[Federal Register Volume 61, Number 177 (Wednesday, September 11, 1996)]
[Notices]
[Pages 47973-47987]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: X96-20911]
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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
[[Page 47974]]
make immediately effective any amendment to an operating license upon a
determination by the Commission that such amendment involves no
significant hazards consideration, notwithstanding the pendency before
the Commission of a request for a hearing from any person.
This biweekly notice includes all notices of amendments issued, or
proposed to be issued from August 19, 1996, through August 29, 1996.
The last biweekly notice was published on August 28, 1996 (61 FR
44353).
Notice Of Consideration Of Issuance Of Amendments To Facility
Operating Licenses, Proposed No Significant Hazards Consideration
Determination, And Opportunity For A Hearing
The Commission has made a proposed determination that the following
amendment requests involve no significant hazards consideration. Under
the Commission's regulations in 10 CFR 50.92, this means that operation
of the facility in accordance with the proposed amendment would not (1)
involve a significant increase in the probability or consequences of an
accident previously evaluated; or (2) create the possibility of a new
or different kind of accident from any accident previously evaluated;
or (3) involve a significant reduction in a margin of safety. The basis
for this proposed determination for each amendment request is shown
below.
The Commission is seeking public comments on this proposed
determination. Any comments received within 30 days after the date of
publication of this notice will be considered in making any final
determination.
Normally, the Commission will not issue the amendment until the
expiration of the 30-day notice period. However, should circumstances
change during the notice period such that failure to act in a timely
way would result, for example, in derating or shutdown of the facility,
the Commission may issue the license amendment before the expiration of
the 30-day notice period, provided that its final determination is that
the amendment involves no significant hazards consideration. The final
determination will consider all public and State comments received
before action is taken. Should the Commission take this action, it will
publish in the Federal Register a notice of issuance and provide for
opportunity for a hearing after issuance. The Commission expects that
the need to take this action will occur very infrequently.
Written comments may be submitted by mail to the Chief, Rules
Review and Directives Branch, Division of Freedom of Information and
Publications Services, Office of Administration, U.S. Nuclear
Regulatory Commission, Washington, DC 20555-0001, and should cite the
publication date and page number of this Federal Register notice.
Written comments may also be delivered to Room 6D22, Two White Flint
North, 11545 Rockville Pike, Rockville, Maryland from 7:30 a.m. to 4:15
p.m. Federal workdays. Copies of written comments received may be
examined at the NRC Public Document Room, the Gelman Building, 2120 L
Street, NW., Washington, DC. The filing of requests for a hearing and
petitions for leave to intervene is discussed below.
By October 11, 1996, the licensee may file a request for a hearing
with respect to issuance of the amendment to the subject facility
operating license and any person whose interest may be affected by this
proceeding and who wishes to participate as a party in the proceeding
must file a written request for a hearing and a petition for leave to
intervene. Requests for a hearing and a petition for leave to intervene
shall be filed in accordance with the Commission's ``Rules of Practice
for Domestic Licensing Proceedings'' in 10 CFR Part 2. Interested
persons should consult a current copy of 10 CFR 2.714 which is
available at the Commission's Public Document Room, the Gelman
Building, 2120 L Street, NW., Washington, DC and at the local public
document room for the particular facility involved. If a request for a
hearing or petition for leave to intervene is filed by the above date,
the Commission or an Atomic Safety and Licensing Board, designated by
the Commission or by the Chairman of the Atomic Safety and Licensing
Board Panel, will rule on the request and/or petition; and the
Secretary or the designated Atomic Safety and Licensing Board will
issue a notice of a hearing or an appropriate order.
As required by 10 CFR 2.714, a petition for leave to intervene
shall set forth with particularity the interest of the petitioner in
the proceeding, and how that interest may be affected by the results of
the proceeding. The petition should specifically explain the reasons
why intervention should be permitted with particular reference to the
following factors: (1) the nature of the petitioner's right under the
Act to be made a party to the proceeding; (2) the nature and extent of
the petitioner's property, financial, or other interest in the
proceeding; and (3) the possible effect of any order which may be
entered in the proceeding on the petitioner's interest. The petition
should also identify the specific aspect(s) of the subject matter of
the proceeding as to which petitioner wishes to intervene. Any person
who has filed a petition for leave to intervene or who has been
admitted as a party may amend the petition without requesting leave of
the Board up to 15 days prior to the first prehearing conference
scheduled in the proceeding, but such an amended petition must satisfy
the specificity requirements described above.
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,
[[Page 47975]]
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.
Baltimore Gas and Electric Company, Docket Nos. 50-317 and 50-318,
Calvert Cliffs Nuclear Power Plant, Unit Nos. 1 and 2, Calvert
County, Maryland
Date of amendments request: August 1, 1996
Description of amendments request: The amendment will allow use of
blind flanges during MODES 1-4 in the Calvert Cliffs Units 1 and 2
Containment Purge Systems. These flanges will establish integrity in
Mode 5, prior to entering Mode 4, and maintain it in Modes 1-4,
functions presently served by the valve.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. Would not involve a significant increase in the probability
or consequences of an accident previously evaluated.
The purpose of the Containment Purge System is to provide
ventilation for the containment while in a shutdown condition.
Valves, which are disabled in the shut position in Modes 1-4, may be
opened in Modes 5 and 6 to allow air flow, are provided in the
supply and exhaust piping, and are automatically shut on a
Containment Radiation Signal to prevent release of radioactive
material in the event of a fuel handling incident. Manual operation
is also provided. In Modes 1-4, the valves are kept shut to provide
containment integrity to withstand a presumed increase in
containment pressure in the event of a loss-of-coolant accident. The
proposed change will allow blind flanges to serve in place of the
purge valves in Modes 1-4 by blocking off the purge penetration on
both the supply and exhaust sides. The blind flanges will provide
the same level of containment integrity previously provided by the
purge valves. The revised Technical Specifications will continue to
verify containment building leakage is maintained within the
allowable limits by requiring the performance of a 10 CFR Part 50,
Appendix J, Type B, leakage test on the blind flanges. The outside
valve in each containment purge penetration will be removed and the
inside valves will be left in place. The remaining inside valves
will no longer by required to provide containment integrity in Modes
1-4. Only one of each pair of valves was credited for containment
closure (Modes 5 and 6); therefore, removing the outside valves and
the associated automatic closure signals is not a modification of
the required capability to close the penetration. The inside valves
will maintain their current safety function to close containment (if
needed) by closing either on a Containment Radiation Signal (Mode 6)
or manually (Modes 5 and 6). The Technical Specification
surveillances associated with the purge valves will be changed to
reflect the proposed modification to the plant. Since the blind
flanges will limit radiological releases in Modes 1-4, and the purge
valves will limit radiological releases in Modes 5 and 6, the
proposed change will not increase the consequences of an accident
previously evaluated.
The Containment Purge System is not an accident initiator but
acts to limit the consequences of accidents. The system will provide
containment isolation in Modes 1-4 as before, and the inside valves
will still be available to close in Modes 5 and 6. Therefore, the
proposed change does not increase the probability of an accident
previously evaluated.
As stated above, the proposed changes do not involve a
significant increase in the probability or consequences of an
accident previously evaluated.
2. Would not create the possibility of a new or different type
of accident from any accident previously evaluated.
This requested change does not involve a significant alteration
of the operation of the plant, and no new accident initiation
mechanism is created by the modification. Four purge valves per unit
currently provide containment closure in Modes 5 and 6. The outside
valve in the supply and the exhaust lines will be removed to allow
for installation of a blind flange in each line. The remaining
supply and exhaust valves inside containment will continue to
provide containment closure. The function currently performed by the
four purge valves in Modes 1, 2, 3 and 4 will be performed by the
blind flanges. Other, similar, blind flanges have been in service in
the plant for a number of years, and have proven reliable. The
Technical Specification surveillances associated with the testing of
the purge valves and flanges will be changed to reflect the proposed
modification to the plant. Therefore, this change does not create
the possibility of a new or different type of accident from any
accident previously evaluated.
3. Would not involve a significant reduction in the margin of
safety.
The valves in the Containment Purge System currently provide
containment integrity during Modes 1, 2, 3 and 4, and containment
closure during Modes 5 and 6. The function currently performed by
the purge valves in Modes 1, 2, 3 and 4 will be performed by the
blind flanges. Because of their design and mounting method, the
blind flanges will perform the containment integrity function as
well as, or better than, the purge valves. In Modes 1-4, the double
o-rings in the blind flanges will provide single-failure protection
similar to the other existing Type B penetrations. The established
allowable containment building leakage rate will be maintained by
the implementation of a requirement to perform 10 CFR Part 50,
Appendix J, Type B, leakage rate on the installed blind flanges. The
outside valve in each purge containment penetration will be removed.
Single failure is not assumed in the fuel handling accident
analysis, therefore, removing the outside valves and their
Containment Radiation Signal channels is not a modification of the
required capability to close the penetration. The remaining inside
valves will continue to provide automatic and manual containment
closure in Mode 6 to mitigate the effects of a fuel handling
accident. The Technical Specification surveillances associated with
purge valve testing will be changed to reflect the proposed
modification to the plant. Therefore, this 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 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendments request
[[Page 47976]]
involves no significant hazards consideration.
Local Public Document Room location: Calvert County Library, Prince
Frederick, Maryland 20678.
Attorney for licensee: Jay E. Silbert, Esquire, Shaw, Pittman,
Potts and Trowbridge, 2300 N Street, NW., Washington, DC 20037.
NRC Project Director: Jocelyn A. Mitchell, Acting Director
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: August 7, 1996
Brief description of amendment: The amendment proposes revising the
Technical Specifications (TSs) to allow the use of 10 CFR Part 50,
Appendix J, Option B, Performance-Based Containment Leakage Rate
Testing. This performance-based Option B may be used as an alternative
to the requirements in Appendix J, ``Primary Reactor Containment
Leakage Testing for Water-Cooled Power Reactors,'' of 10 CFR Part 50.
To implement Option B to Appendix J, the amendment proposes modifying
TSs to eliminate reference to the prescriptive Appendix J requirements
and instead reference NRC Regulatory Guide 1.163, ``Performance-Based
Containment Leak-Test Program.'' The amendment also proposes an
editorial correction to the mathematical formula minimum testing
frequency in the basis for TS 4.1.
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 proposed change does not involve a significant increase
in the probability or consequences of an accident previously
evaluated.
For Indian Point Unit No. 2, the integrated leak rate testing
[ILRT] as-found measured leakage rate acceptance criteria is changed
from 0.75 La to 1.0 La. This change is consistent with the revised
10 CFR 50 Appendix J, NEI 94-01, ``Industry Guidelines for
Implementing Performance-Based Option of 10 CFR Part 50, Appendix
J.'' In addition, an as-found leakage rate acceptance criteria of
1.0 LA for Type A tests is consistent with the design basis and
accident analysis assumptions. The as-left acceptance criteria
remains unchanged at 0.75 La in accordance with the NEI guidance.
Therefore, prior to entering an operating mode where containment
integrity is required the as-left leakage rate will not exceed 0.75
La. The combined leakage rate for containment isolation valves
listed in Technical Specification Table 4.4-1 subject to gas or
nitrogen pressurization testing, air lock testing, and portions of
the sensitive leakage rate test which pertain to containment
penetrations and double-gasketed seals shall be less than 0.6 La.
The extensive operations and testing experience derived from
industry show that risk to the general population is generally
insensitive to changes in the allowable leakage rate. It has been
determined that the allowable containment leakage can be increased
by one to two orders of magnitude without significantly impacting
the estimates of population dose in the event of an accident.
Furthermore, the Indian Point Unit No. 2 ILRT test history provides
substantial justification for the proposed changes.
Test results demonstrate that IP2 [Indian Point 2] has a low
leakage containment and that the proposed changes would not
jeopardize the ability of the containment to maintain the leakage
rate at or below the required limits. The proposed change to
Technical Specification 4.1 Basis represent a minor editorial
correction to the mathematical formula for minimum testing frequency
which does not change the formula. Therefore, the probability and
the consequence of a design basis accident are not being increased
by the proposed changes.
2. The proposed change does not create the possibility of a new
or different kind of accident from any accident previously
evaluated.
Plant systems and components will not be operated in a different
manner as a result of the proposed Technical Specification change.
The proposed change permits a performance-based approach to
determining the leakage-rate test frequency for the containment and
containment penetrations (Type A, B, and C tests). There are no
plant modifications, or changes in methods of operation. Therefore,
the changes in testing intervals for the containment and containment
penetrations have no affect on the probability of occurrence of a
LOCA [loss-of-coolant-accident]. The Limiting Conditions for
Operation are not being changed. Changing the as-found leakage-rate
acceptance criterion to 1.0 La does not increase the probability or
consequences of an accident. Changing the test interval for the
containment and containment penetrations does not create any new
accident precursors or methods of operation. The proposed change to
Technical Specification 4.1 Basis represent a minor editorial
correction to the mathematical formula for minimum testing frequency
which does not change the formula. Therefore, the possibility for an
accident of a different type than was previously evaluated in the
safety analysis report is not created by the proposed Technical
Specification.
3. The proposed change does not involve a significant reduction
in a margin of safety.
While the proposed changes do increase the probability for
malfunction of equipment important to safety due to the longer
intervals between leakage tests, it has been estimated that the
longer test intervals will have an insignificant increase in the
overall accident risk to the public. This increase has been reviewed
and found to be acceptable by the NRC as documented in NUREG-1493
and the recent rulemaking to 10 CFR 50 Appendix J. We also agree
that this increase in accident risk is insignificant. Changing the
as-found acceptance criterion to 1.0 La does not increase the
consequences of an accident, since the accident analysis assume a
leakage rate of La for design basis accidents. The as-left Type A
test acceptance criterion remains at less than 0.75 La. Given that
the Indian Point Unit No. 2 ILRT test history show no failures
during plant life, the proposed changes should not lead to a
significant probability of creating new leakage paths or increased
leakage rates. The proposed change to Technical Specification 4.1
Basis represent a minor editorial correction to the mathematical
formula for minimum testing frequency which does not change the
formula. Therefore, the accident analysis assumptions for design
basis accidents are unaffected and the margin of safety is not
decreased by the proposed Technical Specification change.
Public Document Room location: White Plains Public Library, 100
Martine Avenue, White Plains, New York 10610.
Consumers Power Company, Docket No. 50-255, Palisades Plant, Van
Buren County, Michigan
Date of amendment request: January 18, 1996
Description of amendment request: The proposed amendment would
delete the requirement to perform inservice inspections of the primary
coolant pump (PCP) flywheels.
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 following evaluation supports the finding that operation of
the facility in accordance with the proposed change to the Technical
Specifications would not:
1. Involve a significant increase in the probability or
consequences of an accident previously evaluated.
The proposed change to the Technical Specifications would delete
the requirement to perform non-destructive examination of the upper
flywheel on the PCPs. The fracture mechanics analyses conducted to
support the change show that a preexisting crack sized just below
detection level will not grow to the flaw size necessary to result
in flywheel failure within the life of the plant. This analysis
conservatively assumes minimum material properties, maximum flywheel
accident speed, location of the flaw in the highest stress area and
a number of startup/shutdown cycles eight times greater than
expected. Since an existing flaw in the flywheel will not grow to
the allowable flaw size under normal operating conditions or to the
critical flaw size under LOCA [loss-of-coolant accident] conditions
over the life of the plant, elimination of inservice inspection for
such cracks during the plant's life will not involve a significant
increase in the
[[Page 47977]]
probability of an accident previously considered.
The proposed changes do not increase the amount of radioactive
material available for release or modify any systems used for
mitigation of such releases during accident conditions. Therefore,
operation of the facility in accordance with the proposed change to
the Technical Specifications would not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
2. Create the possibility of a new or different kind of accident
from any previously evaluated.
The proposed change to the Technical Specifications would not
change the design, configuration, or method of operation of the
plant and therefore, operation of the facility in accordance with
the proposed change to the Technical Specifications would not create
the possibility of a new or different kind of accident from any
previously evaluated.
3. Involve a significant reduction in a margin of safety.
The proposed change to the Technical Specifications would not
result in a significant reduction in the margin of safety.
Significant conservatisms have been used for calculating the
allowable flaw size, critical flaw size and crack growth rate in the
PCP flywheels. These include minimum material properties, maximum
flywheel accident speed, location of the postulated flaw in highest
stress area and a number of startup/shutdown cycles eight times
greater than expected. Since an existing flaw in the flywheel will
not grow to the maximum allowable flaw size under normal operating
conditions or to the critical flaw size under LOCA conditions over
the life of the plant, elimination of inservice inspections for such
cracks during the plant's life will 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: Van Wylen Library, Hope
College, Holland, Michigan 49423
Attorney for licensee: Judd L. Bacon, Esquire, Consumers Power
Company, 212 West Michigan Avenue, Jackson, Michigan 49201
NRC Project Director: John Hannon
Duke Power Company, Docket Nos. 50-413 and 50-414, Catawba Nuclear
Station, Units 1 and 2, York County, South Carolina
Date of amendment request: August 8, 1996
Description of amendment request: The proposed amendments would
change the Technical Specifications (TS) of each unit to reference
updated or recently approved methodologies used to calculate cycle-
specific limits contained in the Core Operating Limits Report (COLR).
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 will not involve a significant increase
in the probability or consequences of an accident previously
evaluated.
The proposed changes are administrative in nature, and do not
affect any system, procedure, or manipulation of any equipment which
could affect the probability or consequences of any accident.
(2) The proposed changes will not create the possibility of a
new or different kind of accident from any accident previously
evaluated.
The proposed changes are administrative in nature, and cannot
introduce any new failure mode or transient which could create any
accident.
(3) The proposed changes will not involve a significant
reduction in a margin of safety.
The proposed changes are administrative in nature, and will not
affect any operating parameters or limits which could result in a
reduction in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
proposed amendments involve no significant hazards consideration.
Local Public Document Room location: York County Library, 138 East
Black Street, Rock Hill, South Carolina 29730
Attorney for licensee: Mr. Albert Carr, Duke Power Company, 422
South Church Street, Charlotte, North Carolina 28242
NRC Project Director: Herbert N. Berkow
Duke Power Company, Docket Nos. 50-369 and 50-370, McGuire Nuclear
Station, Units 1 and 2, Mecklenburg County, North Carolina
Date of amendment request: June 21, 1996
Description of amendment request: The proposed amendments would
revise the term ``lifting loads'' used in Technical Specification
3.9.6b.2, Manipulator Crane, to ``lifting force.'' This revision will
clarify that the static loads associated with the lifting tool, drive
rod and control rod weights are not included in the lifting force
limit.
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:
Will the change involve a significant increase in the
probability or consequences of an accident previously evaluated?
No. The proposed change is administrative in nature and does not
represent any changes to the refueling process in the field. It more
accurately describes the components for which the LCO's [Limiting
Condition for Operation] protection is intended as well as giving a
more accurate description of the auxiliary hoist's minimum capacity.
It also broadens the domain of activities for which protective
measures are taken by including drag load testing into monitored
activities. At CNS [Catawba Nuclear Station], the auxiliary hoists
and the manipulator cranes are rated at [greater than or equal to]
3000 pounds and are surveillance tested to greater than 1000 pounds.
This brackets the limit force lifting value change from 600 to 1000
pounds in the amendment proposal.
Will the change create the possibility of a new or different
kind of accident from any accident previously evaluated?
No. This proposed administrative change reflects no changes in
the refueling processes, or any systems, structures or components
connected with the refueling process.
Will the change involve a significant reduction in a margin of
safety?
No. The proposed administrative change has no impact on
refueling processes, systems, structures or components, and does not
result in any significant reduction in a margin of safety. The
subject change only clarifies the original intent of the
specification and more accurately describes the involved components,
component capacities and the domain of activities for which measures
are taken to protect the reactor internals.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: Atkins Library, University of
North Carolina, Charlotte (UNCC Station), North Carolina 28223
Attorney for licensee: Mr. Albert Carr, Duke Power Company, 422
South Church Street, Charlotte, North Carolina 28242
NRC Project Director: Herbert N. Berkow
GPU Nuclear Corporation, et al., Docket No. 50-219, Oyster Creek
Nuclear Generating Station, Ocean County, New Jersey
Date of amendment request: August 23, 1996 (TSCR 245)
Description of amendment request: The amendment request proposes
new pressure-temperature (P-T) limits up to
[[Page 47978]]
22, 27, and 32 effective full power years (EFPY). The new sets of P-T
curves would be used beyond 17 EFPY in the future as the corresponding
EFPY of operation is completed.
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:
We have determined that this change request with respect to P-T
limits involves no significant hazards considerations in that
operation of the Oyster Creek Plant in accordance with the proposed
amendment, will not:
1. Involve a significant increase in the probability of an
accident because the new limits account for the increase in RT
NDT, including statistical uncertainty, due to neutron
irradiation of the reactor vessel as well as establishing initial RT
NDT on the basis of current Code requirements, also including
statistical uncertainty, in accordance with Reg. Guide 1.99, Rev. 2.
The new P-T curves will assure that brittle fracture of the reactor
vessel is prevented.
2. Create the probability of a new or different kind of accident
from any accident previously evaluated. These new limits are the
result of the calculation methodology in Reg. Guide 1.99, Rev. 2
[Radiation Embrittlement of Reactor Vessel Materials], as required
by Generic Letter 88-11 [NRC Position on Radiation Embrittlement of
Reactor Materials and its Impact on Plant Operations]. Primary
system configuration and function remain unchanged.
3. Involve a significant reduction in margin of safety because
the bases for the margin of safety remain the same as current
limits, i.e., ASME [American Society of Mechanical Engineers], Sect.
XI, App. G for available fracture toughness and applied stress
intensity, Reg. Guide 1.99, Rev. 2 for calculating applied stress
intensity, Reg. Guide 1.99, Rev. 2 for calculating adjusted RT
NDT and 10 CFR 50, App. G, for criticality conditions.
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: Ocean County Library,
Reference Department, 101 Washington Street, Toms River, NJ 08753
Attorney for licensee: Ernest L. Blake, Jr., Esquire. Shaw,
Pittman, Potts & Trowbridge, 2300 N Street, NW., Washington, DC 20037.
NRC Project Director: John F. Stolz
Illinois Power Company and Soyland Power Cooperative, Inc., Docket
No. 50-461, Clinton Power Station, Unit No. 1, DeWitt County,
Illinois
Date of amendment request: August 15, 1996
Description of amendment request: The proposed amendment would
modify the Clinton Power Station Technical Specifications to
incorporate the revised Safety Limit Minimum Critical Power Ratio
(SLMCPR) as calculated by General Electric (GE) for Cycle 7 operation.
The need to change the SLMCPR resulted from the 10 CFR Part 21
condition reported by GE in their letter to the NRC dated May 24, 1996.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration which is presented below:
(1) This change does not involve a significant increase in the
probability or consequences of any accident previously evaluated. In
lieu of utilizing a potentially nonconservative generic value, this
change revises the SLMCPR to be appropriately conservative as it has
been specifically calculated on a plant- and cycle-specific basis.
Although the SLMCPR does not apply (i.e., is not assumed or required
to be met) during any analyzed accident, the MCPR fuel cladding
Safety Limit ensures that during normal operation and during
anticipated operational occurrences (AOOs), at least 99.9% of the
fuel rods in the core do not experience transition boiling. The
revised value for the SLMCPR is determined using the same
methodology as the previous SLMCPR with the exception that it
utilizes plant specific conditions to determine the safety limit.
The revised SLMCPR, therefore, accounts for actual expected power
distributions in the Clinton Power Station (CPS) core as well as
CPS-specific uncertainties. This provides a more conservative SLMCPR
than the generic value used previously.
The proposed change does not affect any of the parameters or
conditions that contribute to initiation of any accidents previously
evaluated. In addition, the proposed change does not affect the
ability of any plant systems or equipment to operate as assumed in
the safety analyses. The revised SLMCPR will continue to ensure that
the fuel cladding integrity is not lost as a result of over-heating
during normal plant operation or any AOO. As a result, the proposed
change will not result in a significant increase in the consequences
of any accident previously evaluated.
(2) The proposed change does not involve any new modes or
operation, any changes to setpoints, or any plant modifications.
Further, the incorporation of a revised MCPR safety limit, which has
been determined to be acceptable for CPS Cycle 7 operation, does not
result in the creation of any new failure modes or potential
precursors to an accident. Therefore, the proposed change does not
create the possibility of a new or different type of accident from
any accident previously evaluated.
(3) The proposed SLMCPR has been evaluated to ensure that during
normal operation and during AOOs, at least 99.9% of the fuel rods in
the core do not experience transition boiling. As noted above, the
revised SLMCPR has been determined using the same methodology as
used previously with the exception of using CPS Cycle 7 specific
core and fuel design data. This change ensures that the margin of
safety for fuel cladding integrity is maintained by providing a CPS
specific MCPR safety limit as opposed to utilizing a potentially
less conservative generic limit. Therefore, the implementation of
the proposed change to the SLMCPR 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: Vespasian Warner Public
Library, 120 West Johnson Street, Clinton, Illinois 61727
Attorney for licensee: Leah Manning Stetzner, Vice President,
General Counsel, and Corporate Secretary, 500 South 27th Street,
Decatur, Illinois 62525
NRC Project Director: Gail H. Marcus
Southern California Edison Company, et al., Docket Nos. 50-361 and
50-362, San Onofre Nuclear Generating Station, Unit Nos. 2 and 3,
San Diego County, California
Date of amendment requests: May 8, 1996
Description of amendment requests: The licensee proposes to revise
improved Technical Specifications (TS) 3.9.4 and 3.9.5 to facilitate
testing of low pressure safety injection system components and permit
additional flexibility in scheduling maintenance on the shutdown
cooling system.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. The proposed change will not involve a significant increase
in the probability or consequences of an accident previously
evaluated.
Limiting Conditions for Operation (LCO) in Technical
Specifications (TSs) 3.9.4 and 3.9.5 define the operability
requirements for the Shutdown Cooling (SDC) system during refueling
operations (Mode 6) while the water level above the top of the
reactor vessel
[[Page 47979]]
flange is at least 23 feet and less than 23 feet, respectively. The
objective of these TSs is to ensure that 1) sufficient cooling is
available to remove decay heat, 2) the water in the reactor vessel
is maintained below 140 deg.F, and 3) sufficient coolant circulation
is maintained in the reactor core to minimize boron stratification
leading to a boron dilution incident.
The proposed TS changes affect the current limits imposed while
ensuring adherence to the bases of the TS. No plant modifications
are being made. The reactor cavity water level limitations and SDC
system required operating times are being changed based on plant
specific calculations and the objectives of the TSs are being
maintained.
1) Reduce the water level where two loops of SDC are required
from 23 feet to 20 feet above the reactor vessel flange,
Prior to the approval of Unit 2 Amendment No. 127 and Unit 3
Amendment No. 116, Technical Specification Bases Section 3/4.9.8 has
stated that ``With the reactor vessel head removed and 23 feet of
water above the reactor vessel flange, a large heat sink is
available for core cooling, thus in the event of a failure of the
operating shutdown cooling loop, adequate time is provided to
initiate emergency procedures to cool the core.''
In the Bases for the New Standard Technical Specifications,
``NUREG 1432, Revision 0, dated September 30, 1992, Section B 3.9.4
it is stated that; ``The 23 ft level was selected because it
corresponds to the 23 ft requirement established for fuel movement
in LCO 3.9.6, ``Refueling Water Level.''
Southern California Edison (Edison) calculations show that there
is an insignificant difference in the time to boil due to the 3-foot
change in required water level. Therefore, adequate water is still
available to mitigate the consequences of losing SDC.
2) Increase the time a required loop of the SDC system may be
removed from service from up to 1 hour per 8-hour period to up to 2
hours per 8-hour period, provided the upper guide structure has been
removed from the reactor vessel,
The proposed TS changes the time the SDC loop may be removed
from operation from up to 1 hour per 8-hour period to up to 2 hours
per 8-hour period, and allows removal of the SDC loop from operation
for testing of the Low Pressure Safety Injection (LPSI) system
components as well as for core alterations in the vicinity of the
hot legs. The proposed TS change also imposes certain restrictions
to ensure operating the SDC system in accordance with this proposed
TS change is of no safety significance. These [r]estrictions are
discussed separately below.
Specifically stating that the upper guide structure will be
removed assures that natural heat transfer is not impeded.
When securing the only operating loop of the SDC system the
maximum Reactor Coolant System (RCS) temperature is maintained [less
than or equal to] 140 deg.F. The initial conditions and heatup rate
are selected such that the RCS temperature remains [less than or
equal to] 140 deg.F during the test. Therefore, there is ample
margin to boiling. Typical initial temperatures are less than
100 deg.F.
The water being injected by the LPSI system test is cool water
from the Refueling Water Storage Tank (RWST) and will increase the
reactor cavity water level by several inches, providing more cool
water to the heat sink. The two hours is sufficient time to align
the system to test, perform the test, and restore the loop of SDC to
operation prior to exceeding 140 deg.F.
No operations are permitted that would cause a reduction of the
RCS boron concentration. This minimizes the probability of an
inadvertent boron dilution event. The use of adequately borated
water for injection into the RCS during the test provides assurance
that the test itself cannot lead to a boron dilution event. When the
SDC system is operating, the minimum SDC flow rate of 2200 gpm
imposed by Surveillance Requirements SR 3.9.4.1 and SR 3.9.5.1 is
sufficient to ensure complete mixing of the boron within the RCS.
Securing SDC flow is only allowed when the reactor cavity water
level is maintained greater than or equal to 20 feet above the
reactor vessel flange. This level ensures an adequate heat sink to
perform the LPSI pump suction header check valve test.
3) Allow for running 1 loop of shutdown cooling with additional
requirements when the water level is less than 20 feet but greater
than or equal to 12 feet above the reactor vessel flange,
4) Add an action to be taken when operating 1 loop of SDC with
less than 20 feet of water above the reactor vessel flange when the
specified requirements are not met,
In the event of a loss of SDC the time to boil is reduced from
approximately 4.0 hours when the water level is 23 feet above the
reactor vessel flange to approximately 2.3 hours at 12 feet,
assuming the reactor has only been shutdown for 6 days. However,
this is ample time to close containment (less than 1 hour) and to
restore SDC or initiate alternative cooling (e.g., add water to the
cavity (approximately 1 hour)). The reactor pressure vessel flange
is approximately 11' above the top of the fuel. Therefore, the water
level will be a minimum of 23' above the fuel, which still maintains
a large volume of water to provide a heat sink.
Requiring the reactor to be shutdown for at least 6 days to have
only one loop of SDC operable when the reactor cavity level is
between 20 feet and 12 feet above the reactor vessel flange ensures
that the time to boil is greater than twice the time it would take
to establish containment closure and to commence reactor cavity fill
with the required standby equipment.
One loop of SDC operating with a containment spray pump allows
for the high capacity LPSI pump to be the main standby pump capable
of filling the reactor cavity to at least 20 feet above the reactor
pressure vessel flange in the event SDC is lost. The high pressure
safety injection pump will also be maintained OPERABLE to increase
the water level if needed. In support of this contingency the RWST
will be required to contain the volume of water needed to raised
[raise] the level to 20 feet above the reactor pressure vessel
flange. As discussed above, the reactor cavity can be filled at a
rate of approximately 4.0 inches per minute with the LPSI pump.
If operating one loop of the SDC system with less than 20 feet
of water above the reactor vessel flange and any of the required
conditions are not met, requiring immediate action to establish
greater than or equal to 20 feet of water above the reactor vessel
flange ensures no time is wasted trying to restore the required
condition not met. By taking action to restore the level to 20 feet
above the reactor vessel flange the plant will be placed in TS
3.9.4, which only requires one loop of SDC to be operable.
Additionally, the core will not heat up while the water level in the
reactor cavity is being raised with cool water from the RWST. This
will provide additional time to either restore the one loop of SDC
or take other actions to provide core cooling as required by TS
3.9.4.
A Probabilistic Risk Assessment (PRA), with a) one loop of the
SDC system operable with the reactor cavity water level greater than
or equal to 12 feet above the reactor vessel flange, and b) one loop
of the SDC system operable with the reactor cavity water level
greater than or equal to 20 feet above the reactor vessel flange,
showed that the operations in accordance with the proposed TS would
not significantly increase the probabilities of inventory boiling
and core damage.
5) Item 6 adds wording to the notes in LCOs 3.9.4 and 3.9.5 that
was unintentionally deleted by the Unit 2 Amendment No. 127 and Unit
3 Amendment No. 116.
This is an editorial change.
Therefore, proposed changes 1 through 5 do not involve a
significant increase in the probability or consequences of an
accident.
2. The proposed change will not create the possibility of a new or
different kind of accident from any accident previously evaluated.
1) Reduce the water level where two loops of SDC are required
from 23 feet to 20 feet above the reactor vessel flange,
2) Increase the time a required loop of the SDC system may be
removed from service from up to 1 hour per 8-hour period to up to 2
hours per 8-hour period, provided the upper guide structure has been
removed from the reactor vessel,
3) Allow for running 1 loop of shutdown cooling with additional
requirements when the water level is less than 20 feet but greater
than or equal to 12 feet above the reactor vessel flange,
4) Add an action to be taken when operating 1 loop of SDC with less
than 20 feet of water above the reactor vessel flange when the
specified requirements are not met,
The Limiting Conditions for Operation (LCO) in Technical
Specifications (TSs) 3.9.4 and 3.9.5 define the operability
requirements for the SDC system during refueling operations (Mode 6)
while the water level above the top of the reactor vessel flange is
at least 23 feet and less than 23 feet, respectively. The objective
of the proposed TS changes is to ensure that the intent of the Bases
is maintained. [i.e., 1) sufficient cooling is available to remove
decay heat, 2) water in the reactor vessel is maintained below
140 deg.F, and 3) sufficient coolant
[[Page 47980]]
circulation is maintained in the reactor core to minimize boron
stratification leading to a boron dilution incident.]
The proposed TS changes affect the current limits imposed while
ensuring adherence to the bases of the TS. No plant modifications
are being made. The reactor cavity water level limitations and SDC
system required operating times are being changed based on plant
specific calculations, and the objective of the TSs are being
maintained. The added requirements and action statement facilitate
safe operation.
5) Item 6 adds wording to the notes in LCOs 3.9.4 and 3.9.5 that
was unintentionally deleted by the Unit 2 Amendment No. 127 and Unit 3
Amendment No. 116.
This is an editorial change.
Therefore, the operation of the facility in accordance with
proposed changes 1 through 5 does not create the possibility of a
new or different kind of accident from any accident previously
evaluated.
3. The proposed change will not involve a significant reduction
in a margin of safety.
Limiting Conditions for Operation (LCO) in TSs 3.9.4 and 3.9.5
define the operability requirements for the SDC system during
refueling operations (Mode 6) while the water level above the top of
the reactor vessel flange is at least 23 feet and less than 23 feet,
respectively. The objectives of these TSs are to ensure that 1)
sufficient cooling is available to remove decay heat, 2) the water
in the reactor vessel is maintained below 140 deg.F, and 3)
sufficient coolant circulation is maintained in the reactor core to
minimize boron stratification leading to a boron dilution incident.
1) Reduce the water level where two loops of SDC are required
from 23 feet to 20 feet above the reactor vessel flange,
Prior to the approval of Unit 2 Amendment No. 127 and Unit 3
Amendment No. 116, Technical Specification Bases Section 3/4.9.8 has
stated that ``With the reactor vessel head removed and 23 feet of
water above the reactor vessel flange, a large heat sink is
available for core cooling, thus in the event of a failure of the
operating shutdown cooling loop, adequate time is provided to
initiate emergency procedures to cool the core.''
In the Bases for the New Standard Technical Specifications,
NUREG 1432, Revision 0, dated September 30, 1992, Section B 3.9.4 it
is stated that ``The 23 ft level was selected because it corresponds
to the 23 ft requirement established for fuel movement in LCO 3.9.6,
``Refueling Water Level.''
Edison calculations show that there is a minimal difference in
the time to boil due to the 3-foot change in required water level.
Therefore, the margin of safety has not been significantly reduced.
2) Increase the time a required loop of the SDC system may be
removed from service from up to 1 hour per 8-hour period to up to 2
hours per 8-hour period, provided the upper guide structure has been
removed from the reactor vessel,
The proposed TS changes the time the SDC loop may be removed
from operation from up to 1 hour per 8-hour period to up to 2 hours
per 8-hour period, and allows removal of the SDC loop from operation
for testing of the LPSI system components as well as for core
alterations in the vicinity of the hot legs. The proposed TS change
also imposes certain restrictions to ensure operating the SDC system
in accordance with this proposed TS change is of no safety
significance. These restrictions are discussed separately below.
Specifically stating that the upper guide structure will be
removed assures that natural heat transfer is not impeded.
When securing the only operating loop of the SDC system, the
maximum RCS temperature is maintained [less than or equal to]
140 deg.F. The initial conditions and heatup rate are selected such
that RCS temperature remains [less than or equal to] 140 deg.F
during the test. Therefore, there is ample margin to boiling.
Typical initial temperatures are less than 100 deg.F.
The water being injected by the LPSI system test is cool borated
water from the RWST and will increase the level of the reactor
cavity by several inches. The two hours is sufficient time to align
the system to test, perform the test, and restore the loop of SDC to
operation prior to exceeding 140 deg.F.
No operations are permitted that would cause a reduction of the
RCS boron concentration. This minimizes the probability of an
inadvertent boron dilution event. The use of adequately borated
water for injection into the RCS during the test provides assurance
that the test itself cannot lead to a boron dilution event. When the
SDC system is operating, the minimum SDC flow rate of 2200 gpm is
sufficient to ensure complete mixing of the boron within the RCS.
Securing SDC flow is only allowed when the reactor cavity water
level is maintained greater than or equal to 20 feet above the
reactor vessel flange. This level ensures an adequate heat sink to
perform the LPSI pump suction header check valve test.
The added requirements and the nature of the test provide
assurances that the water temperature will be maintained less than
140 deg.F and that boron stratification is prevented.
3) Allow for running 1 loop of shutdown cooling with additional
requirements when the water level is less than 20 feet but greater
than or equal to 12 feet above the reactor vessel flange,
4) Add an action to be taken when operating 1 loop of SDC with less
than 20 feet of water above the reactor vessel flange when the
specified requirements are not met,
In the event of a loss of SDC, the time to boil is reduced from
approximately 4.0 hours when the water level is 23 feet above the
reactor vessel flange to approximately 2.3 hours at 12 feet, when
the reactor has only been shutdown for 6 days. However, this is
ample time to close containment (less than 1 hour), and to restore
SDC or initiate alternative cooling (e.g., add water to the cavity
(approximately 1 hour)).
Requiring the reactor to be shutdown for at least 6 days to have
only one loop of SDC operable when the reactor cavity level is
between 20 feet and 12 feet above the reactor vessel flange ensures
that the time to boil is greater than twice the time it would take
us to establish containment closure and to commence reactor cavity
fill with the required standby equipment.
One loop of SDC operating with a containment spray pump allows
for the high capacity LPSI pump to be the main standby pump capable
of filling the reactor cavity to at least 20 feet above the reactor
pressure vessel flange in the event SDC is lost. The high pressure
safety injection pump will also be maintained OPERABLE to increase
the water level if needed. In support of this contingency the RWST
will be required to contain the volume of water needed to raised
[raise] the level to 20 feet above the reactor pressure vessel
flange. As discussed above, the reactor cavity can be filled at a
rate of approximately 4.0 inches per minute with the LPSI pump.
If operating one loop of the SDC system with less than 20 feet of
water above the reactor vessel flange and any of the required
conditions are not met, requiring immediate action to establish greater
than or equal to 20 feet of water above the reactor vessel flange
ensures no time is wasted trying to restore the required condition not
met. By taking action to restore the level to 20 feet above the reactor
vessel flange the plant will be placed in TS 3.9.4, which only requires
one loop of SDC to be operable. Additionally, the core will not heat up
while the reactor cavity water level is being raised with cool water
from the RWST. This will provide additional time to either restore the
one loop of SDC or take other actions to provide core cooling as
required by TS 3.9.4.
A PRA showed that operations in accordance with the proposed TS
did not significantly increase the probabilities of inventory
boiling and core damage.
5) Item 6 adds wording to the notes in LCOs 3.9.4 and 3.9.5 that
was unintentionally deleted by the Unit 2 Amendment No. 127 and Unit 3
Amendment No. 116.
This is an editorial change.
Therefore, operation of the facility in accordance with proposed
changes 1 through 5 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.
Temporary
Local Public Document Room location: Science Library, University of
California, Irvine, California 92713
Attorney for licensee: T. E. Oubre, Esquire, Southern California
Edison Company, P. O. Box 800, Rosemead, California 91770
NRC Project Director: William H. Bateman
[[Page 47981]]
Southern California Edison Company, et al., Docket Nos. 50-361 and
50-362, San Onofre Nuclear Generating Station, Unit Nos. 2 and 3,
San Diego County, California
Date of amendment requests: May 9, 1996, as supplemented by letter
dated June 27, 1996.
Description of amendment requests: The licensee proposes to add a
requirement to maintain a Barrier Control Program to Section 5 of the
improved 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:
1. The proposed change does not involve a significant increase in the
probability or consequences of an accident previously evaluated.
The proposed change will allow a passive support system, plant
barriers, to be taken out of service for a specific allowed outage
time. Since the allowed outage times are to limit the average annual
cumulative increase in fuel damage risk to less than 1.0E-6, there
will not be a significant increase in either the probability or
consequences of any accident previously evaluated. Additionally, the
proposed change will allow barrier impairments if allowed by a 10
CFR 50.59 evaluation and also if the equipment is declared
inoperable or is not needed. Since these two conditions are already
a part of the San Onofre Units 2 and 3 Licensing Basis, there will
be no change 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.
Barriers have been analyzed for specific hazards. The nature of
these hazards will not change due to this amendment, and therefore
no new or different kind of accident will be created from any
accident previously evaluated.
3. The proposed change does not involve a significant reduction
in a margin of safety.
Since allowing barrier impairments in accordance with 10 CFR
50.59 or declaring affected equipment inoperable is part of the
SONGS Units 2 and 3 Licensing Basis, there will be no reduction in
the margin of safety from these two criteria.
Allowing allowed outage times for barrier impairments does not
have a significant effect on a margin of safety because the average
annual cumulative increase in fuel damage risk is limited to less
than 1.0E-6/yr. This small increase is about 3% of the San Onofre
Units 2 and 3 core damage risk as reported in the Individual Plant
Examination (IPE).
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.
Temporary
Local Public Document Room location: Science Library, University
of California, Irvine, California 92713
Attorney for licensee: T. E. Oubre, Esquire, Southern California
Edison Company, P. O. Box 800, Rosemead, California 91770
NRC Project Director: William H. Bateman
Southern California Edison Company, et al., Docket Nos. 50-361 and
50-362, San Onofre Nuclear Generating Station, Unit Nos. 2 and 3,
San Diego County, California
Date of amendment requests: May 29, 1996
Description of amendment requests: The licensee proposes to revise
the acceptance criteria for the Agastat time delay relays used in the
engineered safety features (ESF) load sequencer in Surveillance
Requirement (SR) 3.8.1.18, ``A.C. Sources - Operating'' of Technical
Specification (TS) 3.8.1, ``A.C. Sources - Operating.''
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 will not involve a significant increase
in the probability or consequences of any accident previously
evaluated.
The proposed change would expand the current surveillance
acceptance criteria to more accurately reflect the characteristics
of the installed plant equipment. The diesel generators (DG's) have
sufficient capacity to maintain adequate voltage and frequency
during load sequencing with the expanded tolerance. The overall
Engineered Safety Features (ESF) response times in the Technical
Specifications and safety analyses are maintained even though the
timer tolerance is increased, therefore, the consequences of any
accident previously evaluated are not increased. The DG load
sequence timers are not of themselves a credible initiator of any
accident, so the probability of an accident has not been increased.
The timers will function acceptably to support the equipment needed
for accident mitigation, so the consequences of an accident are not
increased. Therefore, the probability or consequences of any
accident previously evaluated is not increased.
2. The proposed change will not create the possibility of a new
or different kind of accident from any accident previously
evaluated.
This amendment request does not involve any change to plant
equipment or operation. In the event of a loss of preferred power,
the ESF electrical loads are automatically connected to the DG's in
sufficient time to provide for safe reactor shutdown and to mitigate
the consequences of a Design Basis Accident (DBA) such as a loss of
coolant accident (LOCA). Increasing the timer tolerance will not
create the possibility of a new or different kind of accident from
any previously evaluated.
3. The proposed change will not involve a significant reduction
in a margin of safety.
This amendment does not change the manner in which safety
limits, limiting safety settings, or limiting conditions for
operations are determined. The actual response times have not been
altered by this amendment, therefore, operations will not be
affected. Accordingly, this amendment will not involve a significant
reduction in the margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment requests involve no significant hazards consideration.
Temporary
Local Public Document Room location: Science Library, University of
California, Irvine, California 92713
Attorney for licensee: T. E. Oubre, Esquire, Southern California
Edison Company, P. O. Box 800, Rosemead, California 91770
NRC Project Director: William H. Bateman
Southern California Edison Company, et al., Docket Nos. 50-361 and
50-362, San Onofre Nuclear Generating Station, Unit Nos. 2 and 3,
San Diego County, California
Date of amendment requests: May 30, 1996
Description of amendment requests: The licensee proposes to revise
Surveillance Requirements (SR) 3.6.1.1, 3.6.2.1, and 3.6.3.6, of the
improved Technical Specifications. The proposed change will allow
implementation of the recently approved Option B to 10 CFR Part 50,
Appendix J. This new rule allows for a performance-based option for
determining the test frequency for containment leakage rate testing.
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.
Since the interval between containment leakage rate tests is not
related in any way to conditions which cause accidents, and plant
structures, systems, and components
[[Page 47982]]
will not be operated in a different manner as a result of the
proposed Technical Specification (TS) change, the proposed changes
will not increase the probability of an accident previously
evaluated.
Containment leakage may result from accidents which are
evaluated in the Updated Final Safety Analysis Report. The proposed
TS changes may result in an acceptably small increase in post-
accident containment leakage. Using a statistical approach, NUREG-
1493 determined that the increase in hypothetical dose to the public
resulting from extending the testing interval is extremely small.
NUREG-1493 concluded that such small hypothetical dose increases to
the public are justifiable due to the real reduction in occupational
exposure resulting from interval extension. Therefore, the proposed
change does not significantly increase the consequences of an
accident previously evaluated.
2. The proposed change does not create the possibility of a new
or different kind of accident from any accident previously
evaluated.
The proposed change only incorporates the performance based
approach for containment leak rate testing authorized in the new
Option B to Appendix J of 10 CFR Part 50. The interval extensions
allowed, through this approach, do not have the potential for
creating the possibility of new or different kinds of accidents from
those previously evaluated because plant structures, systems, and
components will not be operated in a different manner as a result of
the TS change and, therefore, will not introduce any new or
different failure modes or initiators. 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 change does not involve a significant reduction
in a margin of safety.
The proposed Technical Specification does not alter the
allowable containment leakage rate. The proposed change replaces the
current, prescriptive testing requirements with a new performance
based approach for establishing the testing intervals. Therefore,
the proposed change does not involve a significant reduction in a
margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment requests involve no significant hazards consideration.
Temporary
Local Public Document Room location: Science Library, University of
California, Irvine, California 92713
Attorney for licensee: T. E. Oubre, Esquire, Southern California
Edison Company, P. O. Box 800, Rosemead, California 91770
NRC Project Director: William H. Bateman
Southern Nuclear Operating Company, Inc., Docket Nos. 50-348 and
50-364, Joseph M. Farley Nuclear Plant, Units 1 and 2, Houston
County, Alabama
Date of amendments request: August 23, 1996
Description of amendments request: The proposed amendments would
revise the Technical Specifications to allow installation of laser
welded elevated tubesheet sleeves in Farley, Units 1 and 2.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. Operation of the Farley Nuclear Plant Units 1 and 2 steam
generators in accordance with the proposed license amendment does
not involve a significant increase in the probability or
consequences of an accident previously evaluated.
The installation of elevated tubesheet laser welded sleeves as
described below, can be used to repair degraded tubes by returning
the condition of the tubes to their original design condition (for
tube integrity, stress and fatigue considerations, and leaktightness
during all plant conditions). Tube bundle overall structural and
leakage integrity will be increased with the installation of the
laser welded sleeves. The performance history of Westinghouse
sleeves has shown that, to date, no domestic laser welded sleeves
have been removed from service due to corrosion degradation of the
sleeve or parent tube in the joint area.
Any hypothetical sleeve failure is bounded by the consequences
of a postulated steam generator tube rupture event. The use of
elevated tubesheet laser welded sleeves will not increase the amount
of primary-to-secondary leakage anticipated during a postulated
steam linebreak and other analyzed accidents. Leak rate tests show
only negligible primary-to-secondary leakage through the non-welded
elevated tubesheet sleeve lower joints during normal or accident
conditions such that any consequences are insignificant with regard
to offsite doses. Sleeve installation will result in an increase in
resistance to primary coolant flow through the tube. Depending on
the assumed steam generator tube rupture location, the primary
coolant flow through the ruptured tube is reduced by the influence
of sleeves installed below the break location, thereby reducing the
consequences to the public due to a steam generator tube rupture
event. Steam generator tube sleeving has as a basis that the
analyzed steam generator tube plugging level and associated minimum
measured flow rate, is not exceeded. Therefore, primary coolant flow
area assumptions in the accident analyses are not affected and any
consequences of a postulated loss of coolant accident would not be
increased.
2. The proposed license amendment does not create the
possibility of a new or different kind of accident from any accident
previously evaluated.
Installation of elevated tubesheet laser welded sleeves will
increase the leaktightness of the tube bundle in addition to
enhancing overall steam generator tube bundle integrity by isolating
localized tube wall degradation. Isolation of the tube degradation
is provided by attachment between the tube and sleeve at each end of
the sleeve. Following the installation of the sleeves, steam
generator tube integrity is restored to its original design bases.
Testing has shown that once installed, there is no mechanism for
the sleeves to affect any portion of the steam generator other than
the tubes in which they are installed. No other system or component
connecting with the steam generator is adversely affected by the
operation of the steam generator following installation of laser
welded tube sleeves.
Structural analyses of the tube, sleeve and sleeve joints show
the stress limits defined in the ASME [American Society of
Mechanical Engineers] Code are not exceeded during all plant
conditions. The effect of any hypothetical failure of the sleeve
would be bounded by existing tube rupture analyses. No increase in
leakage is anticipated during a postulated steam line break event.
Therefore, operation of the steam generators following installation
of elevated tubesheet laser welded sleeves in the tubes of the
Farley steam generators will not result in an accident previously
not analyzed in the FSAR [Final Safety Analysis Report].
Therefore, SNC [Southern Nuclear Operating Company] concludes
that the proposed license amendment does not create the possibility
of a new or different kind of accident from any accident previously
evaluated.
3. The proposed license amendment does not involve a significant
reduction in a margin of safety.
The margin of safety with respect to maintenance of the
integrity of the tube bundle is provided, in part, by the safety
factors included in the ASME Code, and is not reduced.
Nondestructive examination of the sleeve and non-sleeved tube length
still can be performed; therefore, the recommendations of Regulatory
Guide 1.83, Revision 1 can be implemented. The installation process
of the elevated tubesheet laser welded sleeves has been shown to
provide an essentially leaktight bond between the sleeve and the
tube during all plant conditions, and, as such, would not
significantly contribute to the radiological consequences of a
postulated steam line break event. Any combination of sleeving and
plugging utilized at Farley Units 1 and 2 up to the level that
analyzed minimum measured reactor coolant flow rate is maintained
per Technical Specification requirements, will be bounded by the
accident analyses supporting the analyzed flow level.
Therefore, SNC, concludes that the proposed change does not
result in a significant reduction in a loss of margin with respect
to plant safety as defined in the Final Safety Analysis Report or
the bases of the Farley technical specifications.
Based on the preceding analysis, it is concluded that operation
of the Farley
[[Page 47983]]
Nuclear Plant steam generators in accordance with the proposed
amendment does not involve a significant hazards consideration as
defined in 10 CFR 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 request involves no significant hazards consideration.
Local Public Document Room location: Houston-Love Memorial Library,
212 W. Burdeshaw Street, Post Office Box 1369, Dothan, Alabama 36302
Attorney for licensee: M. Stanford Blanton, Esq., Balch and
Bingham, Post Office Box 306, 1710 Sixth Avenue North, Birmingham,
Alabama 35201
NRC Project Director: Herbert N. Berkow
Tennessee Valley Authority, Docket Nos. 50-390 Watts Bar Nuclear
Plant, Unit 1, Rhea County, Tennessee
Date of amendment request: June 29, 1996 (TS 5.2.2.f)
Description of amendment request: The proposed amendment would
revise the Watts Bar (WBN) Unit 1 Technical Specification (TS)
requirements to delete the first sentence of TS Section 5.2.2.f which
reads, ``The Operations Manager shall hold or have held an SRO [Senior
Reactor Operator] license on a similar unit.'' The remaining sentence
of this section is being revised to indicate that the Operations
Superintendent will hold an SRO license for WBN Unit 1. This change is
consistent with the Tennessee Valley Authority's (TVA) commitment to
ANSI N18.1-1971 regarding the qualification of this position and is
consistent with the Standard TS.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration. The NRC staff has reviewed the licensee's analysis
against the standards of 10 CFR 50.92(c). The NRC staff's review is
presented below.
Operation of the plant in accordance with the proposed amendment
will not:
(1) Involve a significant increase in the probability or
consequences of an accident previously evaluated.
As explained in the June 29, 1996 submittal, the proposed change is
considered to be administrative in nature. The proposed change affects
an administrative control, which was based on the guidance of ANSI
N18.-1971. ANSI N18.1-1971 recommended that the Operations Manager hold
an SRO license. The ANSI N18.1-1971 Standard defines the positions of
Plant Manager, Operations Manager, Supervisors and Operators. A
subsequent update of this standard, ANSI/ANS 3.1-1987, also defines the
position of Operations Middle Manager. The correlating named positions
in the TVA management structure at WBN are: WBN Operations Manager
correlates to ANSI Plant Manager, WBN Operations Superintendent
correlates to ANSI Operations Manager or Operations Middle Manager, WBN
Shift Operations Supervisor correlates to ANSI Shift Supervisor, and
WBN Senior and Licensed Operators correlate to ANSI operators. The
guidance in Section 4.2.2 of ANSI/ANS 3.1-1987 recommends that ``If the
Operations Manager does not hold an NRC License, then the Operations
Middle Manager shall hold an NRC Senior Operator's License. This would
be consistent with TVA's proposal that the WBN Operations
Superintendent (ANSI Operations Middle Manager) continue to be required
to maintain an SRO license.
The proposed change does not alter the design of any system,
structure, or component, nor does it change the way plant systems are
operated. It does not reduce the knowledge, qualifications, or skills
of licensed operators. The control room operators will continue to be
supervised by the licensed Shift Supervisors and the first level of
off-shift WBN managemet directing the activities of licensed operators
will continue to hold an SRO license. In summary, the proposed change
does not affect the ability of the Operations Superintendent to provide
the plant oversight required of his position. Thus, it does not involve
a significant increase in the probability or consequence of an accident
previously evaluated.
(2) Create the possibility of a new or different kind of accident
from any previously evaluated.
The proposed change to TS 5.2.2.f does not affect the design or
function of any plant system, structure, or component, nor does it
change the way plant systems are operated. It does not affect the
performance of NRC licensed operators. Operation of the plant will
continue to be supervised by personnnel who hold an NRC SRO license.
Based on the above, the proposed change does not create the possibility
of a new or different kind of accident from any previously evaluated.
(3) Involve a significant reduction in a margin of safety.
The proposed change involves an administrative control. The
proposed change does not reduce the level of knowledge or experience
required of an individual who fills the Operations Superintendent
position. The control room operators will continue to be supervised by
personnel who hold an SRO license. Thus, the proposed change does not
ivnolve a significant reduction in a margin of safety.
Based on this review, it appears that the three standards of 10 CFR
50.92(c) are satisfied. Therefore, the NRC staff proposes to determine
that the amendment request involves no significant hazards
consideration.
Local Public Document Room location: 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
Wisconsin Electric Power Company, Docket Nos. 50-266 and 50-301,
Point Beach Nuclear Power Plant, Unit Nos. 1 and 2, Town of Two
Creeks, Manitowoc County, Wisconsin
Date of amendment request: November 17, 1995, as supplemented July
29, 1996
Description of amendment request: The proposed amendment would
revise Technical Specification Section 15.6.3, ``Facility Staff
Qualifications.'' The title of the responsible health physicist would
be changed, and a requirement for this individual to be a supervisor
would be added.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration which is presented below:
1. Operation of the Point Beach Nuclear Plant in accordance with
the proposed amendments does not result in a significant increase in
the probability or consequences of an accident previous evaluated.
The proposed changes separate the qualifications requirements of
the Technical Specifications from the Health Physics Manager, while
requiring that the same qualifications be fulfilled by a designated
Health Physicist position within the organization. This change
maintains the present knowledge requirements of the PBNP [Point
Beach Nuclear Plant] staff. The personnel holding the health physics
qualifications are not considered in the probability of any
accident. By ensuring the appropriate expertise remains on the staff
to advise management on issues related to radiological safety,
appropriate action is assured during analyzed events to assess and
mitigate the radiological consequences. Therefore, this change does
not affect the probability or consequences of any accident
previously evaluated.
[[Page 47984]]
2. Operation of the Point Beach Nuclear Plant in accordance with
the proposed amendments will not result in a new or different kind
of accident from any accident previously evaluated.
The proposed change separates the Health Physics Manager
qualifications from the position while maintaining the requirements
for that expertise to be maintained within the organization. This is
an administrative change only and does not affect any plant
structures, systems or components. Therefore, a new or different
kind of accident from any accident previously evaluated cannot
result.
3. Operation of the Point Beach Nuclear Plant in accordance with
the proposed amendments will not result in a significant reduction
in a margin of safety.
The proposed changes are administrative only. The required
levels of expertise and experience will be maintained within the
Health Physics organization. Therefore, there is no reduction in a
margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: Joseph P. Mann Library, 1516
Sixteenth Street, Two Rivers, Wisconsin 54241
Attorney for licensee: Gerald Charnoff, Esq., Shaw, Pittman, Potts,
and Trowbridge, 2300 N Street, NW., Washington, DC 20037
NRC Project Director: Gail H. Marcus
Previously Published Notices Of Consideration Of Issuance Of
Amendments To Facility Operating Licenses, Proposed No Significant
Hazards Consideration Determination, And Opportunity For A Hearing
The following notices were previously published as separate
individual notices. The notice content was the same as above. They were
published as individual notices either because time did not allow the
Commission to wait for this biweekly notice or because the action
involved exigent circumstances. They are repeated here because the
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.
Consumers Power Company, Docket No. 50-255, Palisades Plant, Van
Buren County, Michigan
Date of amendment request: January 5, 1996, as supplemented July
12, 1996
Description of amendment request: The proposed amendment would
revise the requirements of technical specification 3.1.9.3 to permit a
filled refueling cavity to serve as a back-up means of decay heat
removal.
Date of individual notice in the Federal Register: August 28, 1996
(61 FR 44348)
Expiration date of individual notice: September 27, 1996
Local Public Document Room location: Van Wylen Library, Hope
College, Holland, Michigan 49423
Notice Of Issuance Of Amendments To Facility Operating Licenses
During the period since publication of the last biweekly notice,
the Commission has issued the following amendments. The Commission has
determined for each of these amendments that the application complies
with the standards and requirements of the Atomic Energy Act of 1954,
as amended (the Act), and the Commission's rules and regulations. The
Commission has made appropriate findings as required by the Act and the
Commission's rules and regulations in 10 CFR Chapter I, which are set
forth in the license amendment.
Notice of Consideration of Issuance of Amendment to Facility
Operating License, Proposed No Significant Hazards Consideration
Determination, and Opportunity for A Hearing in connection with these
actions was published in the Federal Register as indicated.
Unless otherwise indicated, the Commission has determined that
these amendments satisfy the criteria for categorical exclusion in
accordance with 10 CFR 51.22. Therefore, pursuant to 10 CFR 51.22(b),
no environmental impact statement or environmental assessment need be
prepared for these amendments. If the Commission has prepared an
environmental assessment under the special circumstances provision in
10 CFR 51.12(b) and has made a determination based on that assessment,
it is so indicated.
For further details with respect to the action see (1) the
applications for amendment, (2) the amendment, and (3) the Commission's
related letter, Safety Evaluation and/or Environmental Assessment as
indicated. All of these items are available for public inspection at
the Commission's Public Document Room, the Gelman Building, 2120 L
Street, NW., Washington, DC, and at the local public document rooms for
the particular facilities involved.
Baltimore Gas and Electric Company, Docket Nos. 50-317 and 50-318,
Calvert Cliffs Nuclear Power Plant, Unit Nos. 1 and 2, Calvert
County, Maryland
Date of application for amendments: March 15, 1995, as supplemented
June 29, 1995, May 1, 1996 and May 15, 1996.
Brief description of amendments: The amendments revise the
Technical Specification (TS) Section 6.0, ``Administrative Controls''
to be consistent with the guidance provided in the Improved Standard
Technical Specifications (STSs) for Combustion Engineering Plants.
Additionally, the amendments (a) allow the Shift Technical Advisory to
perform dual roles, (b) establishes a TS Bases Control Program, (c)
provides for a reduction in the reporting requirements, and (d)
provides an option for estimating occupational doses.
Date of issuance: August 26, 1996
Effective date: As of the date of issuance to be implemented within
30 days.
Amendment Nos.: 216 and 193
Facility Operating License Nos. DPR-53 and DPR-69: Amendments
revised the Technical Specifications.
Date of initial notice in Federal Register: August 16, 1995 (60 FR
42598) The Commission's related evaluation of these amendments is
contained in a Safety Evaluation dated August 26, 1996. No significant
hazards consideration comments received: No
Local Public Document Room location: Calvert County Library, Prince
Frederick, Maryland 20678.
Connecticut Yankee Atomic Power Company, Docket No. 50-213, Haddam
Neck Plant, Middlesex County and Northeast Nuclear Energy Company,
et al., Docket Nos. 50-245, 50-336, and 50-423, Millstone Nuclear
Power Station, Units 1, 2, and 3, New London County, Connecticut
Date of application for amendments: November 22, 1995
Brief description of amendments: The amendments replace the title-
specific designation of members representing specific functional areas
on the Plant Operating Review Committee (PORC) for the Haddam Neck
Plant and Millstone Units 1, 2, and 3 with a functional area-specific
designation that stipulates membership qualification and experience
requirements. The amendments also clarify the composition of the Site
Operations Review Committee (SORC) at Millstone.
Date of issuance: July 16, 1996
Effective date: As of the date of issuance, to be implemented
within 60 days.
[[Page 47985]]
Amendment Nos.: 190, 95, 200, 130
Facility Operating License Nos. DPR-61, DPR-21, DPR-65, AND NPF-49:
Amendments revised the Technical Specifications.
Date of initial notice in Federal Register: February 28, 1996 (61
FR 7549) The Commission's related evaluation of the amendments is
contained in a Safety Evaluation dated July 16, 1996 No significant
hazards consideration comments received: No.
Local Public Document Room location: Russell Library, 123 Broad
Street Middletown, Connecticut 06457, for the Haddam Neck Plant, and
the Learning Resources Center, Three Rivers Community-Technical
College, 574 New London Turnpike, Norwich, Connecticut 06360, and
Waterford Library, ATTN: Vince Juliano, 49 Rope Ferry Road, Waterford,
Connecticut 06385, for Millstone 1, 2, and 3.
Duke Power Company, Docket Nos. 50-269, 50-270, and 50-287, Oconee
Nuclear Station, Units 1, 2, and 3, Oconee County, South Carolina
Date of application of amendments: June 6, 1996; supplemented
August 1, 1996
Brief description of amendments: The amendments revise the
Technical Specification requirements related to testing of the Low
Pressure Service Water pumps and valves, LPSW-4 and LPSW-5, to reflect
a design change to remove the Engineered Safeguards signal from the
valves.
Date of Issuance: August 19, 1996
Effective date: As of the date of issuance to be implemented within
30 days
Amendment Nos.: 217, 217, 214
Facility Operating License Nos. DPR-38, DPR-47, and DPR-55: The
amendments revised the Technical Specifications.
Date of initial notice in Federal Register: July 17, 1996 (61 FR
37298) The Commission's related evaluation of the amendments is
contained in a Safety Evaluation dated August 19, 1996. No significant
hazards consideration comments received: No
Local Public Document Room location: Oconee County Library, 501
West South Broad Street, Walhalla, South Carolina
Entergy Operations, Inc., System Energy Resources, Inc., South
Mississippi Electric Power Association, and Entergy Mississippi,
Inc., Docket No. 50-416, Grand Gulf Nuclear Station, Unit 1,
Claiborne County, Mississippi
Date of application for amendment: May 31, 1996, as supplemented by
letter dated May 2, 1996
Brief description of amendment: The amendment revised the schedule
for withdrawing capsules with reactor vessel material specimens in
accordance with the reactor vessel material surveillance program for
the Grand Gulf Nuclear Station, Unit 1 and Section III.B.3 of Appendix
H, ``Reactor Vessel Material Surveillance Program Requirements,'' of 10
CFR Part 50.
Date of issuance: August 21, 1996
Effective date: August 21, 1996
Amendment No: 127
Facility Operating License No. NPF-29: Amendment revises the
license.
Date of initial notice in Federal Register: June 19, 1996 (61 FR
31179) The Commission's related evaluation of the amendment is
contained in a Safety Evaluation dated August 21, 1996. No significant
hazards consideration comments received: No
Local Public Document Room location: Judge George W. Armstrong
Library, 220 S. Commerce Street, Natchez, MS 39120.
Florida Power and Light Company, et al., Docket Nos. 50-335 and 50-
389, St. Lucie Plant, Unit Nos. 1 and 2, St. Lucie County, Florida
Date of application for amendments: May 17, 1995, as supplemented
July 15, 1996.
Brief description of amendments: These amendments improve
consistency between the Technical Specifications (TS) and the improved
Combustion Engineering Standard Technical Specifications (STS) and
resolve other inconsistencies in the TS.
Date of Issuance: August 14, 1996
Effective Date: August 14, 1996
Amendment Nos.: 146 and 85
Facility Operating License Nos. DPR-67 and NPF-16: Amendments
revised the Technical Specifications.
Date of initial notice in Federal Register: June 21, 1995 (60 FR
32363). The July 15, 1996, letter made a minor change to the proposed
definition of core alteration which made it more closely match the
wording in the STS and did not change the scope of the May 17, 1995,
application and initial proposed no significant hazards consideration
determination. The Commission's related evaluation of the amendments is
contained in a Safety Evaluation dated August 14, 1996. No significant
hazards consideration comments received: No
Local Public Document Room location: Indian River Junior College
Library, 3209 Virginia Avenue, Fort Pierce, Florida 34954-9003
Florida Power and Light Company, et al., Docket Nos. 50-335 and 50-
389, St. Lucie Plant, Unit Nos. 1 and 2, St. Lucie County, Florida
Date of application for amendments: August 16, 1996
Brief description of amendments: Relocates selected Technical
Specifications (TS) related to instrumentation to the Updated Final
Safety Analysis Report, in accordance with the Commissions Final Policy
Statement on TS Improvement for Nuclear Power Reactors (58 FR 39132,
July 22, 1993). Also relocates review requirements related to the
Emergency Plan and the Security Plan from the TS to the respective
plans.
Date of Issuance: August 20, 1996
Effective Date: August 20, 1996
Amendment Nos.: 147 and 86
Facility Operating License Nos. DPR-67 and NPF-16: Amendments
revised the Technical Specifications.
Date of initial notice in Federal Register: September 27, 1995 (60
FR 49938) The Commission's related evaluation of the amendments is
contained in a Safety Evaluation dated August 20, 1996. No significant
hazards consideration comments received: No.
Local Public Document Room location: Indian River Junior College
Library, 3209 Virginia Avenue, Fort Pierce, Florida 34954-9003
Georgia Power Company, Oglethorpe Power Corporation, Municipal
Electric Authority of Georgia, City of Dalton, Georgia, Docket Nos.
50-424 and 50-425, Vogtle Electric Generating Plant, Units 1 and 2,
Burke County, Georgia
Date of application for amendments: June 17, 1996
Brief description of amendments: The amendments revise Technical
Specification 5.3.1, Fuel Assemblies, to remove the restriction on the
number of fuel rods clad with ZIRLOTM that can be loaded into the
core.
Date of issuance: August 19, 1996
Effective date: As of the date of issuance to be implemented within
30 days
Amendment Nos.: 94, 72
Facility Operating License Nos. NPF-68 and NPF-81: Amendments
revised the Technical Specifications.
Date of initial notice in Federal Register: July 17, 1996 (61 FR
37299)
[[Page 47986]]
The Commission's related evaluation of the amendments is contained in a
Safety Evaluation dated August 19, 1996. No significant hazards
consideration comments received: No
Local Public Document Room location: Burke County Library, 412
Fourth Street, Waynesboro, Georgia 30830
Georgia Power Company, Oglethorpe Power Corporation, Municipal
Electric Authority of Georgia, City of Dalton, Georgia, Docket Nos.
50-424 and 50-425, Vogtle Electric Generating Plant, Units 1 and 2,
Burke County, Georgia
Date of application for amendments: June 17, 1996
Brief description of amendments: The amendments revise Technical
Specification 3/4.8.1, A.C. Sources, and its associated Bases, by
changing Surveillance Requirement 4.8.1.1.2.j(2) to limit the 10-year
pressure test of certain portions of the diesel fuel oil system to the
isolable portions of the fuel oil piping.
Date of issuance: August 28, 1996
Effective date: As of the date of issuance to be implemented within
30 days
Amendment Nos.: 95 and 73
Facility Operating License Nos. NPF-68 and NPF-81: Amendments
revised the Technical Specifications.
Date of initial notice in Federal Register: July 17, 1996 (61 FR
37300) The Commission's related evaluation of the amendments is
contained in a Safety Evaluation dated August 28, 1996. No significant
hazards consideration comments received: No
Local Public Document Room location: Burke County Library, 412
Fourth Street, Waynesboro, Georgia 30830
Niagara Mohawk Power Corporation, Docket No. 50-410, Nine Mile
Point Nuclear Station, Unit 2, Oswego County, New York
Date of application for amendment: March 15, 1996, as supplemented
July 18, 1996.
Brief description of amendment: The amendment revised TS 4.6.2.1
``Containment Systems - Depressurization Systems - Suppression Pool''
to extend the time interval for performing the containment drywell-to-
suppression chamber bypass leakage tests consistent with schedules for
containment integrated leak rate testing under Option B to 10 CFR Part
50, Appendix J.
Date of issuance: August 27, 1996
Effective date: As of the date of issuance to be implemented within
30 days.
Amendment No.: 75
Facility Operating License No. NPF-69: Amendment revises the
Technical Specifications.
Date of initial notice in Federal Register: May 8, 1996 (61 FR
20851) The Commission's related evaluation of the amendment is
contained in a Safety Evaluation dated August 27, 1996 No significant
hazards consideration comments received: No
Local Public Document Room location: Reference and Documents
Department, Penfield Library, State University of New York, Oswego, New
York 13126.
Niagara Mohawk Power Corporation, Docket No. 50-410, Nine Mile
Point Nuclear Station, Unit 2, Oswego County, New York
Date of application for amendment: March 20, 1996
Brief description of amendment: The amendment revises Technical
Specification 3/4.3.1 ``Reactor Protection System Instrumentation'' to
modify operability requirements for the Average Power Range Monitor for
operational conditions 3, 4, and 5.
Date of issuance: August 28, 1996
Effective date: As of the date of issuance to be implemented within
30 days.
Amendment No.: 76
Facility Operating License No. NPF-69: Amendment revises the
Technical Specifications.
Date of initial notice in Federal Register: May 8, 1996 (61 FR
20852) The Commission's related evaluation of the amendment is
contained in a Safety Evaluation dated August 28, 1996 No significant
hazards consideration comments received: No
Local Public Document Room location: Reference and Documents
Department, Penfield Library, State University of New York, Oswego, New
York 13126.
Northeast Nuclear Energy Company, Docket No. 50-245, Millstone
Nuclear Power Station, Unit 1, New London County, Connecticut
Date of application for amendment: April 25, 1996
Brief description of amendment: The amendment modifies the
calibration requirement for the source range monitors and intermediate
range monitors by noting that the sensors are excluded.
Date of issuance: August 19, 1996
Effective date: As of the date of issuance, to be implemented
within 30 days.
Amendment No.: 96
Facility Operating License No. DPR-21. Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: June 19, 1996 (61 FR
31183) The Commission's related evaluation of the amendment is
contained in a Safety Evaluation dated August 19, 1996. No significant
hazards consideration comments received: No.
Local Public Document Room location: Learning Resources Center,
Three Rivers Community-Technical College, 574 New London Turnpike,
Norwich, CT 06360, and the Waterford Library, ATTN: Vince Juliano, 49
Rope Ferry Road, Waterford, CT 06385
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 application for amendments: November 14, 1994, as
supplemented by letters dated December 7, 1995, February 2, 1996, May
28, 1996, and July 30, 1996.
Brief description of amendments: The amendment revised the combined
Technical Specifications (TS) for the Diablo Canyon Nuclear Power
Plant, Unit Nos. 1 and 2, for the slave relay test frequency from
quarterly (Q) to refueling (R). The request also removed table notation
4 from Table 4.3-2. The associated Bases were revised.
Date of issuance: August 19, 1996
Effective date: August 19, 1996, to be implemented within 30 days
of date of issuance.
Amendment Nos.: Unit 1 - 115; Unit 2 - 113
Facility Operating License Nos. DPR-80 and DPR-82: The amendments
revised the Technical Specifications.
Date of initial notice in Federal Register: December 6, 1995 (60 FR
62495). The supplemental letters provided additional clarifying
information and did not change the original no significant hazards
consideration determination. The Commission's related evaluation of the
amendments is contained in a Safety Evaluation dated August 19, 1996.
No significant hazards consideration comments received: No.
Local Public Document Room location: California Polytechnic State
University, Robert E. Kennedy Library, Government Documents and Maps
Department, San Luis Obispo, California 93407
[[Page 47987]]
PECO Energy Company, Public Service Electric and Gas Company,
Delmarva Power and Light Company, and Atlantic City Electric
Company, Docket No. 50-277, Peach Bottom Atomic Power Station, Unit
No. 2, York County, Pennsylvania
Date of application for amendment: June 13, 1996, as supplemented
by letter dated August 7, 1996.
Brief description of amendment: This amendment will permit a one
time performance of TS surveillance requirement 3.3.1.1.12 for the
Average Power Range Monitor Flow Biased High Scram function with a
delayed entry into associated TS Conditions and Required Actions for up
to six hours provided core flow is maintained at or above eighty-two
percent. This change is in effect until the end of refueling outage
2R11.
Date of issuance: August 16, 1996
Effective date: Unit 2, as of the date of issuance, to be
implemented within 30 days.
Amendment No.: 216
Facility Operating License No. DPR-44: Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: July 3, 1996 (61 FR
34895) The Commission's related evaluation of the amendment is
contained in a Safety Evaluation dated August 16, 1996. No significant
hazards consideration comments received: No
Local Public Document Room location: Government Publications
Section, State Library of Pennsylvania, (REGIONAL DEPOSITORY) Education
Building, Walnut Street and Commonwealth Avenue, Box 1601, Harrisburg,
PA 17105.
Power Authority of the State of New York, Docket No. 50-333, James
A. FitzPatrick Nuclear Power Plant, Oswego County, New York
Date of applications for amendment: June 15, September 15, October
25, and November 30, 1995.
Brief description of amendment: The amendments change the Technical
Specifications regarding the Control Rod System, the Auxiliary
Electrical Systems, the Containment Systems and the Standby Liquid
Control System to reflect changes to the length of the operating cycle
of 24 months.
Date of issuance: August 16, 1996
Effective date: As of the date of issuance to be implemented within
30 days.
Amendment No.: 232
Facility Operating License No. DPR-59: Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: September 13, 1995 (60
FR 47623), January 22, 1996 (61 FR 1633, 61 FR 1634, 61 FR 1635) The
Commission's related evaluation of the amendment is contained in a
Safety Evaluation dated August 16, 1996. No significant hazards
consideration comments received: No
Local Public Document Room location: Reference and Documents
Department, Penfield Library, State University of New York, Oswego, New
York 13126.
Dated at Rockville, Maryland, this 4th day of September 1996.
For the Nuclear Regulatory Commission
Steven A. Varga,
Director, Division of Reactor Projects - I/II, Office of Nuclear
Reactor Regulation
[Doc. 96-23032 Filed 9-10-96; 8:45 am]
BILLING CODE 7590-01-F
