[Federal Register Volume 60, Number 187 (Wednesday, September 27, 1995)]
[Notices]
[Pages 49929-49963]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: X95-10927]
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NUCLEAR REGULATORY COMMISSION
Biweekly Notice
Applications and Amendments to Facility Operating
LicensesInvolving No Significant Hazards Considerations
I. Background
Pursuant to Public Law 97-415, the U.S. Nuclear Regulatory
Commission (the Commission or NRC staff) is publishing this regular
biweekly notice. Public Law 97-415 revised section 189 of the Atomic
Energy Act of 1954, as amended (the Act), to require the Commission to
publish notice of any amendments issued, or proposed to be issued,
under a new provision of section 189 of the Act. This provision grants
the Commission the authority to issue and make immediately effective
any amendment to an operating license upon a determination by the
Commission that such amendment involves no significant hazards
consideration, notwithstanding the pendency before the Commission of a
request for a hearing from any person.
This biweekly notice includes all notices of amendments issued, or
proposed to be issued from August 30, 1995, through September 15, 1995.
The last biweekly notice was published on Wednesday, September 13, 1995
(60 FR 47613).
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
[[Page 49930]]
accident previously evaluated; or (2) create the possibility of a new
or different kind of accident from any accident previously evaluated;
or (3) involve a significant reduction in a margin of safety. The basis
for this proposed determination for each amendment request is shown
below.
The Commission is seeking public comments on this proposed
determination. Any comments received within 30 days after the date of
publication of this notice will be considered in making any final
determination.
Normally, the Commission will not issue the amendment until the
expiration of the 30-day notice period. However, should circumstances
change during the notice period such that failure to act in a timely
way would result, for example, in derating or shutdown of the facility,
the Commission may issue the license amendment before the expiration of
the 30-day notice period, provided that its final determination is that
the amendment involves no significant hazards consideration. The final
determination will consider all public and State comments received
before action is taken. Should the Commission take this action, it will
publish in the Federal Register a notice of issuance and provide for
opportunity for a hearing after issuance. The Commission expects that
the need to take this action will occur very infrequently.
Written comments may be submitted by mail to the Rules Review and
Directives Branch, Division of Freedom of Information and Publications
Services, Office of Administration, U.S. Nuclear Regulatory Commission,
Washington, DC 20555, and should cite the publication date and page
number of this Federal Register notice. Written comments may also be
delivered to Room 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 27, 1995, the licensee may file a request for a hearing
with respect to issuance of the amendment to the subject facility
operating license and any person whose interest may be affected by this
proceeding and who wishes to participate as a party in the proceeding
must file a written request for a hearing and a petition for leave to
intervene. Requests for a hearing and a petition for leave to intervene
shall be filed in accordance with the Commission's ``Rules of Practice
for Domestic Licensing Proceedings'' in 10 CFR Part 2. Interested
persons should consult a current copy of 10 CFR 2.714 which is
available at the Commission's Public Document Room, the Gelman
Building, 2120 L Street, NW., Washington, DC and at the local public
document room for the particular facility involved. If a request for a
hearing or petition for leave to intervene is filed by the above date,
the Commission or an Atomic Safety and Licensing Board, designated by
the Commission or by the Chairman of the Atomic Safety and Licensing
Board Panel, will rule on the request and/or petition; and the
Secretary or the designated Atomic Safety and Licensing Board will
issue a notice of a hearing or an appropriate order.
As required by 10 CFR 2.714, a petition for leave to intervene
shall set forth with particularity the interest of the petitioner in
the proceeding, and how that interest may be affected by the results of
the proceeding. The petition should specifically explain the reasons
why intervention should be permitted with particular reference to the
following factors: (1) the nature of the petitioner's right under the
Act to be made a party to the proceeding; (2) the nature and extent of
the petitioner's property, financial, or other interest in the
proceeding; and (3) the possible effect of any order which may be
entered in the proceeding on the petitioner's interest. The petition
should also identify the specific aspect(s) of the subject matter of
the proceeding as to which petitioner wishes to intervene. Any person
who has filed a petition for leave to intervene or who has been
admitted as a party may amend the petition without requesting leave of
the Board up to 15 days prior to the first prehearing conference
scheduled in the proceeding, but such an amended petition must satisfy
the specificity requirements described above.
Not later than 15 days prior to the first prehearing conference
scheduled in the proceeding, a petitioner shall file a supplement to
the petition to intervene which must include a list of the contentions
which are sought to be litigated in the matter. Each contention must
consist of a specific statement of the issue of law or fact to be
raised or controverted. In addition, the petitioner shall provide a
brief explanation of the bases of the contention and a concise
statement of the alleged facts or expert opinion which support the
contention and on which the petitioner intends to rely in proving the
contention at the hearing. The petitioner must also provide references
to those specific sources and documents of which the petitioner is
aware and on which the petitioner intends to rely to establish those
facts or expert opinion. Petitioner must provide sufficient information
to show that a genuine dispute exists with the applicant on a material
issue of law or fact. Contentions shall be limited to matters within
the scope of the amendment under consideration. The contention must be
one which, if proven, would entitle the petitioner to relief. A
petitioner who fails to file such a supplement which satisfies these
requirements with respect to at least one contention will not be
permitted to participate as a party.
Those permitted to intervene become parties to the proceeding,
subject to any limitations in the order granting leave to intervene,
and have the opportunity to participate fully in the conduct of the
hearing, including the opportunity to present evidence and cross-
examine witnesses.
If a hearing is requested, the Commission will make a final
determination on the issue of no significant hazards consideration. The
final determination will serve to decide when the hearing is held.
If the final determination is that the amendment request involves
no significant hazards consideration, the Commission may issue the
amendment and make it immediately effective, notwithstanding the
request for a hearing. Any hearing held would take place after issuance
of the amendment.
If the final determination is that the amendment request involves a
significant hazards consideration, any hearing held would take place
before the issuance of any amendment.
A request for a hearing or a petition for leave to intervene must
be filed with the Secretary of the Commission, U.S. Nuclear Regulatory
Commission, Washington, DC 20555, 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
[[Page 49931]]
should also be sent to the Office of the General Counsel, U.S. Nuclear
Regulatory Commission, Washington, DC 20555, and to the attorney for
the licensee.
Nontimely filings of petitions for leave to intervene, amended
petitions, supplemental petitions and/or requests for a hearing will
not be entertained absent a determination by the Commission, the
presiding officer or the Atomic Safety and Licensing Board that the
petition and/or request should be granted based upon a balancing of
factors specified in 10 CFR 2.714(a)(1)(i)-(v) and 2.714(d).
For further details with respect to this action, see the
application for amendment which is available for public inspection at
the Commission's Public Document Room, the Gelman Building, 2120 L
Street, NW., Washington, DC, and at the local public document room for
the particular facility involved.
Carolina Power & Light Company, et al., Docket No. 50-324,
Brunswick Steam Electric Plant, Unit 2, Brunswick County, North
Carolina
Date of amendment request: August 4, 1995
Description of amendment request: The proposed amendment will allow
the loading and use of GE13 fuel assemblies in the Brunswick Steam
Electric Plant (BSEP), Unit 2, during Cycle 12 operation. The use of
GE13 fuel assemblies requires that the safety limit value for minimum
critical power ratio be revised. This safety limit is established to
maintain fuel cladding integrity. Use of GE13 fuel also requires an
increase in the concentration of sodium pentaborate solution required
by the Technical Specifications (TS) for the standby liquid control
system. This change provides the additional shutdown reactivity
necessary to permit use of this fuel.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. The proposed amendment does not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
Proposed Change 1:
The proposed amendment will allow the loading and use of GE13
fuel assemblies in the Brunswick Unit 2 reactor core. The use of
GE13 fuel assemblies requires that the safety limit minimum critical
power ratio value also be revised. The safety limit minimum critical
power ratio is established to maintain fuel cladding integrity. The
GE13 fuel assembly design has been analyzed using methods that have
been previously approved by the Nuclear Regulatory Commission and
documented in General Electric Nuclear Energy's reload licensing
methodology Topical Report (NEDE-24011-P-A-10, ``General Electric
Standard Application for Reactor Fuel (GESTAR II)'' dated February
1991).
The proposed revision of the safety limit minimum critical power
ratio does not alter any plant safety-related equipment, safety
function, or plant operations that could change the probability of
an accident. The change does not affect the design, materials, or
construction standards applicable to the fuel bundles in a manner
that could change the probability of an accident.
A methodology that has been previously reviewed and accepted by
the Nuclear Regulatory Commission was used to derive the both
existing and updated safety limit minimum critical power ratio
value. The same methodology criteria have been applied to derive the
existing safety limit minimum critical power ratio of 1.07 as that
used to derive the updated safety limit minimum critical power ratio
value of 1.09. The updated safety limit minimum critical power ratio
assures that fuel cladding protection equivalent to that provided
with the existing safety limit minimum critical power ratio value is
maintained. This ensures that the consequences of previously
evaluated accidents are not significantly increased.
Proposed Change 2:
The standby liquid control system provides a means of reactivity
control that is independent of the normal reactivity control system.
The standby liquid control system must be capable of assuring that
the reactor core can be placed in a subcritical condition at any
time during reactor core life. Technical Specification Figure 3.1.5-
1 specifies the acceptable range of concentrations and volumes for
sodium pentaborate solution used as a neutron absorber (i.e., for
reactivity control). The portion of the sodium pentaborate
concentration range shown in Technical Specification Figure 3.1.5-1
applicable to the lower range of tank volumes is being revised to
increase the required concentration of sodium pentaborate solution.
This change is needed to account for the additional shutdown
reactivity needed based on the planned use of GE13 fuel assemblies
as reload fuel for the Unit 2 reactor core. Since the standby liquid
control system is independent from the normal means of controlling
reactor core reactivity and not used to control core reactivity
during normal plant operations, the proposed revision to the sodium
pentaborate concentration curve for the standby liquid control
system does not alter any plant safety-related equipment, safety
function, or plant operations that could change the probability of
an accident.
The current volume-concentration range of sodium pentaborate
used in the standby liquid control system will achieve a sufficient
concentration of boron in the reactor vessel to ensure reactor
shutdown. Based on the increased reactivity of the new GE13 reload
fuel assemblies, the required sodium pentaborate volume-
concentration range is being revised to ensure sufficient neutron
absorbing solution is available to achieve reactor shutdown;
therefore, the consequences of an accident previously evaluated are
not significantly increased.
2. The proposed amendment would not create the possibility of a
new or different kind of accident from any accident previously
evaluated.
Proposed Change 1:
The GE13 fuel assembly has been designed and complies with the
acceptance criteria contained in General Electric Nuclear Energy's
standard application for reactor fuel (GESTAR-II), which provides
the latest acceptance criteria for new General Electric fuel
designs. The GE13 fuel assembly complies with GESTAR-II acceptance
criteria that have been previously reviewed and accepted by the
Nuclear Regulatory Commission. The similarity of the GE13 fuel
design to the previously accepted GE11 fuel design, in conjunction
with the increased critical power capability of the GE13 fuel
design, ensure that no new mode or condition of plant operation is
being authorized by the loading and use of the GE13 fuel type. The
proposed revision of the safety limit minimum critical power ratio
from 1.07 to 1.09 does not modify any plant controls or equipment
that will change the plant's responses to any accident or transient
as given in any current analysis. Therefore, the proposed change to
allow the loading and use of the GE13 fuel type and the revision of
the safety limit minimum critical power ratio value from 1.07 to
1.09 will not create the possibility for a new or different kind of
accident from any accident previously evaluated.
Proposed Change 2:
As discussed above, the standby liquid control system provides a
means of reactivity control that is independent of the normal
reactivity control system and is capable of assuring that the
reactor core can be placed in a subcritical condition at any time
during reactor core life. The proposed revision to the sodium
pentaborate concentration range does not modify the standby liquid
control system or its controls, does not modify other plant systems
and equipment, and does not permit a new or different mode of plant
operation. As such, the proposed revision to the minimum pentaborate
concentration value 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.
Proposed Change 1:
As previously discussed, the GE13 fuel assembly design has been
analyzed using methods that have been previously approved by the
Nuclear Regulatory Commission and documented in General Electric
Nuclear Energy's reload licensing methodology Topical Report (NEDE-
24011-P-A-10, ``General Electric Standard Application for Reactor
Fuel (GESTAR II)'' dated February 1991). The safety limit minimum
critical power ratio value is selected to maintain the fuel cladding
integrity safety limit (i.e., that 99.9 percent of all fuel rods in
the core be expected to avoid boiling transition).
[[Page 49932]]
Appropriate operating limit minimum critical power ratio values are
established, based on the safety limit minimum critical power ratio
value, to ensure that the fuel cladding fuel integrity safety limit
is maintained. The operating limit minimum critical power ratio
values are incorporated in the Core Operating limits Report as
required by Technical Specification 6.9.3.1. The new GE13 safety
limit minimum critical power ratio value of 1.09 is based on the
same fuel cladding integrity safety limit criteria at that for the
GE11 safety limit minimum critical power ratio value of 1.07 (i.e.,
that 99.9 percent of all fuel rods in the core be expected to avoid
boiling transition); therefore, the proposed change does not result
in a significant reduction in the margin of safety.
Proposed Change 2:
As previously stated, the purpose of the standby liquid control
is to inject a neutron absorbing solution into the reactor in the
event that a sufficient number of control rods cannot be manually
inserted to maintain subcriticality. Sufficient solution is to be
injected such that the reactor will be brought from maximum rated
power conditions to subcritical over the entire reactor temperature
range from maximum operating to cold shutdown conditions. General
Electric reactor fuel methodology establishes a fuel type dependent
standby liquid control system shutdown margin to account for
calculational uncertainties. General Electric calculations show that
an in-vessel concentration of 660 ppm will provide an estimated
standby liquid control system minimum shutdown margin of 4.1% delta
k. To achieve an in-vessel concentration of 660 ppm, the acceptable
range of standby liquid control system tank concentrations is being
revised for the lower range of tank volumes. Thus, proposed revision
of the standby liquid control system sodium pentaborate volume-
concentration range ensures that there will not be a significant
reduction in the amount of available shutdown margin and, therefore,
not a significant reduction in the margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: University of North Carolina
at Wilmington, William Madison Randall Library, 601 S. College Road,
Wilmington, North Carolina 28403-3297.
Attorney for licensee: R. E. Jones, General Counsel, Carolina Power
& Light Company, Post Office Box 1551, Raleigh, North Carolina 27602
NRC Project Director: David B. Matthews
Duke Power Company, et al., Docket Nos. 50-413 and 50-414, Catawba
Nuclear Station, Units 1 and 2, York County, South Carolina
Date of amendment request: April 10, 1995
Description of amendment request: The requested amendment would
modify Technical Specification 4.6.4.3 to allow a reduction in the
number of hydrogen mitigation system igniters that must be maintained
Operable. This would allow removal of the hydrogen igniters in the
incore instrument tunnel.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
Criterion 1
The requested amendments will not involve a significant increase
in the probability or consequences of an accident previously
evaluated. No impact upon accident probabilities will be created,
since the EHM System is not an accident initiating system. In
addition, it has been demonstrated that based on the results of
computer analysis, and the review of results of an external study
performed for a similar type containment, that hydrogen
concentrations in the cavity during degraded core accidents will
remain within acceptable limits. No impact on the plant response to
any accident will be created (either design basis or beyond-design
basis).
Criterion 2
The requested amendments will not create the possibility of a
new or different kind of accident from any accident previously
evaluated. As stated previously, the EHM System is not an accident
initiating system. No new accident causal mechanisms will be created
as a result of deleting the affected igniters. Plant operation will
not be affected by the proposed amendments and no new failure modes
will be created.
Criterion 3
The requested amendments will not involve a significant
reduction in a margin of safety. No adverse impact upon any plant
safety margins will be created. As shown previously, applicable
computer analysis has successfully demonstrated that the affected
igniters could be removed with no adverse consequences. No fission
product barriers are being degraded. No change to the manner in
which the units are operated is being made.
Based upon the preceding analyses, Duke Power Company concludes
that the requested amendments do not involve a significant hazards
consideration.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: 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, et al., Docket Nos. 50-413 and 50-414, Catawba
Nuclear Station, Units 1 and 2, York County, South Carolina
Date of amendment request: September 1, 1995
Description of amendment request: Generic Letter 88-16 provided
guidance on removing cycle-specific parameters which are calculated
using NRC approved methodologies from Technical Specifications (TS).
The parameters are replaced in TS with a reference to a named report
which contains the parameters, and a requirement that the parameters
remain within the limits specified in the report. The proposed changes
incorporate NRC approved methodologies, approved revisions to
previously approved methodologies, or republished versions of
previously approved methodologies into Section 6.9 of the Catawba TS.
For Catawba, the limits to which these methodologies are applied are
explicitly listed in the TS. Since the proposed changes only
incorporate NRC approved methodologies into the TS the licensee
proposed that the changes are administrative in nature and can be
assumed to have no impact, or potential impact, on the health and
safety of the public.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
The proposed changes will not create a significant hazards
consideration, as defined by 10 CRF 50.92, because:
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 any
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.
[[Page 49933]]
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
amendment request involves 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, et al., Docket Nos. 50-413 and 50-414, Catawba
Nuclear Station, Units 1 and 2, York County, South Carolina
Date of amendment request: September 13, 1995
Description of amendment request: The proposed amendments modify
the notation for the overpower delta-temperature (OPDT) reactor trip
heatup setpoint penalty coefficient to be consistent with NUREG-0452,
Revision 4, ``Standard Technical Specifications for Westinghouse
Pressurized Water Reactors'' (STS). This change is necessary in order
to allow implementation of the modification to reduce the reactor
coolant system hot leg temperature as planned during the Unit 2 end-of-
cycle 7 refueling outage.
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:
As required by 10CFR50.91, this analysis is provided concerning
whether the requested amendments involve significant hazards
considerations, as defined by 10CFR50.92. Standards for
determination that an amendment request involves no significant
hazards considerations are if operation of the facility in
accordance with the requested 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.
Criterion 1
The proposed amendments will not involve a significant increase
in the probability or consequences of an accident previously
evaluated. The amendments will have no impact whatsoever upon the
probability of any accident being initiated, since the reactor trip
system is an accident mitigating system. The amendments will have no
adverse impact upon any accident consequences or upon the function
of the OPDT setpoint. The reactor trip heatup setpoint penalty will
continue to be applied anytime T-avg is greater than T [double
prime] and will not be applied when T-avg is less than or equal to T
[double prime]. This is consistent with the intent of this function.
Criterion 2
The proposed amendments will not create the possibility of a new
or different kind of accident from any accident previously
evaluated. The function of the OPDT setpoint will not be altered by
the proposed changes. As stated previously, the reactor trip system
is an accident mitigating system, so no new failure modes can be
created. No change to any aspect of plant operation will result from
NRC approval of the proposed amendments.
Criterion 3
The proposed amendments will not involve a significant reduction
in a margin of safety. The changes are necessary to allow full
implementation of the T-hot reduction modification on Catawba Unit
2. The proposed changes are consistent with the terminology of both
NUREG-0452, Revision 4 and NUREG-1431, Revision 1. OPDT setpoint
behavior will not be adversely impacted by the proposed changes;
therefore, no impact upon any plant safety margins will result.
Based upon the preceding analyses, Duke Power Company concludes
that the requested amendments do not involve a significant hazards
consideration.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: 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: March 29, 1995
Description of amendment request: The amendments would revise the
Technical Specification 3.4.9.3 requirements for the Low Temperature
Overpressure Protection (LTOP) system and update the heatup and
cooldown curves. The intent of the proposed amendments is to enhance
overpressure protection during low temperature operations. These
enhancements can be fully implemented, improving startup and shutdown
operation of McGuire Units 1 and 2.
Specifically, these changes are categorized into five groups
identified as follows:
1) Revisions to the LCO requirements, the Action Statements and the
SR for the Reactor Coolant System Overpressure Protection System during
low temperature conditions,
2) A reduction in the Reactor Coolant System (RCS) vent requirement
from 4.5 square inches to 2.75 square inches,
3) The use of the Residual Heat Removal suction relief valve (1ND3
and 2ND3) for overpressure protection under restricted conditions. (RCS
greater than 107 deg.F and cooldown rate less than 20 deg.F/hr; or RCS
greater than 167 deg.F),
4) Revisions of the Pressure/Temperature curves to 16 EFPY,
including the incorporation of the latest radiation surveillance
capsule results and removal of instrumentation margins from the
Technical Specification figures, and
5) Changes to format and consistency.
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 for each of the five groups listed above.
FIRST STANDARD
(Amendment would not) involve a significant increase in the
probablility or consequences of an accident previously evaluated.
1) Revised LCO [limiting conditions for operation] and SR
[surveillance requirements] for LTOP:
The reduced maximum setpoint will prevent the violation of the
10 CFR 50 Appendix G pressure/temperature curves (as modified by the
provisions of ASME Code Case N-514) during overpressure transients
at low temperatures. Since the maximum setpoint is reduced, the peak
pressure for LTOP [low-temperature overpressure protection] events
will be reduced as well. Accordingly, the consequences of an LTOP
event would not change as result of the proposed changes.
The analysis performed to determine the setpoint is, in
accordance with the methods used in previous evaluations, found
acceptable by the NRC. The three possible transients evaluated are;
1) a mass input from an operable safety injection pump; 2) a mass
input from an operable centrifugal charging pump; and 3) a heat
input from a 50 deg.F temperature difference between the steam
generators and the NC system. The LTOP setpoint of the PORV [power-
operated relief valve] proposed by this technical specification
change is not considered to be an initiator of any of these three
transients. As such, the probability of an accident
[[Page 49934]]
previously evaluated would not be increased as a result of the proposed
changes.
Two additional conditions for operability of the LTOP system are
defined (accumulator isolation and only one NV or NI pump operable)
and new surveillance requirements are specified as well. They
provide additional limitations, requirements and restrictions that
currently do not exist within the technical specifications for
McGuire. The incorporation of these proposed changes are consistent
with what is specified within NUREG-1341. Therefore, these changes
do not increase the probability of consequences of an accident
previously evaluated.
2) Reduction in NC vent opening:
The bases for the size of the vent to be established per the
technical specifications is to ensure that the 10 CFR 50, Appendix G
pressure/temperature limits are not exceeded during an LTOP event.
The determination of the size of the opening continues to preserve
the above design basis. The evaluation performed demonstrated that a
2.75 square inch opening would provide adequate overpressure
protection for the combined capacity of a centrifugal charging pump
and a safety injection pump.
The only time that the vent path is to be established is when
the PORVs may not be available. Defining the size of the vent is not
considered to be an initiator of any LTOP events that have been
previously evaluated. As such, this change in the size of the vent
opening does not increase the probability of an overpressure event
during low temperature conditions. The analysis performed verifies
that the size opening specified is sufficient to mitigate the
consequences of an LTOP event. Accordingly, the change in the size
of the opening for the vent will not impact the consequences of LTOP
events.
3) Use of RHR [residual heat removal] suction relief valves:
By letter dated September 11, 1990, the NRC authorized the
deletion of the RHR autoclosure interlock circuitry. A modification
which removed the RHR system suction isolation valve autoclosure
interlocks has been completed. As such, the RHR suction relief valve
can be exposed to NC system pressure and would be available to
mitigate LTOP events.
The proposed amendments specify the necessary requirements and
controls to ensure proper ND system alignments and conditions will
exist to protect the pressure/temperature limits. This added
relieving capacity will enhance the current LTOP system at McGuire
in mitigating overpressure events at low temperatures. As such, the
mitigation of previously evaluated LTOP events would be improved by
the proposed technical specification changes. Further, the proposed
changes would not esult in the initiation of an LTOP event or cause
an overpressure transient. Accordingly, the proposed amendment would
not involve an increase in the consequences or the probability of an
accident previously evaluated.
4) Revised pressure/temperature curves to 16 EFPY [effective
full-power year]:
The proposed pressure/temperature curves, provided by this
amendment request, satisfy all regulatory required material
embrittlement considerations including: ASME Section XI Appendix G,
10 CFR 50 Appendix G, and Regulatory Guide 1.99, Revision 2. In
addition, the margins for instrument error have been removed from
the curves. Instrument error will be administratively handled by
incorporating them into the LTOP system setpoint selection
calculations and into appropriate controlling procedures for unit
operations.
The proposed changes to the pressure/temperature curves are not
considered to be an initiator of LTOP events. The changes to the
curves proposed by this amendment request will not cause an LTOP
event. The curves define the new limits that have been defined in
accordance with regulatory requirements by which both units are to
be operated within. Accordingly, the proposed amendment will not
increase the probability or the consequences of previously evaluated
accidents.
5) Format and consistency:
The changes associated within this group are considered to be
administrative in nature. They do not affect station operability or
require any modifications to the facility. Accordingly, the proposed
amendment request does not increase the probability or consequences
of any previously evaluated accident.
SECOND STANDARD
(Amendment would not) create the possibility of a new or
different kind of accident from any kind of accident previously
evaluated.
1) Revised LCO and SR for LTOP:
The only potential impact to plant systems, structures and
components, as a result of the proposed changes associated with this
group, would be the setting of the PORV low pressure setpoint. No
other changes to plant systems, structures or components would
occur. The proposed amendments, also, would not impact the plant
operation. Although the value for the PORV pressure setting
specified within the technical specification would be reduced per
the proposed amendment, the actual settings of the PORV are now
currently set low enough to comply with the proposed lower setpoint
value. As such, the proposed lower setpoint would not require any
changes to the plant nor how the plant is operated.
The additional requirements for LTOP operability will not
require any modifications to the plant nor how the plant is
operated. Currently, when entering LTOP conditions, the accumulators
are isolated and only one NV or NI pump is capable of injecting into
the reactor vessel. these actions are currently controlled and are
specified within the operating procedures for heatup and cooldown of
the respective units. The proposed changes will now specify these
current operating requirements within the technical specifications
as well.
Accordingly, the proposed revisions will not create a new or
different kind of accident than what has already been previously
evaluated.
2) Reduction in NC vent opening:
The proposed changes to the technical specifications associated
with this group involves the size of the vent opening. The proposed
amendment reduces the size of the vent opening from 4.5 square
inches to 2.75 square inches. The analysis that was performed has
determined that the proposed size for the vent opening is adequate
for overpressure events. Therefore, this proposed revision to the
technical specifications will not result in a new or different kind
of accident from any kind of accident previously evaluated.
3) Use of RHR suction relief valves;
The proposed amendment associated with this group will specify
the necessary requirements and controls to ensure the appropriate
use of the RHR suction relief valve for overpressure protection.
This added relieving capacity will enhance the current LTOP system
in mitigating overpressure events during low temperature conditions.
The analysis that has been performed demonstrates the adequacy of
the RHR suction relief valve, in conjunction with a PORV, in
mitigating overpressure events at low temperatures, assuming a worst
case single failure as well. As such, the use of the RHR suction
relief valve in the manner prescribed by the proposed technical
specification amendment will not create a new or different kind of
accident from those accidents that have been previously evaluated.
4) Revised pressure/temperature curves to 16 EFPY:
The changes associated with this group, provide new heatup and
cooldown curves for both Units 1 and 2, which will extend the
service period from 10 EFPY to 16 EFPY and will remove the
instrument error as well. The proposed [heatup] and cooldown curves
were developed in accordance with all regulatory required material
embrittlement criteria. Thus, operation of the units in accordance
with the proposed new pressure/temperature curves will not create
the possibility of a new or different kind of accident from those
accident[s] that have been previously evaluated.
5) Format and consistency:
The changes associated within this group are considered to be
administrative in nature. They do not affect station operability or
require any modifications to the facility. Accordingly, the proposed
amendment will create the possibility of a new or different kind of
accident from that previously evaluated.
THIRD STANDARD
(Amendment would not) involve a significant reduction in a
margin of safety.
1) Revised LCO and SR for LTOP:
This proposed change will reduce the maximum PORV setpoint such
that, for LTOP events, the maximum pressure in the vessel would not
exceed 110% of the pressure/temperature limits that have been
established in accordance with ASME Appendix G. This is congruous
with the provisions of ASME Code Case N-514. Currently, the maximum
PORV setpoint for LTOP events ensure that the maximum pressure would
not exceed 100% of the pressure/temperature curves. As such, the
proposed change appears to involve a slight reduction in a margin of
safety.
Although the proposed change may involve a slight reduction in a
margin of safety, the proposed change will provide an
[[Page 49935]]
equivalent margins of safety to the reactor vessel during LTOP
transients and will satisfy the underlying purpose of 10 CFR 50.60
for fracture toughness requirements. By letter dated June 28, 1994,
an exemption request and authorization to use ASME Code Case N-514
at McGuire was submitted to the NRC for review and approval.
Approval for the use of the code case was granted on September 30,
1994. The proposed change to reduce the maximum PORV setpoint,
coupled with the September 30, 1994 NRC approval for the use of Code
Case N-514 satisfies current regulatory acceptance criteria.
Therefore, the proposed change would not involve a significant
reduction in a margin of safety.
This change group, also, defines two additional conditions for
the operability of the LTOP system (accumulator isolation and only
one NV or NI pump operable) and proposes new surveillance
requirements and restrictions that currently do not exist within the
technical specifications for McGuire. The incorporation of these
proposed changes are consistent with what is specified within NUREG-
1341. Therefore, these changes do not involve a significant
reduction in a margin of safety.
2) Reduction in NC vent opening:
The proposed changes to the technical specifications associated
with this group involves the size of the vent opening. The proposed
amendment reduces the size of the vent opening from 4.5 square
inches to 2.75 square inches. The basis for the size of the vent to
be established per the technical specifications is to ensure that
the 10 CFR 50, Appendix G pressure/temperature limits are not
exceeded during an LTOP event. The determination of the size of the
opening continues to preserve the above design basis. The evaluation
performed demonstrated that a 2.75 square inch opening would provide
adequate overpressure protection for the combined capacity of a
centrifugal charging pump and a safety injection pump. Accordingly,
the proposed changes would not involve a significant reduction in a
margin of safety.
3) Use of RHR suction relief valves:
The proposed amendment associated with this group will specify
the necessary requirements and controls to ensure the appropriate
use of the RHR suction relief valves for overpressure protection.
This added relieving capacity will enhance the current LTOP system
in mitigating overpressure events during low temperature conditions.
The analysis that has been performed demonstrates the adequacy of
the RHR suction relief valve, in conjunction with a PORV, in
mitigating overpressure events at low temperatures.
Further, by letter dated September 11, 1990, the NRC approved
amendments to delete a portion of the surveillance requirements
regarding periodic verification that the RHR suction isolation
valves automatically close on a RCS [reactor coolant system] signal
less than or equal to 560 psig. This action, in effect, authorizes
the removal of the RHR autoclosure interlock circuitry. As discussed
within the NRC Safety evaluation for the amendment, the Commission
and industry have recognized the safety benefits of removing the ACI
[automatic closure and interlock] circuitry from the RHR system to
minimize, and thus reduce the risk associated with loss of decay
heat removal events.
Therefore, the proposed amendments associated with this change
group will not involve a significant reduction in a margin of
safety.
4) Revised pressure/temperature curves to 16 EFPY:
The changes associated with this group provide new heatup and
cooldown curves for both Units 1 and 2, which will extend the
service period from 10 EFPY to 16 EFPY and will relocate the
instrument error as well. The proposed pressure/temperature curves
provided by this amendment request satisfy all regulatory required
material embrittlement considerations including; ASME Section XI
Appendix G, 10 CFR 50 Appendix G, and Regulatory Guide 1.99,
Revision 2. The instrument error will be administratively handled by
incorporating them into the LTOP system setpoint selection
calculations and into the controlling procedures for unit
operations.
The relocation of the instrument error to licensee controlled
documents is consistent with the NRC actions proposed within NUREG-
1431, new standard technical specifications for Westinghouse plants.
As prescribed within NUREG-1431, the pressure/temperature limit
curves are to be relocated to a licensee controlled document
entitled ``Pressure Temperature Limit Report (PTLR)''. Changes to
the heatup and cooldown curves would then be performed in accordance
with 10 CFR 50.59 criteria. For the situation proposed by this
amendment, updates and revisions of the instrument error associated
with the pressure/temperature limit curves will be processed in a
similar fashion. Thus, the proposed change to relocate the
instrument error to licensee controlled documents is analogous with
NRC acceptable practices.
Accordingly, the proposed changes will not reduce a margin of
safety.
5) Format and consistency:
The changes associated within this group are considered to be
administrative in nature. They do not affect station operability or
require any modifications to the facility. Accordingly, there is no
reduction in the margin of safety of the LTOP system due to the
incorporation of these editorial/administrative changes.
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
Florida Power and Light Company, et al., Docket Nos. 50-335 and 50-
389, St. Lucie Plant, Unit Nos. 1 and 2, St. Lucie County, Florida
Date of amendment request: June 21, 1995
Description of amendment request: The proposed amendments will
revise the action statements for a single inoperable Emergency Diesel
Generator (EDG), TS 3.8.1.1.b, to extend the allowed outage time (AOT)
from 72 hours to 7 days, and permit a 10 day AOT to be used once per
refueling cycle. This proposal is a result of a cooperative study by
participating Combustion Engineering Owners Group members which
concluded that the proposed AOT extension improves plant operational
flexibility while adequately controlling overall plant risk.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
(1) Operation of the facility in accordance with the proposed
amendment would not involve a significant increase in the
probability or consequences of an accident previously evaluated.
The proposed amendments for St. Lucie Unit 1 and Unit 2 will
extend the action completion/allowed outage time (AOT) for a single
inoperable Emergency Diesel Generator (EDG) from 72 hours to 7 days,
with provisions for a 10 day AOT once per refueling cycle. The EDGs
are designed as backup AC power sources for essential safety systems
in the event of a loss of offsite power. As such, the EDGs are not
accident initiators, and an extended AOT to restore operability of
an inoperable diesel generator would not increase the probability of
occurrence of accidents previously analyzed.
The proposed technical specification revisions involve the AOT
for a single inoperable EDG, and do not change the conditions,
operating configuration, or minimum amount of operating equipment
assumed in the plant safety analyses for accident mitigation. In
addition, a Probability Safety Assessment (PSA) was performed to
quantitatively assess the risk impact of the proposed amendment. The
impact on the early radiological release probability for design
basis events was also evaluated. It was concluded that the risk
contribution from this proposed AOT is very small, and that the
impact will be negligible.
Therefore, operation of either facility in accordance with its
proposed amendment would not involve a significant increase in the
probability or consequences of an accident previously evaluated.
(2) Operation of the facility in accordance with the proposed
amendment would not
[[Page 49936]]
create the possibility of a new or different kind of accident from any
accident previously evaluated.
The proposed amendments will not change the physical plant or
the modes of plant operation defined in either Facility License. The
changes do not involve the addition or modification of equipment,
nor do they alter the design of plant systems. Therefore, operation
of either facility in accordance with its proposed amendment would
not create the possibility of a new or different kind of accident
from any accident previously evaluated.
(3) Operation of the facility in accordance with the proposed
amendment would not involve a significant reduction in a margin of
safety.
The proposed amendments are designed to improve EDG reliability
by providing flexibility in the scheduling and performance of
preventive and corrective maintenance activities. The surveillance
intervals or the operability requirements are not changed by the
proposal; only the AOT for a single inoperable EDG will be extended.
The proposed changes do not alter the basis for any technical
specification that is related to the establishment of, or the
maintenance of, a nuclear safety margin. Moreover, an integrated
assessment of the risk impact of extending the AOT for a single
inoperable EDG has determined that the risk contribution is very
small and can be offset by improvements in EDG reliability.
Therefore, operation of either facility in accordance with its
proposed amendment would not involve a significant reduction in a
margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: Indian River Junior College
Library, 3209 Virginia Avenue, Fort Pierce, Florida 34954-9003
Attorney for licensee: J. R. Newman, Morgan, Lewis & Bockius, 1800
M Street, NW., Washington, DC 20036
NRC Project Director: David B. Matthews
Florida Power and Light Company, et al., Docket Nos. 50-335 and 50-
389, St. Lucie Plant, Unit Nos. 1 and 2, St. Lucie County, Florida
Date of amendment request: June 21, 1995
Description of amendment request: The proposed amendments will
revise TS 3.5.2 to allow up to 7 days to restore an inoperable Low
Pressure Safety Injection train to operable status. This proposal is a
result of a cooperative study by participating Combustion Engineering
Owners Group members which concluded that an extension of the allowed
outage time (AOT) from 72 hours to 7 days can improve plant operational
flexibility and is risk beneficial.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
(1) Operation of the facility in accordance with the proposed
amendment would not involve a significant increase in the
probability or consequences of an accident previously evaluated.
The proposed amendments for St. Lucie Unit 1 and Unit 2 will
extend the action completion/allowed outage time (AOT) for a single
inoperable Low Pressure Safety Injection (LPSI) train from 72 hours
to 7 days. A LPSI train is designed as a part of each Emergency Core
Cooling System (ECCS) subsystem to supplement Safety Injection Tank
(SIT) inventory during the early stages of mitigating a Design Basis
Accident. As such, components of the LPSI system are not accident
initiators, and an extended AOT to restore operability of an
inoperable LPSI train would not increase the probability of
occurrence of accidents previously analyzed.
The safety analyses for both St. Lucie Units demonstrate that
ECCS performance acceptance criteria are satisfied with only one of
the two redundant ECCS subsystems operating during the postulated
Design Basis Accident. The proposed technical specification
revisions involve the AOT for a single inoperable LPSI train, and do
not change the conditions assumed for the minimum amount of
operating equipment needed for accident mitigation. Therefore, the
consequences of an accident previously evaluated will not be
significantly increased.
In addition to the preceding evaluation, a Probabilistic Safety
Analysis (PSA) was performed to quantitatively assess the risk
impact of the proposed amendments. It was concluded from the results
of that assessment that the risk contribution of the AOT extension
is very small, and that the net impact of the proposed amendment can
be risk beneficial.
Therefore, operation of either facility in accordance with its
proposed amendment would not involve a significant increase in the
probability or consequences of an accident previously evaluated.
(2) Operation of the facility in accordance with the proposed
amendment would not create the possibility of a new or different
kind of accident from any accident previously evaluated.
The proposed amendments will not change the physical plant or
the modes of plant operation defined in either Facility License. The
changes do not involve the addition or modification of equipment nor
do they alter the design of plant systems. Therefore, operation of
either facility in accordance with its proposed amendment would not
create the possibility of a new or different kind of accident from
any accident previously evaluated.
(3) Operation of the facility in accordance with the proposed
amendment would not involve a significant reduction in a margin of
safety.
The margin of safety associated with the ECCS system is
established by acceptance criteria for system performance defined in
10 CFR 50.46. The proposed amendments will not change this
acceptance criteria nor the operability requirements for equipment
that is used to achieve such performance as demonstrated in the
plant safety analyses. Moreover, an integrated assessment of the
risk impact of extending the AOT for a single inoperable LPSI train
has concluded that the risk contribution is very small, LPSI system
reliability can potentially be improved, and the net impact of the
proposed change can be risk beneficial. Therefore, operation of
either facility in accordance with its proposed amendment would not
involve a significant reduction in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: Indian River Junior College
Library, 3209 Virginia Avenue, Fort Pierce, Florida 34954-9003
Attorney for licensee: J. R. Newman, Morgan, Lewis & Bockius, 1800
M Street, NW., Washington, DC 20036
NRC Project Director: David B. Matthews
Florida Power and Light Company, et al., Docket Nos. 50-335 and 50-
389, St. Lucie Plant, Unit Nos. 1 and 2, St. Lucie County, Florida
Date of amendment request: June 21, 1995
Description of amendment request: The proposed amendments will
revise the action statements and certain surveillances of TS 3/4.5.1,
Safety Injection Tanks (SIT). This proposal is based on the results of
a cooperative study performed by participating Combustion Engineering
Owners Group members which investigated the impact of a risk-based
allowed outage time (AOT) extension, and also included recommendations
for line-item TS improvements from NUREG-1366 and Generic Letter 93-05.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
(1) Operation of the facility in accordance with the proposed
amendment would not involve a significant increase in the
probability or consequences of an accident previously evaluated.
The license amendments proposed for St. Lucie Units 1 and 2
incorporate certain line-
[[Page 49937]]
item Technical Specifications (TS) improvements for the Safety
Injection Tanks (SIT), and include an extension of the required
action completion/allowed outage time (AOT) from one hour to 72
hours to restore an inoperable SIT (that is still able to perform
its safety function) to operable status. In addition, an AOT of 24
hours, based on risk assessment techniques, is proposed for an SIT
that may be unable to perform its design function.
The SITs are passive components of the Emergency Core Cooling
System (ECCS). As such, they are not accident initiators for any
transient evaluated in the plant safety analyses, and an extension
of the AOTs for restoring an inoperable SIT to operable status would
not increase the probability of occurrence of accidents previously
analyzed.
The SITs, in combination with other ECCS components, are used to
mitigate the consequences of a loss of coolant accident. The TS
revisions will provide a longer AOT for a single inoperable SIT, but
do not involve a change to the ECCS configuration or method of
operation. The proposed amendments will not change the conditions
assumed for the minimum amount of operating equipment needed for
accident mitigation. Therefore, the consequences of an accident
previously evaluated will not be significantly increased.
In addition to the preceding evaluation, a Probability Safety
Assessment (PSA) was performed to quantitatively assess the risk
impact of the 24 hour AOT proposal. The impact on the early
radiological release probability for design basis events was also
evaluated. It was concluded that the risk contribution from this AOT
is very small, and that the impact is negligible.
Therefore, operation of either facility in accordance with its
proposed amendment would not involve a significant increase in the
probability or consequences of an accident previously evaluated.
(2) Operation of the facility in accordance with the proposed
amendment would not create the possibility of a new or different
kind of accident from any accident previously evaluated.
The proposed amendments will not change the physical plant or
the modes of plant operation defined in either Facility License. The
changes do not involve the addition or modification of equipment,
nor do they alter the design of plant systems. Therefore, operation
of either facility in accordance with its proposed amendment would
not create the possibility of a new or different kind of accident
from any accident previously evaluated.
(3) Operation of the facility in accordance with the proposed
amendment would not involve a significant reduction in a margin of
safety.
The margin of safety associated with the ECCS system is
established by acceptance criteria for system performance defined in
10 CFR 50.46. The proposed amendments will not change this criteria
nor the operability requirements for equipment that is used to
achieve such performance as demonstrated by the plant safety
analyses. Moreover, an integrated assessment of the risk impact of
allowing 24 hours to restore an inoperable SIT to operable status
has concluded that this impact is very small, and can be offset by
averting an unnecessary transition to the shutdown modes. Therefore,
operation of either facility in accordance with its proposed
amendment would not involve a significant reduction in a margin of
safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: Indian River Junior College
Library, 3209 Virginia Avenue, Fort Pierce, Florida 34954-9003
Attorney for licensee: J. R. Newman, Morgan, Lewis & Bockius, 1800
M Street, NW., Washington, DC 20036
NRC Project Director: David B. Matthews
Florida Power and Light Company, et al., Docket Nos. 50-335 and 50-
389, St. Lucie Plant, Unit Nos. 1 and 2, St. Lucie County, Florida
Date of amendment request: August 16, 1995
Description of amendment request: The revisions will modify
Technical Specification 3.6.6.1, Shield Building Ventilation System
(SBVS), to more effectively address the design functions performed by
the SBVS for both the Shield Building (secondary containment) and the
Fuel Handling Building.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
(1) Operation of the facility in accordance with the proposed
amendment would not involve a significant increase in the
probability or consequences of an accident previously evaluated.
The proposed license amendment for St. Lucie Unit 2 will clarify
the Applicability and the Actions required by Technical
Specification (TS) 3.6.6.1, and explicitly account for the dual
purpose of the Shield Building Ventilation System (SBVS) to perform
design functions for both the Shield Building (secondary
containment) and the Fuel Handling Building. The proposed amendment
is administrative in nature.
The SBVS only operates when actuated by automatic control
signals generated by systems detecting postulated accident
conditions. The SBVS is not an accident initiator, the proposed TS
changes do not involve any assumptions relative to accident
initiators used in the plant safety analyses, and the amendment,
therefore, will not impact the probability of occurrence for
accidents previously analyzed. Relative to accident consequences, at
least one train of the SBVS must operate to fulfill the design
function of evacuating filtered air from the Shield Building during
the postulated Loss of Coolant Accident; and likewise assumed in the
analysis for the Fuel Handling Building during a fuel handling
accident. The proposed changes simply remove elements of ambiguity
from TS 3.6.6.1; do not reduce the existing operability requirements
for the system; and provide further assurance that proper
compensatory measures will be taken in the event one or both SBVS
trains become inoperable.
Therefore, operation of the facility in accordance with the
proposed amendment will not involve a significant increase in the
probability or consequences of an accident previously evaluated.
(2) Operation of the facility in accordance with the proposed
amendment would not create the possibility of a new or different
kind of accident from any accident previously evaluated.
The proposed amendment is administrative in nature and will not
change the physical plant or the modes of plant operation defined in
the facility license. The changes do not involve the addition or
modification of equipment, nor do they alter the design or methods
of operation of plant systems. Plant configurations that are
prohibited by TS will not be created by this amendment. Therefore,
operation of the facility in accordance with the proposed amendment
would not create the possibility of a new or different kind of
accident from any accident previously evaluated.
(3) Operation of the facility in accordance with the proposed
amendment would not involve a significant reduction in a margin of
safety.
The proposed amendment will not change the SBVS operability
requirements nor otherwise alter the basis for any technical
specification that is related to the establishment of, or the
maintenance of, a nuclear safety margin. The proposed changes are
administrative in nature, and are designed to provide assurance that
the SBVS capability to perform design functions assumed available in
the safety analyses will remain available during the various plant
operating modes. Therefore, operation of the facility in accordance
with the proposed amendment would not involve a significant
reduction in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: Indian River Junior College
Library, 3209 Virginia Avenue, Fort Pierce, Florida 34954-9003
Attorney for licensee: J. R. Newman, Morgan, Lewis & Bockius, 1800
M Street, NW., Washington, DC 20036
NRC Project Director: David B. Matthews
[[Page 49938]]
Florida Power and Light Company, et al., Docket Nos. 50-335 and 50-
389, St. Lucie Plant, Unit Nos. 1 and 2, St. Lucie County, Florida
Date of amendment request: August 16, 1995
Description of amendment request: The proposed amendments revise
St. Lucie Units 1 and 2 Technical Specifications to relocate selected
Technical Specification Monitoring Instrumentation utilizing the Final
Policy Statement on Technical Specification Improvement for Nuclear
Power Reactors, 58 FR 39132, July 22, 1993. The proposed amendments
also include relocation of Technical Specifications related to the
Emergency and Security Plan review process utilizing the guidance
contained in NRC Generic Letter 93-07, ``Modification of the Technical
Specification Administrative Requirements for Emergency and Security
Plans.''
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
(1) Operation of the facility in accordance with the proposed
amendments would not involve a significant increase in the
probability or consequences of an accident previously evaluated.
The proposed changes to the Selected Technical Specification
Requirements Related to Instrumentation are administrative in nature
in that the specifications for operation and surveillance of the
selected Technical Specification instrumentation will be relocated
from Appendix A of the facility operating license to the Updated
Final Safety Analysis Report (UFSAR) for each unit. Once relocated,
future changes will be controlled by 10 CFR 50.59 and the UFSARs
updated pursuant to 10 CFR 50.71(e). Relocation of these
requirements to the UFSAR is consistent with the NRC ``Final Policy
Statement on Technical Specifications Improvements for Nuclear Power
Reactors'' published in the Federal Register (58 FR 39132) dated
July 22, 1993.
The selected Technical Specification instruments are not
accident initiators nor a part of the success path(s) which function
to mitigate accidents evaluated in the plant safety analyses. The
proposed Technical Specification change does not involve any change
to the configuration or method of operation of any plant equipment
that is used to mitigate the consequences of an accident, nor do the
changes alter any assumptions or conditions in any of the plant
accident analyses. Therefore, operation of the facility in
accordance with the proposed amendment would not involve a
significant increase in the probability or consequences of an
accident previously evaluated.
The Technical Specifications changes associated with Emergency
Plan and Security Plan requirements are proposed in accordance with
Generic Letter 93 07. The changes being proposed are administrative
in nature and do not affect assumptions contained in plant safety
analyses, the physical design and/or operation of the plant, nor do
they affect Technical Specifications that preserve safety analysis
assumptions. Therefore, operation of the facility in accordance with
the proposed amendments would not affect the probability or
consequences of an accident previously analyzed.
(2) Use of the modified specification would not create the
possibility of a new or different kind of accident from any
previously evaluated.
The proposed amendment to relocate the existing Technical
Specification requirements for selected Technical Specification
instrumentation to the UFSAR will not change the physical plant or
the modes of plant operation defined in the Facility License. The
change does not involve the addition or modification of equipment
nor does it alter the design or operation of plant systems.
Therefore, operation of the facility in accordance with the proposed
amendment would not create the possibility of a new or different
kind of accident from any accident previously evaluated.
The proposed amendments, in accordance with Generic Letter 93-
07, change the Technical Specifications to remove the audit of the
emergency and security plans and implementing procedures from the
list of responsibilities of the Facility Review Group. The changes
being proposed are administrative in nature and will not change the
physical plant or the modes of operation defined in the Facility
License. The change does not involve the addition or modification of
equipment nor does it alter the design or operation of plant
systems. Therefore, operation of the facility in accordance with the
proposed amendments would not create the possibility of a new or
different kind of accident from any accident previously evaluated.
(3) Use of the modified specification would not involve a
significant reduction in a margin of safety.
The proposed changes are administrative in nature in that
operating and surveillance requirements for the selected Technical
Specification instrumentation will be relocated from Appendix A of
the facility license to the appropriate Updated Final Safety
Analysis Report for each unit. These selected instruments are not
used to actuate safety-related equipment, provide interlocks, or
otherwise perform plant control functions. Conditions evaluated in
plant accident and transient analyses do not involve these selected
instruments. The proposed changes do not alter the basis for any
technical specification that is related to the establishment of, or
the maintenance of, a nuclear safety margin. Therefore, operation of
the facility in accordance with the proposed amendment would not
involve a significant reduction in a margin of safety.
The proposed amendments, in accordance with Generic Letter 93-
07, change the Technical Specifications to remove the audit of the
emergency and security plans and implementing procedures from the
list of responsibilities of the Facility Review Group. The changes
being proposed are administrative in nature and do not alter the
bases for assurance that safety-related activities are performed
correctly or the basis for any Technical Specification that is
related to the establishment of or maintenance of a safety margin.
Therefore, operation of the facility in accordance with the proposed
amendments would not involve a significant reduction in a margin of
safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: Indian River Junior College
Library, 3209 Virginia Avenue, Fort Pierce, Florida 34954-9003
Attorney for licensee: J. R. Newman, Morgan, Lewis & Bockius, 1800
M Street, N.W., Washington, DC 20036
NRC Project Director: David B. Matthews
IES Utilities Inc., Docket No. 50-331, Duane Arnold Energy Center,
Linn County, Iowa
Date of amendment request: July 21, 1995
Description of amendment request: The proposed amendment would make
administrative changes to various sections of the Duane Arnold Energy
Center (DAEC) Technical Specifications (TS). These changes replace a
conditional surveillance if one emergency service water (ESW) pump or
loop is determined to be inoperable (TS 4.8.E.2); credit successful
emergency diesel generator (EDG) tests performed in the previous 24
hours (TS 4.8.E.2); clarify the requirements governing spent and new
fuel storage in Section 5.5 of the DAEC TS; and eliminate the
Operations Committee reviews of procedures in support of the DAEC
Emergency Plan and Security Plan, as specified in Sections 6.5 and 6.8
of the TS. DAEC TS Section 4.8.E.2 states the surveillance requirement
applicable when one ESW pump or loop is determined to be inoperable.
This amendment request deletes the surveillance requirement to
physically test the opposite train's EDG and replaces it with a
requirement to verify OPERABILITY of the opposite train low pressure
core and containment cooling systems and EDG.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the
[[Page 49939]]
licensee has provided its analysis of the issue of no significant
hazards consideration, which is presented below:
1. The proposed revision does not involve a significant increase
in the probability or consequences of an accident previously
evaluated. The changes are administrative in nature and are
consistent with previously-published NRC guidance. The proposed
revision does not change any accident analysis, plant safety
analysis or calculations; degrade existing plant programs; or modify
any functions of safety related systems or accident mitigation
functions for which the DAEC has previously been credited. The
proposed revision to the Surveillance Requirements will continue to
assure OPERABILITY as required, but eliminate unnecessary operation
of an EDG.
2. The proposed revision does not create the possibility of a
new or different kind of accident from any accident previously
evaluated. The proposed revision does not alter any plant
parameters, revise any safety limit setpoint, or provide any new
release pathways. In addition, the proposed revision does not modify
the operation or function of any safety-related equipment, nor
introduce any new modes of operation, failure modes, or physical
changes to the plant.
3. The proposed revision does not involve a significant
reduction in a margin of safety. The proposed revision does not
alter any plant parameters, revise any safety limit setpoint, or
provide any new release pathways. In addition, the proposed revision
does not modify the operation or function of any safety-related
equipment, nor introduce any new modes of operation, failure modes,
or physical changes to the plant.
The NRC staff has reviewed the licensee's analysis and, based on
thisreview, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: Cedar Rapids Public Library,
500 First Street, S.E., Cedar Rapids, Iowa 52401
Attorney for licensee: Jack Newman, Kathleen H. Shea, Morgan, Lewis
& Bockius, 1800 M Street, NW., Washington, DC 20036-5869
NRC Project Director: Gail H. Marcus
Nebraska Public Power District, Docket No. 50-298, Cooper Nuclear
Station, Nemaha County, Nebraska
Date of amendment request: May 5, 1995, as revised by letter dated
July 14, 1995
Description of amendment request: The proposed changes would amend
the Cooper Nuclear Station (CNS) Technical Specifications (TS) sections
3/4.5.F.1, 3.5.F.2, 3.9.B.1, 3.9.B.2, 4.9.A.2, and the associated
bases. These changes would revise the TS to: 1) verify that the
redundant diesel generator is operable upon the loss of one diesel
generator, and implement provisions to verify that the operable diesel
generator does not have a common cause failure; 2) incorporate
provisions to allow a modified start for the diesel generators; and 3)
remove the requirement that the reactor power level be reduced to 25%
of rated power upon loss of both diesel generator units or both
incoming power sources (start-up and emergency transformers). In
addition, the period of time allowed for continued reactor operation
with both diesels inoperable would be reduced from 24 to two hours.
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:
10 CFR 50.91(a)(1), requires that licensee requests for
operating license amendments be accompanied by an evaluation of
significant hazards posed by the issuance of the amendment. NPPD has
reviewed the proposed changes in accordance with 10CFR50.92 and
concludes that the changes do not involve a significant hazards
consideration (SHC). The basis for this conclusion is that the three
criteria of 10CFR50.92(c) are not compromised. The proposed changes
do not involve a SHC because the changes would not:
1. Involve a significant increase in the probability or
consequences of an accident previously analyzed.
Proposed Revision 1:
This proposed revision serves to ensure that an emergency diesel
generator is always available to perform on demand and that lowering
the number of demands to demonstrate operability reduces the
probability of equipment failure. The required action no longer
requires the redundant emergency diesel generator to be demonstrated
operable immediately. Therefore, this requirement has been deleted
from TS 4.5.F.1.
The proposed change includes provisions to determine if the
redundant diesel generator has been made inoperable by a common
cause failure or perform a demonstration test. The redundant
emergency diesel generator will remain in service during the entire
period of inoperability of the out of service emergency diesel
generator. If a common cause failure cannot be ruled out, the
redundant diesel generator will be tested in accordance with the
surveillance requirements of TS section 4.9.A.2.a.1 to assure
operability.
Since this proposed revision does not affect the design or
negatively affect the performance of the diesel generators, the
change will not result in an increase in the consequences or
probability of an accident previously analyzed. This proposed
revision will increase diesel generator reliability and
availability, thereby increasing overall plant safety.
Proposed Revision 2:
This proposed revision only affects emergency diesel generator
periodic testing. The diesel generators are not accident initiators
and the method of testing the diesel generators cannot initiate an
accident and therefore will not increase the probability of an
accident. This change to the diesel generator testing method does
not impact any Updated Safety Analysis Report (USAR) safety
analysis. The proposed surveillances will still provide assurance
that the diesel generators are available to mitigate the
consequences of accidents previously evaluated. Thus the
consequences of an accident previously evaluated are not increased.
The revised periodic testing will still demonstrate that the
emergency diesel generators are ready to perform their safety
function. An overall improvement in diesel engine reliability and
availability can be gained by performing diesel generator starts for
surveillance testing using engine prelubes, warmups and other
manufacturer recommended practices to reduce engine stress and wear.
Since this proposed revision does not affect the design or
negatively affect the performance of the diesel generators, the
change will not result in an increase in the consequences or
probability of an accident previously analyzed. This proposed
revision will increase diesel generator reliability, thereby
increasing overall plant safety.
Proposed Revision 3:
This proposed revision does not affect the operation of the
emergency diesel generators or the incoming power sources (start-up
and emergency transformers). Both the diesel generators and the
incoming power sources function to mitigate the consequences of
postulated accidents. As such, removing the requirement to reduce
power level upon the loss of both redundant components in either of
these systems does not create an increase in the probability of an
accident. By eliminating this requirement, the potential for plant
transients during power reduction to 25% are also eliminated.
Eliminating this requirement will not increase the consequences of a
postulated accident because the redundant components will remain
available. Additionally, the loss of both offsite power sources
condition becomes more restrictive by requiring a plant shutdown
instead of notification within 24 hours.
The proposed changes do not alter the conditions or assumptions
in any of the Updated Safety Analysis Report (USAR) accident
analyses. Since the USAR accident analyses remains bounding, the
radiological consequences previously evaluated are not adversely
affected by the proposed changes. Therefore, no significant increase
in the probability or consequences of an accident previously
analyzed would occur.
The proposed rearrangement of information, and rewording of some
the TS requirements are included to enhance usability and alleviate
any possible confusion. These changes are strictly editorial have no
impact, and do not alter technical content or meaning of the
specifications. These editorial changes do not involve a significant
increase in the probability or consequences of an accident
previously analyzed.
[[Page 49940]]
2. Create the possibility of a new or different kind of accident
from any previously analyzed.
Proposed Revision 1:
Accidents involving loss of off-site power and single failure
have been previously evaluated, and this proposed change does not
impact any of those assumptions. This proposed revision does not
introduce any new mode of plant operation or new accident
precursors, involve any physical alterations to plant
configurations, or make changes to system setpoints which could
initiate a new or different kind of accident. Operation of the
facility in accordance with the proposed revised changes does not
create the possibility of a new or different kind of accident from
any previously evaluated.
Proposed Revision 2:
This proposed revision only affects emergency diesel generator
periodic testing. The diesel generators are not accident initiators
and the method of testing the diesel generators cannot initiate an
accident. This revision does not relieve the operation of the diesel
generator from existing requirements and the diesel generators
remain bounded by the assumptions in the USAR accident analysis. The
method of testing provides assurance that the diesel generators are
available when needed. The proposed revision does not involve any
changes in setpoints, plant equipment, plant operation, protective
functions, or the design basis of the plant. Therefore, a change in
the method of starting the diesel generators during periodic testing
would not create a different kind of accident than previously
evaluated.
Proposed Revision 3:
This proposed revision does not add or change any equipment or
logic, nor do the changes associated with this revision alter any
system operability requirements. The proposed changes for this
revision do not introduce any new failure modes for any plant system
or component important to safety nor has any new limiting failure
been identified as a result of the proposed revision. Since there
are no changes to the function, or operation of any system,
equipment, or component, the possibility of a new or different kind
of accident is not created.
The proposed rearrangement of information, and rewording of some
[of] the TS requirements are included to enhance usability and
alleviate any possible confusion. These changes are strictly
editorial have no impact, and do not alter technical content or
meaning of the specifications. These editorial changes do not create
the possibility of a new or different kind of accident from any
previously analyzed.
3. Involve a significant reduction in the margin of safety.
Proposed Revision 1:
This proposed revision does not result in an overall reduction
in the margin of safety. The reduction in margin going from
``immediately'' testing an operable diesel generator to 24 hours to
determine no common cause, is offset by the increase in margin
resulting from increased diesel generator reliability and
availability associated with implementing the vendor recommendations
for testing and not exposing the diesel generator to potential grid
disturbances when a diesel generator is found to be inoperable. No
physical modification to the plant or change in the procedurally
prescribed operator actions result from the proposed changes
associated with this revision. Operation of the facility in
accordance with the proposed revision does not involve a significant
reduction in a margin of safety.
Proposed Revision 2:
This proposed revision is made to increase the reliability and
availability of the emergency diesel generators thus enhancing the
safety of the plant. Changing the way periodic testing of the diesel
generators is conducted does not involve a reduction in safety. The
test still demonstrates the ability of the diesel generator to start
within the time required, and reach rated voltage and frequency as
required in the accident analysis. The test also demonstrates the
ability of the diesel generator to start reliably, carry the
required load, and ensures the capabilities of the cooling system
and other support systems are operable. Therefore, assurance that
the diesel generators operate within the limits determined to be
acceptable continues to be provided. Implementing manufacturer's
recommendations to minimize stress and wear of the diesel engine
does not involve a significant reduction in the margin of safety,
but rather enhances safety.
Proposed Revision 3:
This proposed revision deletes the requirement to reduce reactor
power level to 25% of rated power upon the loss of either both
diesel generators or both incoming power sources. The elimination of
this requirement will allow the plant to maintain the existing power
level rather than subject the plant to an unnecessary transient.
Maintaining the plant at the existing power level provides a more
stable operating environment. The equipment and components of the
diesel generators or the incoming power sources are not impacted in
any way as a result of the proposed revisions. The margin of safety
for the diesel generators and the incoming power sources are not
significantly reduced since these systems are not altered in any
way, and will continue to be surveillance tested as required.
Assurance of operability is provided by the normal, scheduled
surveillances which have been established at a sufficient interval
to provide reasonable assurance of operability. Therefore, the
proposed changes do not involve a significant reduction in the
margin of safety.
The proposed rearrangement of information, and rewording of some
[of] the TS requirements are included to enhance usability and
alleviate any possible confusion. These changes are strictly
editorial have no impact, and do not alter technical content or
meaning of the specifications. These editorial changes do not
involve a significant reduction in the margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. The licensee's July 14, 1995, letter revised the proposed
changes in their letter of May 5, 1995, to further limit the period of
time that continued reactor operation would be allowed with both
emergency diesel generators inoperable from 24 to two hours. This
revision to the proposed changes is more restrictive and does not
impact the licensee's analysis of the criteria of 10 CFR 50.92(c).
Therefore, the NRC staff proposes to determine that the amendment
request involves no significant hazards consideration.
Local Public Document Room location: Auburn Public Library, 118
15th Street, Auburn, Nebraska 68305
Attorney for licensee: Mr. John R. McPhail, Nebraska Public Power
District, Post Office Box 499, Columbus, Nebraska 68602-0499
NRC Project Director: William D. Beckner
Northeast Nuclear Energy Company (NNECO), Docket No. 50-245,
Millstone Nuclear Power Station, Unit 1, New London County,
Connecticut
Date of amendment request: August 31, 1995
Description of amendment request: The proposed amendment modifies
the definition of HOT SHUTDOWN and COLD SHUTDOWN to specify that the
definitions are not applicable during the performance of an inservice
hydrostatic and leak test (IHLT). Technical Specification Section 3.6.B
and 4.6.B would be modified by adding Section 3.6.B.1.b and 4.6.B.1.b
to identify the requirements that must be satisfied to consider the
reactor in COLD SHUTDOWN during the performance of an IHLT. In
addition, the proposed amendment will change temperature specific
requirements on several pages to mode or condition specific
requirements; make several editorial changes; and change the associated
Bases.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has rovided
its analysis of the issue of no significant hazards consideration,
which is presented below:
NNECO has reviewed the proposed changes in accordance with
10CFR50.92 and concluded that the changes do not involve a
significant hazards consideration (SHC). The bases for this
conclusion is that the three criteria of 10CFR50.92(c) are not
compromised. The proposed changes do not involve an SHC because the
changes would not:
[[Page 49941]]
1. Involve a significant increase in the probability or
consequences of an accident previously analyzed.
The proposed changes will allow the reactor to be considered in
COLD SHUTDOWN during an IHLT with the average reactor coolant
temperature greater than 212 deg.F but less than 280 deg.F. The
change to allow the reactor to be in COLD SHUTDOWN during the
performance of IHLT will not increase the probability or
consequences of an accident. The probability of a leak in the
reactor pressure boundary during this testing is not increased by
considering the reactor to be in COLD SHUTDOWN. The IHLT is
performed near water solid, all control rods inserted, and with an
appropriate availability of engineering safety features. The stored
energy in the reactor core will be very low and the potential for
failed fuel and a subsequent increase in coolant activity are
minimal. In addition, secondary containment will be operable and
capable of handling airborne radioactivity from leaks that could
occur during the performance of an IHLT. Requiring secondary
containment to be operable will further ensure that potential
airborne radiation from leaks will be filtered by one or both trains
of SBGT [standby gas treatment], thereby limiting releases to the
environment. Therefore, the changes will not significantly increase
the consequences of an accident.
In the unlikely event of a large pressure boundary leak, the
reactor vessel would rapidly depressurize, allowing one or both of
the operable core spray systems to operate. Small system leaks would
be detected by leakage inspections before significant inventory loss
occurred, since leakage inspections are an integral part of the IHLT
program.
Based upon the above, the proposed changes do 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 analyzed.
The IHLT conditions remain unchanged. The potential for a system
leak remains unchanged since the reactor coolant system is designed
for temperatures exceeding 500 deg.F with similar pressures. The
change in operable engineered safety features available to mitigate
a postulated accident does not reduce the ability to
safely mitigate a postulated accident. Adequate ECCS [emergency
core cooling system] equipment will be available to mitigate a LOCA
[loss of coolant accident] with an assumed single failure.
Therefore, this will not create the possibility of a new or
different kind of accident from any previously evaluated.
3. Will not involve a significant reduction in a margin of
safety.
The proposed changes will not have any significant impact on any
design basis accident or safety limit. The various engineered safety
features which are required by the proposed change will ensure
appropriate mitigation of postulated events. Since the test is
performed at a near water solid condition and at low decay heat
values, no fuel damage is expected in case of an accident such as a
LOCA. Nevertheless, secondary containment and the SBGT system will
be maintained operable to process air-borne radioactivity from a
steam leak that could occur during the performance of the IHLT.
Therefore, the proposed change does not constitute 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: Learning Resources Center,
Three Rivers Community-Technical College, 574 New London Turnpike,
Norwich, CT 06360.
Attorney for licensee: Lillian M. Cuoco, Esq., Senior Nuclear
Counsel, Northeast Utilities Service Company, P.O. Box 270, Hartford,
CT 06141-0270.
NRC Project Director: Phillip F. McKee
Northeast Nuclear Energy Company, et al., Docket No. 50-336,
Millstone Nuclear Power Station, Unit No. 2, New London,
Connecticut
Date of amendment request: August 31, 1995
Description of amendment request: The proposed change to the
Millstone 2 Technical Specifications would remove the phrase ``other
than Millstone Unit No. 2'' from Section 6.3.1 on page 6-2.
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:
NNECO has reviewed the proposed change in accordance with
10CFR50.92 and concluded that the change does not involve a
significant hazards consideration (SHC). The basis for this
conclusion is that the three criteria of 10CFR50.92(c) are not
compromised. The proposed change does not involve an SHC because the
change would not:
1. Involve a significant increase in the probability or
consequences of an accident previously analyzed.
The proposed change does not affect any system or equipment of
Millstone Unit No. 2. The proposed change does not affect the
qualification of any of the licensed individuals involved in the
day-to-day operation of Millstone Unit No. 2. The proposed change
corrects a statement which could be interpreted such that an
individual who once held a Millstone Unit No. 2 SRO [Senior Reactor
Operator] license would not be eligible to be Operations Manager.
Since this change does not affect any equipment or operating
procedures, does not affect the level of expertise and
training required for on-shift personnel, and does not reduce
the level of expertise required of operations management, this
change does not involve a significant increase in the probability or
consequences of an accident previously analyzed.
2. Create the possibility of a new or different kind of accident
from any previously analyzed.
This change does not affect any equipment or operating
procedures, does not affect the level of expertise and training
required for on-shift personnel, and does not reduce the level of
expertise required of operations management. Therefore, this change
does not create the possibility of a new or different kind of
accident.
3. Involve a significant reduction in the margin of safety.
This change eliminates a phrase which could be interpreted to
prevent an individual who had possessed a Millstone Unit No. 2 SRO
license from becoming the Operations Manager. The training and
experience necessary to possess a Millstone Unit No. 2 SRO license
is equivalent to that of other PWRs. Therefore, this proposed change
does not involve a significant reduction in the margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: Learning Resources Center,
Three Rivers Community-Technical College, 574 New London Turnpike,
Norwich, CT 06360.
Attorney for licensee: Lillian M. Cuoco, Esq., Senior Nuclear
Counsel, Northeast Utilities Service Company, P.O. Box 270, Hartford,
CT 06141-0270.
NRC Project Director: Phillip F. McKee
Philadelphia Electric Company, Docket Nos. 50-352 and 50-353,
Limerick Generating Station, Units 1 and 2, Montgomery County,
Pennsylvania
Date of amendment request: July 28, 1995
Description of amendment request: The proposed amendment would
eliminate the Technical Specifications requirements to perform 10 CFR
50, Appendix J, Type C hydrostatic testing on certain valves that are
within closed systems and are assured a water seal following a Design
Basis Accident.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. The proposed TS changes do not involve a significant increase
in the probability or
[[Page 49942]]
consequences of an accident previously evaluated.
The primary containment (drywell and suppression pool) and the
affected closed systems are accident mitigators not accident
initiators. The proposed change to the scope of Appendix J, Type C
testing does not affect the probability of the DBA [Design Basis
Accident]. The valves will continue to be maintained in an operable
state, and in their current design configuration. There is no
correlation between the scope of Appendix J, Type C testing and
accident probability. There are no physical or operational changes
to the containment structure, system or components being made as a
result of the proposed changes. Therefore, the consequences of a
malfunction of equipment important to safety is not increased from
those previously evaluated.
The consequences of loss-of-coolant accidents (LOCAs) under the
proposed change were considered where a single active failure of a
containment isolation valve (CIV) or a passive failure of the closed
system were reviewed, within the limits of the existing licensing
basis. Under the existing licensing basis, a pipe rupture of the
seismically qualified ECCS piping does not have to be assumed
concurrent with the LOCA, except if it is a consequence of the LOCA.
Consideration of consequential failures can be eliminated, since a
LOCA inside containment is separated from the affected piping by the
containment structure. Consideration of consequential failures of
the ECCS piping from LOCAs outside containment are outside the
Appendix J design considerations. A single active failure of the
CIV, under the LOCA condition, can be accommodated since the closed
and water sealed system piping remains as the leakage barrier. The
ECCS passive failure criterion does require consideration of system
leaks, but not pipe breaks, beyond the initiating LOCA. The
capability to make-up water inventory to the suppression pool is
adequate to ensure that postulated seat leakage and pipe leakage
does not result in a condition that jeopardizes pool level. Make-up
capability exists for the suppression pool via the Condensate
Storage Tank and Ultimate Heat Sink Spray Pond. Operator actions to
make-up the suppression pool are delineated in existing Operating
Procedures.
The subject valves are single isolation valves associated with
lines that penetrate the primary containment, but are not connected
directly to the primary containment atmosphere or the reactor
coolant pressure boundary. This configuration is described in the
LGS UFSAR, Section 6.2.4.3.1.3.1, which states ``the systems which
the lines from the suppression pool connect to outside containment
are closed systems meeting the appropriate requirements of closed
systems.'' The integrity of these closed systems are also monitored
and controlled in accordance with TS Section 6.8.4.a. Any leakage
that may escape the confines of the closed system will be contained
within the Reactor Building, treated by standby gas and radwaste
systems, and, therefore, are within the existing LGS licensing
bases.
Finally, the affected penetrations will continue to be subjected
to the periodic 10 CFR 50, Appendix J, Type A test (Integrated
Containment Leakage Rate Test).
The suppression pool level is designed and operated so that
water level is maintained in accordance with current TS, and the
associated bases. The supply of water in the suppression pool is
assured for 30 days during all DBA, post-accident modes of
operation. The lowest water level which the suppression pool will
reach was analyzed, and it was determined that the affected lines
will remain below this minimum level, thereby assuring a water seal.
The valves will continue to be tested and maintained to ensure their
operability, and the closed systems' integrity will continue to be
monitored and controlled in accordance with TS 6.8.4.a and the
performance of the periodic 10 CFR 50, Appendix J, Type A test.
Therefore, the proposed changes will not increase the
probability or consequences of an accident previously evaluated.
2. The proposed TS changes do not create the possibility of a
new or different kind of accident from any accident previously
evaluated.
The proposed changes do not change the plant response to
accident scenarios, and do not introduce new or different scenarios.
The primary containment (drywell and suppression pool) and the
affected closed systems are accident mitigators not accident
initiators. The proposed change to the scope of Appendix J, Type C
hydrostatic testing maintains the existing barriers to primary
containment bypass leakage by the assurance that a water seal will
be maintained for 30 days during all DBA, post-accident modes of
operation. The valves will continue to be tested and maintained to
ensure their operability, and the closed systems' integrity will
continue to be monitored and controlled in accordance with TS
6.8.4.a. Therefore, the proposed changes cannot cause an accident,
and the plant response to the design basis events is unchanged,
whereby the change does not create the possibility of a new or
different kind of accident from any accident previously evaluated.
3. The proposed TS changes do not involve a significant
reduction in a margin of safety.
The water seal provided by the assurance of a minimum
suppression pool level will prevent post-accident containment bypass
leakage. Appendix J does not require air leak testing of the valves
since the 30 day post-accident supply of water is maintained. In
addition, the closed systems' integrity is monitored and controlled
in accordance with TS 6.8.4.a. Any leakage that may escape the
confines of the closed system will be contained within the Reactor
Building, and is within the existing LGS licensing bases. Therefore,
the proposed TS changes do not involve a significant reduction in a
margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 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: Pottstown Public Library, 500
High Street, Pottstown, Pennsylvania 19464.
Attorney for licensee: J. W. Durham, Sr., Esquire, Sr. V. P. and
General Counsel, Philadelphia Electric Company, 2301 Market Street,
Philadelphia, Pennsylvania 19101
NRC Project Director: John F. Stolz
Philadelphia Electric Company, Docket Nos. 50-352 and 50-353,
Limerick Generating Station, Units 1 and 2, Montgomery County,
Pennsylvania
Date of amendment request: July 28, 1995
Description of amendment request: The proposed amendments, which
are consistent with the Improved Standard Technical Specifications
(NUREG-1433), delete the operability and surveillance requirements
involving secondary containment differential pressure instrumentation.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. The proposed Technical Specifications (TS) changes do not
involve a significant increase in the probability or consequences of
an accident previously evaluated.
Deleting the operability and surveillance requirements for the
secondary containment differential pressure instrumentation does not
involve any changes to the design, function, or operation of any
plant components or safety-related systems. There are no changes to
the separation, redundancy, qualification, quality assurance or fire
protection requirements for the associated components and systems,
nor are there any new failure modes created. This activity only
removes operability and surveillance requirements from the Technical
Specifications for selected plant components associated with the
secondary containment differential pressure trip functions. No
credit for operation of these trip functions is taken in any design
basis accidents valuated in the SAR [Safety Analysis Report].
Thesecomponents will be maintained in accordance with the plant
preventive maintenance program. The failure of any of these
components does not result in the occurrence of an accident.
Consequently, there is no increase in the probability of occurrence
of an accident previously evaluated in the SAR.
The Outside Atmosphere to Reactor Enclosure Delta Pressure-Low
and Outside Atmosphere To Refueling Area Delta Pressure-Low trip
functions are not symptomatic of a design basis accident. No credit
for operation of the trip functions is taken in any design basis
accidents evaluated in the SAR. Neither failure of the differential
pressure components nor failure to generate the associated trip
functions affects the consequences of an accident previously
evaluated in the SAR. The appropriate
[[Page 49943]]
accident prevention and mitigation actions are generated from other
plant parameters symptomatic of an accident. Sufficient plant
parameters symptomatic of a design basis accident are monitored to
initiate the appropriate actions as evaluated in the SAR.
Furthermore, all safety-related systems will still be able to
perform all of their design basis safety-related functions.
Consequently, there is no increase in the consequences of an
accident previously evaluated in the SAR.
Therefore, the proposed TS changes do not involve an increase in
the probability or consequences of an accident previously evaluated.
2. The proposed TS changes do not create the possibility of a
new or different kind of accident from any accident previously
evaluated.
The failure of the differential pressure automatic isolation
instrumentation components does not result in the occurrence of an
accident. The failure to generate the associated trip functions does
not result in the occurrence of an accident. This activity does not
involve any changes to the design, function, or operation of any
plant components or safety-related systems. There are no changes to
the separation, redundancy, qualification, quality assurance or fire
protection requirements for the associated components and systems.
These components will be maintained in accordance with the plant
preventative maintenance program. Consequently, there is no
possibility of an accident of a different type than previously
evaluated in the SAR.
Therefore, the proposed TS changes do not create the possibility
of a new or different kind of accident from any accident previously
evaluated.
3. The proposed TS changes do not involve a significant
reduction in a margin of safety.
The ability of secondary containment to minimize any ground
level release of radioactive material which may result from any
accident is not affected. Surveillance and operability requirements
for secondary containment SGTS [Standby Gas Treatment System] and
RERS [Reactor Enclosure Recirculation System] are not changed by
this activity. Draw down time, leakage factors, secondary
containment system ratings, and secondary containment system
response to a LOCA [Loss of Coolant Accident] or refueling accident
are not affected by this activity. SGTS and RERS initiation will
continue to occur when plant parameters symptomatic of a LOCA or
refueling accident exceed predetermined values. There are no changes
to the inputs for the post-LOCA offsite dose analysis.
Therefore, the proposed TS changes do not involve a significant
reduction in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 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: Pottstown Public Library, 500
High Street, Pottstown, Pennsylvania 19464.
Attorney for licensee: J. W. Durham, Sr., Esquire, Sr. V. P. and
General Counsel, Philadelphia Electric Company, 2301 Market Street,
Philadelphia, Pennsylvania 19101
NRC Project Director: John F. Stolz
Philadelphia Electric Company, Docket Nos. 50-352 and 50-353,
Limerick Generating Station, Units 1 and 2, Montgomery County,
Pennsylvania
Date of amendment request: July 28, 1995
Description of amendment request: The proposed amendment would
modify Technical Specifications (TS) Surveillance Requirements 4.9.1.1,
4.9.1.2, 4.9.3, 4.9.5, and 4.9.8 to delete specific requirements to
perform surveillances just prior to beginning or resuming core
alterations or control rod withdrawal associated with refueling
activities. This proposed TS change would delete the phrase ``incore
instrumentation'' from the footnote in TS Section 3/4.9.5,
``Communications.''
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 Technical Specifications (TS) changes do not
involve a significant increase in the probability or consequences of
an accident previously evaluated.
The proposed TS changes do not involve any physical changes to
plant systems or equipment. The proposed TS changes only delete
those Surveillance Requirements (SRs) pertaining to the performance
of tests just prior to beginning or resuming core alterations or
control rod withdrawal, and revises a footnote description to be
consistent with the current TS definition of ``Core Alteration.''
The proposed TS changes do not revise any of the other applicable
periodic SRs, or modify any procedural controls currently in place
governing fuel handling operations. The periodic surveillance test
frequencies provide adequate assurance that the equipment will
remain in an operable condition. The normal periodic surveillance
intervals bound those surveillance intervals for the tests that are
being altered by this proposed TS change. In the event that one of
the periodic surveillances has not been performed within the
specified time interval, entry into the specified condition (i.e.,
performance of core alterations, control rod withdrawal, or handling
of fuel or control rods) is not permitted as required by TS 4.0.4
until the surveillance has been satisfactorily completed.
The consequences of an accident are not increased by the
proposed TS changes, since the changes only involve revising the
frequency of conducting surveillance tests. The method of operation
or performance of plant structures, systems, or plant components are
not affected by the proposed TS changes. The proposed TS changes
will not impact the operation of any fuel handling equipment, and
therefore, the potential for a Fuel Handling Accident as described
in Section 15.7.4 of the LGS [Limerick Generating Station] Updated
Final Safety Analysis Report (UFSAR) is not increased.
In addition, any unexpected reduction of water level in the
reactor cavity or fuel pool at the start of fuel handling or control
rod handling will be immediately apparent to operators by direct
observation. Plant procedures utilized by the refueling personnel
require the suspension of core component transfers in the event of
loss of water inventory.
Therefore, the proposed TS change does not involve an increase
in the probability or consequences of an accident previously
evaluated.
2. The proposed TS changes do not create the possibility of a
new or different kind of accident from any accident previously
evaluated.
The proposed TS changes only involve changes to the frequency in
which the specified surveillance tests are performed. The proposed
TS changes do not revise any of the other applicable periodic SRs,
or modify any procedural controls currently in place governing fuel
handling operations. The periodic surveillance test frequencies
provide adequate assurance that the equipment will remain in
operable condition. The periodic surveillance intervals bound those
surveillance intervals for the tests that are being altered by this
proposed TS change. The refueling interlock system combined with
strict procedural controls provide multiple barriers to preclude an
inadvertent criticality.
The proposed TS changes do not involve any physical changes to
plant systems or equipment. The proposed TS changes do not alter the
configuration of the plant or the way that the plant is operated.
The associated plant equipment will continue to function as
designed. This equipment is not designed to perform any other
function than it is presently capable of, and therefore, will not
affect the operation of any other plant equipment.
Therefore, the proposed TS change does not create the
possibility of a new or different kind of accident from any
previously evaluated.
3. The proposed TS changes do not involve a significant
reduction in a margin of safety.
The proposed TS changes do not involve any physical changes to
plant systems or equipment. The reactor will continue to be
maintained subcritical during refueling operations and reactor water
level will be maintained at the required level (i.e., above the
vessel flange). The proposed TS changes do not affect the operation
of other plant systems and equipment essential in maintaining
reactor water temperature during refueling operations, or the
capability in responding to a postulated Fuel Handling Accident.
The proposed changes do not adversely affect reliability of the
refueling interlocks or refuel platform communications equipment.
[[Page 49944]]
Since the proposed changes only impact the frequency in which certain
surveillance tests are performed, and do not change the plant
configuration or setpoints, there is substantial assurance that the
reactor will be maintained subcritical during refueling.
Therefore, the proposed TS changes do not involve a reduction in
a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: Pottstown Public Library, 500
High Street, Pottstown, Pennsylvania 19464.
Attorney for licensee: J. W. Durham, Sr., Esquire, Sr. V. P. and
General Counsel, Philadelphia Electric Company, 2301 Market Street,
Philadelphia, Pennsylvania 19101
NRC Project Director: John F. Stolz
Philadelphia Electric Company, Docket Nos. 50-352 and 50-353,
Limerick Generating Station, Units 1 and 2, Montgomery County,
Pennsylvania
Date of amendment request: July 28, 1995
Description of amendment request: The proposed amendment would
revise Technical Specifications Table 4.3.1.1-1, ``Reactor Protection
System Instrumentation Surveillance Requirements'', to reflect changes
to the surveillance test frequency requirements for various Reactor
Protection System [RPS] instrumentation.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. The proposed Technical Specifications (TS) changes do not
involve a significant increase in the probability or consequences of
an accident previously evaluated.
In all of the applicable SAR [Safety Analysis Report] evaluated
events, the IRM [Intermediate Range Monitor] and APRM [Average Range
Power Monitor] instrumentation is credited for performing a
mitigating function (i.e., initiating a scram), to terminate the
transient prior to a safety limit being exceeded. The proposed TS
changes do not alter the RPS configuration, or RPS instrumentation
setpoints, nor do they change the manner in which the IRM and APRM
instrumentation carry out the scram functions. Therefore the
consequences of any potential malfunction of equipment important to
safety will remain unchanged.
In each case where a startup surveillance test requirement is
proposed to be deleted, (i.e., IRM and APRM), the normal
surveillance test frequency specified for the required Operational
Condition remains unchanged (except for the APRM Upscale Setdown
functional test). The startup surveillance requirement is
conservatively bounded by the normal surveillance test interval
which is greater than or equal to any interval associated with the
startup surveillance requirement and ensures that the IRM and APRM
instrumentation reliability is unchanged. This is in accordance with
the Improved Standard Technical Specifications, NUREG-1433, issued
September 28, 1992.
The reliability of the APRM Upscale Setdown scram function will
not be decreased due to changing the functional test frequency from
Weekly (W), to Quarterly (Q), in Operational Conditions 2, 3, and 5
(Startup, Hot Shutdown and Refueling, respectively). Plant
operational data taken from each of the APRM calibration/functional
tests performed since August 1992 until present at LGS Units 1 and
2, shows that setpoint reliability will be maintained if the
functional test frequency is increased to quarterly, as proposed.
Presently, each time an APRM calibration/functional test is
performed, both the Upscale Setdown and the Flow Reference scram
circuits are tested. The results of the quarterly tests confirm that
the APRM Upscale Setdown function already has over 2.5 years of
performance without failure in Operational Condition 1, thus being
extremely reliable.
Therefore, the proposed TS changes do not involve an increase in
the probability or consequences of an accident previously evaluated.
2. The proposed TS changes do not create the possibility of a
new or different kind of accident from any accident previously
evaluated.
The proposed TS changes affect only the required surveillance
test intervals, not the RPS configuration or RPS instrumentation
setpoints. The proposed TS changes do not introduce a new failure
mode for the IRM or APRM instrumentation. Plant operating experience
data confirms that at LGS Units 1 and 2, the IRM and APRM
instrumentation will continue to perform their safety function as
currently designed, with the same degree of reliability.
The proposed TS changes do not alter the configuration of the
plant, nor the way the plant is operated.
Therefore, the proposed TS changes do not create the possibility
of a new or different kind of accident, from any accident previously
evaluated.
3. The proposed TS changes do not involve a significant
reduction in a margin of safety.
The following TS Bases were reviewed for potential reduction in
the margin of safety:
B 2.2.1 Reactor Protection System Instrumentation Setpoints
B 3/4.1.4 Control Rod Program Controls
B 3/4.2 Power Distribution Limits
B 3/4.3.1 Reactor Protection System Instrumentation
B 3/4.3.6 Control Rod Block Instrumentation
The surveillance test frequency changes proposed for the RPS
instrumentation section of TS do not adversely affect the IRM or
APRM instrumentation, which will continue to perform the RPS
functions required to maintain the present margin of safety. Changes
to the IRM instrumentation startup surveillance intervals are
already bounded by the existing surveillance requirements, and are
in accordance with the Improved Standard Technical Specifications,
NUREG-1433, issued September 28, 1992. The same statement applies to
the APRM instrumentation, with respect to deletion of the startup
surveillance requirement. The change of the APRM Upscale Setdown
Channel functional test surveillance interval from Weekly to
Quarterly was evaluated to ensure that the APRM instrumentation
would perform that function, with the same degree of reliability as
presently experienced. A review of the plant operating experience
data at LGS Units 1 and 2 shows that APRM instrumentation is
extremely reliable for a quarterly surveillance test interval. The
proposed TS changes do not modify plant configuration, RPS
instrumentation setpoints, or RPS operation. The margin of safety
remains unchanged.
Therefore, the proposed TS changes do not involve a reduction in
a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: Pottstown Public Library, 500
High Street, Pottstown, Pennsylvania 19464.
Attorney for licensee: J. W. Durham, Sr., Esquire, Sr. V. P. and
General Counsel, Philadelphia Electric Company, 2301 Market Street,
Philadelphia, Pennsylvania 19101
NRC Project Director: John F. Stolz
Philadelphia Electric Company, Docket Nos. 50-352 and 50-353,
Limerick Generating Station, Units 1 and 2, Montgomery County,
Pennsylvania
Date of amendment request: August 1, 1995
Description of amendment request: The proposed amendment would
modify Technical Specifications Section 3/4.9.1, ``Reactor Mode
Switch,'' in order to provide alternate actions to allow the
continuation of core alterations in the event certain Reactor Manual
Control System (RMCS) and refueling interlocks are inoperable, while
preserving the intended function of the inoperable interlocks.
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 Technical Specifications (TS) changes do not
involve a significant
[[Page 49945]]
increase in the probability or consequences of an accident previously
evaluated.
The refueling and one-rod-out interlocks impose barriers to
preclude an inadvertent criticality during refueling operations.
Section 7.7.2.15.1 of the LGS Updated Final Safety Analysis Report
(UFSAR) clearly delineates the functions of the interlocks and the
criteria used in assessing correct refueling and one-rod-out
interlock operation in the following statement.
In all cases, correct operation of the refueling interlock
prevents either the operation of loaded refueling equipment over the
core when any control rod is withdrawn, or the withdrawal of any
control rod when fuel-loaded refueling equipment is operating over
the core. In addition, when the reactor mode switch is in REFUEL
position, only one rod can be withdrawn, and selection of a second
control rod initiates a rod block.
The proposed TS changes provide operational flexibility while
strictly conforming to, and preserving, the intended function of the
refueling and one-rod-out interlocks. The proposed TS changes that
could affect interlock capabilities are identified below, along with
the appropriate justification to substantiate that the proposed TS
changes will not result in an increase in the probability or
consequences of an accident previously evaluated.
a.TS Section 3.9.1, ACTION Statement b. The proposed change to
this existing TS ACTION will add a verification that all control
rods are fully inserted, and then disabled from being withdrawn as a
suitable alternative to placing the reactor mode switch in the
SHUTDOWN position when the one-rod-out interlock is not operable. In
addition, the proposed change to this TS section includes a caveat
of non-applicability for those control rods already removed in
accordance with requirements stipulated in TS Sections 3.9.10.1 and
3.9.10.2. As indicated in LGS UFSAR which is described in the
statement above, it is expected that the refuel and one-rod-out
interlocks will permit the withdrawal of only one (1) control rod at
a time with the reactor mode switch in the REFUEL position, and no
control rods can be moved when fuel-loaded refueling equipment is
operating over the core. By verifying all control rods are inserted,
then disabling withdraw capabilities of all rods, as requested, the
most limiting requirements for control rod motion will be met. The
potential for having more than one (1) control rod out at a time, or
having any control rod not fully inserted while fuel-loaded
refueling equipment is operating over the core, does not exist when
applying the alternative. Therefore, the intended functions of the
refuel and one-rod-out interlocks are operationally preserved. Since
TS Sections 3.9.10.1 and 3.9.10.2 have specific requirements for
removing surrounding fuel prior to control rod blade removal, the
control rods already removed are no longer required to carry out a
safety function in the defueled cell, and as a result would not
apply for this specific proposed TS change. From a control rod
withdrawal perspective, there is no functional difference between
the proposed TS change and the existing, and still remaining, TS
ACTION of locking the reactor mode switch in SHUTDOWN position.
b. TS Section 3.9.1, ACTION Statement c. This existing TS ACTION
requires that core alterations be suspended in the event that a
refueling interlock is not operable. The proposed TS change to this
TS ACTION leaves this requirement in place, but makes this ACTION
specifically applicable to the refueling platform, and adds three
(3) new additional ACTION alternatives. The wording for changes to
this TS section are such that implementation of any one of the three
(3) new alternatives can be substituted for suspending core
alterations. The proposed wording for these three (3) new
alternatives and justification is provided below.
1) Verify all control rods are fully inserted and disable
withdraw capabilities of all control rods***.
Since this alternative ensures all control rods are, and will
remain fully inserted, all required conditions of the associated
refueling and one-rod-out interlocks are met. The refueling
interlock is satisfied since a fuel-loaded refueling platform
operating over the core would be assured that all control rods are
fully inserted and prevented from being withdrawn. The one-rod-out
interlock is satisfied since control rod withdrawal is disabled for
all control rods, which is an even more conservative requirement
than the one-rod-out interlock itself. While operating in this
configuration, there will be no associated travel or hoist
restrictions for the refueling platform over the core, which is
normal for the current refuel interlock design. The potential for
having any control rod not fully inserted while a fuel-loaded
refueling platform is operating over the core, does not exist when
applying this proposed alternative. Therefore, the intended function
of the refueling platform refuel interlocks are operationally
preserved with the implementation of this proposed alternative, and
there will be no increase in the probability of occurrence of an
accident. This proposed alternative also maintains an exclusion (via
a reference to the proposed *** footnote) for control rods removed
in accordance [with] TS Sections 3.9.10.1 and 3.9.10.2. This
exclusion does not apply to inadvertent criticality concerns, as
previously discussed in Item 1.a above.
2) Verify Refuel Platform is not over core (limit switches not
reached) and disable refuel platform travel over core.
As previously stated above, LGS UFSAR Section 7.7.2.15.1
stipulates that the refueling platform position interlocks initiate
a control rod block whenever a fuel-loaded refueling platform is
over the core, and stop a fuel-loaded refueling platform from moving
over the core if a control rod is already withdrawn. This specific
proposed TS change satisfies both these requirements by precluding
the possibility of the platform from being over the core. If a
control rod is being withdrawn, the platform will not be over the
core, and the withdrawal will be in accordance with the current
design. If a control rod is already withdrawn, disabling platform
travel over the core, before reaching the over-core limit switches,
is performing the same function as the existing refueling platform
reverse and forward motion blocks. Therefore, the potential for
having any control rod not fully inserted while a fuel-loaded
refueling platform is operating over the core, does not exist when
applying this proposed alternative. The intended refueling interlock
functions are operationally preserved with the implementation of
this proposed alternative.
3) Verify that no Refuel Platform hoist is loaded and disable
all Refuel Platform hoists from picking up (grappling) a load.
As previously stated above, UFSAR Section 7.7.2.15.1 stipulates
that blocking control rod withdrawal with a refueling platform over
the core, and restricting refueling platform travel from going over
the core with a control rod already withdrawn, are based on the
refueling platform hoist being fuel-loaded. An unloaded platform
without grappling capabilities poses no threat to erroneous fuel
bundle or control rod removal, and eliminates the potential for
having any control rod not fully inserted while a fuel-loaded
refueling platform is operating over the core. Therefore,
implementing this proposed alternative operationally preserves the
intended interlock functions.
c. TS Section 3.9.1, ACTION Statement d. The proposed TS change
adds this new TS ACTION section to specify the refueling interlock
requirements for the service platform, since the applicability of
ACTION Statement c above is being revised to specifically address
refueling interlocks associated with the refueling platform. The
proposed TS changes for new this TS section retain the existing
requirement to suspend core alterations if the service platform
associated refueling interlock is not operable, unless the service
platform is not installed over vessel. The specific proposed TS
changes add two (2) new additional ACTION alternatives. The proposed
wording for these two (2) new ACTION statements are such that
implementation of any one of the two (2) new alternatives can be
substituted for suspending core alterations. Not enforcing
operability requirements on the service platform refueling
interlocks when the service platform is not over the vessel does not
pose an inadvertent criticality concern since there is no associated
hoist to manipulate fuel bundles or control rods. These two (2) new
alternatives are not applicable unless the service platform is
installed over the vessel, and are described below.
1) Verify all control rods are fully inserted and disable
withdraw capabilities of all control rods***.
This alternative ensures that all control rods are, and will
remain, fully inserted which meets the required conditions for
proper refueling and one-rod-out interlock operation. The refueling
interlock is satisfied since a fuel-loaded service platform hoist
operating over-core is assured that all control rods are fully
inserted and prevented from being withdrawn. The one-rod-out
interlock is satisfied since all control rods are disabled, an even
more conservative requirement than the one-rod-out interlock itself.
While operating in this configuration, there will be no associated
hoist restrictions for the service
[[Page 49946]]
platform, which is normal for the current refuel interlock design. The
potential for having any control rod not fully inserted while a
service platform hoist is fuel-loaded over the core, does not exist
when utilizing this proposed alternative. Therefore, the intended
function of the service platform refuel interlocks are operationally
preserved with the implementation of this proposed alternative. This
proposed alternative also maintains an exclusion (via a reference to
the proposed *** footnote) for control rods removed in accordance
with the requirements of TS Sections 3.9.10.1 and 3.9.10.2. This
exclusion is not applicable to inadvertent criticality concerns as
discussed in Item 1.a above.
2) Verify Service Platform hoist is not loaded and disable
Service Platform hoist from picking up (grappling) a load.
As previously described above, UFSAR Section 7.7.2.15.1
stipulates that blocking control rod withdrawal with the service
platform over the core is based on the service platform hoist being
fuel-loaded. An unloaded hoist without grappling capabilities poses
no threat to erroneous fuel bundle or control rod removal and
eliminates the potential for having any control rod not fully
inserted while a fuel-loaded service platform is operating over the
core. Therefore, implementing this proposed alternative
operationally preserves the intended refueling interlock functions.
As discussed in the LGS UFSAR, the use of the refueling and one-
rod-out interlocks are evaluated from a prevention, not a
mitigation, perspective. A Rod Withdrawal Error (RWE) transient
event during refueling is concerned with an inadvertent criticality,
and assumes the reactor vessel head is off, and the plant is
shutdown (i.e, Operating State A). As described in the LGS UFSAR
under Nuclear Safety Operational Analysis (NSOA) Event 16, it is
assumed that the Reactor Protection System (RPS) terminates the
event should the reactor actually reach Operating State B (i.e.,
head off and not shut down), which is conditional on the reactor
mode switch being in the STARTUP position. The proposed TS changes
only pertain to the refueling and one-rod-out interlocks. Since
these interlocks act only in a preventive mode, the consequences of
an inadvertent criticality accident during refueling remain
unchanged.
Since the proposed TS changes are limited to the one-rod-out and
refueling interlocks, they do not affect the reliability of the
associated equipment. The proposed TS changes specify alternative
actions that can be taken in the event that an interlock is
inoperable. These alternative actions serve to ensure the failed
interlock function is preserved, and do not affect the probability
of malfunction of the interlocks.
The one-rod-out and refueling interlocks, as evaluated in the
LGS UFSAR, are designed to preclude an inadvertent criticality
during refueling operations by placing strict controls on fuel
bundle and control rod manipulations, using the following methods.
a. Preventing operation of a fuel-loaded refueling platform or
service platform hoist while over the core if a control rod is
already withdrawn.
b. Preventing a fuel-loaded refueling platform from traveling
over the core if a control rod is already withdrawn.
c. Preventing any control rod from being withdrawn if a fuel-
loaded refueling platform or service platform is already operating
over the core.
d. Preventing the withdrawal of more than one control rod at a
time with the reactor mode switch in the REFUEL position.
The LGS UFSAR indicates that a single component failure does not
cause an interlock failure and that a single interlock failure does
not cause an accident. The proposed TS changes provide alternative
actions that can be taken in the event of an associated component or
interlock malfunction. Implementing the proposed TS changes will
continue to ensure that the intended interlock functions are
maintained and operationally preserved, as described in the LGS
UFSAR.
Therefore, the proposed TS changes do not involve an increase in
the probability or consequences of an accident previously evaluated.
2. The proposed TS changes do not create the possibility of a
new or different kind of accident from any accident previously
evaluated.
The proposed TS changes only pertain to the refueling and one-
rod-out interlocks. The refueling and one-rod-out interlocks impose
barriers to preclude an inadvertent criticality during refueling
operations. The proposed TS changes provide operational flexibility,
while strictly conforming to, and preserving, the intended function
of the refueling and one-rod-out interlocks. There is no other
potential failure mode for these interlocks than has already been
evaluated and described in the LGS UFSAR. Implementation of these
proposed changes will maintain and operationally preserve the
intended interlock functions. Therefore, the malfunction of any
associated component or interlock will not adversely impact the
plant and any other equipment important to safety, directly or
indirectly.
Therefore, the proposed TS changes do not create the possibility
of a new or different kind of accident from any previously
evaluated.
3. The proposed TS changes do not involve a significant
reduction in a margin of safety.
The proposed TS changes only affect the TS associated with the
one-rod-out and refueling interlocks. The associated TS Bases
Section 3/4.9, ``Refueling Operations,'' states that the one-rod-out
and refueling interlocks maintain conditions during refueling
activities that reinforce refueling procedures and reduce the
potential for the probability of occurrence of each of the following
conditions:
a. Inadvertent criticality,
b. Damage to reactor internals or fuel assemblies, and
c. Exposure of personnel to excessive radioactivity.
The proposed TS changes do not adversely affect the one-rod-out
or refueling interlocks. The associated interlocks will continue to
perform the refueling functions required to maintain the present
margin of safety. The proposed TS changes only contain alternative
actions that can be taken in the event an interlock is inoperable.
These proposed alternative actions ensure that the intent of the
interlocks is preserved, and that there is no reduction in the
ability of the interlocks to maintain adequate refueling conditions.
The proposed TS changes will preserve the intended interlock
functions, and maintain the existing level of protection against
refueling errors that could lead to an inadvertent criticality,
damage to reactor internals or fuel assemblies, or excessive
personnel radiation exposure. The one-rod-out and refueling
interlocks will continue to function with their present degree of
reliability. The proposed TS changes will continue to maintain
strict controls on fuel bundle and control rod manipulations to
avoid inadvertent criticality. The proposed TS changes provide the
same level of assurance regar[d]ing the manipulation of control rods
during refueling operations as that currently described in the LGS
UFSAR, and as discussed below.
a. Preventing operation of a fuel-loaded refueling platform or
service platform hoist while over the core if a control rod is
already withdrawn.
b. Preventing a fuel-loaded refueling platform from traveling
over the core if a control rod is already withdrawn.
c. Preventing any control rod from being withdrawn if fuel-
loaded refueling platform or service platform is already operating
over the core.
d. Preventing the withdrawal of more than one control rod at a
time with the reactor mode switch in the REFUEL position.
Therefore, the proposed TS changes do not involve a reduction in
a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this eview, 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: Pottstown Public Library, 500
High Street, Pottstown, Pennsylvania 19464.
Attorney for licensee: J. W. Durham, Sr., Esquire, Sr. V. P. and
General Counsel, Philadelphia Electric Company, 2301 Market Street,
Philadelphia, Pennsylvania 19101
NRC Project Director: John F. Stolz
South Carolina Electric & Gas Company (SCE&G), South Carolina
Public Service Authority, Docket No. 50-395, Virgil C. Summer
Nuclear Station, Unit No. 1, Fairfield County, South Carolina
Date of amendment request: February 21, 1995, as revised on August
31, 1995
Description of amendment request: The proposed amendment would
revise the Technical Specifications (TS) to reflect changes to 10 CFR
Part 20 (including Appendix B, Table 2 concentrations) and provide
additional administrative corrections.
[[Page 49947]]
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. The probability or consequences of an accident previously
evaluated does not involve a significant increase.
The proposed TS changes showing the relocation of the old 10 CFR
20.106 requirements to the new 10 CFR 20.1302, the old 10 CFR
20.203(c)(2) requirements to the new 10 CFR 20.1601(a), and the old
10 CFR 20.407 requirements to the new 10 CFR 20.2206(b) will not
involve a significant increase in the probability or consequences of
an accident previously evaluated because there will be no change in
the types and amounts of effluents that will be released, nor will
there be an increase in individual or cumulative occupational
radiation exposures.
The proposed revision to the liquid and gaseous release rate
limits will not involve a significant increase in the probability or
consequences of an accident previously evaluated because there will
be no change in the types and amounts of effluents that will be
released, nor will there be an increase in individual or cumulative
occupational radiation exposures. This is only a change to the
method of (algorithm) determining release rate limits and will not
change net limits or change the more restrictive 10 CFR 50 Appendix
I dose limits.
The proposed revision to the radioactive material quantity in
the settling pond and its associated TS Bases will not involve a
significant increase in the probability or consequences of an
accident previously evaluated because there will be no change in the
types of effluents that will be released, nor will there be an
increase in individual or cumulative occupational radiation
exposures. This is only a change to the quantity of radioactive
material in the settling pond and will conservatively lower net
limits.
The proposed revision to the TS bases for the liquid holdup tank
activity limit will not involve a significant increase in the
probability or consequences of an accident previously evaluated
because there will be no change in the types and amounts of
effluents that will be released, nor will there be an increase in
individual or cumulative occupational radiation exposures. The curie
limit is not affected, therefore, the change does not represent a
decrease in the level of control previously evaluated.
The proposed revision to the distance at which dose rates are
measured from the radiation source or surface will not involve a
significant increase in the probability or consequences of an
accident previously evaluated because there will be no increase in
the individual or cumulative occupational radiation exposures. The
change in distance is conservative in its effect on worker
protection and is in conformance with 10 CFR 20.1601 requirements.
2. The possibility of a new or different kind of accident from
any previously evaluated is not created.
The proposed TS changes showing the relocation of the old 10 CFR
20.106 requirements to the new 10 CFR 20.1302, relocation of the old
10 CFR 20.203(c)(2) requirements to the new 10 CFR 20.1601(a), and
relocation of the old 10 CFR 20.407 requirements to the new 10 CFR
20.2206(b) will not create the possibility of a new or different
kind of accident from any previously evaluated because the revisions
are administrative and will not change the types and amounts of
effluents that will be released.
The proposed revision to the liquid and gaseous release rate
limits will not create the possibility of a new or different kind of
accident from any previously evaluated because the revision is
administrative and will not change the types and amounts of
effluents that will be released.
The proposed revision to the quantity of radioactive material in
the settling pond and its associated TS Bases will not create the
possibility of a new or different kind of accident from any
previously evaluated because the revision will not change the types
of effluents that will be released. This is only a change to the
quantity of radioactive material in the settling pond and will
conservatively lower net limits.
The proposed revision to the TS bases for the liquid holdup tank
activity limit will not create the possibility of a new or different
kind of accident from any previously evaluated because the revision
is administrative and will not change the types and amounts of
effluents that will be released.
Implementation of the more conservative distance at which dose
rates are measured will not create the possibility of a new or
different kind of accident from any previously evaluated.
3. A significant reduction in a margin of safety is not
involved.
The proposed revisions due to the location of requirements will
not reduce a margin of safety because they are administrative in
nature. No equipment or procedural changes are postulated. There is
no impact on any margin of safety.
The proposed revision to liquid and gaseous release rate limits
will not reduce a margin of safety because it is administrative in
nature. These revisions preserve the existing level of effluent
control. No changes to the more restrictive 10 CFR 50 Appendix I
dose limits are made. There are no equipment or operational
procedure changes, therefore, no accidents of any kind will be
created by this change.
The proposed revision to the quantity of radioactive material in
the settling pond and its associated TS Bases will not reduce a
margin of safety because it is conservative in nature and preserves
the existing level of effluent control. There are no equipment or
operational procedure changes required, therefore, no accidents of
any kind will be created by this change.
The proposed revision to the TS bases for the liquid holdup tank
activity limit will not reduce a margin of safety because it is
administrative in nature and preserve[s] the existing level of
effluent control. No equipment or procedural changes are postulated.
There is no impact on any margin of safety.
The change in distance for a High Radiation Area classification
from 18 in.(45 cm) to (30 cm)12 in. from the radiation source or
surface will not reduce the margin of safety because this change
will reduce the worker's stay time in the area and therefore
minimize exposure.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: Fairfield County Library, 300
Washington Street, Winnsboro, SC 29180
Attorney for licensee: Randolph R. Mahan, South Carolina Electric &
Gas Company, Post Office Box 764, Columbia, South Carolina 29218
NRC Project Director: Frederick J. Hebdon
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: July 19, 1995
Description of amendment requests: The licensee proposes to revise
technical specifications (TSs) to (1) support modifications to the
containment area radiation monitors, to either upgrade or replace
existing equipment with state-of-the-art equipment, (2) relocate the
setpoint and allowable values for the control room airborne radiation
monitors to be consistent with the containment airborne radiation
monitors TS, and (3) make minor editorial changes to the TS pertaining
to the control room airborne radiation monitors and the containment
airborne radiation monitors. The proposed changes affect TS Tables 3.3-
3, 3.3-4, 3.3-5, 3.3-6, 4.3-2, and 4.3-3.
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.
Control Room Airborne Radiation Monitors
The proposed change would permit relocation of the setpoint and
allowable values for the monitors from the Technical Specifications
(TSs) to the administrative control procedures. This change is
consistent with the existing Containment Airborne Radiation Monitor
TSs. This change will not prevent the radiation monitors from
[[Page 49948]]
performing their intended function following a design basis accident.
Therefore, operation of the facility in accordance with this change
will not involve a significant increase in the probability or
consequences of an accident previously evaluated.
Containment Area Radiation Monitors
The proposed change deletes the existing Containment Area
Radiation Monitors RE-7856-1 and RE-7857-2 and their Engineered
Safety Feature Actuation System (ESFAS) function to initiate
containment purge isolation on high radiation in containment. The
deletion of this ESFAS function does not create a precursor to any
analyzed accident since these monitors are for accident mitigation
only.
Currently, no release of radioactivity is assumed during a Fuel
Handling Accident in containment since the Containment Area
Radiation Monitors detect and isolate containment purge prior to
release. The proposed deletion will cause some release prior to
detection and isolation of purge by the remaining noble gas
containment monitors. The consequences of a Fuel Handling Accident
inside containment were previously re-evaluated, assuming no
containment purge isolation, to resolve inconsistencies in the
original analysis assumptions and methodology. The results of the
calculation indicated off-site doses well within the limits of 10
CFR 100 and Control Room doses that met the limits of 10 CFR 50
Appendix A General Design Criterion 19. Containment purge isolation
on high gaseous activity during a Fuel Handling Accident will still
be available with this proposed change but is not required for the
dose consequences to remain within the dose criteria. Therefore, the
proposed change will not significantly increase the consequences of
a Fuel Handling Accident inside containment.
The Loss of Coolant Accident (LOCA) function of the Containment
Purge Isolation System (CPIS) signal will be essentially unaffected
by this proposed change. Currently, containment purge isolation
(containment minipurge) on high radiation signals is a diverse
signal with Safety Injection Actuation System (SIAS) and Containment
Isolation Actuation System (CIAS). In a LOCA event, containment
purge isolation is expected to occur on either SIAS or CIAS prior to
a CPIS signal on high radiation in containment. While this proposed
change reduces the diversity of radiation monitoring inputs, the
diversity of parameters measured (pressure and radiation) is still
preserved. Therefore, the proposed change will not increase the
consequences of a LOCA.
2. The proposed change does not create the possibility of a new
or different kind of accident from any accident previously
evaluated.
Control Room Airborne Radiation Monitors
Relocating the monitor setpoint and allowable values from the
TSs to the administrative procedures would not alter the design and
operational interface between the Control Room Isolation System
instrumentation and existing plant equipment. As such, the monitors
would continue to operate and perform their intended safety function
to isolate the control room following a design basis accident as
before. Therefore, operation of the facility in accordance with this
proposed change will not create the possibility of a new or
different kind of accident from any accident previously evaluated.
Containment Area Radiation Monitors
The deletion of the Containment Area Radiation Monitors will not
alter the operation of CPIS. The remaining interface between CPIS
and existing plant equipment will continue to perform their intended
safety function to isolate containment purge by closing the
containment purge valves. This function will continue to be
performed by Containment Airborne Radiation Monitors 2(3) RT-7804-1
and 2(3) RT-7807-2. Therefore, operation of the facility in
accordance with this proposed change will not create the possibility
of a new or different kind of accident from any accident previously
evaluated.
3. The proposed change does not involve a significant reduction
in a margin of safety.
Control Room Airborne Radiation Monitors
Relocating the monitor setpoint and allowable values to the
administrative procedures would not alter the existing margin of
safety. The relocation would only relinquish control of the setpoint
and allowable values from the TSs to quality-affecting (changes will
require a 10 CFR 50.59 evaluation) procedures. Therefore, operation
of the facility will not involve a significant reduction in a margin
of safety.
Containment Area Radiation Monitors
The proposed change does not affect the margin of safety in
Modes 1 through 4 since the diversity of the parameters measured is
maintained for minipurge isolation. Either SIAS, CIAS, CPIS, or
manual operation will close the containment mini purge valves. The
main purge is sealed closed during Modes 1 through 4 with the purge
valves closed and deactivated.
The diversity of the parameters measured is not maintained for
the containment main purge isolation. The main purge is only
applicable during Modes 5 and 6 and main purge isolation is
initiated only by either CPIS or manual operation. This proposed
change along with the previously submitted PCN-299 reduces the
diversity of radiation sensing in containment for CPIS generation
from four types (gaseous, iodine, particulate, and gamma) to one
type (gaseous activity). Since the consequences of a Fuel Handling
Accident inside containment without purge isolation have been
calculated to be well within 10 CFR 100 dose limits, the loss of
diversity for this accident does not result in a significant
reduction in a margin of safety. Therefore, this proposed change
will not involve a significant reduction in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment requests involve no significant hazards consideration.
Local Public Document Room location: Main Library, University of
California, P. O. Box 19557, 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
Tennessee Valley Authority, Docket No. 50-328, Sequoyah Nuclear
Plant, Unit 2, Hamilton County, Tennessee
Date of amendment request: May 19, 1995; revised September 11, 1995
(TS 95-13)
Description of amendment request: The proposed change would revise
License Condition 2.C.(17) to extend the required surveillance interval
to May 18, 1996, for Surveillance Requirement 4.3.2.1.3. The proposed
change would extend the Engineered Safety Features Response Time
instrument tests required at 36-month intervals shown in Table 3.3-3
associated with safety injection, feedwater isolation, containment
isolation Phase A, auxiliary feedwater pump, essential raw cooling
water system, emergency gas treatment system, containment spray,
containment isolation Phase B, turbine trip, 6.9-kilovolt shutdown
board-degraded voltage or loss of voltage, and automatic switchover to
containment sump actuations. The proposed extension will limit the
interval past the allowable extension provided by TS 4.0.2 to 5 months.
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:
TVA has evaluated the proposed technical specification (TS)
change and has determined that it does not represent a significant
hazards consideration based on criteria established in 10 CFR
50.92(c). Operation of Sequoyah Nuclear Plant (SQN) in accordance
with the proposed amendment will not:
1. Involve a significant increase in the probability or
consequences of an accident previously evaluated.
The proposed change is temporary and allows a one-time extension
of Surveillance Requirement 4.3.2.1.3 for Cycle 7 to allow
surveillance testing to coincide with the seventh refueling outage.
The proposed surveillance interval extension will not cause a
significant reduction in system reliability nor affect the ability
of the systems to perform their design function. Current monitoring
of plant conditions and continuation of the surveillance testing
required during normal plant operation will continue to be performed
to ensure conformance with TS operability requirements. Therefore,
this change does not involve a significant increase in the
[[Page 49949]]
probability or consequences of an accident previously evaluated.
2. Create the possibility of a new or different kind of accident
from any previously analyzed.
Extending the surveillance interval for the performance of
specific testing will not create the possibility of a new or
different kind of accidents. No changes are required to any system
configurations, plant equipment, or analyses. Therefore, this change
will 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.
Surveillance interval extensions will not impact any plant
safety analyses since the assumptions used will remain unchanged.
The safety limits assumed in the accident analyses and the design
function of the equipment required to mitigate the consequences of
any postulated accidents will not be changed since only the
surveillance test interval is being extended. Historical performance
generally indicates a high degree of reliability, and surveillance
testing performed during normal plant operation will continue to be
performed to verify proper performance. Therefore, the plant will be
maintained within the analyzed limits, and the proposed extension
will not significantly reduce the margin of saety.
The NRC has reviewed the licensee's analysis and, based on
thisreview, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: Chattanooga-Hamilton County
Library, 1101 Broad Street, Chattanooga, Tennessee 37402
Attorney for licensee: General Counsel, Tennessee Valley Authority,
400 West Summit Hill Drive, ET 11H, Knoxville, Tennessee 37902
NRC Project Director: Frederick J. Hebdon
Virginia Electric and Power Company, Docket Nos. 50-338 and 50-339,
North Anna Power Station, Units No. 1 and No. 2, Louisa County,
Virginia
Date of amendment request: September 1, 1995
Description of amendment request: The proposed changes to the
Technical Specifications (TS) for the North Anna Power Station, Units
1&2 (NA-1&2) would allow a single outage of up to 14 days for each
emergency diesel generator (EDG) once every 18 months. The purpose of
the outage is the performance of a preventive maintenance inspection,
appropriate for diesels used for this class of standby service, which
requires disassembly of the EDG. Currently this maintenance inspection
is performed during refueling outages. The proposed changes would
permit this maintenance inspection to be performed during Modes 1 to 4
in addition to the current allowance during Modes 5 or 6.
A probabilistic safety analysis (PSA) has been performed which
demonstrates that a fourteen (14) day maintenance inspection outage,
once every eighteen (18) months for each EDG, results in no significant
change in core damage frequency assuming adequate compensatory measures
are in place. The compensatory measures include requirements that the
other EDGs, off-site power supply, and the alternate A.C. diesel (AAC
DG) be operable during the preventive maintenance inspection outage.
The effect of the proposed change has been calculated to be an
increase in core damage frequency of approximately 1E-6 per year, which
is not considered to be a significant change (i.e., an acceptable
change in risk, or a non-risk significant change) from the baseline
core damage frequency of 4.1E-5.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
Specifically, operation of North Anna Power Station in
accordance with the [proposed] Technical Specifications changes will
not:
a. involve a significant increase in the probability or
consequences of an accident previously evaluated. The probabilistic
safety analysis (PSA) demonstrates that the increase in core damage
frequency due to performing the EDG maintenance inspection over a
fourteen day period once every 18 months is not significant as long
as the AAC DG is operable to act as a source of emergency power to
replace the EDG. The period of time during which the EDG is
unavailable is short enough to limit the impact of using the
manually operated AAC DG as a replacement for the automatically
operated EDG.
b. create the possibility of a new or different kind of accident
from any accident previously evaluated. The proposed Technical
Specifications changes only modify the operability of an EDG for a
limited and defined period of time. The UFSAR [Updated Final Safety
Analysis Report] accidents are analyzed assuming that the EDG is the
worst single failure. This assumption is more severe than the
proposed Technical Specifications changes which replaces the EDG
with the AAC DG. Similarly, the PSA performed to evaluate the
proposed Technical Specifications changes considered all of the
initiating events defined for the PSA performed for the Individual
Plant Examination. No new initiators were defined as a result of a
review of the PSA model. Therefore, it is concluded that no new or
different kind of accident from any previously evaluated has been
created.
c. The proposed Technical Specifications changes do not result
in a reduction in margin of safety as defined in the basis for any
Technical Specifications. The PSA was performed to evaluate the
concept of a one time outage. The results of the analyses show no
significant change in the core damage frequency. As described above
the proposed Technical Specifications changes only modify the
operability of an EDG for a limited and defined period of time.
Thus, operation with slightly increased EDG unavailability due to
maintenance, and the AAC DG operable is acceptable.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: The Alderman Library, Special
Collections Department, University of Virginia, Charlottesville,
Virginia 22903-2498.
Attorney for licensee: Michael W. Maupin, Esq., Hunton and
Williams, Riverfront Plaza, East Tower, 951 E. Byrd Street, Richmond,
Virginia 23219.
NRC Project Director: David B. Matthews
Wolf Creek Nuclear Operating Corporation, Docket No. 50-482, Wolf
Creek Generating Station, Coffey County, Kansas
Date of amendment request: March 24, 1995, as supplemented by
letter dated August 16, 1995.
Description of amendment request: This request proposes to revise
Technical Specification 1.7, ``Containment Integrity,'' Technical
Specification 3/4.6.1, ``Containment Integrity,'' Technical
Specification 3/4.6.3, ``Containment Isolation Valves,'' and their
associated Bases. These proposed changes will remove Technical
Specification Table 3.6-1 ``Containment Isolation Valves,'' to Wolf
Creek Generating Station (WCGS) procedures. This proposed change is in
accordance with the guidance provided in Generic Letter 91-08,
``Removal of Component Lists from Technical Specifications,'' dated May
6, 1991. In addition, this request proposes to add a footnote to
Technical Specification 3.6.3 extending the allowed outage time for the
component cooling water (CCW) system reactor coolant pump seal water
supply and return valves. This determination supersedes the staff's
proposed no significant hazards consideration determination evaluation
for the requested changes that was published on April 26, 1995 (60 FR
20532).
[[Page 49950]]
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 changes simplify the technical specifications, meet
the regulatory requirements for control of containment isolation,
and are consistent with the guidelines of GL 91-08. The procedural
details of Technical Specification Table 3.6-1 have not been
changed, but only relocated to a different controlling document. The
proposed changes are administrative in nature, should result in
improved administrative practices, and do not affect plant
operations. The addition of the footnote to allow up to 12 hours for
valve testing the CCW MOVs [motor-operated valves] does not affect
the severity of any accident previously evaluated. This footnote
does not impact plant safety since the second isolation device in
the affected penetrations would still be available to provide
isolation between the RCS and the outside atmosphere.
The probability of occurrence of a previously evaluated accident
is not increased because this change does not introduce any new
potential accident initiating conditions. The consequences of an
accident previously evaluated is not increased because the ability
of containment to restrict the release of any fission product
radioactivity to the environment will not be degraded by this
change.
2. The proposed change does not create the possibility of a new
or different kind of accident from any accident previously
evaluated.
The proposed changes are administrative in nature, do not result
in physical alterations or changes to the operation of the plant,
and cause no change in the method by which any safety-related system
performs its function. The addition of the footnote to allow up to
12 hours for valve testing the CCW MOVs does not affect the severity
of any accident previously evaluated. The additional time provides
assurance that the inoperable valve is in proper working order prior
to returning it to OPERABLE status. Therefore, this proposed change
will not create the possibility of a new or different kind of
accident from any previously evaluated.
3. The proposed change does not involve a significant reduction
in a margin of safety.
The administrative change to relocate Technical Specification
Table 3.6-1 to appropriate plant procedures does not alter the basic
regulatory requirements for containment isolation and will not
adversely affect containment isolation capability for credible
accident scenarios. Adequate control of the content of the table is
assured by existing plant procedures. The additional footnote to
extend the allowed outage time to 12 hours for the CCW MOVs does not
affect containment isolation capability since the second isolation
device in the affected penetrations would still be available to
provide isolation between the RCS and the outside atmosphere, and to
ensure that a release of radioactive material to the environment
following an accident will not exceed the assumptions used in the
LOCA Analyses.
The proposed relocation of the Technical Specification Table
3.6-1 does not alter the requirements for containment isolation
valve operability currently in the technical specifications. The LCO
and Surveillance Requirements would be retained in the revised
technical specifications. Therefore, the proposed change will not
affect the meaning, application, and function of the current
technical specification requirements for the valves in Table 3.6-1.
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 locations: Emporia State University,
William Allen White Library, 1200 Commercial Street, Emporia, Kansas
66801 and Washburn University School of Law Library, Topeka, Kansas
66621
Attorney for licensee: Jay Silberg, Esq., Shaw, Pittman, Potts and
Trowbridge, 2300 N Street, N.W., Washington, D.C. 20037
NRC Project Director: William H. Bateman
Wolf Creek Nuclear Operating Corporation, Docket No. 50-482, Wolf
Creek Generating Station, Coffey County, Kansas
Date of amendment request: August 22, 1995
Description of amendment request: The proposed license amendment
request would relocate Technical Specification Tables 3.3-2, ``Reactor
Trip System Instrumentation Response Times,'' and 3.3-5, ``Engineered
Safety Features Response Times,'' and applicable Bases discussions, to
Updated Safety Analysis Report (USAR) Chapter 16. The NRC has already
implemented this line-item technical specification improvement in the
new Standard Technical Specifications (NUREG-1431 for Westinghouse
plants). This amendment request follows the guidance provided by the
NRC in Generic Letter 93-08, ``Relocation of Technical Specification
Tables of Instrument Response Time Limits,'' for relocating instrument
response time tables.
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.
This license amendment request does not change any Reactor Trip
System (RTS) or Engineered Safety Features Actuation System (ESFAS)
instrument response times or surveillance intervals currently
prescribed in Technical Specification Tables 3.3-2 and 3.3-5. The
RTS and ESFAS will continue to function in a manner consistent with
the assumptions in the Updated Safety Analysis Report Chapter 15
accident analyses and the plant design basis. Therefore, overall
protection system performance will remain within the bounds of the
accident analyses documented in USAR Chapter 15. As such, there will
be no degradation in system performance, nor will there be an
increase in the number of challenges to equipment assumed to
function during an accident situation.
The proposed technical specification revision does not involve
any hardware changes or changes to any instrumentation setpoints,
system operating parameters, or system accident mitigation
capabilities, nor do the changes affect the probability of any event
initiators. Thus, the proposed change will not result in an increase
in the consequences of or the probability of occurrence of any
accident or safety-related equipment malfunction.
2. The proposed change does not create the possibility of a new
or different kind of accident from any accident previously
evaluated.
As discussed above, there are no hardware changes associated
with this proposed amendment request, nor are there any changes in
the method by which any safety-related plant system performs its
safety function. The normal manner of plant operation is not
affected by this proposed change.
No new accident scenarios, transient precursors, failure
mechanisms, or limiting single failures are introduced as a result
of the proposed changes. There will be no adverse effect or
challenges imposed on any safety-related system as a result of these
changes. Therefore, the possibility of a new or different kind of
accident is not created by the proposed changes.
3. The proposed change does not involve a significant reduction
in a margin of safety.
No response times will be changed by this amendment request. The
proposed request only changes the document where the response times
will be listed. This proposed amendment request will not affect the
manner in which safety limits or limiting safety system settings are
determined, nor will there by [be] any effect on plant systems
necessary to assure the accomplishment of protection functions. The
proposed change will not impact any margin of safety defined in the
basis for any Technical Specification.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
[[Page 49951]]
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room locations: Emporia State University,
William Allen White Library, 1200 Commercial Street, Emporia, Kansas
66801 and Washburn University School of Law Library, Topeka, Kansas
66621
Attorney for licensee: Jay Silberg, Esq., Shaw, Pittman, Potts and
Trowbridge, 2300 N Street, N.W., Washington, D.C. 20037
NRC Project Director: William H. Bateman
Previously Published Notices Of Nonsideration 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.
Indiana Michigan Power Company, Docket No. 50-315, Donald C. Cook,
Nuclear Plant, Unit No. 1, Berrien County, Michigan
Date of application for amendment: August 4, 1995 (AEP:NRC:1129E)
Description of amendment request: The proposed amendment would
modify Technical Specification 4.4.5.4 and 4.4.5.5, on steam
generators, to allow for repair of hybrid expansion joint sleeves under
redefined repair boundary limits.
Date of publication of individual notice in the Federal Register:
August 14, 1995 (60 FR 41904)
Expiration date of individual notice: For comments: August 29,
1995; hearing requests: September 13, 1995
Local Public Document Room location: Maud Preston Palenske Memorial
Library, 500 Market Street, St. Joseph, Michigan 49085
Tennessee Valley Authority, Docket Nos. 50-327 and 50-328, Sequoyah
Nuclear Plant, Units 1 and 2, Hamilton County, Tennessee
Date of application for amendments: August 21, 1995Description of
amendments request: Amend technical specification 3.7.5.c to allow an
increase in the average essential raw cooling water supply header
temperature from 84.5 deg.F to 87 deg.F until September 30, 1995.
Date of publication of individual notice in the Federal Register:
August 28, 1995 (60 FR 44517)
Expiration date of individual notice: September 12, 1995
Local Public Document Room location: Chattanooga-Hamilton County
Library, 1101 Broad Street, Chattanooga, Tennessee 37402
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.
Arizona Public Service Company, et al., Docket Nos. STN 50-528, STN
50-529, and STN 50-530, Palo Verde Nuclear Generating Station,
Units 1, 2, and 3, Maricopa County, Arizona
Date of application for amendments: June 17, 1994
Brief description of amendments: These amendments revise the
surveillance requirement and Bases section of TS 4.7.1.6 to increase
the minimum nitrogen accumulator pressure for the atmospheric dump
valves (ADVs).
Date of issuance: September 6, 1995
Effective date: September 6, 1995
Amendment Nos.: Unit 1 - Amendment No. 99; Unit 2 - Amendment No.
87; Unit 3 - Amendment No. 70
Facility Operating License Nos. NPF-41, NPF-51, and NPF-74: The
amendments revised the Technical Specifications.
Date of initial notice in Federal Register: August 17, 1994 (59 FR
42333) The Commission's related evaluation of the amendments is
contained in a Safety Evaluation dated September 6, 1995.No significant
hazards consideration comments received: No
Local Public Document Room location: Phoenix Public Library, 12
East McDowell Road, Phoenix, Arizona 85004
Arizona Public Service Company, et al., Docket Nos. STN 50-528, STN
50-529, and STN 50-530, Palo Verde Nuclear Generating Station,
Units 1, 2, and 3, Maricopa County, Arizona
Date of application for amendments: March 31, 1995
Brief description of amendments: The amendments clarify the
shutdown margin definition, change the shutdown margin applicability
and surveillance requirements to comply with the safety analysis
assumptions for subcritical inadvertent control element assembly
withdrawal (UFSAR Section 15.4, and expand the applicability for core
protection calculator (CPC) operability. In addition, the amendments
add a reference to the Core Operating Limits Report for the MODE 6
refueling boron concentration limit. The amendments also change the
power calibration requirements for the linear power level, the CPC
delta T power, and CPC nuclear power signals to allow more conservative
settings than previously required.
Date of issuance: September 1, 1995
[[Page 49952]]
Effective date: September 1, 1995
Amendment Nos.: Unit 1 - Amendment No. 98; Unit 2 - Amendment No.
86; Unit - Amendment No. 69
Facility Operating License Nos. NPF-41, NPF-51, and NPF-74: The
amendments revised the Technical Specifications.
Date of initial notice in Federal Register: June 6, 1995 (60 FR
29871) The Commission's related evaluation of the amendments is
contained in a Safety Evaluation dated September 1, 1995.No significant
hazards consideration comments received: No
Local Public Document Room location: Phoenix Public Library, 1221
N. Central, Phoenix, Arizona 85004
Carolina Power & Light Company, Docket No. 50-261, H. B. Robinson
SteamElectric Plant, Unit No. 2, Darlington County, South Carolina
Date of application for amendment: June 3, 1995, as supplemented on
August 7, 1995. The supplemental submittal did not expand the scope of
the original Federal Register notice or change the no significant
hazards determination.
Brief description of amendment: The amendment clarifies the
definition of operability of the charging pumps by adding a footnote to
TS Section 3.2.2.a that states that the connectibility of the emergency
power sources is not required for charging pump operability. The bases
statement for TS 3.2.2 is also changed for clarification.
Date of issuance: September 5, 1995
Effective date: September 5, 1995
Amendment No.: 166
Facility Operating License No. DPR-23. Amendment revises the
Technical Specifications.
Date of initial notice in Federal Register: July 5, 1995 (60 FR
35063) The Commission's related evaluation of the amendment is
contained in a Safety Evaluation dated September 5, 1995.No significant
hazards consideration comments received: No
Local Public Document Room location: Hartsville Memorial Library,
147 West College Avenue, Hartsville, South Carolina 29550
Commonwealth Edison Company, Docket Nos. STN 50-454 and STN 50-455,
Byron Station, Unit Nos. 1 and 2, Ogle County, IllinoisDocket Nos.
STN 50-456 and STN 50-457, Braidwood Station, Unit Nos. 1 and 2,
Will County, Illinois
Date of application for amendments: February 21, 1995
Brief description of amendments: The amendments revise the
technical specifications to permit replacement of the reactor coolant
resistance temperature detector (RTD) bypass manifold system with fast
response RTDs mounted in thermowells welded directly into the reactor
coolant system piping.
Date of issuance: September 5, 1995
Effective date: September 5, 1995
Amendment Nos.: 74 and 66
Facility Operating License Nos. NPF-37, NPF-66, NPF-72 and NPF-77:
The amendments revised the Technical Specifications.
Date of initial notice in Federal Register: July 5, 1995 (60 FR
35063) The Commission's related evaluation of the amendments is
contained in a Safety Evaluation dated September 5, 1995.No significant
hazards consideration comments received: No
Local Public Document Room location: For Byron, the Byron Public
Library District, 109 N. Franklin, P.O. Box 434, Byron, Illinois 61010;
for Braidwood, the Wilmington Public Library, 201 S. Kankakee Street,
Wilmington, Illinois 60481
Commonwealth Edison Company, Docket Nos. 50-254 and 50-265, Quad
Cities Nuclear Power Station, Units 1 and 2, Rock Island County,
Illinois
Date of application for amendments: February 23, 1995
Brief description of amendments: The amendment revises the Quad
Cities Nuclear Power Station, Units 1 and 2, operating licenses to
reflect the transfer of the Iowa-Illinois Gas and Electric Company's 25
percent undivided ownership to MidAmerican Energy Company.
Date of issuance: September 11, 1995
Effective date: As of the consummation of the merger between Iowa-
Illinois Gas and Electric Company, Midwest Power Systems, Inc.,
MidAmerican Energy Company, and Midwest Resources, Inc.
Amendment Nos.: 159 and 155
Facility Operating License Nos. DPR-29 and DPR-30: The amendments
revised the operating licenses.
Date of initial notice in Federal Register: July 5, 1995 (60 FR
35054) The Commission's related evaluation of the amendments is
contained in an Environmental Assessment and Finding of No Significant
Impact dated March 21, 1995, and in a Safety Evaluation dated September
11, 1995.No significant hazards consideration comments received: No
Local Public Document Room location: Dixon Public Library, 221
Hennepin Avenue, Dixon, Illinois 61021
Commonwealth Edison Company, Docket Nos. 50-295 and 50-304, Zion
Nuclear Power Station Units 1 and 2, Lake County, Illinois
Date of application for amendments: May 31, 1995
Brief description of amendments: The amendments authorize an
alternative repair criteria for defects found in the tube expansion
region within the tubesheet.
Date of issuance: September 11, 1995
Effective date: September 11, 1995
Amendment Nos.: 168 and 155
Facility Operating License Nos. DPR-39 and DPR-48: The amendments
revised the Technical Specifications.
Date of initial notice in Federal Register: July 5, 1995 (60 FR
35067) The Commission's related evaluation of the amendments is
contained in a Safety Evaluation dated August 11, 1995.No significant
hazards consideration comments received: No
Local Public Document Room location: Waukegan Public Library, 128
N. County Street, Waukegan, Illinois 60085
Duke Power Company, et al., Docket Nos. 50-413 and 50-414, Catawba
Nuclear Station, Units 1 and 2, York County, South Carolina
Date of application for amendments: September 19, 1994, as
supplemented April 26 and June 19, 1995
Brief description of amendments: These changes to the Technical
Specifications (TS) increase the enrichment limits for fuel stored in
the fuel pools and establish restricted loading patterns and associated
burnup criteria for qualifying fuel in the spent fuel pools. In
addition, several administrative changes have been included in order to
provide clarity to the TS and bring them more in line with the Standard
Technical Specifications format. These changes are as follows: (1) The
TS index is changed to add TS 3/4.9.12 and 3/4.9.13, Tables 3.9-1 and
3.9-2 and Figure 3.9-1; (2) TS 3/4.9.12, Spent Fuel Pool (SFP) Boron
Concentration is added to establish a boron concentration limit and to
establish a Limiting Condition for Operation (LCO) for all modes of
operation and to allow the numerical value of the limit to be specified
in the Core Operating Limits Report (COLR); (3) TS 3/4.9.13, Tables
3.9-1 and 3.9-2 and Figure 3.9-1 are being added to establish
restricted loading patterns for spent fuel storage and associated
burnup criteria; (4) Corresponding BASES for TS 3/4.9.12 and 3/4.9.13
are added to explain the basis for each LCO, Action Statement and
Surveillance
[[Page 49953]]
Requirement covered by the subject TS; (5) TS 5.6, Fuel Storage, is
changed to reflect limits for criticality analysis for fuel storage;
and (6) TS 6.9, Reporting Requirements, is changed to reflect the
inclusion of the SFP boron concentration limit values in the COLR as
established by TS 3/4.9.12.
Date of issuance: August 31, 1995
Effective date: As of the date of issuance to be implemented within
30 days
Amendment Nos.: 134 and 128
Facility Operating License Nos. NPF-35 and NPF-52: Amendments
revised the Technical Specifications.
Date of initial notice in Federal Register: May 23, 1995 (60 FR
27338) The June 19, 1995, letter provided clarifying information that
did not change the scope of the September 19, 1994, application and the
initial proposed no significant hazards consideration determination.
The Commission's related evaluation of the amendments is contained in a
Safety Evaluation dated August 31, 1995, and Environmental Assessment
dated August 15, 1995.No significant hazards consideration comments
received: No
Local Public Document Room location: York County Library, 138 East
Black Street, Rock Hill, South Carolina 29730
Duquesne Light Company, et al., Docket Nos. 50-334 and 50-412,
Beaver Valley Power Station, Unit Nos. 1 and 2, Shippingport,
Pennsylvania
Date of application for amendments: August 31, 1994, as
supplemented May 18, 1995.
Brief description of amendments: These amendments delete Beaver
Valley Power Station, Unit 2, License Conditions 2.C.(3), 2.C.(5),
2.C.(7), 2.C.(8), 2.C.(9) and 2.C.(10) to reflect completion of
activities required by these license conditions and make the following
revisions to the Beaver Valley Power Station, Units 1 and 2, TSs:
1. Eliminate references to specific frequencies for each of the TS
required audits (TS 6.2.2.8).
2. Eliminate references to reviews and audits of the Emergency plan
and Security Plant (TSs 6.5.2.8 and 6.8.1).
3. Include Offsite Dose Calculation Manual and Process Control
Program and associated implementing procedures into the list of
required audits (TS 6.5.2.8).
4. Editorial changes which were necessitated by a reorganization
(TS 6.2.1, 6.2.3.1, 6.2.3.4, 6.5.2.2, 6.5.2.8, 6.5.2.9, and 6.5.2.10).
5. Eliminate reference to Appendix A of 10 CFR Part 55 (TS 6.4.1).
6. Separate the Inservice Inspection (ISI) and Inservice Testing
(IST) Programs surveillance requirements and remove the requirement
that relief requests be granted before they are implemented for both
IST and ISI (TS 4.0.5).
The May 18, 1995, letter requested withdrawal of the proposed
changes to TS 6.5.2.8 dealing with audits of the Beaver Valley Power
Station, Units 1 and 2, fire protection program and withdrawal of a
proposed 25-percent grace period for all audit frequencies (Item 6 in
August 31, 1994 application).
Date of issuance: August 31, 1995
Effective date: Units 1 and 2, as of the date of issuance and shall
be implemented within 60 days.
Amendment Nos.: 191 and 74
Facility Operating License Nos. DPR-66 and NPF-73: Amendments
revised the Units 1 and 2 Technical Specifications, and the Unit 2
License.
Date of initial notice in Federal Register: (59 FR 65812) December
21, 1994. The May 18, 1995, letter did not change the original no
significant hazards consideration determination or expand the scope of
the December 21, 1994, Federal Register notice. The Commission's
related evaluation of the amendments is contained in a Safety
Evaluation dated August 31, 1995.No significant hazards consideration
comments received: No
Local Public Document Room location: B. F. Jones Memorial Library,
663 Franklin Avenue, Aliquippa, Pennsylvania 15001
Duquesne Light Company, et al., Docket Nos. 50-334 and 50-412,
Beaver Valley Power Station, Unit Nos. 1 and 2, Shippingport,
Pennsylvania
Date of application for amendments: October 11, 1994, as
supplemented June 23, 1995, and August 24, 1995
Brief description of amendments: These amendments revise Beaver
Valley Power Station Technical Specifications (TSs) 1.18, ``Quadrant
Power Tilt Ratio,'' 3/4.2.4, ``Quadrant Power Tilt Ratio,'' the table
Notation of TS Table 3.3-1, ``Reactor Trip System Instrumentation,''
and associated Bases to incorporate the guidance provided in the NRC's
Improved Standard Technical Specifications (NUREG-1431, Revision 1) to
these TSs. The amendments clarify the requirements of the subject TSs
with regard to the use of excore power range neutron flux detectors to
monitor quadrant power tilt ratio when an excore power range neutron
flux instrument is inoperable. The changes also make several minor
editorial changes in the subject TSs.
Date of issuance: September 15, 1995
Effective date: As of date of issuance, to be implemented within 60
days.
Amendment Nos.: 192 and 75
Facility Operating License Nos. DPR-66 and NPF-73: Amendments
revised the Technical Specifications.
Date of initial notice in Federal Register: August 2, 1995 (60 FR
39436) The August 24, 1995, letter provided typed final TS pages, with
minor editorial changes, for issuance of these amendments. The August
24, 1995, letter did not change the initial proposed no significant
hazards consideration determination or expand the scope of the August
2, 1995, Federal Register notice. The Commission's related evaluation
of the amendments is contained in a Safety Evaluation dated September
15, 1995. No significant hazards consideration comments received: No
Local Public Document Room location: B. F. Jones Memorial Library,
663 Franklin Avenue, Aliquippa, Pennsylvania 15001
Entergy Operations, Inc., Docket No. 50-382, Waterford Steam
ElectricStation, Unit 3, St. Charles Parish, Louisiana
Date of amendment request: June 22, 1994
Brief description of amendment: The amendment changes the Appendix
A Technical Specifications by removing the seismic and meteorological
monitoring instrumentation requirements. These requirements are to be
relocated in the Updated Final Safety Analysis Report.
Date of issuance: September 5, 1995
Effective date: September 5, 1995
Amendment No.: 112
Facility Operating License No. NPF-38. Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: August 3, 1994 (59 FR
39585) The Commission's related evaluation of the amendment is
contained in a Safety Evaluation dated September 5, 1995.No significant
hazards consideration comments received: No
Local Public Document Room location: University of New Orleans
Library, Louisiana Collection, Lakefront, New Orleans, LA 70122
Entergy Operations, Inc., Docket No. 50-382, Waterford Steam
ElectricStation, Unit 3, St. Charles Parish, Louisiana
Date of amendment request: June 22, 1994, and December 9, 1994
Brief description of amendment: The amendment changes the Appendix
A TSs by revising the plant protection system trip setpoints and
allowable
[[Page 49954]]
values such that they will be consistent with the current setpoint/
uncertainty methodology being implemented at Waterford 3.
Date of issuance: September 5, 1995
Effective date: September 5, 1995
Amendment No.: 113
Facility Operating License No. NPF-38. Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: August 3, 1994 (59 FR
39586) and February 1, 1995 (60 FR 6300)The Commission's related
evaluation of the amendment is contained in a Safety Evaluation dated
September 5, 1995.No significant hazards consideration comments
received: No
Local Public Document Room location: University of New Orleans
Library, Louisiana Collection, Lakefront, New Orleans, LA 70122
Entergy Operations, Inc., Docket No. 50-382, Waterford Steam
ElectricStation, Unit 3, St. Charles Parish, Louisiana
Date of amendment request: December 9, 1994, as supplemented by
letter dated July 25, 1995
Brief description of amendment: The requested changes revised the
allowable opening tolerances on the pressurizer safety valves (PSVs)
and the main steam line safety valves (MSSVs) from plus or minus 1
percent to plus or minus 3 percent. However, following testing, the as-
left lift setting of the PSVs and MSSVs will be within plus or minus 1
percent of the pressure specified in the Technical Specifications.
Date of issuance: September 11, 1995
Effective date: September 11, 1995
Amendment No.: 111
Facility Operating License No. NPF-38. Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: February 1, 1995 (60 FR
6300) The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated September 11, 1995.No significant hazards
consideration comments received: No
Local Public Document Room location: University of New Orleans
Library, Louisiana Collection, Lakefront, New Orleans, Louisiana 70122
Florida Power and Light Company, Docket No. 50-335, St. Lucie
Plant, Unit No. 1, St. Lucie County, Florida
Date of application for amendment: May 23, 1994
Brief description of amendment: The amendment revises Technical
Specification 3.5.2 for Emergency Core Cooling Systems (ECCS) by
removing the option that allows High Pressure Safety Injection (HPSI)
Pump 1C to be used as an alternative to the preferred pump for
subsystem operability. HPSI pump 1C is an installed spare which is not
required to be maintained in an operable status, and this change
upgrades the ECCS operability requirements consistent with actual plant
operating needs.
Date of issuance: September 11, 1995
Effective date: September 11, 1995
Amendment No.: 139
Facility Operating License No. DPR-67: Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: July 6, 1994 (59 FR
34663) The Commission's related evaluation of the amendment is
contained in a Safety Evaluation dated September 11, 1995.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 No. 50-389, St.
Lucie Plant, Unit No. 2, St. Lucie County, Florida
Date of application for amendment: February 27, 1995
Brief description of amendment: This amendment will change Table
3.3-3 and 3.3-4 to accommodate an improved coincidence logic and relay
replacement for the 4.16 kV Loss of Voltage Relays. Actions required
for certain trip units with the number of operable channels one less
than the total number of channels will also be changed. In addition,
the format used to state the time delay for the 4.16 kV Degraded
Voltage trip unit will be revised.
Date of issuance: September 1, 1995
Effective date: September 1, 1995
Amendment No.: 79
Facility Operating License No. NPF-16: Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: March 29, 1995 (60 FR
16187) The Commission's related evaluation of the amendment is
contained in a Safety Evaluation dated September 1, 1995.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.
50-321 and 50-366, Edwin I. Hatch Nuclear Plant, Units 1 and 2,
Appling County, Georgia
Date of application for amendments: January 13, 1995, as
supplemented by letters dated April 5 and June 20, 1995.
Brief description of amendments: The amendments modify
Facility Operating License Nos. DRP-57 and NPF-5 and the
corresponding TS for Hatch Units 1 and 2, respectively, to authorize an
increase in the maximum power level from 2436 megawatts thermal (MWt)
to 2558 MWt. The amendments also approve changes to the Technical
Specification to implement uprated power operation.
Date of issuance: August 31, 1995
Effective date: As of the date of issuance to be implemented prior
to startup in Cycle 17 for Unit 1; and prior to startup in Cycle 13 for
Unit 2
Amendment Nos.: 197 and 138
Facility Operating License Nos. DPR-57 and NPF-5. Amendments
revised the Technical Specifications.
Date of initial notice in Federal Register: July 5, 1995 (60 FR
35072) The Commission's related evaluation of the amendments is
contained in a Safety Evaluation dated August 31, 1995 and an
Environmental Assessment dated July 21, 1995.No significant hazards
consideration comments received: No
Local Public Document Room location: Appling County Public Library,
301 City Hall Drive, Baxley, Georgia 31513
Georgia Power Company, Oglethorpe Power Corporation, Municipal
Electric Authority of Georgia, City of Dalton, Georgia, Docket Nos.
50-321 and 50-366, Edwin I. Hatch Nuclear Plant, Units 1 and 2,
Appling County, Georgia
Date of application for amendments: June 6, 1995, as supplemented
August 9, 1995.
Brief description of amendments: The amendments revise Technical
Specification Surveillance Requirements (SR) 3.6.4.1.3 and 3.6.4.1.4
for the secondary containment drawdown. The revision reduces the SR
acceptance criteria to greater than or equal to 0.20 inch water gauge
(wg) negative pressure from greater than or equal to O.25 inch wg
negative pressure. The appropriate TS Bases pages are also changed to
reflect the TS revision.
Date of issuance: September 11, 1995
Effective date: As of the date of issuance to be implemented within
60 days
Amendment Nos.: 198 and 139
[[Page 49955]]
Facility Operating License Nos. DPR-57 and NPF-5. Amendments
revised the Technical Specifications.
Date of initial notice in Federal Register: June 21, 1995 (60 FR
32364) The August 9, 1995, letter provided clarifying information that
did not change the scope of the June 6, 1995, application and the
initial proposed no significant hazards consideration determination.
The Commission's related evaluation of the amendments is contained in a
Safety Evaluation dated September 11, 1995.No significant hazards
consideration comments received: No
Local Public Document Room location: Appling County Public Library,
301 City Hall Drive, Baxley, Georgia 31513
GPU Nuclear Corporation, et al., Docket No. 50-219, Oyster Creek
Nuclear Generating Station, Ocean County, New Jersey
Date of application for amendment: June 26, 1995
Brief description of amendment: The amendment revises the snubber
visual inspection intervals to match the schedule developed by the NRC
staff for use with a 24-month refueling interval. This schedule was
documented in Generic Letter 90-09. The amendment also revises the
bases for the snubber visual inspection interval to be consistent with
the bases described in Generic Letter 90-09.
Date of issuance: September 6, 1995
Effective date: September 6, 1995
Amendment No.: 182
Facility Operating License No. DPR-16. Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: August 2, 1995 (60 FR
39440). The Commission's related evaluation of this amendment is
contained in a Safety Evaluation dated September 6, 1995. No
significant hazards consideration comments received: No
Local Public Document Room location: Ocean County Library,
Reference Department, 101 Washington Street, Toms River, NJ 08753
Gulf States Utilities Company, Cajun Electric Power Cooperative,
and Entergy Operations, Inc., Docket No. 50-458, River Bend
Station, Unit 1, West Feliciana Parish, Louisiana
Date of amendment request: May 25, 1995
Brief description of amendment: The amendment revises the Physical
Security Plan vital island requirements.
Date of issuance: September 12, 1995
Effective date: September 12, 1995
Amendment No.: 83
Facility Operating License No. NPF-47. The amendment revised the
operating license.
Date of initial notice in Federal Register: July 19, 1995 (60 FR
37091) The Commission's related evaluation of the amendment is
contained in a Safety Evaluation dated September 12, 1995.No
significant hazards consideration comments received. No
Local Public Document Room location: Government Documents
Department, Louisiana State University, Baton Rouge, LA 70803
Houston Lighting & Power Company, City Public Service Board of San
Antonio, Central Power and Light Company, City of Austin, Texas,
Docket Nos. 50-498 and 50-499, South Texas Project, Units 1 and 2,
Matagorda County, Texas
Date of amendment request: May 25, 1995, as supplemented by letter
dated August 3, 1995.
Brief description of amendments: The amendments revised the
technical specifications (TSs) on containment leakage, making the
action statement consistent with the need to perform Type C testing at
power, and replacing the surveillance requirements with a single
requirement to apply the requirements of Appendix J as modified by
approved exemptions. The amendments also revised the TSs on containment
integrity, containment leakage, and containment air locks, to eliminate
the numerical value of calculated peak containment internal pressure
related to the design basis accident.
Date of issuance: September 7, 1995
Effective date: September 7, 1995
Amendment Nos.: Unit 1 - Amendment No. 80; Unit 2 - Amendment No.
69
Facility Operating License Nos. NPF-76 and NPF-80. The amendments
revised the Technical Specifications.
Date of initial notice in Federal Register: July 19, 1995 (60 FR
37092) The August 3, 1995, supplement provided 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 September 7, 1995.No significant
hazards consideration comments received: No
Local Public Document Room location: Wharton County Junior College,
J. M. Hodges Learning Center, 911 Boling Highway, Wharton, Texas 77488
Houston Lighting & Power Company, City Public Service Board of San
Antonio, Central Power and Light Company, City of Austin, Texas,
Docket Nos. 50-498 and 50-499, South Texas Project, Units 1 and 2,
Matagorda County, Texas
Date of amendment request: May 31, 1995, as supplemented by letter
dated August 2, 1995
Brief description of amendments: The amendments modified (by
relocation to the Technical Requirements Manual) TS 3/4.1.2.1, Boration
Systems/Flow Paths - Shutdown, TS 3/4.1.2.2, Boration Systems/Flow
Paths - Operating, TS 3/4.1.2.3, Charging Pumps - Shutdown, TS 3/
4.1.2.4, Charging Pumps - Operating, TS 3/4.1.2.5, Borated Water
Sources - Shutdown, TS 3/4.1.2.6, Borated Water Sources - Operating, TS
3/4.4.2.1, Safety Valves - Shutdown, and the associated Bases.
Date of issuance: September 5, 1995
Effective date: September 5, 1995
Amendment Nos.: Unit 1 - Amendment No. 79; Unit 2 - Amendment No.
68
Facility Operating License Nos. NPF-76 and NPF-80. The amendments
revised the Technical Specifications.
Date of initial notice in Federal Register: August 2, 1995 (60 FR
39441) The additional information contained in the supplemental letter
dated August 2, 1995, was clarifying in nature and thus, within the
scope of the initial notice and did not affect the staff's proposed no
significant hazards consideration determination.The Commission's
related evaluation of the amendments is contained in a Safety
Evaluation dated September 5, 1995.No significant hazards consideration
comments received: No
Local Public Document Room location: Wharton County Junior College,
J. M. Hodges Learning Center, 911 Boling Highway, Wharton, TX 77488
Indiana Michigan Power Company, Docket No. 50-315, Donald C. Cook,
Nuclear Plant, Unit No. 1, Berrien County, Michigan
Date of application for amendment: February 3, 1995, as
supplemented April 25, 1995.
Brief description of amendment: The amendment modifies the
technical specifications to extend the interim steam generator tube
plugging criteria used in Cycle 14 to the next operating cycle (Cycle
15).
Date of issuance: September 13, 1995
Effective date: September 13, 1995
Amendment No.: 200
[[Page 49956]]
Facility Operating License No. DPR-58. Amendment revises the
Technical Specifications.
Date of initial notice in Federal Register: July 19, 1995 (60 FR
37093) The Commission's related evaluation of the amendment is
contained in a Safety Evaluation dated September 13, 1995.No
significant hazards consideration comments received: No
Local Public Document Room location: Maud Preston Palenske Memorial
Library, 500 Market Street, St. Joseph, Michigan 49085
Indiana Michigan Power Company, Docket No. 50-316, Donald C. Cook,
Nuclear Plant, Unit No. 2, Berrien County, Michigan
Date of application for amendment: February 15, 1994, as
supplemented June 29, 1995
Brief description of amendment: The amendment deletes Technical
Specification section 3/4.3.4, associated bases, and associated index
listings for the Unit 2 turbine overspeed protection system.
Date of issuance: September 1, 1995
Effective date: September 1, 1995
Amendment No.: 185
Facility Operating License No. DPR-74. Amendment revises the
Technical Specifications.
Date of initial notice in Federal Register: March 30, 1994 (59 FR
14890) The licensee's submittal of June 29, 1995, did not change the
basis for the proposed no significant hazards consideration
determination.The Commission's related evaluation of the amendment is
contained in a Safety Evaluation dated September 1, 1995.No significant
hazards consideration comments received: No
Local Public Document Room location: Maud Preston Palenske Memorial
Library, 500 Market Street, St. Joseph, Michigan 49085
Nebraska Public Power District, Docket No. 50-298, Cooper Nuclear
Station, Nemaha County, Nebraska
Date of amendment request: June 15, 1995
Brief description of amendment: The amendment changes the Technical
Specifications to revise the definition for logic system functional
test and revises the surveillance interval for emergency core cooling
system logic system functional testing from 6 months to 18 months.
Date of issuance: September 7, 1995
Effective date: September 7, 1995
Amendment No.: 171
Facility Operating License No. DPR-46. Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: July 19, 1995 (60 FR
37096) The Commission's related evaluation of the amendment is
contained in a Safety Evaluation dated September 7, 1995.No significant
hazards consideration comments received: No
Local Public Document Room location: Auburn Public Library, 118
15th Street, Auburn, NE 68305
Niagara Mohawk Power Corporation, Docket No. 50-410, Nine Mile
PointNuclear Station, Unit 2, Oswego County, New York
Date of application for amendment: January 6, 1995
Brief description of amendment: The amendment incorporates Limiting
Condition for Operation 3.3.3.1 from Standard Technical Specifications
into Technical Specification (TS) 3/4.3.7.5, Accident Monitoring
Instrumentation and make associated changes in TS 3/4.4.2, Safety
Relief Valves.
Date of issuance: September 11, 1995
Effective date: As of the date of issuance to be implemented within
30 days.
Amendment No.: 69
Facility Operating License No. NPF-69: Amendment revises the
Technical Specifications.
Date of initial notice in Federal Register: February 15, 1995 (60
FR 8748)The Commission's related evaluation of the amendment is
contained in a Safety Evaluation dated September 11, 1995.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
PointNuclear Station, Unit 2, Oswego County, New York
Date of application for amendment: January 6, 1995, as supplemented
April 18, 1995
Brief description of amendment: The amendment revises Technical
Specifications (TSs) Sections 3.8.1.1 and 3.8.1.2; TS Surveillance
Requirements Section 4.8.1.1.2; TS Bases Section 3/4.8.1.3; and TS
Administrative Controls Section 6.8.4. The changes include: updating
the minimum day tank and storage tank oil inventory, specific actions
required if oil level fall below minimum required, revising and
relocating the fuel oil sampling and testing criteria to the associated
Bases, and specific action to be taken if the fuel oil properties do
not meet the specified limits. In addition, a requirement was added for
a diesel fuel oil testing program. These changes are consistent with
guidance provided in NUREG-1434.
Date of issuance: September 15, 1995
Effective date: As of the date of issuance to be implemented within
30 days.
Amendment No.: 70
Facility Operating License No. NPF-69: Amendment revises the
Technical Specifications.
Date of initial notice in Federal Register: February 15, 1995 (60
FR 8747) The Commission's related evaluation of the amendment is
contained in a Safety Evaluation dated September 15, 1995.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
North Atlantic Energy Service Corporation, Docket No. 50-443,
Seabrook Station, Unit No. 1, Rockingham County, New Hampshire
Date of amendment request: April 16, 1995.
Description of amendment request: The amendment revises the
Appendix A Technical Specifications (TS) relating to containment
building penetrations. Specifically, the amendment modifies Limiting
Conditions for Operation 3.9.4 to permit both doors of one personnel
airlock to be open during core alterations or irradiated fuel movement
if certain conditions are met and to add equivalent and alternate
penetration closure methodologies. Surveillance Requirement 4.9.4 is
changed to reflect that the penetrations are to be verified to be in
the condition required. Bases Section 3/4 9.4 also is revised to
reflect the changes described above.
Date of issuance: August 31, 1995
Effective date: As of the date of issuance, to be implemented
within 60 days.
Amendment No.: 40
Facility Operating License No. NPF-86. Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: June 21, 1995 (60 FR
32369) The Commission's related evaluation of the amendment is
contained in a Safety Evaluation dated August 31, 1995.No significant
hazards consideration comments received: No
Local Public Document Room location: Exeter Public Library,
Founders Park, Exeter, NH 03833
[[Page 49957]]
North Atlantic Energy Service Corporation, Docket No. 50-443,
Seabrook Station, Unit No. 1, Rockingham County, New Hampshire
Date of amendment request: May 30, 1995.
Description of amendment request: The amendment revises the
Appendix A Technical Specifications (TS) relating to Moderator
Temperature Coefficient. The amendment changes the upper limit for the
moderator temperature coefficient (MTC) for certain operating
conditions. Additionally, a reference for the analytical method used to
determine the cycle-specific MTC upper limit is added to TS 6.8.1.6.b.
Date of issuance: September 14, 1995
Effective date: As of the date of issuance to be implemented within
60 days.
Amendment No.: 41
Facility Operating License No. NPF-86. Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: July 5, 1995 (60 FR
35082). The Commission's related evaluation of the amendment is
contained in a Safety Evaluation dated September 14, 1995. No
significant hazards consideration comments received: No
Local Public Document Room location: Exeter Public Library,
Founders Park, Exeter, NH 03833
North Atlantic Energy Service Corporation, Docket No. 50-443,
Seabrook Station, Unit No. 1, Rockingham County, New Hampshire
Date of amendment request: June 16, 1995
Description of amendment request: The amendment revises the
Appendix A Technical Specifications (TS) relating to core reactivity
control available from borated water sources. The amendment changes the
minimum boron concentration specified for the refueling water storage
tank (RWST) in Limiting Condition for Operation (LCO) in TS 3.1.2.5 and
replaces the minimum specified concentration for boron with an
acceptable range of boron concentration for the RWST and the
accumulators in the LCOs for TS 3.1.2.6, 3.5.1.1, and 3.5.4.
Date of issuance: September 14, 1995
Effective date: As of the date of issuance, to be implemented prior
to entering MODE 4 following the fourth refueling outage.
Amendment No.: 42
Facility Operating License No. NPF-86. Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: August 2, 1995 (60 FR
39442). The Commission's related evaluation of the amendment is
contained in a Safety Evaluation dated September 14, 1995. No
significant hazards consideration comments received: No
Local Public Document Room location: Exeter Public Library,
Founders Park, Exeter, NH 03833
Northeast Nuclear Energy Company, Docket No. 50-245,
MillstoneNuclear Power Station, Unit 1, New London County,
Connecticut
Date of application for amendment: July 11, 1995
Brief description of amendment: The amendment modifies Technical
Specification 3.5.F.7 to also allow the use of pull-to-lock switches to
defeat the automatic initiation of the emergency core cooling system
while in the refuel condition. The amendment also makes editorial
corrections and makes changes to the associated Bases section.
Date of issuance: September 13, 1995
Effective date: As of the date of issuance to be implemented within
30 days.
Amendment No.: 88
Facility Operating License No. DPR-21. Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: August 2, 1995 (60 FR
39442). The Commission's related evaluation of the amendment is
contained in a Safety Evaluation dated September 13, 1995.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
Northeast Nuclear Energy Company, Docket No. 50-245,
MillstoneNuclear Power Station, Unit 1, New London County,
Connecticut
Date of application for amendment: July 18, 1995
Brief description of amendment: The amendment adds operability and
surveillance requirements for reactor pressure vessel overfill
protection instrumentation. The amendment also adds the associated
Bases.
Date of issuance: September 13, 1995
Effective date: As of the date of issuance to be implemented within
60 days.
Amendment No.: 87
Facility Operating License No. DPR-21. Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: August 2, 1995 (60 FR
39443) The Commission's related evaluation of the amendment is
contained in a Safety Evaluation dated September 13, 1995.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
Omaha Public Power District, Docket No. 50-285, Fort Calhoun
Station,Unit No. 1, Washington County, Nebraska
Date of amendment request: April 7, 1995
Brief description of amendment: The amendment revises the technical
specifications (TS) to relocate the axial power distribution limits to
the Core Operating Limits Report (COLR).
Date of issuance: September 1, 1995
Effective date: September 1, 1995
Amendment No.: 170
Facility Operating License No. DPR-40. Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: May 23, 1995 (60 FR
27339) The Commission's related evaluation of the amendment is
contained in a Safety Evaluation dated September 1, 1995.No significant
hazards consideration comments received: No
Local Public Document Room location: W. Dale Clark Library, 215
South 15th Street, Omaha, Nebraska 68102
Omaha Public Power District, Docket No. 50-285, Fort Calhoun
Station,Unit No. 1, Washington County, Nebraska
Date of amendment request: May 8, 1995, as supplemented by letter
dated July 11, 1995.
Brief description of amendment: The amendment changes Sections 2.3,
3.1, 3.2, 3.3, and 3.6 of the Technical Specifications in accordance
with the guidance of Generic Letter (GL) 93-05, ``Line Item Technical
Specifications Improvements to Reduce Surveillance Requirements for
Testing During Power Operation,'' dated September 27, 1993. The changes
are consistent with Station operating experience and NUREG-1366,
``Improvements to Technical Specifications Surveillance Requirements,''
dated December 1992. In addition, a change was made to TS Section 3.1
in accordance with the Commission's Final Policy Statement on Technical
Specifications Improvements for Nuclear Power Reactors. Also, changes
were made to the TS sections identified above for clarity and to
correct administrative errors.
Date of issuance: September 7, 1995
Effective date: September 7, 1995
Amendment No.: 171
[[Page 49958]]
Facility Operating License No. DPR-40. Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: June 6, 1995 (60 FR
29883) The Commission's related evaluation of the amendment is
contained in a Safety Evaluation dated September 7, 1995.No significant
hazards consideration comments received: No
Local Public Document Room location: W. Dale Clark Library, 215
South 15th Street, Omaha, Nebraska 68102
Pennsylvania Power and Light Company, Docket No. 50-387,
Susquehanna Steam Electric Station, Unit 1, Luzerne County,
Pennsylvania
Date of application for amendment: April 11, 1995
Brief description of amendment: This amendment extends on a one-
time basis the allowed outage time from 3 to 7 days for one offsite
circuit being out of service.
Date of issuance: August 31, 1995
Effective date: As of the date of issuance and is to be implemented
within 30 days.
Amendment No.: 153
Facility Operating License No. NPF-14: This amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: June 6, 1995 (60 FR
29886). The Commission's related evaluation of the amendment is
contained in a Safety Evaluation dated August 31, 1995.No significant
hazards consideration comments received: No
Local Public Document Room location: Osterhout Free Library,
Reference Department, 71 South Franklin Street, Wilkes-Barre,
Pennsylvania 18701
Pennsylvania Power and Light Company, Docket Nos. 50-387 and 50-388
Susquehanna Steam Electric Station, Units 1 and 2, Luzerne County,
Pennsylvania
Date of application for amendments: February 2, 1995
Brief description of amendments: These amendments change the
Technical Specifications for the two Susquehanna units to increase the
licensed discharge fuel assembly for SPC 9X9-2 fuel from 40 to 45 GWD/
MTU. This change is consistent with the Commissions approval of Topical
Report PL-NF-94-005-P, ``Technical Basis for SPC 9X9-2 Extended Fuel
Exposure at Susquehanna SES,'' documented in a letter to PP&L dated
December 15, 1994.
Date of issuance: September 12, 1995
Effective date: September 12, 1995
Amendment Nos.: 154 and 124
Facility Operating License Nos. NPF-14 and NPF-22. The amendments
revised the Technical Specifications.
Date of initial notice in Federal Register: March 29, 1995 (60 FR
16194) The Commission's related evaluation of the amendments is
contained in a Safety Evaluation dated September 12, 1995.No
significant hazards consideration comments received: No
Local Public Document Room location: Osterhout Free Library,
Reference Department, 71 South Franklin Street, Wilkes-Barre,
Pennsylvania 18701
Pennsylvania Power and Light Company, Docket No. 50-388,
Susquehanna Steam Electric Station, Unit 2, Luzerne County,
Pennsylvania
Date of application for amendment: March 31, 1995
Brief description of amendment: This amendment changes Technical
Specification Section 6.9.3.2 to allow four GE demonstration assemblies
to be loaded into Susquehanna Unit 2, Cycle 8 core.
Date of issuance: September 13, 1995
Effective date: As of date of issuance and shall be implemented
within 30 days.
Amendment No.: 125
Facility Operating License No. NPF-22. This amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: April 26, 1995 (60 FR
20523) The Commission's related evaluation of the amendment is
contained in a Safety Evaluation dated September 13, 1995.No
significant hazards consideration comments received: No
Local Public Document Room location: Osterhout Free Library,
Reference Department, 71 South Franklin Street, Wilkes-Barre,
Pennsylvania 18701
Pennsylvania Power and Light Company, Docket No. 50-388,
Susquehanna Steam Electric Station, Unit 2, Luzerne County,
Pennsylvania
Date of application for amendment: May 5, 1995, and supplemented by
letter dated August 18, 1995
Brief description of amendment: This amendment deletes from SSES
Technical Specification Table 3.6.3-1, ``Primary Containment Isolation
Valves,'' three relief valves in the residual heat removal system.
These specific valves which were originally intended to support the
steam condensing mode, were previously eliminated from the plant
design. The valves are being replaced during the September Unit 2
refueling outage and will be replaced by blind flanges.
Date of issuance: September 11, 1995
Effective date: September 11, 1995
Amendment No.: 123
Facility Operating License No. NPF-22. This amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: July 5, 1995 (60 FR
35083 and July 17, 1995 (60 FR 36449)The supplemental letter provided
clarifying information that did not change the initial proposed no
significant hazards consideration determination or expand the scope of
the original Federal Register notice. The Commission's related
evaluation of the amendment is contained in a Safety Evaluation dated
September 11, 1995.No significant hazards consideration comments
received: No
Local Public Document Room location: Osterhout Free Library,
Reference Department, 71 South Franklin Street, Wilkes-Barre,
Pennsylvania 18701
Power Authority of The State of New York, Docket No. 50-286, Indian
Point Nuclear Generating Unit No. 3, Westchester County, New York
Date of application for amendment: February 23, 1995, as
supplemented July 28, 1995
Brief description of amendment: The amendment revised the minimum
emergency diesel generator (EDG) fuel oil requirements, as indicated in
Technical Specification (TS) Section 3.7 (Auxiliary Electrical
Systems), from 7056 to 6671 gallons. The actual minimum fuel oil level
had always been 6671 gallons; however, the previous TS limit of 7056
gallons was based on a level indicator that had an accuracy of +/- 385
gallons. This revision clarified the TS such that any level indicator
can now be used as long as an actual minimum level of 6671 gallons is
assured.
Date of issuance: August 30, 1995
Effective date: As of the date of issuance to be implemented within
30 days.
Amendment No.: 161
Facility Operating License No. DPR-64: Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: March 29, 1995 (60 FR
16196) The Commission's related evaluation of the amendment is
contained in a Safety Evaluation dated August 30, 1995.No significant
hazards consideration comments received: No
Local Public Document Room location: White Plains Public Library,
100 Martine Avenue, White Plains, New York 10610
[[Page 49959]]
Power Authority of The State of New York, Docket No. 50-286, Indian
Point Nuclear Generating Unit No. 3, Westchester County, New York
Date of application for amendment: March 2, 1995
Brief description of amendment: The amendment revised the titles of
several management positions as described in Technical Specifications
Section 6.0 (Administrative Controls). Specifically, the title of
Executive Vice President and Chief Nuclear Officer and the title of
Shift Supervisor were changed to Chief Nuclear Officer and Shift
Manager, respectively. In addition, the position titles of Senior
Reactor Operator and Reactor Operator were deleted and replaced with
qualification requirements.
Date of issuance: August 31, 1995
Effective date: As of the date of issuance to be implemented within
30 days.
Amendment No.: 162
Facility Operating License No. DPR-64: Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: March 29, 1995 (60 FR
16197) The Commission's related evaluation of the amendment is
contained in a Safety Evaluation dated August 31, 1995.No significant
hazards consideration comments received: No
Local Public Document Room location: White Plains Public Library,
100 Martine Avenue, White Plains, New York 10610
Power Authority of the State of New York, Docket No. 50-333, James
A. FitzPatrick Nuclear Power Plant, Oswego County, New York
Date of application for amendment: October 3, 1994
Brief description of amendment: The amendment proposed changes to
FitzPatrick TSs which will extend the instrumentation functional test
interval and allowable out-of-service times, remove the average power
range monitor downscale scram function and the instrument response time
values, and incorporate several editorial, clarification, and
correction changes.
Date of issuance: September 11, 1995
Effective date: As of the date of issuance to be implemented within
30 days.
Amendment No.: 227
Facility Operating License No. DPR-59: Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: November 9, 1994 (59 FR
55887) The Commission's related evaluation of the amendment is
contained in a Safety Evaluation dated September 11, 1995.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
Public Service Electric & Gas Company, Docket No. 50-354, Hope
Creek Generating Station, Salem County, New Jersey
Date of application for amendment: September 29, 1994
Brief description of amendment: This amendment revises Table 4.3.6-
1, ``Control Rod Block Instrumentation Surveillance,'' of the Hope
Creek TS. The channel calibration frequencies for the Source Range
Monitor (SRM) and the Intermediate Range Monitor (IRM), in TS Table
4.3.6-1, are changed for the up-scale and the down-scale trip functions
on each instrument from ``SA'' (once-per-184 days) to ``R'' (once-per-
refueling cycle).
Date of issuance: September 12, 1995
Effective date: As of the date of issuance and shall be implemented
within 60 days.
Amendment No.: 78
Facility Operating License No. NPF-57: This amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: June 18, 1995 (60 FR
3676). The Commission's related evaluation of the amendment is
contained in a Safety Evaluation dated September 12, 1995. No
significant hazards consideration comments received: No
Local Public Document Room location: Pennsville Public Library, 190
S. Broadway, Pennsville, New Jersey 08070
Public Service Electric & Gas Company, Docket Nos. 50-272 and 50-
311, Salem Nuclear Generating Station, Unit Nos. 1 and 2, Salem
County, New Jersey
Date of application for amendments: May 2, 1995
Brief description of amendments: The amendments eliminate the
monthly manual initiation of auxiliary feedwater from Technical
Specification Tables 3.3-3, 3.3.-4 and 4.3-2.
Date of issuance: September 6, 1995
Effective date: Units 1 and 2, as of the date of issuance, to be
implemented within 60 days.
Amendment Nos. 175 and 156
Facility Operating License Nos. DPR-70 and DPR-75. The amendments
revised the Technical Specifications.
Date of initial notice in Federal Register: June 6, 1995 (60 FR
29887)The Commission's related evaluation of the amendments is
contained in a Safety Evaluation dated September 6, 1995.No significant
hazards consideration comments received: No
Local Public Document Room location: Salem Free Public Library, 112
West Broadway, Salem, New Jersey 08079
South Carolina Electric & Gas Company, South Carolina Public
ServiceAuthority, Docket No. 50-395, Virgil C. Summer Nuclear
Station, Unit No. 1, Fairfield County, South Carolina
Date of application for amendment: June 19, 1995
Brief description of amendment: The amendment restructures the
primary containment and primary containment leakage technical
specifications to reduce the repetition of those requirements contained
in NRC regulations such as Appendix J to 10 CFR Part 50.
Date of issuance: September 5, 1995
Effective date: September 5, 1995
Amendment No.: 126
Facility Operating License No. NPF-12: Amendment revises the
Technical Specifications.
Date of initial notice in Federal Register: July 19, 1995 (60 FR
37099)The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated September 5, 1995.No significant hazards
consideration comments received: No
Local Public Document Room location: Fairfield County Library, 300
Washington Street, Winnsboro, SC 29180
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 application for amendments: May 20, 1994
Brief description of amendments: The amendments revise Technical
Specification 3/4.7.3, ``Component Cooling Water System,'' and the
corresponding Bases to support the addition of the component cooling
water surge tank backup nitrogen supply (BNS) system.
Date of issuance: September 13, 1995
Effective date: September 13, 1995
Amendment Nos.: Unit 2 - Amendment No. 125; Unit 3 - Amendment No.
114
Facility Operating License Nos. NPF-10 and NPF-15: The amendments
revised the Technical Specifications.
Date of initial notice in Federal Register: August 31, 1994 (59 FR
[[Page 49960]]
45034)The Commission's related evaluation of the amendments is
contained in a Safety Evaluation dated September 13, 1995.No
significant hazards consideration comments received: No
Local Public Document Room location: Main Library, University of
California, P.O. Box 19557, Irvine, California 92713
Tennessee Valley Authority, Docket Nos. 50-259, 50-260, and 50-296,
Browns Ferry Nuclear Plant, Units 1, 2, and 3, Limestone County,
Alabama
Date of application for amendments: March 31, 1995, supplemented
July 14, 1995 (TS 349)
Brief description of amendment: These amendments revise the Browns
Ferry Nuclear Plant (BFN) Units 1, 2, and 3 reactor vessel pressure-
temperature curves and bolt-up temperatures.
Date of issuance: September 13, 1995
Effective date: September 13, 1995
Amendment Nos.: 224, 239, 198
Facility Operating License Nos. DPR-33, DPR-52 and DPR-68:
Amendments revised the Technical Specifications.
Date of initial notice in Federal Register: June 6, 1995 (60 FR
29888)The July 14, 1995 letter provided clarifying information that did
not change the initial proposed no significant hazards consideration
determination. The Commission's related evaluation of the amendment is
contained in a Safety Evaluation dated September 13, 1995.No
significant hazards consideration comments received: No
Local Public Document Room location: Athens Public library, South
Street, Athens, Alabama 35611
Tennessee Valley Authority, Docket Nos. 50-259, 50-260, and 50-296,
Browns Ferry Nuclear Plant, Units 1, 2, and 3, Limestone County,
Alabama
Date of application for amendments: May 11, 1995, supplemented
June 30, 1995 (TS 359)
Brief description of amendment: The amendments provide for the
addition of a reactor trip on low scram pilot air header pressure for
BFN Unit 3, and revise a note regarding instrumentation requirements
for all three BFN reactors.
Date of issuance: August 29, 1995
Effective date: August 29, 1995
Amendment Nos.: 223, 228 and 197
Facility Operating License Nos. DPR-33, DPR-52 and DPR-68:
Amendments revised the Technical Specifications.
Date of initial notice in Federal Register: June 6, 1995 (60 FR
29889)The June 30, 1995 letter provided clarifying information that did
not change the initial proposed no significant hazards consideration
determination. The Commission's related evaluation of the amendment is
contained in a Safety Evaluation dated August 29, 1995.No significant
hazards consideration comments received: No
Local Public Document Room location: Athens Public library, South
Street, Athens, Alabama 35611
Tennessee Valley Authority, Docket Nos. 50-327 and 50-328, Sequoyah
Nuclear Plant, Units 1 and 2, Hamilton County, Tennessee
Date of application for amendments: June 29, 1995 (TS 95-14)
Brief description of amendments: The amendments revise Technical
Specification 3.9.4, Containment Building Penetrations, to allow both
sets of containment personnel airlock doors to be open during core
alterations and fuel movement provided one door is capable of closure
and one train of auxiliary building gas treatment remains operable.
Date of issuance: September 6, 1995
Effective date: September 6, 1995
Amendment Nos.: 209 and 199
Facility Operating License Nos. DPR-77 and DPR-79: Amendments
revise the technical specifications.
Date of initial notice in Federal Register: July 19, 1995 (60 FR
37100)The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated September 6, 1995. No significant hazards
consideration comments received: No
Local Public Document Room location: Chattanooga-Hamilton County
Library,1101 Broad Street, Chattanooga, Tennessee 37402
Tennessee Valley Authority, Docket Nos. 50-327 and 50-328, Sequoyah
Nuclear Plant, Units 1 and 2, Hamilton County, Tennessee
Date of application for amendments: August 7, 1995 (TS 95-11)
Brief description of amendments: The amendments revise the time
constant used in the overtemperature delta temperature and the
overpower delta temperature trip equations of Technical Specification
Table 2.2-1.
Date of issuance: September 15, 1995
Effective date: September 15, 1995
Amendment Nos.: 211 and 201
Facility Operating License Nos. DPR-77 and DPR-79: Amendments
revise the technical specifications.
Date of initial notice in Federal Register: April 26, 1995 (60 FR
20527); superseded August 15, 1995 (60 FR 42187) The Commission's
related evaluation of the amendment is contained in a Safety Evaluation
dated September 15, 1995.No significant hazards consideration comments
received: No
Local Public Document Room location: Chattanooga-Hamilton County
Library,1101 Broad Street, Chattanooga, Tennessee 37402
Tennessee Valley Authority, Docket Nos. 50-327 and 50-328, Sequoyah
Nuclear Plant, Units 1 and 2, Hamilton County, Tennessee
Date of application for amendments: August 21, 1995 (TS 95-21)
Brief description of amendments: The amendments change Technical
Specification 3.7.5.c to allow an increase in the average essential raw
cooling water supply header temperature from 84.5 deg.F to 87 deg.F
untilSeptember 30, 1995.
Date of issuance: September 13, 1995
Effective date: September 13, 1995
Amendment Nos.: 210 and 200
Facility Operating License Nos. DPR-77 and DPR-79: Amendments
revise the technical specifications.Public comments requested as to
proposed no significant hazards consideration: Yes (August 28, 1995, 60
FR 44517). That notice provided an opportunity to submit comments on
the Commission's proposed no significant hazards determination. No
comments have been received. The notice also provided an opportunity to
request a hearing, by September 12, 1995, but indicated that if the
Commission makes a final no significant hazards consideration
determination before the expiration of the notice period, any such
hearing would take place after issuance of the amendments.The
Commission's related evaluation of the amendment, finding of exigent
circumstances, and final determination of no significant hazards
consideration are contained in a Safety Evaluation dated September 13,
1995.
Local Public Document Room location: Chattanooga-Hamilton County
Library,1101 Broad Street, Chattanooga, Tennessee 37402
The Cleveland Electric Illuminating Company, Centerior Service
Company, Duquesne Light Company, Ohio Edison Company, Pennsylvania
Power Company, Toledo Edison Company, Docket No. 50-440, Perry
Nuclear Power Plant, Unit No. 1, Lake County, Ohio
Date of application for amendment: April 28, 1995
Brief description of amendment: The amendment extends for one
additional operating cycle the exception to Limiting Condition for
Operation 3.0.4 as it applies to the main steam isolation
[[Page 49961]]
valve leakage control system Technical Specification.
Date of issuance: September 8, 1995
Effective date: September 8, 1995
Amendment No.: 71
Facility Operating License No. NPF-58: This amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: May 23, 1995 (60 FR
27344)The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated September 8, 1995. No significant hazards
consideration comments received: No
Local Public Document Room location: Perry Public Library, 3753
Main Street, Perry, Ohio 44081
Toledo Edison Company, Centerior Service Company, and The Cleveland
Electric Illuminating Company, Docket No. 50-346, Davis-Besse
Nuclear Power Station, Unit No. 1, Ottawa County, Ohio
Date of application for amendment: April 10, 1995
Brief description of amendment: This amendment changes auxiliary
feedwater system, motor driven feedwater pump, and condensate system
Technical Specifications to increase clarity and changes format to more
closely follow improved standard technical specifications and increases
content of Bases discussions.
Date of issuance: September 5, 1995
Effective date: September 5, 1995
Amendment No.: 200
Facility Operating License No. NPF-3. Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: August 2, 1995 (60 FR
39453) The Commission's related evaluation of the amendment is
contained in a Safety Evaluation dated September 5, 1995.No significant
hazards consideration comments received: No
Local Public Document Room location: University of Toledo Library,
Documents Department, 2801 Bancroft Avenue, Toledo, Ohio 43606
Virginia Electric and Power Company, Docket Nos. 50-280 and 50-281,
Surry Power Station, Unit Nos. 1 and 2, Surry County, Virginia.
Date of application for amendments: November 10, 1994
Brief description of amendments: Clarify the surveillance
requirementsfor the Reactor Protection and Engineered Safeguards
Systems instrumentation and actuation logic.
Date of issuance: September 14, 1995
Effective date: September 14, 1995
Amendment Nos.: 205 and 205
Facility Operating License Nos. DPR-32 and DPR-37: Amendments
revised the Technical Specifications.
Date of initial notice in Federal Register: April 12, 1995 (60 FR
18630) The Commission's related evaluation of the amendment is
contained in a Safety Evaluation dated September 14, 1995.No
significant hazards consideration comments received: No
Local Public Document Room location: Swem Library, College of
William and Mary, Williamsburg, Virginia 23185
Virginia Electric and Power Company, Docket Nos. 50-280 and 50-281,
Surry Power Station, Unit Nos. 1 and 2, Surry County, Virginia.
Date of application for amendments: July 14, 1995
Brief description of amendments: These amendments would provide a
2-hour allowed outage time for one residual heat removal loop to
accommodate plant safety and emergency power systems surveillance
testing, permit depressurizing safety injection accumulators in lieu of
accumulator isolation, and make administrative changes.
Date of issuance: September 1, 1995
Effective date: September 1, 1995
Amendment Nos.: 204 and 204
Facility Operating License Nos. DPR-32 and DPR-37: Amendments
revised the Technical Specifications.
Date of initial notice in Federal Register: August 2, 1995 (60 FR
39455) The Commission's related evaluation of the amendment is
contained in a Safety Evaluation dated September 1, 1995.No significant
hazards consideration comments received: No
Local Public Document Room location: Swem Library, College of
William and Mary, Williamsburg, Virginia 23185
Wisconsin Public Service Corporation, Docket No. 50-305, Kewaunee
NuclearPower Plant, Kewaunee County, Wisconsin
Date of application for amendment: December 16, 1994.
Brief description of amendment: The amendment revises the Kewaunee
Nuclear Power Plant (KNPP) Technical Specifications Sections 3.4 and
4.1 by removing the limiting conditions for operation (LCO) and the
surveillance requirements for the turbine overspeed protection system
(TOPS). The TOPS requirements will be relocated to the Updated Safety
Analysis Report (USAR).
Date of issuance: August 31, 1995
Effective date: August 31, 1995
Amendment No.: 121
Facility Operating License No. DPR-43. Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: January 18, 1995 (60 FR
3676). The Commission's related evaluation of the amendment is
contained in a Safety Evaluation dated August 31, 1995.No significant
hazards consideration comments received: No
Local Public Document Room location: University of Wisconsin
Library Learning Center, 2420 Nicolet Drive, Green Bay, Wisconsin 54301
Notice Of Issuance Of Amendments To Facility Operating Licenses And
Final Determination Of No Significant Hazards Consideration And
Opportunity For A Hearing (Exigent Public Announcement Or Emergency
Circumstances)
During the period since publication of the last biweekly notice,
the Commission has issued the following amendments. The Commission has
determined for each of these amendments that the application for the
amendment complies with the standards and requirements of the Atomic
Energy Act of 1954, as amended (the Act), and the Commission's rules
and regulations. The Commission has made appropriate findings as
required by the Act and the Commission's rules and regulations in 10
CFR Chapter I, which are set forth in the license amendment.
Because of exigent or emergency circumstances associated with the
date the amendment was needed, there was not time for the Commission to
publish, for public comment before issuance, its usual 30-day Notice of
Consideration of Issuance of Amendment, Proposed No Significant Hazards
Consideration Determination, and Opportunity for a Hearing.
For exigent circumstances, the Commission has either issued a
Federal Register notice providing opportunity for public comment or has
used local media to provide notice to the public in the area
surrounding a licensee's facility of the licensee's application and of
the Commission's proposed determination of no significant hazards
consideration. The Commission has provided a reasonable opportunity for
the public to comment, using its best efforts to make available to the
public means of communication for the public to respond quickly, and in
the case of telephone comments, the comments have been recorded or
transcribed as appropriate and the licensee has been informed of the
public comments.
In circumstances where failure to act in a timely way would have
resulted, for
[[Page 49962]]
example, in derating or shutdown of a nuclear power plant or in
prevention of either resumption of operation or of increase in power
output up to the plant's licensed power level, the Commission may not
have had an opportunity to provide for public comment on its no
significant hazards consideration determination. In such case, the
license amendment has been issued without opportunity for comment. If
there has been some time for public comment but less than 30 days, the
Commission may provide an opportunity for public comment. If comments
have been requested, it is so stated. In either event, the State has
been consulted by telephone whenever possible.
Under its regulations, the Commission may issue and make an
amendment immediately effective, notwithstanding the pendency before it
of a request for a hearing from any person, in advance of the holding
and completion of any required hearing, where it has determined that no
significant hazards consideration is involved.
The Commission has applied the standards of 10 CFR 50.92 and has
made a final determination that the amendment involves no significant
hazards consideration. The basis for this determination is contained in
the documents related to this action. Accordingly, the amendments have
been issued and made effective as indicated.
Unless otherwise indicated, the Commission has determined that
these amendments satisfy the criteria for categorical exclusion in
accordance with 10 CFR 51.22. Therefore, pursuant to 10 CFR 51.22(b),
no environmental impact statement or environmental assessment need be
prepared for these amendments. If the Commission has prepared an
environmental assessment under the special circumstances provision in
10 CFR 51.12(b) and has made a determination based on that assessment,
it is so indicated.
For further details with respect to the action see (1) the
application for amendment, (2) the amendment to Facility Operating
License, and (3) the Commission's related letter, Safety Evaluation
and/or Environmental Assessment, as indicated. All of these items are
available for public inspection at the Commission's Public Document
Room, the Gelman Building, 2120 L Street, NW., Washington, DC, and at
the local public document room for the particular facility involved.
The Commission is also offering an opportunity for a hearing with
respect to the issuance of the amendment. By October 27, 1995, the
licensee may file a request for a hearing with respect to issuance of
the amendment to the subject facility operating license and any person
whose interest may be affected by this proceeding and who wishes to
participate as a party in the proceeding must file a written request
for a hearing and a petition for leave to intervene. Requests for a
hearing and a petition for leave to intervene shall be filed in
accordance with the Commission's ``Rules of Practice for Domestic
Licensing Proceedings'' in 10 CFR Part 2. Interested persons should
consult a current copy of 10 CFR 2.714 which is available at the
Commission's Public Document Room, the Gelman Building, 2120 L Street,
NW., Washington, DC and at the local public document room for the
particular facility involved. If a request for a hearing or petition
for leave to intervene is filed by the above date, the Commission or an
Atomic Safety and Licensing Board, designated by the Commission or by
the Chairman of the Atomic Safety and Licensing Board Panel, will rule
on the request and/or petition; and the Secretary or the designated
Atomic Safety and Licensing Board will issue a notice of a hearing or
an appropriate order.
As required by 10 CFR 2.714, a petition for leave to intervene
shall set forth with particularity the interest of the petitioner in
the proceeding, and how that interest may be affected by the results of
the proceeding. The petition should specifically explain the reasons
why intervention should be permitted with particular reference to the
following factors: (1) the nature of the petitioner's right under the
Act to be made a party to the proceeding; (2) the nature and extent of
the petitioner's property, financial, or other interest in the
proceeding; and (3) the possible effect of any order which may be
entered in the proceeding on the petitioner's interest. The petition
should also identify the specific aspect(s) of the subject matter of
the proceeding as to which petitioner wishes to intervene. Any person
who has filed a petition for leave to intervene or who has been
admitted as a party may amend the petition without requesting leave of
the Board up to 15 days prior to the first prehearing conference
scheduled in the proceeding, but such an amended petition must satisfy
the specificity requirements described above.
Not later than 15 days prior to the first prehearing conference
scheduled in the proceeding, a petitioner shall file a supplement to
the petition to intervene which must include a list of the contentions
which are sought to be litigated in the matter. Each contention must
consist of a specific statement of the issue of law or fact to be
raised or controverted. In addition, the petitioner shall provide a
brief explanation of the bases of the contention and a concise
statement of the alleged facts or expert opinion which support the
contention and on which the petitioner intends to rely in proving the
contention at the hearing. The petitioner must also provide references
to those specific sources and documents of which the petitioner is
aware and on which the petitioner intends to rely to establish those
facts or expert opinion. Petitioner must provide sufficient information
to show that a genuine dispute exists with the applicant on a material
issue of law or fact. Contentions shall be limited to matters within
the scope of the amendment under consideration. The contention must be
one which, if proven, would entitle the petitioner to relief. A
petitioner who fails to file such a supplement which satisfies these
requirements with respect to at least one contention will not be
permitted to participate as a party.
Those permitted to intervene become parties to the proceeding,
subject to any limitations in the order granting leave to intervene,
and have the opportunity to participate fully in the conduct of the
hearing, including the opportunity to present evidence and cross-
examine witnesses. Since the Commission has made a final determination
that the amendment involves no significant hazards consideration, if a
hearing is requested, it will not stay the effectiveness of the
amendment. Any hearing held would take place while the amendment is in
effect.
A request for a hearing or a petition for leave to intervene must
be filed with the Secretary of the Commission, U.S. Nuclear Regulatory
Commission, Washington, DC 20555, 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
[[Page 49963]]
Regulatory Commission, Washington, DC 20555, and to the attorney for
the licensee.
Nontimely filings of petitions for leave to intervene, amended
petitions, supplemental petitions and/or requests for a hearing will
not be entertained absent a determination by the Commission, the
presiding officer or the Atomic Safety and Licensing Board that the
petition and/or request should be granted based upon a balancing of the
factors specified in 10 CFR 2.714(a)(1)(i)-(v) and 2.714(d).
Niagara Mohawk Power Corporation, Docket No. 50-410, Nine Mile
PointNuclear Station, Unit 2, Oswego County, New York
Date of application for amendment: August 28, 1995
Brief description of amendment: The amendment revises Primary
Containment Purge System Technical Specification Section 3.6.1.7,
Limiting Condition for Operation. The revision extends the amount of
time the 12-inch and 14-inch purge system supply and exhaust lines may
be used for venting or purging from 90 to 135 hours per 365 days. In
addition, expired footnotes were deleted as an editorial change and the
associated Bases section was revised.
Date of issuance: August 31, 1995
Effective date: As of the date of issuance to be implemented upon
receipt.
Amendment No.: 68
Facility Operating License No. NPF-69: Amendment revises the
Technical Specifications.Public comments requested as to proposed no
significant hazards consideration: NoThe Commission's related
evaluation of the amendment, emergency circumstances and consultation
with the State, and final determination of no significant hazards
consideration are contained in a Safety Evaluation dated August 31,
1995.
Local Public Document Room location: Reference and Documents
Department, Penfield Library, State University of New York, Oswego, New
York 13126.
Attorney for licensee: Mark J. Wetterhahn, Esquire, Winston &
Strawn, 1400 L Street, NW., Washington, DC 20005-3502.
NRC Project Director: Ledyard B. Marsh
For the Nuclear Regulatory Commission
John N. Hannon,
Acting Director, Division of Reactor Projects - III/IV Office of
Nuclear Reactor Regulation
[Doc. 95-23806 Filed 9-26-95; 8:45 am]
BILLING CODE 7590-01-F