. No problems have recurred
following identification of root causes and implementation of
corrective actions. All identified potential time related
degradations have insignificant effects in the timeframe of
interest. The proposed surveillance interval increases would not
affect the type of accident possible.
Deletion of the phrase during shutdown for the applicable
electrical TS does not result in any physical modifications, affect
safety function performance or the manner in which the plant is
operated, or alter the intent or method by which surveillance tests
are performed. DCPP has administrative programs in place which
require evaluation of risk and suitability of surveillance and
maintenance activities to ensure that performance during plant
operation does not adversely affect safety.
The administrative change for one PORV TS regarding channel
calibration only maintains the existing surveillance frequency. This
revision does not result in any physical modifications, affect
safety performance or the manner in which the plant is operated, or
alter the intent or method by which surveillance tests are
performed.
Therefore, the proposed changes do not create the possibility of
a new or different kind of accident from any accident previously
evaluated.
3. The proposed change does not involve a significant reduction
in a margin of safety.
For the proposed TS changes involving surveillance interval
increases to 24 months, evaluation of historical surveillance and
maintenance data indicates there have been few problems experienced
with the evaluated equipment. There are no indications that
potential problems would be cycle ength dependent or that potential
degradation would be significant for the timeframe of interest;
therefore, increasing the surveillance interval will have little, if
any, impact on safety. There is no safety analysis impact since
these changes will have no effect on any safety limit, protection
system setpoint, or limiting condition for operation, and there are
no hardware changes that would impact existing safety analysis
acceptance criteria. Safety margins would not be significantly
affected by the proposed surveillance interval increases.
Deletion of the phrase ``during shutdown'' for the applicable
electrical TS has no safety analysis impact since these changes will
have no effect on any safety limit, protection system setpoint, or
limiting condition for operation, and there are no hardware changes
that would impact existing safety analysis acceptance criteria. DCPP
has administrative programs in place which require evaluation of
risk and suitability of surveillance and maintenance activities to
ensure that performance during plant operation does not adversely
affect safety.
The administrative change for one PORV TS regarding channel
calibration only maintains the existing surveillance frequency. This
revision has no safety analysis impact.
Therefore, the proposed changes do not involve a significant
reduction in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment requests involve no significant hazards consideration.
Local Public Document Room location: California Polytechnic State
University, Robert E. Kennedy Library, Government Documents and Maps
Department, San Luis Obispo, California 93407
Attorney for licensee: Christopher J. Warner, Esq., Pacific Gas and
Electric Company, P.O. Box 7442, San Francisco, California 94120
[[Page 14467]]
NRC Project Director: William H. Bateman
Portland General Electric Company, et al., Docket No. 50-344,
Trojan Nuclear Plant, Columbia County, Oregon
Date of amendment request: January 16, 1997, as supplemented
February 24, 1997.
Description of amendment request: The proposed amendment would
allow pre-operational testing and load handling of spent fuel transfer
and storage casks in the Trojan Fuel Building.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensees have
provided their analysis of the issue of no significant hazards
consideration. The NRC staff has reviewed the licensee's analysis
against the standards of 10 CFR 50.92(c). The staff's review is
presented below:
The proposed changes would not involve a significant increase in
the probability or consequences of an accident previously evaluated.
With the permanent cessation of operations at the Trojan Plant, the
number of potential accidents was reduced to those types of
accidents associated with the storage of irradiated fuel and
radioactive waste storage and handling. Additional events were
postulated for decommissioning activities due to the difference in
the types of activities that were to be performed. The postulated
accidents described in the Defueled Safety Analysis Report (DSAR)
are generally classified as: 1) radioactive release from a subsystem
or component, 2) fuel handling accident, and 3) loss of spent fuel
decay heat removal capability. The postulated events described in
the Decommissioning Plan are grouped as: 1) decontamination,
dismantlement, and materials handling events, 2) loss of support
systems (offsite power, cooling water, and compressed air), 3) fire
and explosions, and 4) external events (earthquake, external
flooding, tornadoes, extreme winds, volcanoes, lightning, toxic
chemical release). These types of accidents are discussed below.
Radioactive release from a subsystem or component involves failure
of a radioactive waste gas decay tank (WGDT) or failure of a chemical
and volume control system holdup tank (HUT). For a failure of a WGDT,
the radioactive contents are assumed to be principally noble gases
krypton and xenon, the particulate daughters of some of the krypton and
xenon isotopes, and trace quantities of halogens. For the failure of a
HUT, the assumptions were full power operation with 1-percent failed
fuel, 40 weeks elapsed since power operation, and 60,000 gallons of
120 deg.F liquid released over a 2-hour period. However, the WGDTs and
HUTs are no longer active and have been drained. Therefore, pre-
operational testing and load handling activities cannot increase the
probability of occurrence of a failure of a WGDT or HUT. Since the
failure of a WGDT or HUT is no longer credible, the consequences of
failure of a WGDT or HUT cannot significantly increase as a result of
pre-operational testing and load handling.
The fuel handling accident involves a stuck or dropped fuel
assembly that results in damage of the cladding of the fuel rods in one
assembly and the release of gaseous fission products. Pre-operational
testing and load handling do not involve the movement of irradiated
fuel. A dummy assembly will be used for fit-up testing. The fuel
handling equipment will be the same as previously analyzed with the
exception of special tools that may be used to manipulate the dummy
fuel assembly. These special tools will be similar in size and weight
to other tools used for underwater manipulation, and therefore, would
not present a new hazard. In addition, the same administrative controls
and physical limitations imposed on any fuel handling operation will be
used for pre-operational testing and load handling. Thus, there is no
increase in the probability of occurrence of a fuel handling accident
over what would be expected for any routine fuel handling operation. If
a dummy fuel assembly were dropped in the spent fuel pool, then only
one fuel assembly could be damaged. Therefore, the consequences of a
dummy fuel assembly drop would be the same as the consequences of the
analysis described in the DSAR. Therefore, the consequences of a dummy
fuel assembly drop are not significantly increased as a result of pre-
operational testing and load handling.
The loss of spent fuel decay heat removal capability involves the
loss of forced spent fuel cooling with and without concurrent spent
fuel pool (SFP) inventory loss. The only requirement to assume adequate
decay heat removal capability for the spent fuel is to maintain the
water level in the SFP so that the spent fuel assemblies remain covered
(i.e., the capability to makeup water to the SFP must be available when
required). The potential events that could result in a loss of spent
fuel decay heat removal capability include external events (explosions,
toxic chemicals, fires, ship collision with the intake structure, oil
or corrosive liquid spills in the river, cooling tower collapse,
seismic events, severe meteorological events), and internal events,
including SFP makeup water system malfunctions. Pre-operational testing
and load handling will not require the use of explosive materials,
toxic chemicals, or flammable materials. The probability of other
external events (e.g., cooling tower collapse) would be unaffected by
the pre-operational testing and load handling activities inside the
fuel building. Pre-operational testing and load handling activities
will not directly interface with the SFP makeup water systems, and
therefore could not affect their probability of failure. The safe load
path and handling height limitations will ensure that a load drop does
not adversely affect the SFP or makeup water systems. Therefore, there
is no significant increase in the probability of a loss of spent fuel
decay heat removal capability. There are no credible adverse
consequences of the loss of spent fuel decay heat removal as the DSAR
demonstrates that adequate time is available to establish a source of
makeup water to the SFP such that uncovering the fuel and an actual
loss of spent fuel cooling is not credible. The postulated events that
could affect the SFP (liner tear/breach and heavy load drop) do not
have a significant adverse effect. In addition, establishment of the
makeup water path and recovery of spent fuel cooling would not be
affected because postulated off-normal events and accidents could not
affect the capability to provide makeup water to the SFP by various
water sources. Therefore, pre-operational testing and load handling
cannot significantly increase the consequences of the loss of spent
fuel decay heat removal.
The events postulated in the Decommissioning Plan are similar to
the DSAR with the exception of decontamination, dismantlement, and
materials handling events. Decontamination events involve gross liquid
leakage from in-situ decontamination equipment or accidental spraying
of liquids containing concentrated contamination. Dismantlement events
include segmentation of components and structures, or removal of
concrete by rock splitting, explosives, or electric and/or pneumatic
hammers. Dismantlement events potentially result in airborne
contamination. Materials handling events involve dropping contaminated
components, concrete rubble, or filters or packages of particulate
materials. Pre-operational testing and load handling activities are
material handling activities and are therefore, within the bounds of
the existing analysis. Therefore, the probability and consequences of
decontamination, dismantlement, and materials handling events would not
be significantly increased.
[[Page 14468]]
Based on the above, the pre-operational testing and load handling
activities do not involve a significant increase in the probability or
consequences of an accident previously evaluated.
The proposed changes would not create the possibility of a new or
different kind of accident from any accident previously evaluated. As
described in the licensee's safety evaluation of the proposed pre-
operational testing and load handling activities, no types of off-
normal events/accidents were determined to have radiological
consequences greater than currently evaluated in the DSAR and
Decommissioning Plan.
The postulated dummy fuel assembly drop is considered the same type
or kind of event as the previously analyzed fuel handling accident,
mainly because the initiator for this postulated event is the same
(i.e., a (non-specified) failure of the fuel handling equipment or the
fuel handling bridge crane. During pre-operational testing and load
handling, a dummy fuel assembly could be dropped in the SFP or the cask
loading pit. As the cask loading pit is similar in construction to the
SFP and the cask loading pit will be flooded with borated water of the
same concentration as the SFP, the differences between the two events
are negligible and the two events may be considered the same type or
kind of accident. Therefore the dummy fuel assembly drop is not a new
or different type or kind of accident.
The postulated transfer cask drop or mishandling event is similar
to a materials handling event. Therefore, the consequences of a
transfer cask drop or mishandling event would not represent a new or
different type or kind of accident.
Based on the above, the pre-operational testing and load handling
activities do not create the possibility of a new or different kind of
accident.
The proposed changes do not involve a significant reduction in the
margin of safety. The Trojan Permanently Defueled Technical
Specifications (PDTS) contain four limiting conditions of operation
that address SFP water level, SFP boron concentration, SFP temperature,
and SFP load restrictions. These PDTS will remain in effect as long as
spent fuel is stored in the SFP, which is in accordance with their
applicability statements. The pre-operational testing and load handling
activities will not affect these PDTS or their bases.
The cask loading pit (CLP) is immediately adjacent to the SFP. The
gate between the CLP and the SFP may be opened to allow a dummy fuel
assembly to moved from the spent fuel storage racks in the SFP to the
basket in the CLP. Opening the gate will allow free exchange of water
between the CLP and the SFP. The water in the CLP must be at
essentially the same level, boron concentration, and temperature as the
SFP prior to the first opening of the gate to ensure that the limited
conditions of operation are continuously satisfied for the SFP.
Therefore, the CLP will be initially filled to about the same level as
the SFP with water that is about the same boron concentration and
temperature as the SFP. With these precautions, the limiting conditions
of operation for SFP level, boron concentration, and temperature will
be continuously maintained and the margin of safety will be unaffected.
Pre-operational testing and load handling activities will involve
lifting and moving heavy loads (e.g., transfer casks). Loads that will
be carried over fuel in the SFP racks and the heights at which they may
be carried will be limited in accordance with LCO 3.1.4, ``Spent Fuel
Pool Load Restrictions,'' in such a way as to preclude impact energies
over 240,000 in-lbs. With this precaution, the limiting condition of
operation pertaining to load restrictions over the SFP will be
satisfied for fuel stored in the SFP racks and the margin of safety
will be unaffected. The safe load path for heavy loads being lifted and
moved outside the SFP will be located sufficiently far from the SFP as
to not have an adverse effect on the SFP in the unlikely event of a
load drop. In addition, the mechanical stops and electrical interlocks
on the fuel building overhead crane will provide additional assurance
that heavy loads are not carried over the fuel in the SFP racks.
Based on the above, the pre-operational testing and load handling
activities will not reduce the margin of safety.
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: Branford Price Millar
Library, Portland State University, 934 S.W. Harrison Street, P.O. Box
1151, Portland, Oregon 97207
Basis for proposed no significant hazards consideration
determination:
Attorney for licensees: Leonard A. Girard, Esq., Portland General
Electric Company, 121 S.W. Salmon Street, Portland, Oregon 97204
NRR Project Director: Seymour H. Weiss
Power Authority of The State of New York, Docket No. 50-286, Indian
Point Nuclear Generating Unit No. 3, Westchester County, New York
Date of amendment request: January 2, 1997
Description of amendment request: The proposed amendment would
allow a change to the current functional testing frequency for
Inservice Inspection of American Society of Mechanical Engineers Code
Class 1, 2, and 3 pumps and valves from the current monthly to a
quarterly testing frequency.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
(1) Does the proposed license amendment involve a significant
increase in the probability or consequences of an accident
previously analyzed?
Response: Operation of Indian Point 3 in accordance with the
proposed license does not involve a significant increase in the
probability or consequences of an accident previously evaluated.
The proposed changes involve no hardware changes, no changes to
the operation of any systems or components, and no changes to
existing structures. 10 CFR 50.55a(g) requires that safety related
components (e.g. - pumps and valves) be tested according to the
requirements of Section XI of the American Society of Mechanical
Engineers (ASME) Boiler and Pressure Vessel Code (Code) and
applicable addenda. The revision of functional test frequencies for
pumps and valves, which are categorized as Code Class 1, 2, or 3,
from a monthly to a quarterly test interval is consistent with NRC
guidance provided in NUREG-1366 and in accordance with recommended
test intervals in the ASME Code. These changes will reduce component
degradation resulting from unnecessary tests and provide better
system availability from not having to remove a system/component
from operability while performing a surveillance. Such changes will
not alter the probability or consequences of any previously analyzed
accidents.
(2) Does the proposed license amendment create the possibility
of a new or different kind of accident from any accident previously
evaluated?
Response: The proposed change does not create the possibility of
a new or different kind of accident from any accident previously
evaluated.
The proposed changes are procedural in nature concerning the
functional testing frequencies of pumps and valves that have
historically shown a high percentage of successfully meeting
surveillance requirements. The methodology of testing these pumps
and valves will remain unchanged. The proposed changes, while
slightly increasing the possibility of an
[[Page 14469]]
undetected pump or valve defect, will not create a new or
unevaluated accident or operating condition.
(3) Does the proposed license amendment involve a significant
reduction in a margin of safety?
Response: The proposed license amendment does not involve a
significant reduction in a margin of safety.
The proposed changes are in accordance with recommendations
provided by the NRC regarding the improvement of Technical
Specifications. These changes will result in the perpetuation of
current safety margins while reducing the testing burden and
decreasing equipment degradation.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: White Plains Public Library,
100 Martine Avenue, White Plains, New York 10601.
Attorney for licensee: Mr. Charles M. Pratt, 10 Columbus Circle,
New York, New York 10019.
NRC Project Director: S. Singh Bajwa, Acting Director
Public Service Electric & Gas Company, Docket No. 50-354, Hope
Creek Generating Station, Salem County, New Jersey
Date of amendment request: February 11, 1997
Description of amendment request: The proposed change to Hope Creek
Technical Specification (TS) Sections 3/4.8.1 ``A.C. Sources,'' 6.8
``Procedures and Programs,'' and the Bases for Section 3/4.8,
``Electrical Power Systems,'' would include: 1) the relocation of
existing surveillance requirements related to diesel fuel oil
chemistry; 2) the introduction of a new program under TS 6.8.4.e,
``Diesel Fuel Oil Testing Program;'' 3) revisions to the TS Bases for
Section 3/4.8 to incorporate information associated with the TS
changes; and 4) editorial changes to implement required corrections.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. The proposed changes do not involve a significant increase in
the probability or consequences of an accident previously evaluated.
The proposed changes involve: 1) no hardware changes; 2) no
significant changes to the operation of any systems or components in
normal or accident operating conditions; and 3) no changes to
existing structures, systems or components. Therefore these changes
will not increase the probability of an accident previously
evaluated.
Establishment of [Emergency Diesel Generator] EDG fuel oil
testing requirements in TS 6.8.4.e is a change that is consistent
with changes made in the improved STS [Standard Technical
Specifications] as contained in Specification 5.5.10 of that
document. These changes establish a new requirement to test for
particulates in the EDG fuel oil, but establish a 92 day test
frequency (as opposed to 31 days in the improved STS) and a 3.0
micron acceptance criteria (as opposed to 0.8 micron in the improved
STS) for particulate testing. [Public Service Electric and Gas
Company] PSE&G concludes that these changes are acceptable based
upon past EDG fuel oil tests for particulates and acceptable
performance of the EDG with 5.0 micron filters. In addition, PSE&G
will utilize more objective test criteria for water and sediment in
the EDG fuel oil than established by the ``clear and bright''
acceptance criteria contained in the improved STS.
Since the EDG fuel oil will still: 1) meet all of the
requirements established for fuel oil specified in the improved STS;
and 2) retain the capability to mitigate the consequences of
accidents described in the [Hope Creek Generating Station] HC Safety
Analysis Report, the proposed changes were determined to be
justified. Based on established fuel oil quality history, the
proposed testing methods and frequencies will not significantly
decrease confidence in fuel oil quality and EDG operability, nor
will they have any negative effect on established plant practices in
regards to the testing of EDG fuel oil. Therefore, these changes
will not involve a significant increase in the consequences of an
accident previously evaluated.
The revisions proposed to the TS Bases are being made to provide
additional information supporting the proposed EDG TS. With the
approval of the proposed TS changes, the associated Bases changes
would be editorial in nature. Therefore, these changes will not
involve a significant increase in the consequences of an accident
previously evaluated.
In addition, the proposed change to [Limiting Condition for
Operation] LCO 3.8.1.1, ACTION c., is considered to be editorial in
nature and will not result in a significant increase in the
consequences of an accident previously evaluated.
2. The proposed change does not create the possibility of a new
or different kind of accident from any accident previously
evaluated.
The HC EDGs are designed to mitigate the consequences of
accidents by providing electrical power to safety-related equipment.
Failure of the EDGs are not considered to initiate any of the
accidents described in the HC Safety Analysis Report. The proposed
changes concern fuel oil system surveillances and testing frequency.
The proposed changes will not adversely impact the operation of any
safety related component or equipment. Since the proposed changes
involve: 1) no hardware changes; 2) no significant changes to the
operation of any systems or components; and 3) no changes to
existing structures, systems or components, there can be no impact
on the occurrence of any accident. 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.
In addition, the proposed change to LCO 3.8.1.1, ACTION c., is
considered to be editorial in nature and will not result in a new or
different kind of accident from any previously evaluated.
3. The proposed change does not involve a significant reduction
in a margin of safety.
Establishment of EDG fuel oil testing requirements in TS 6.8.4.e
is a change that is consistent with changes made in the improved
STS. The proposed changes address: 1) how EDG fuel oil quality is to
be determined; 2) how frequently this determination is to be
performed; and 3) how to control the process for determining fuel
oil acceptability and resultant EDG operability. With the exception
of particulate testing (which is being added) all acceptance
criteria for fuel oil testing remain unchanged. Based on historical
data, EDG fuel oil quality will not be adversely affected or
impacted by the proposed changes. Therefore, the proposed amendment
does not involve any significant reduction in a safety margin.
The revisions proposed to the TS Bases are being made to provide
additional information supporting the proposed EDG TS. With the
approval of the proposed TS changes, the associated Bases changes
would be editorial in nature. Therefore, these changes will not
involve a significant reduction in a safety margin.
In addition, the proposed change to LCO 3.8.1.1, ACTION c., is
considered to be editorial in nature and will not involve a
significant reduction in a safety margin.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: Pennsville Public Library, 190
S. Broadway, Pennsville, New Jersey 08070
Attorney for licensee: M. J. Wetterhahn, Esquire, Winston and
Strawn, 1400 L Street, NW., Washington, DC 20005-3502
NRC Project Director: John F. Stolz
Tennessee Valley Authority, Docket No. 50-390 Watts Bar Nuclear
Plant, Unit 1, Rhea County, Tennessee
Date of amendment request: October 23, 1996, January 31, February
10 and 24 and March 11, 1997.
Description of amendment request: The proposed amendment would
revise
[[Page 14470]]
the Watts Bar Nuclear Plant (WBN) Unit 1 Technical Specifications to
increase the enrichment and storage capacity of the spent fuel pool
racks. The proposed modification increases the (Watts Bar Nuclear
Plant) WBN spent fuel storage capacity from 484 fuel assemblies to 1835
fuel assemblies. The initial enrichment of the fuel to be stored in the
spent fuel storage racks will be increased from 3.5 weight percent
(wt%) to 5.0 wt%. This modification would also change the spacing of
stored fuel assembly center-to-center spacing from a nominal 10.72
inches to 10.375 inches in 24 PaR flux trap rack modules and 8.972
inches in ten smaller burnup credit rack modules to be installed
peripherally along the south and west pool walls and in a single 15 x
15 burnup credit rack to be installed in the cask pit.
In addition to the above proposed revisions, two limiting
conditions for operation will be added to require that the combination
of initial enrichment and burnup of each spent fuel assembly to be
stored is in the acceptable region and to require boron concentration
of the cask pit to be greater than or equal to 2000 parts per million
(ppm) during fuel movement in the flooded cask pit. As an added
protection to the fuel stored in the cask pit area, the Technical
Requirements Manual (TRM) is being revised to require that an impact
shield be in place over the fuel when heavy loads are moved near or
across the cask pit area.
The WBN Unit 1 Technical Specification Bases and the TRM would be
revised to support these changes.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
The Nuclear Regulatory Commission has provided standards for
determining whether a significant hazards consideration exists (10
CFR 50.92(c)). A proposed amendment to an operating license for a
facility involves no significant hazards consideration if operation
of the facility in accordance with the proposed amendment would not
(1) involve a significant increase in the probability or
consequences of an accident previously evaluated; or (2) create the
possibility of a new or different kind of accident from any accident
previously evaluated; or (3) involve a significant reduction in a
margin of safety. Each standard is discussed below for the proposed
amendment.
(1) Operation of the facility in accordance with the proposed
amendment would not involve a significant increase in the
probability or consequences of an accident previously evaluated.
The following potential scenarios were considered:
1. A spent fuel assembly drop.
2. Drop of the transfer canal gate or the cask pit divider gate.
3. A seismic event.
4. Loss-of-cooling flow in the spent fuel pool.
5. Installation activities.
The effect of additional spent fuel pool storage cells fully
loaded with fuel on the first four potential accident scenarios
listed above has been considered. It was concluded that after
installation activities have been completed, the presence of
additional fuel in the pool does not increase the probability of
occurrence of these four events. Also, based on evaluations of bulk
pool temperature, rack seismic responses, and refueling accidents,
it is reasonable to conclude that there is no significant increase
in the consequences of these events after installation is complete
(See Reference 1). During the installation activities, the following
considerations support a conclusion that neither the probability or
consequences of these four scenarios would be significantly
increased.
A spent fuel assembly cannot be dropped during installation of
the 24 Programmed and Remote System Corporation (PaR) flux trap rack
modules because this activity will take place before the end of
operating cycle one and there will be no spent fuel in the WBN pool
to be moved or shuffled. Before installing the ten smaller burnup
credit racks in the pool, some fuel will be moved to create a three
foot lateral free zone clearance from stored fuel. This would
involve a one-time movement of an estimated maximum of 225 fuel
assemblies, which is less that half the fuel movements during one
refueling outage. This does not significantly increase the
probability of dropping a fuel assembly, particularly when the many
administrative controls and physical limitations imposed on fuel
handling operations are considered. The fuel handling system
consists of equipment and structures utilized for safely
implementing refueling operations in accordance with requirements of
General Design Criteria 61 and 62 of 10 CFR 50, Appendix A. The
radiological dose consequences of dropping a 5.0 wt% fuel assembly
are different from the previous FSAR [Final Safety Analysis Report]
evaluation for the 3.5 wt% fuel assembly. The Beta and Gamma doses
decrease and the maximum thyroid dose increase is less than 9%.
Therefore, the change in calculated dose values is insignificant and
remains well within regulatory guidelines.
It may be necessary to move the transfer canal gate and the cask
pit divider gate between their gated and stored positions during
installation of the burnup credit ``baby'' rack modules along the
south and west walls. During rack installation, the previously
mentioned three foot lateral free zone clearance to stored fuel
would exist. Therefore, no heavy load would be carried directly over
irradiated fuel during installation of the racks. There are numerous
design features which comply with NUREG-0612 to preclude these gates
from dropping on spent fuel. These features include design of the
lifting devices, design of the crane, and use of written procedures.
Also, the evaluation results for a gate drop on the racks indicates
that permanent damage to a fuel storage cell is limited to a maximum
depth of less than six inches below the top of the rack with no
effect on the subcriticality of fuel stored in adjacent cells. Based
on the foregoing, it is reasonable to conclude that gate handling
during the installation of the ``baby'' racks would not involve a
significant increase in the probability or consequences of an
accident.
The probability of a seismic event is not related to
installation activities. The worst consequence resulting from a
seismic event during installation activities would occur during
handling of a rack. The consequences would be insignificant because
the Auxiliary Building crane is seismically qualified and both
handling equipment and operations meet the criteria of NUREG-0612.
Nevertheless, if the seismic event resulted in a rack drop, the
consequences are insignificant, i.e., localized damage to the pool
liner and a minor leak rate which would be small in comparison to
available installed makeup capacity. The cooling and shielding of
the spent fuel would remain unaffected. Also the racks being moved
are empty during installation and therefore, the criticality
consequences of seismic events are bounded by evaluations for loaded
racks.
Rack installation activities cannot cause an accidental loss-of-
cooling flow in the spent fuel pool. The vital components of the
spent fuel pool cooling and cleanup system (SFPCCS) are not located
proximate to the pool installation activities. Coolant flow may be
deliberately curtailed to facilitate installation of the ``baby''
racks directly beneath the discharge piping in the southwest corner
of the pool. The effects of such an action would be readily
minimized and made inconsequential during the detailed installation
planning phase by selecting a time when decay heat input from stored
fuel is relatively constant. Also careful preplanning of the work
would minimize out-of-service time and provide for intermittent
coolant flow restart, if necessary, to maintain acceptable bulk
coolant temperatures. Similarly, the effect of an independently
initiated loss-of-coolant flow incident on reracking activities can
be easily accommodated by stopping work, as necessary, to mitigate
any adverse effects on the installation process. The consequences of
loss-of-cooling flow in the spent fuel pool during installation are
bounded by the analysis in Chapter 5 of the report which includes
the situation in which ``baby'' racks and the 15 x 15 cask pit rack
are installed, and the pool is filled to capacity with spent fuel.
With regard to the actual installation activities, the existing
WBN TRM prohibits loads in excess of 2059 pounds from travel over
fuel assemblies in the storage pool and requires the associated
crane interlocks and physical stops be periodically demonstrated
operable. During installation, racks and associated handling tools
will be moved over the spent fuel pool, however there will be no
fuel in the pool when the 24 flux trap rack modules are installed. A
three foot lateral free zone clearance from stored spent fuel
[[Page 14471]]
will be maintained during installation of the ten smaller burnup
credit rack modules. Installation work in the spent fuel pit area
will be controlled and performed in strict accordance with specific
written instructions.
NUREG-0612 states that in lieu of providing a single failure-
proof crane system, the control-of-heavy-loads guidelines can be
satisfied by establishing that the potential for a heavy load drop
is extremely small. Storage rack movements to be accomplished with
the WBN Auxiliary Building crane will conform with NUREG-0612
guidelines in that the probability of a drop of a storage rack is
extremely small. The crane has a tested capacity of 125 tons. The
maximum weight of any existing, replacement, or new storage rack and
its associated handling tool is less than 20 tons. Therefore, there
is ample safety factor margin for movements of the storage racks by
the Auxiliary Building crane. Special lifting devices, which have
redundancy or a rated capacity sufficient to maintain adequate
safety factors, will also be utilized in the movements of the
storage racks. In accordance with NUREG-0612, Appendix B, the safety
margin ensures that the probability of a load drop is extremely low.
Future load travel over fuel stored in a rack specifically
designed for the cask loading area of the cask pit will be
prohibited unless an impact shield, which has been specifically
designed for this purpose, is covering the area. Loads that are
permitted when the shield is in place must meet analytically
determined weight, travel height, and cross-sectional area criteria
that preclude penetration of the shield. A Technical Requirement
(TR) has been proposed that incorporates the previously mentioned
load criteria.
Also a rack change-out sequence is being developed that
addresses removal of the existing racks, movement of the new racks
into the Auxiliary Building, initial staging on the refueling floor,
and final installation in the pool. The change-out sequence
objectives include establishing lift heights, travel distances, and
number of lifts to be as low as reasonably achievable. Accordingly,
it is concluded that the proposed installation activities will not
significantly increase the probability of a load-handling accident.
The consequences of a load-handling accident are unaffected by the
proposed installation activities.
The consequences of a spent fuel assembly drop were evaluated,
and it was determined that the racks will not be distorted such that
the racks would not perform their safety function. The criticality
acceptance criterion, Keff less than or equal to 0.95, is not
violated, and the calculated doses are well within 10 CFR Part 100
guidelines. The radiological consequences of the fuel assembly drop
accident evaluated for WBN, have changed, however, the changes do
not involve a significant increase in consequences and are well
within the 10 CFR 100 requirements.
A TRM change has been proposed that would permit the transfer-
canal gate and the divider gate for the cask pit to travel over fuel
assemblies in the spent fuel pool during movement between their
gated and stored position. Rack damage is restricted to an area
above the active fuel region, therefore, neither criticality nor
radiological concerns exist.
The consequences of a seismic event have been evaluated. The
replacement racks are designed and fabricated and the new racks will
be fabricated to meet the requirements of applicable portions of the
NRC regulatory guides and published standards. Design margins have
been provided for rack tilting, deflection, and movement such that
the racks do not impact each other or the spent fuel pool walls in
the active fuel region during the postulated seismic events. The
free-standing racks will maintain their integrity during and after a
seismic event. The fuel assemblies also remain intact and therefore
no criticality concerns exist.
The spent fuel pool system is a passive system with the
exception of the fuel pool cooling train and heating, ventilating,
and air-conditioning (HVAC) equipment. Redundancies in the cooling
train and HVAC hardware are not reduced by the planned fuel storage
modification. The potential increased heat load resulting from any
additional storage of spent fuel is well within the existing system
cooling capacity. Therefore, the probability of occurrence or
malfunction of safety equipment leading to the loss-of-cooling flow
in the spent fuel pool is not significantly affected. Furthermore,
the consequences of this type incident are not significantly
increased from previously evaluated cooling system loss of flow
malfunctions. Thermal-hydraulic scenarios assume the reracked pool
is approximately 90% full with spent fuel assemblies. From this
starting point, the remaining storage capacity is utilized by
analyzing both normal and unplanned full core off loads using
conservative assumptions and previously established methods.
Calculated values include maximum pool water bulk temperature,
coincident maximum pool water local temperature, the maximum fuel
cladding temperature, time-to-boil after loss-of-cooling paths, and
the effect of flow blockage in a storage cell.
Although the proposed modification increases the pool heat load,
results from the above analyses yield a maximum bulk temperature
less than 160 degrees Fahrenheit which is below the bulk boiling
temperature. Also the maximum local water temperature is below
nucleate boiling condition values. Associated results from
corresponding loss-of-cooling evaluations give minimums of 5.3 hours
before boiling begins and 45 hours before the pool water level drops
to the minimum required for shielding spent fuel. This is sufficient
time to begin utilization of available alternate sources of makeup
cooling water. Also, the effect of the increased thermal loading on
the pool structure, associated cooling system, and components was
evaluated and determined to establish an acceptable design basis
with the new storage configuration. No modifications were necessary
because of the increased temperature.
(2) Operation of the facility in accordance with the proposed
amendment would not create the possibility of a new or different
kind of accident from any accident previously analyzed.
The proposed modification has been evaluated in accordance with
the guidance of the NRC position paper entitled, ``OT Position for
Review and Acceptance of Spent-Fuel Storage and Handling
Applications'', appropriate NRC regulatory guidelines; appropriate
NRC standard review plans; and appropriate industry codes and
standards. Proven analytical technology was used in designing the
planned fuel storage expansion and will be utilized in the
installation process. Basic reracking technology has been developed
and demonstrated in applications for fuel pool capacity increases
that have already received NRC staff approval.
Proposed TSs for the spent fuel storage racks use burnup credit
and fuel assembly administrative placement restrictions for
criticality control. These restrictions are described in the
proposed change to the design features section of the TSs by
reference to the Spent Fuel Pool Modifications report. Additional
evaluations were required to ensure that the criticality criterion,
keff less than 0.95, is maintained. These include evaluation
for the abnormal placement of unirradiated (fresh) fuel assemblies
of 5.0 wt% enrichment into a storage cell location designed for
lower enrichment or irradiated fuel. Soluble boron, for which credit
is permitted under these abnormal conditions, ensures that
reactivity is maintained substantially less than the design
requirement. For example, if the PaR flux trap racks are
inadvertently all loaded with fresh assemblies of the maximum 5.0
wt% fuel instead of observing the 3.8 wt% and 6.75 MWD/KgU controls,
the worth of the 2000 ppm borated water is sufficient to lower the
keff of the storage racks to 0.83. The existing and proposed
TSs require boron concentration in the pool and cask pit to be more
than or equal to 2000 ppm during fuel movement. An analytical
determination of the reactivity worth of 2000 ppm borated water in
the spent fuel storage pool predicted the change in keff to be
approximately 17 percent keff. Although no credit for soluble
boron was proposed in the TSs, it was also determined by an
independent calculation that a minimum concentration of 520 ppm
soluble boron allows the unrestricted storage of 5.0 wt% enriched
fuel in the PaR flux trap racks.
The Holtec-designed peripheral ``baby'' racks and the 15 x 15
racks in the cask loading area can safely and conservatively store
fuel of 5 wt% initial enrichment burned to 41 MWD/kgU or lower
enriched fuel with lower burnup, i.e., fuel of equivalent
reactivity. Evaluations have confirmed that, for the abnormal
placement of a fresh fuel assembly of 5.0 wt% in these racks, the
criticality criterion is maintained with the existing and proposed
TS requirements of 2000 ppm soluble boron.
Although these changes required addressing additional aspects of
a previously analyzed accident, the possibility of a previously
unanalyzed accident is not created.
The impact shield design together with its attendant
administrative controls and NUREG-0612 heavy load lift compliance,
renders the possibility of a heavy load drop
[[Page 14472]]
on fuel as not credible in accordance with the NUREG-0612 single-
failure-proof criteria. Accordingly, since this particular part of
the proposed reracking modification is not a change that could
malfunction by a new single failure, the movement of heavy loads
over the cask pit does not create the possibility of a new or
different kind of accident.
It is therefore concluded that the proposed reracking does not
create the possibility of a new or different kind of accident from
any previously analyzed.
(3) Operation of the facility in accordance with the proposed
amendment would not involve a significant reduction in a margin of
safety.
The design and technical review process applied to the reracking
modification included addressing the following areas:
1.
Nuclear criticality considerations.
2. Thermal-hydraulic considerations.
3. Mechanical, material, and structural considerations.
The established acceptance criterion for criticality is that the
neutron multiplication factor shall be less than or equal to 0.95,
including all uncertainties. The results of the criticality analyses
for the rack designs demonstrate that this criterion is satisfied.
The methods used in the criticality analysis conform to the
applicable portions of NRC guidance and industry codes, standards,
and specifications. In meeting the acceptance criteria for
criticality in the spent fuel pool and the cask loading area, such
that keff is always less than 0.95 at a 95/95 percent
probability tolerance level, the proposed amendment does not involve
a significant reduction in the margin of safety for nuclear
criticality.
Conservative methods and assumptions were used to calculate the
maximum fuel temperature and the increase in temperature of the
water in the spent fuel pit area. The thermal-hydraulic evaluation
used methods previously employed. The proposed storage modification
will increase the heat load in the spent fuel pool, but the
evaluation shows that the existing spent fuel cooling system will
maintain the bulk pool water temperature at or below 160 degrees
Fahrenheit. Thus it is demonstrated that the worst-case peak value
of the pool bulk temperature is considerably lower than the bulk
boiling temperature. Evaluation also shows that maximum local water
temperatures along the hottest fuel assembly are below the nucleate
boiling condition value. Thus, there is no significant reduction in
the margin of safety for thermal hydraulic or spent fuel cooling
considerations.
The mechanical, material, and structural design of the spent
fuel racks is in accordance with applicable portions of NRCs
position in ``OT Position for Review and Acceptance of Spent-Fuel
Storage and Handling applications,'' dated April 14, 1978 (as
modified January 18, 1979), as well as other applicable NRC guidance
and industry codes. The primary safety function of the spent fuel
racks is to maintain the fuel assemblies in a safe configuration
through normal and abnormal loading conditions. Abnormal loadings
that have been evaluated with acceptable results and discussed
previously include the effect of an earthquake and the impact
because of the drop of a fuel assembly. The rack materials used are
compatible with the fuel assemblies and the environment in the spent
fuel pool. The structural design for the new racks provides tilting,
deflection, and movement margins such that the racks do not impact
each other or the spent fuel pit walls in the active fuel region
during the postulated seismic events. Also the spent fuel assemblies
themselves remain intact and no criticality concerns exist. In
addition, finite element analysis methods were used to evaluate the
continued structural acceptability of the spent fuel pit. The
analysis was performed in accordance with ``Building Code
Requirements for Reinforced Concrete,'' (ACI 318-63,77). Therefore,
with respect to mechanical, material, and structural considerations,
there is no significant reduction in a margin of safety.
Summary
Based on the above analysis, TVA has determined that operation
of WBN, in accordance with the proposed amendment, would not: (1)
involve a significant increase in the probability of consequences of
an accident previously evaluated, (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. Therefore, operations of WBN in accordance with the proposed
amendments as described do not involve significant hazard
considerations as defined in 10 CFR 50.92 and that the criteria of
10 CFR 50.91 have accordingly been met.
TVA has also reviewed the NRC examples of licensing amendments
considered not likely to involve significant hazards considerations
as provided in the final adoption of 10 CFR 50.92 published on page
7751 of the Federal Register, Volume 51, No. 44, March 6, 1986.
Example (X) provides four criteria that, if satisfied by a reracking
request, indicate that it is likely no significant hazards
considerations are involved. The criteria and how TVAs amendment
request for WBN complies are indicated below.
Criterion (1):
The storage expansion method consists of either replacing
existing racks with a design that allows closer spacing between
stored spent fuel assemblies or replacing additional racks of the
original design on the pool floor if space permits.
Proposed Amendment:
The WBN reracking involves replacing the existing racks with a
design that allows slightly closer spacing between stored fuel
assemblies and also provides additional rack storage on the pool
floor where space permits.
Criterion (2):
The storage expansion method does not involve rod consolidation
or double tiering.
Proposed Amendment:
The WBN racks are not double tiered, and the racks will sit on
the floor of the spent fuel pool. Additionally, the amendment
application does not involve consolidation of spent fuel.
Criterion (3):
The keff of the pool is maintained less than or equal to
0.95.
Proposed Amendment
The design of the spent fuel racks contains a neutron absorber,
Boral, to allow close storage of spent fuel assemblies while
ensuring that the keff remains less than 0.95 under normal
operating conditions with unborated water in the pool and less than
0.95 under abnormal conditions with soluble boron in the pool.
Criterion (4):
No new technology or unproven technology is utilized in either
the construction process or the analytical techniques necessary to
justify the expansion.
Proposed Amendment:
The construction processes and analytical techniques used in the
fabrication and design are substantially the same as those of
numerous other rack installations, Thus, no new or unproven
technology is utilized in the construction or analysis of the high
density, spent fuel racks at WBN. TVA's contractor, Holtec
International, has previously supplied licensable racks of several
similar design for about 10 other reracking projects
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: Chattanooga-Hamilton County
Library, 1001 Broad Street, Chattanooga, TN 37402
Attorney for licensee: General Counsel, Tennessee Valley Authority,
400 West Summit Hill Drive, ET llH, Knoxville, Tennessee 37902
NRC Project Director: Frederick J. Hebdon
Virginia Electric and Power Company, Docket Nos. 50-280 and 50-281,
Surry Power Station, Unit Nos. 1 and 2, Surry County, Virginia
Date of amendment request: November 26, 1996
Description of amendment request: The proposed changes would
eliminate the records retention requirements from the administrative
section of the Technical Specifications (TS) in accordance with NRC
Administrative Letter 95-06, ``Relocation of Technical Specifications
Administrative Controls Related to Quality Assurance.''
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:
Specifically, operation of the Surry... Power [Station] in
accordance with the
[[Page 14473]]
proposed Technical Specifications changes will not:
(1) Involve a significant increase in the probability or
consequences of an accident previously evaluated. The proposed
administrative changes do not affect equipment or its operation.
Therefore, the likelihood that an accident will occur is neither
increase nor decreased by relocating record retention requirements
from the Technical Specifications to the Operational Quality
Assurance Program. This TS change will not impact the function or
method of operation of plant equipment. Thus, a significant increase
in the probability of a previously analyzed accident does not result
due to this change. No systems, equipment, or components are
affected by the proposed changes. Thus, the consequences of any
accident previously evaluated in the UFSAR [Updated Final Safety
Analysis Report] are not increased by this change.
(2) Create the possibility of a new or different kind of
accident from any accident previously evaluated. The proposed change
does not alter the design or operations of the physical plant. Since
record retention requirements are administrative in nature, a change
to these requirements does not contribute to accident initiation, an
administrative change related to this activity does not produce a
new accident scenario or produce a new type of equipment
malfunction. [These] changes do not alter any existing accident
scenarios. The proposed administrative change does not affect
equipment or its operation, and, thus, does not create the
possibility of a new or different kind of accident. Therefore, the
proposed change does not create the possibility of a new or
different kind of accident.
(3) Involve a significant reduction in a margin of safety.
Section 6.0 of the...Surry Technical Specifications does not have a
basis description. The proposed administrative change does not
affect equipment or its operation, and, thus, does not involve any
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: Swem Library, College of
William and Mary, Williamsburg, Virginia 23185.
Attorney for licensee: Michael W. Maupin, Esq., Hunton and
Williams, Riverfront Plaza, East Tower, 951 E. Byrd Street, Richmond,
Virginia 23219.
NRC Project Director: Mark Reinhart, Acting
Previously Published Notices Of Consideration Of Issuance Of
Amendments To Facility Operating Licenses, Proposed No Significant
Hazards Consideration Determination, And Opportunity For A Hearing
The following notices were previously published as separate
individual notices. The notice content was the same as above. They were
published as individual notices either because time did not allow the
Commission to wait for this biweekly notice or because the action
involved exigent circumstances. They are repeated here because the
biweekly notice lists all amendments issued or proposed to be issued
involving no significant hazards consideration.
For details, see the individual notice in the Federal Register on
the day and page cited. This notice does not extend the notice period
of the original notice.
Baltimore Gas and Electric Company, Docket No. 50-318, Calvert
Cliffs Nuclear Power Plant, Unit No. 2, Calvert County, Maryland
Date of application for amendment: July 31, 1997, as supplemented
February 13, 1997.
Brief description of amendment: The proposed amendment would revise
the Technical Specifications to reduce the minimum Reactor Coolant
System total flow rate from 370,000 gpm to 340,000 gpm. The proposed
changes are necessary to support a larger number of plugged steam
generator tubes for future operating cycles.
Date of publication of individual notice in Federal Register:
February 26, 1997 (62 FR 8780)
Expiration date of individual notice: March 28, 1997
Local Public Document Room location: Calvert County Library, Prince
Frederick, Maryland 20678.
Consolidated Edison Company of New York, Docket No. 50-247, Indian
Point Nuclear Generating Unit No. 2, Westchester County, New York
Date of application for amendment: February 14, 1997
Brief description of amendment: The proposed amendment would revise
the Technical Specifications to permit a one-time extension of the
current steam generator tube inservice inspection cycle. Date of
publication of individual notice in Federal Register: March 4, 1997 (62
FR 9816)
Expiration date of individual notice: March 28, 1997
Local Public Document Room location: White Plains Public Library,
100 Martine Avenue, White Plains, New York 10610.
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: February 17, 1997
Brief description of amendment: Changes to Technical Specification
to implement 10 CFR 50, Appendix J Option B relating to containment
leakage tests.
Date of publication of individual notice in the Federal Register:
February 28, 1997 (62 FR 9214).
Expiration date of individual notice: March 31, 1997
Local Public Document Room location: Coastal Region Library, 8619
W. Crystal Street, Crystal River, Florida 32629
Toledo Edison Company, Centerior Service Company, and The Cleveland
Electric Illuminating Company, Docket No. 50-346, Davis-Besse
Nuclear Power Station, Unit No. 1, Ottawa County, Ohio
Date of amendment request: February 14, 1997
Brief description of amendment request: The proposed amendment
would revise Technical Specification (TS) Section 3/4.5.2, ``Emergency
Core Cooling Systems, ECCS Subsystems - Tavg more than or equal to
280 deg.F.'' Surveillance requirement 4.5.2.f would be modified to
state that opening and closing of the inspection port on the watertight
enclosure for the decay heat valve pit would not require this
surveillance procedure to be performed. The applicable TS bases would
also be changed. Date of publication of individual notice in Federal
Register: February 26, 1997 (62 FR 8783) Expiration date of individual
notice: March 28, 1997
Local Public Document Room location: University of Toledo, William
Carlson Library, Government Documents Collection, 2801 West Bancroft
Avenue, Toledo, Ohio 43606
Notice Of Issuance Of Amendments To Facility Operating Licenses
During the period since publication of the last biweekly notice,
the Commission has issued the following amendments. The Commission has
determined for each of these amendments that the application complies
with the standards and requirements of the Atomic Energy Act of 1954,
as amended (the Act), and the Commission's rules and regulations. The
Commission has made appropriate findings as required by the Act and the
Commission's rules and regulations in
[[Page 14474]]
10 CFR Chapter I, which are set forth in the license amendment.
Notice of Consideration of Issuance of Amendment to Facility
Operating License, Proposed No Significant Hazards Consideration
Determination, and Opportunity for A Hearing in connection with these
actions was published in the Federal Register as indicated.
Unless otherwise indicated, the Commission has determined that
these amendments satisfy the criteria for categorical exclusion in
accordance with 10 CFR 51.22. Therefore, pursuant to 10 CFR 51.22(b),
no environmental impact statement or environmental assessment need be
prepared for these amendments. If the Commission has prepared an
environmental assessment under the special circumstances provision in
10 CFR 51.12(b) and has made a determination based on that assessment,
it is so indicated.
For further details with respect to the action see (1) the
applications for amendment, (2) the amendment, and (3) the Commission's
related letter, Safety Evaluation and/or Environmental Assessment as
indicated. All of these items are available for public inspection at
the Commission's Public Document Room, the Gelman Building, 2120 L
Street, NW., Washington, DC, and at the local public document rooms for
the particular facilities involved.
Baltimore Gas and Electric Company, Docket Nos. 50-317 and 50-318,
Calvert Cliffs Nuclear Power Plant, Unit Nos. 1 and 2, Calvert
County, Maryland
Date of application for amendments: August 1, 1996
Brief description of amendments: The amendments modify the
Technical Specifications requirements to allow use of blind flanges
during Modes 1-4 in the Calvert Cliffs 1 and 2 Containment Purge system
instead of the two outboard 48-inch isolation valves. Date of issuance:
March 7, 1997
Effective date: As of the date of issuance to be implemented by the
end of the 1998 refueling outage for Unit 1; by the end of the 1997
refueling outage for Unit 2.
Amendment Nos.: 221 and 197
Facility Operating License Nos. DPR-53 and DPR-69: Amendments
revised the Technical Specifications.
Date of initial notice in Federal Register: September 11, 1996 (61
FR 47975) The Commission's related evaluation of these amendments is
contained in a Safety Evaluation dated March 7, 1997 No significant
hazards consideration comments received: No
Local Public Document Room location: Calvert County Library, Prince
Frederick, Maryland 20678.
Carolina Power & Light Company, et al., Docket No. 50-400, Shearon
Harris Nuclear Power Plant, Unit 1, Wake and Chatham Counties,
North Carolina
Date of application for amendment: December 30, 1996
Brief description of amendment: The amendment revises chemistry
data for TS Figures 3.4-2 and 3.4-3 and the associated Bases.
Date of issuance: March 7, 1997
Effective date: March 7, 1997
Amendment No.: 68
Facility Operating License No. NPF-63. Amendment revises the
Technical Specifications
Date of initial notice in Federal Register: January 29, 1997 (62 FR
4342) The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated March 7, 1997. No significant hazards
consideration comments received: No
Local Public Document Room location: Cameron Village Regional
Library, 1930 Clark Avenue, Raleigh, North Carolina 27605.
Commonwealth Edison Company, Docket Nos. 50-237 and 50-249, Dresden
Nuclear Power Station, Units 2 and 3, Grundy County, Illinois
Docket Nos. 50-254 and 50-265, Quad Cities Nuclear Power Station,
Units 1 and 2, Rock Island County, Illinois
Date of application for amendments: September 20, 1996, as
supplemented January 21, 1997.
Brief description of amendments: The amendments would update the
pressure- temperature cures contained in the Dresden and Quad Cities
Technical Specifications to 22 Effective Full Power Years. Date of
issuance: February 28, 1997 Effective date: Immediately, to be
implemented within 30 days.
Amendment Nos.: 153, 148, 172 and 168
Facility Operating License Nos. DPR-19, DPR-25, DPR-29 and DPR-30.
The amendments revised the Technical Specifications.
Date of initial notice in Federal Register: December 18, 1996 (61
FR 66703). The January 21, 1997, submittal provided additional
clarifying information that did not change the original proposed no
significant hazards consideration determination. The Commission's
related evaluation of the amendments is contained in a Safety
Evaluation dated February 28, 1997 No significant hazards consideration
comments received: No
Local Public Document Room location: for Dresden, Morris Area
Public Library District, 604 Liberty Street, Morris, Illinois 60450;
for Quad Cities, Dixon Public Library, 221 Hennepin Avenue, Dixon,
Illinois 61021.
Commonwealth Edison Company, Docket Nos. 50-237 and 50-249, Dresden
Nuclear Power Station, Units 2 and 3, Grundy County, Illinois
Docket Nos. 50-254 and 50-265, Quad Cities Nuclear Power Station,
Units 1 and 2, Rock Island County, Illinois
Date of application for amendments: December 6, 1996
Brief description of amendments: The amendments would change the
Technical Specification (TS) by allowing a single control rod to be
moved when the plant is in the Hot Shutdown or Cold Shutdown condition
provided that the one-rod-out interlock is Operable and the reactor
mode switch is in the refuel position.
Date of issuance: March 4, 1997
Effective date: Immediately, to be implemented within 60 days.
Amendment Nos.: 154, 149, 173, 169
Facility Operating License Nos. DPR-19, DPR-25, DPR-29 and DPR-30.
The amendments revised the Technical Specifications.
Date of initial notice in Federal Register: January 15, 1997 (62 FR
2187). The Commission's related evaluation of the amendments is
contained in a Safety Evaluation dated March 4, 1997. No significant
hazards consideration comments received: No
Local Public Document Room location: For Dresden, Morris Area
Public Library District, 604 Liberty Street, Morris, Illinois 60450;
for Quad Cities, Dixon Public Library, 221 Hennepin Avenue, Dixon,
Illinois 61021.
Commonwealth Edison Company, Docket Nos. 50-237 and 50-249, Dresden
Nuclear Power Station, Units 2 and 3, Grundy County, Illinois
Docket Nos. 50-254 and 50-265, Quad Cities Nuclear Power Station,
Units 1 and 2, Rock Island County, Illinois
Date of application for amendments: January 6, 1997
Brief description of amendments: The amendments would change the
technical specifications to clarify and maintain consistency between
the operability requirements for protective instrumentation and
associated automatic bypass features.
Date of issuance: March 14, 1997
Effective date: Immediately, to be implemented within 30 days.
Amendment Nos.: 155, 150, 174, 170
[[Page 14475]]
Facility Operating License Nos. DPR-19, DPR-25, DPR-29 and DPR-30.
The amendments revised the Technical Specifications.
Date of initial notice in Federal Register: February 12, 1997 (62
FR 6573). The Commission's related evaluation of the amendments is
contained in a Safety Evaluation dated March 14, 1997. No significant
hazards consideration comments received: No
Local Public Document Room location: For Dresden, Morris Area
Public Library District, 604 Liberty Street, Morris, Illinois 60450;
for Quad Cities, Dixon Public Library, 221 Hennepin Avenue, Dixon,
Illinois 61021.
Entergy Operations, Inc., Docket No. 50-313, Arkansas Nuclear One,
Unit No. 1, Pope County, Arkansas
Date of amendment request: November 26, 1996 as supplemented by
letters dated December 17, 1996, March 4, 1997, and March 10, 1997
Brief description of amendment: The amendment changes reactor
coolant systems pressure/temperature limits to incorporate updated
parameters and requirements.
Date of issuance: March 14, 1997
Effective date: March 14, 1997
Amendment No.: 188
Facility Operating License No. DPR-51. Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: January 29, 1997 (62 FR
4346) The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated March 14, 1997. No significant hazards
consideration comments received: No.
Local Public Document Room location: Tomlinson Library, Arkansas
Tech University, Russellville, AR 72801
Entergy Operations, Inc., Docket No. 50-368, Arkansas Nuclear One,
Unit No. 2, Pope County, Arkansas
Date of application for amendment: April 11, 1996 as supplemented
by letters dated June 18, and September 5, 1996.
Brief description of amendment: The amendment adds low-temperature
overpressure protection requirements to the Technical Specifications as
proposed by Generic Letter 90-06.
Date of issuance: March 7, 1997
Effective date: March 7, 1997, to be implemented within 30 days.
Amendment No.: 180
Facility Operating License No. NPF-6. Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: May 8, 1996 (61 FR
20846) The Commission's related evaluation of the amendment is
contained in a Safety Evaluation dated March 7, 1997. No significant
hazards consideration comments received: No.
Local Public Document Room location: Tomlinson Library, Arkansas
Tech University, Russellville, AR 72801
GPU Nuclear Corporation, et al., Docket No. 50-219, Oyster Creek
Nuclear Generating Station, Ocean County, New Jersey
Date of application for amendment: August 23, 1996, as supplemented
January 8, 1997 (TSCR 245)
Brief description of amendment: The amendment updates the pressure-
temperature limits up to 22, 27, and 32 effective full power years.
Date of Issuance: March 6, 1997
Effective date: March 6, 1997, to be implemented within 30 days of
issuance
Amendment No.: 188
Facility Operating License No. DPR-16. Amendment revises the
Technical Specifications.
Date of initial notice in Federal Register: September 11, 1996 (61
FR 47977). The January 8, 1997, letter provided clarifying information
within the scope of the original application and did not change the
staff's initial proposed no significant hazards consideration
determination. The Commission's related evaluation of this amendment is
contained in a Safety Evaluation dated March 6, 1997 No significant
hazards consideration comments received: No.
Local Public Document Room location: Ocean County Library,
Reference Department, 101 Washington Street, Toms River, NJ 08753.
Illinois Power Company and Soyland Power Cooperative, Inc., Docket
No. 50-461, Clinton Power Station, Unit No. 1, DeWitt County,
Illinois
Date of application for amendment: October 17, 1996, as
supplemented and modified on December 13, 1996
Brief description of amendment: The amendment revises the Operating
License to reflect the transfer of Soyland Power Cooperative's 13.21-
percent minority ownership of Clinton Power Station to Illinois Power
Company. The Operating License has been revised to delete Soyland Power
Cooperative as an owner.
Date of issuance: March 13, 1997
Effective date: March 13, 1997
Amendment No.: 114
Facility Operating License No. NPF-62: The amendment revised the
Operating License.
Date of initial notice in Federal Register: November 19, 1996 (61
FR 58897) and January 29, 1997 (62 FR 4337) The Commission's related
evaluation of the amendment is contained in a Safety Evaluation dated
March 13, 1997. No significant hazards consideration comments received:
No
Local Public Document Room location: The Vespasian Warner Public
Library, 120 West Johnson Street, Clinton, Illinois 61727
Indiana Michigan Power Company, Docket No. 50-315, Donald C. Cook
Nuclear Plant, Unit No. 1, Berrien County, Michigan
Date of application for amendment: June 19, 1996, and supplemented
September 19, 1996, and December 20, 1996.
Brief description of amendment: The amendment revises the TS to
allow a permanent extension of the interim steam generator tube
voltage-based repair criteria for steam generator tubes used in Cycles
13, 14 and 15 at the Donald C. Cook Nuclear Power Plant, Unit 1.
Date of issuance: March 13, 1997
Effective date: March 13, 1997, with full implementation within 45
days
Amendment No.: 215
Facility Operating License No. DPR-58. Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: July 31, 1996 (61 FR
40022) The Commission's related evaluation of the amendment is
contained in a Safety Evaluation dated March 13, 1997. No significant
hazards consideration comments received: No. The September 19, 1996,
and December 20, 1996, letters provided additional information within
the scope of the original application and did not change the initial
proposed no significant hazards consideration determination.
Local Public Document Room location: Maud Preston Palenske Memorial
Library, 500 Market Street, St. Joseph, Michigan 49085
Indiana Michigan Power Company, Docket Nos. 50-315 and 50-316,
Donald C. Cook Nuclear Plant, Unit Nos. 1 and 2, Berrien County,
Michigan
Date of application for amendments: May 26, 1995, and supplemented
September 26, 1995, August 2, 1996 and February 6, 1997
Brief description of amendments: The amendments revise the TS to
allow operation of Cook Unit 1 at steam generator tube plugging levels
up to 30%. Additional changes to increase operating margins for both
Unit 1 and Unit 2 are also included.
Date of issuance: March 13, 1997
Effective date: March 13, 1997, with full implementation within 45
days
[[Page 14476]]
Amendment Nos.: 214 and 199
Facility Operating License Nos. DPR-58 and DPR-74. Amendments
revised the Technical Specifications.
Date of initial notice in Federal Register: July 19, 1995 (60 FR
37095) The September 26, 1995, August 2, 1996, and February 6, 1997,
supplements provided clarifying information that did not expand the
scope of the initial application or change the staff's proposed no
significant hazards determination. The Commission's related evaluation
of the amendments is contained in a Safety Evaluation dated March 13,
1997. No significant hazards consideration comments received: No.
Local Public Document Room location: Maud Preston Palenske Memorial
Library, 500 Market Street, St. Joseph, Michigan 49085
North Atlantic Energy Service Corporation, Docket No. 50-443,
Seabrook Station, Unit No. 1, Rockingham County, New Hampshire
Date of amendment request: October 17, 1996
Description of amendment request: The amendment revises the
Appendix A Technical Specifications relating to in-core detector
system, seismic instrumentation, meteorological instrumentation, and
turbine overspeed protection. The amendment deletes Limiting Conditions
for Operation and Surveillance Requirements related to these
instruments. The deleted requirements are to be incorporated into the
Seabrook Station Technical Requirements Manual (SSTR). The associated
Bases Sections are also deleted. In addition, Technical Specification
5.5 is deleted but will not be relocated to the SSTR. The amendment
also redesignates Paragraph 2.J of the Seabrook Operating License as
Paragraph 3, and has added new Paragraph 2.J to document the North
Atlantic commitment to relocate the above mentioned Technical
Specification requirements to the SSTR.
Date of issuance: March 12, 1997
Effective date: March 12, 1997
Amendment No.: 50
Facility Operating License No. NPF-86. Amendment revised the
Technical Specifications and Operating License.
Date of initial notice in Federal Register: December 18, 1996 (61
FR 66713). The Commission's related evaluation of the amendment is
contained in a Safety Evaluation dated March 12, 1997. No significant
hazards consideration comments received: No.
Local Public Document Room location: Exeter Public Library,
Founders Park, Exeter, NH 03833.
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 23, 1996, as supplemented
July 17 and December 4, 1996
Brief description of amendment: The amendment modifies the
description of the time constants associated with the Overtemperature
Delta-T and Overpower Delta-T calculations used to establish the trip
setpoints and the time constant used in the rate-lag controller for
Steam Line Isolation, Steam Line Pressure Negative Rate-High.
Date of issuance: March 11, 1997
Effective date: As of the date of issuance, to be implemented
within 60 days.
Amendment No.: 134
Facility Operating License No. NPF-49: Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: June 17, 1996 (61 FR
30639) The July 17 and December 4, 1996, letters provided additional,
clarifying information that did not change the scope of the May 23,
1996, application and the initial proposed no significant hazards
consideration determination. The Commission's related evaluation of the
amendment is contained in a Safety Evaluation dated March 11, 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
Southern California Edison Company, et al., Docket No. 50-362, San
Onofre Nuclear Generating Station, Unit No. 3, San Diego County,
California
Date of application for amendment: February 7, 1997
Brief description of amendment: The amendment defers implementation
of Surveillance Requirement 3.1.5.4 of Technical Specification 3.1.5,
``Control Element Assembly (CEA) Alignment,'' until the next SONGS Unit
3 shutdown, which will be no later than the upcoming Cycle 9 refueling
outage (currently scheduled for April 12, 1997).
Date of issuance: March 5, 1997
Effective date: March 5, 1997
Amendment No.: 126
Facility Operating License No. NPF-15: The amendments revised the
Technical Specifications. Public comments requested as to proposed no
significant hazards consideration: Yes (62 FR 7477 dated February 19,
1997). The notice provided an opportunity to submit comments on the
Commission's proposed no significant hazards consideration
determination. No comments have been received. The notice also provided
for an opportunity to request a hearing by March 21, 1997, but
indicated that if the Commission makes a final no significant hazards
consideration determination any such hearing would take place after
issuance of the amendment. The Commission's related evaluation of the
amendment is contained in a Safety Evaluation dated March 5, 1997.
Attorney for licensee: T. E. Oubre, Esquire, Southern California
Edison Company, P. O. Box 800, Rosemead, California 91770
Local Public Document Room location: Main Library, University of
California, P. O. Box 19557, Irvine, California 92713
Dated at Rockville, Maryland, this 19th day of March 1997.
For the Nuclear Regulatory Commission
Elinor G. Adensam,
Acting Director, Division of Reactor Projects III/IV, Office of Nuclear
Reactor Regulation
[Doc. 97-7508 Filed 3-25-97; 8:45 am]
BILLING CODE 7590-01-F