[Federal Register Volume 63, Number 9 (Wednesday, January 14, 1998)]
[Notices]
[Pages 2271-2288]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: 98-753]
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NUCLEAR REGULATORY COMMISSION
Biweekly Notice; Applications and Amendments to Facility
Operating Licenses Involving No Significant Hazards Considerations
I. Background
Pursuant to Public Law 97-415, the U.S. Nuclear Regulatory
Commission (the Commission or NRC staff) is publishing this regular
biweekly notice. Public Law 97-415 revised section 189 of the Atomic
Energy Act of 1954, as amended (the Act), to require the Commission to
publish notice of any amendments issued, or proposed to be issued,
under a new provision of section 189 of the Act. This provision grants
the Commission the authority to issue and make immediately effective
any amendment to an operating license upon a determination by the
Commission that such amendment involves no significant hazards
consideration, notwithstanding the pendency before the Commission of a
request for a hearing from any person.
This biweekly notice includes all notices of amendments issued, or
proposed to be issued from December 18, 1997, through January 2, 1998.
The last biweekly notice was published on December 31, 1997 (62 FR
68303).
Notice of Consideration of Issuance of Amendments To Facility Operating
Licenses, Proposed No Significant Hazards Consideration Determination,
and Opportunity for a Hearing
The Commission has made a proposed determination that the following
amendment requests involve no significant hazards consideration. Under
the Commission's regulations in 10 CFR 50.92, this means that operation
of the facility in accordance with the proposed amendment would not (1)
involve a significant increase in the probability or consequences of an
accident previously evaluated; or (2) create the possibility of a new
or different kind of accident from any accident previously evaluated;
or (3) involve a significant reduction in a margin of safety. The basis
for this proposed determination for each amendment request is shown
below.
The Commission is seeking public comments on this proposed
determination. Any comments received within 30 days after the date of
publication of this notice will be considered in making any final
determination.
Normally, the Commission will not issue the amendment until the
expiration of the 30-day notice period. However, should circumstances
change during the notice period such that failure to act in a timely
way would result, for example, in derating or shutdown of the facility,
the Commission may issue the license amendment before the expiration of
the 30-day notice period, provided that its final determination is that
the amendment involves no significant hazards consideration. The final
determination will consider all public and State comments received
before action is taken. Should the Commission take this action, it will
publish in the Federal Register a notice of issuance and provide for
opportunity for a hearing after issuance. The Commission expects that
the need to take this action will occur very infrequently.
Written comments may be submitted by mail to the Chief, Rules and
Directives Branch, Division of Freedom of Information and Publications
Services, Office of Administration, U.S. Nuclear Regulatory Commission,
Washington, DC 20555-0001, and should cite the publication date and
page number of this Federal Register
[[Page 2272]]
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 February 13, 1998, the licensee may file a request for a hearing
with respect to issuance of the amendment to the subject facility
operating license and any person whose interest may be affected by this
proceeding and who wishes to participate as a party in the proceeding
must file a written request for a hearing and a petition for leave to
intervene. Requests for a hearing and a petition for leave to intervene
shall be filed in accordance with the Commission's ``Rules of Practice
for Domestic Licensing Proceedings'' in 10 CFR Part 2. Interested
persons should consult a current copy of 10 CFR 2.714 which is
available at the Commission's Public Document Room, the Gelman
Building, 2120 L Street, NW., Washington, DC and at the local public
document room for the particular facility involved. If a request for a
hearing or petition for leave to intervene is filed by the above date,
the Commission or an Atomic Safety and Licensing Board, designated by
the Commission or by the Chairman of the Atomic Safety and Licensing
Board Panel, will rule on the request and/or petition; and the
Secretary or the designated Atomic Safety and Licensing Board will
issue a notice of a hearing or an appropriate order.
As required by 10 CFR 2.714, a petition for leave to intervene
shall set forth with particularity the interest of the petitioner in
the proceeding, and how that interest may be affected by the results of
the proceeding. The petition should specifically explain the reasons
why intervention should be permitted with particular reference to the
following factors: (1) The nature of the petitioner's right under the
Act to be made a party to the proceeding; (2) the nature and extent of
the petitioner's property, financial, or other interest in the
proceeding; and (3) the possible effect of any order which may be
entered in the proceeding on the petitioner's interest. The petition
should also identify the specific aspect(s) of the subject matter of
the proceeding as to which petitioner wishes to intervene. Any person
who has filed a petition for leave to intervene or who has been
admitted as a party may amend the petition without requesting leave of
the Board up to 15 days prior to the first prehearing conference
scheduled in the proceeding, but such an amended petition must satisfy
the specificity requirements described above.
Not later than 15 days prior to the first prehearing conference
scheduled in the proceeding, a petitioner shall file a supplement to
the petition to intervene which must include a list of the contentions
which are sought to be litigated in the matter. Each contention must
consist of a specific statement of the issue of law or fact to be
raised or controverted. In addition, the petitioner shall provide a
brief explanation of the bases of the contention and a concise
statement of the alleged facts or expert opinion which support the
contention and on which the petitioner intends to rely in proving the
contention at the hearing. The petitioner must also provide references
to those specific sources and documents of which the petitioner is
aware and on which the petitioner intends to rely to establish those
facts or expert opinion. Petitioner must provide sufficient information
to show that a genuine dispute exists with the applicant on a material
issue of law or fact. Contentions shall be limited to matters within
the scope of the amendment under consideration. The contention must be
one which, if proven, would entitle the petitioner to relief. A
petitioner who fails to file such a supplement which satisfies these
requirements with respect to at least one contention will not be
permitted to participate as a party.
Those permitted to intervene become parties to the proceeding,
subject to any limitations in the order granting leave to intervene,
and have the opportunity to participate fully in the conduct of the
hearing, including the opportunity to present evidence and cross-
examine witnesses.
If a hearing is requested, the Commission will make a final
determination on the issue of no significant hazards consideration. The
final determination will serve to decide when the hearing is held.
If the final determination is that the amendment request involves
no significant hazards consideration, the Commission may issue the
amendment and make it immediately effective, notwithstanding the
request for a hearing. Any hearing held would take place after issuance
of the amendment.
If the final determination is that the amendment request involves a
significant hazards consideration, any hearing held would take place
before the issuance of any amendment.
A request for a hearing or a petition for leave to intervene must
be filed with the Secretary of the Commission, U.S. Nuclear Regulatory
Commission, Washington, DC 20555-0001, Attention: Rulemakings and
Adjudications Staff, or may be delivered to the Commission's Public
Document Room, the Gelman Building, 2120 L Street, NW., Washington DC,
by the above date. A copy of the petition should also be sent to the
Office of the General Counsel, U.S. Nuclear Regulatory Commission,
Washington, DC 20555-0001, and to the attorney for the licensee.
Nontimely filings of petitions for leave to intervene, amended
petitions, supplemental petitions and/or requests for a hearing will
not be entertained absent a determination by the Commission, the
presiding officer or the Atomic Safety and Licensing Board that the
petition and/or request should be granted based upon a balancing of
factors specified in 10 CFR 2.714(a)(1)(i)-(v) and 2.714(d).
For further details with respect to this action, see the
application for amendment which is available for public inspection at
the Commission's Public Document Room, the Gelman Building, 2120 L
Street, NW., Washington, DC, and at the local public document room for
the particular facility involved.
Arizona Public Service Company, et al., Docket Nos. STN 50-528, STN 50-
529, and STN 50-530, Palo Verde Nuclear Generating Station, Units Nos.
1, 2, and 3, Maricopa County, Arizona
Date of amendments request: December 17, 1997.
Description of amendments request: The proposed amendment would
modify the Technical Specifications (TS) to replace the current
explicit reference to Exide batteries with a generic reference to low
specific gravity cells. The proposed change would also remove footnotes
for the Unit 2 and Unit 3 TS that referred to one time exemptions that
no longer apply. The proposed change would allow replacement of the
existing Class 1E, 125 volt DC batteries with equivalent batteries
manufactured by different vendors.
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.
[[Page 2273]]
The Class 1E 125V DC system provides DC power to the Class 1E DC
loads for operation, control and switching, including the inverters
which power the Class 1E 120V vital AC busses. This system is not an
accident initiator. It is, however, an accident mitigation system.
The replacement low specific gravity rectangular cell batteries have
been designed to IEEE 485-1978 standards and meet all appropriate
seismic criteria. There is no change in the physical or electrical
separation provisions for the Class 1E 125V DC channels. These
batteries are used extensively throughout the industry and their
failure mechanisms are well understood. The existing high specific
gravity round cell batteries are experiencing premature capacity
loss for which a definitive root cause of failure has not been
determined. Therefore, replacement of the high specific gravity
round cell batteries with low specific gravity rectangular cell
batteries increases the overall reliability of the Class 1E 125 V DC
system. In addition, the design requirements of the replacement
batteries ensures that the batteries will be capable of reliably
performing their design function during all modes of operation and
will serve to mitigate any accident that may occur. The proposed
amendment does not change the performance criteria or cell
parameters for the Class 1E 125V DC sources that are defined in the
current Technical Specifications for each unit. Since this change is
increasing the overall reliability and performance of the system and
is designed to meet the same stringent requirements of the existing
high specific gravity round cell batteries, it does not involve a
significant increase in the probability or consequences of an
accident previously evaluated.
Replacement of the high specific gravity round cell batteries
will occur during acceptable modes of operation as defined in the
current Technical Specifications for each unit, i.e., the work will
be performed during Modes 5 or 6, or with the reactor defueled.
Technical Specification 3.8.2.2, DC Sources--Shutdown, for each unit
requires one Class 1E 125V DC train to be operable in Modes 5 or 6.
With one Class 1E 125V DC train operable, the other train may be
removed from service for battery cell replacement. Since the battery
cell replacement will be performed within the Limiting Condition of
Operation for DC Sources--Shutdown, the replacement sequence of the
battery banks will not involve a significant increase in the
probability or consequences of an accident previously evaluated.
2. The proposed change does not create the possibility of a new
or different kind of accident from any accident previously
evaluated.
Low specific gravity rectangular cell batteries have been used
throughout the industry for many years. Some cells have been in
service for 17 years and have not degraded to the extent that they
require replacement. The low specific gravity rectangular cell
batteries have demonstrated good reliability and the failure
mechanisms associated with these batteries are well understood. The
high specific gravity round cell batteries that are currently
installed are exhibiting premature capacity loss for which a
definitive root cause of failure has not been determined. Replacing
the high specific gravity round cell batteries with low specific
gravity rectangular cell batteries that have seen extensive use in
the industry, are well understood and have been designed to meet the
same stringent requirements as that of the existing batteries
ensures that the overall system reliability is increased. No new or
common mode failures are created since the replacement low specific
gravity rectangular cell batteries have been designed to the same
stringent requirements as the existing batteries. The proposed
amendment does not change the performance criteria or cell
parameters for the Class 1E 125V DC sources that are defined in the
current Technical Specifications for each unit. Therefore,
replacement of the high specific gravity round cell batteries with
low specific gravity rectangular cell batteries does not create the
possibility of a new or different kind of accident from any accident
previously evaluated.
3. The proposed change does not involve a significant reduction
in a margin of safety.
As described previously, replacing the high specific gravity
round cell batteries with low specific gravity rectangular cell
batteries enhances the overall system reliability. The low specific
gravity rectangular cell batteries have been designed to the same
criteria as the existing high specific gravity round cell batteries.
The performance criteria and cell parameters specified in each
unit's Technical Specifications are not affected by this change.
Therefore, the proposed change does not involve a significant
reduction in a margin of safety.
Replacement of the high specific gravity round cell batteries
will occur during acceptable modes of operation as defined in the
current Technical Specifications for each unit, i.e., the work will
be performed during Modes 5 or 6, or with the reactor defueled.
Technical Specification 3.8.2.2, DC Sources--Shutdown, for each unit
requires one Class 1E 125V DC train to be operable in Modes 5 or 6.
With one Class 1E 125V DC train operable, the other train may be
removed from service for battery cell replacement. Since the battery
cell replacement will be performed within the Limiting Condition of
Operation for DC Sources--Shutdown, the work sequence for
replacement of the battery banks will not involve a significant
reduction in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
that review, it appears that the three standards of 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendments request involve no significant hazards consideration.
Local Public Document Room location: Phoenix Public Library, 1221
N. Central Avenue, Phoenix, Arizona 85004.
Attorney for licensee: Nancy C. Loftin, Esq., Corporate Secretary
and Counsel, Arizona Public Service Company, P.O. Box 53999, Mail
Station 9068, Phoenix, Arizona 85072-3999.
NRC Project Director: William H. Bateman.
Commonwealth Edison Company, Docket Nos. 50-237 and 50-249, Dresden
Nuclear Power Station, Units 2 and 3, Grundy County, Illinois
Date of amendment request: October 3, 1996.
Description of amendment request: The proposed amendments would
correct a typographical error which was introduced into the Technical
Specifications (TS) with issuance of Amendment Nos. 150 and 145.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below.
(1) Involve a significant increase in the probability or
consequences of an accident previously evaluated because of the
following:
The proposed change does not alter the manner of operation of
the facility, it merely restores the correspondence between the
applicability of the Limiting Conditions for Operability (LCO) for
the (Source Range) Neutron Monitors and the Drywell Radiation
Monitors and the associated Surveillance Requirements for the same
two instrument functions as described in Tables 3.2.F-1 and 4.2.F-1.
No changes are proposed which will affect the probability of an
accident previously evaluated, since the instruments and their
associated functions are credited to operate during and after a
postulated accident. The function of a device after an event has
occurred cannot affect the probability of that accident occurring.
Similarly, the proposed changes do not effect the operation or
function of structures, systems or components which effect the
probability of any accident previously evaluated.
The proposed changes do not affect the consequence of an
accident previously evaluated since the changes do not decrease the
availability of any functions credited with performing mitigative
actions. The availability requirements of the Drywell Radiation
Monitors is not changed because the associated LCO requires the
monitors to be OPERABLE in the conditions proposed in this change.
The proposed change merely assures that the surveillance
requirements are met in the modes which correspond to the LCO. The
(Source Range) Neutron Monitor surveillance requirements change does
not affect the ability of the system to provide adequate information
to the operators to mitigate the consequences of a postulated
accident, since the system OPERABILITY requirements as specified in
the LCO are not affected.
(2) Create the possibility of a new or different kind of
accident from any accident previously evaluated because:
[[Page 2274]]
The proposed change does not introduce any new or different
types of operation of the plants. No new equipment is introduced as
a result of the implementation of the proposed change. Therefore no
changes are proposed which could introduce a new or different kind
of accident from any previously evaluated.
(3) Involve a significant reduction in the margin of safety
because:
The proposed change does not effect the margin of safety. The
LCO requirements for the two instrument systems which are effected
are not changed; the OPERABILITY requirements remain the same. The
only substantive changes are the modes in which surveillance testing
is required to be performed. The change restores the need to perform
testing of the Drywell Radiation Monitor prior to and during
OPERATIONAL MODE 3 operations, and removes the requirement to
perform testing of the (Source Range) Neutron Monitors prior to and
during operation in MODE 3 when it is not required to be OPERABLE as
described in the associated LCO. Based on this, the availability of
the affected instruments to perform their design function is not
effected by this change and no reduction in the margin of safety is
proposed.
Guidance has been provided in ``Final Procedures and Standards
on No Significant Hazards Considerations,'' Final Rule, 51 FR 7744,
for the application of standards to license change requests for
determination of the existence of significant hazards
considerations. This document provides examples of amendments which
are and are not considered likely to involve significant hazards
considerations.
This proposed amendment does not involve any irreversible
changes, significant relaxation of the criteria used to establish
safety limits, a significant relaxation of the bases for the
limiting safety system settings, or a significant relaxation of the
bases for the limiting conditions for operations. Therefore, based
on the guidance provided in the Federal Register and the criteria
established in 10 CFR 50.92(c), the proposed change does not
constitute 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
requested amendments involve no significant hazards consideration.
Local Public Document Room location: Morris Area Public Library
District, 604 Liberty Street, Morris, Illinois 60450.
Attorney for licensee: Michael I. Miller, Esquire; Sidley and
Austin, One First National Plaza, Chicago, Illinois 60603.
NRC Project Director: Robert A. Capra.
Commonwealth Edison Company, Docket Nos. 50-237 and 50-249, Dresden
Nuclear Power Station, Units 2 and 3, Grundy County, Illinois
Docket Nos. 50-254 and 50-265, Quad Cities Nuclear Power Station, Units
1 and 2, Rock Island County, Illinois
Docket Nos. 50-373 and 50-374, LaSalle County Station, Units 1 and 2,
LaSalle County, Illinois
Date of application for amendment request: August 29, 1997.
Description of amendment request: The proposed amendments would
change the Dresden, Quad Cities and LaSalle Technical Specifications
(TS) to reflect the use of Siemens Power Corporation (SPC) ATRIUM-9B
fuel. Specifically the proposed amendments incorporate the following
into the TS: (a) new Siemens' methodologies that will enhance
operational flexibility and reduce the likelihood of future plant
derates, (b) administrative changes that both eliminate the cycle
specific implementation of Atrium-9B fuel, and (c) changes to the
Dresden and Quad Cities Minimum Critical Power Ratio (MCPR).
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below.
1. Involve a significant increase in the probability or
consequences of an accident previously evaluated.
The probability of an evaluated accident is derived from the
probabilities of the individual precursors to that accident. The
consequences of an evaluated accident are determined by the
operability of plant systems designed to mitigate those
consequences. Limits have been established consistent with NRC
approved methods to ensure that fuel performance during normal,
transient, and accident conditions is acceptable. These changes do
not affect the operability of plant systems, nor do they compromise
any fuel performance limits.
Addition of SPC Revised Jet Pump Methodology (LaSalle Units 1
and 2)
The Reference 1 [ANF-91-048(P), Supplement 1, ``BWR Jet Pump
Model Revision for RELAX''. Submitted to the NRC by SPC letter, ANF-
91-048(P), Supplement 1 and ANF-91-048(NP), Supplement 1, ``BWR Jet
Pump Model Revision for RELAX,'' RAC:96-042, R.A. Copeland to US
NRC, May 6, 1996] methodology to be added to the Technical
Specifications is used as part of the LOCA [loss-of-coolant
accident] analysis and does not introduce physical changes to the
plant. The Reference 1 revised jet pump model changes the
calculational behavior of the jet pump under reversed drive flow
conditions. The revised jet pump model methodology makes the LOCA
model behave more realistically and calculates small break LOCA PCTs
[peak cladding temperature] that are comparable to the large break
LOCA results. Therefore, this change only affects the methodology
for analyzing the LOCA event and determining the protective APLHGR
[average planar linear heat generation rate] limits. The Technical
Specification requirements for monitoring APLHGR are not affected by
this change. The revised method will result in higher APLHGR limits,
thus the SPC fuel will be allowed to operate at higher nodal powers.
The approved methodology, however, still protects the fuel
performance limits specified by 10 CFR 50.46. Therefore, the
probability or consequences of an accident previously evaluated will
not change.
Addition of SPC Generic Methodology for Application of ANFB
Critical Power Correlation to Non-SPC Fuel (Quad Cities Units 1 and
2 and LaSalle Units 1 and 2)
The probability or consequences of a previously evaluated
accident are not increased by adding Reference 3 [EMF-1125(P)(A),
Supplement 1 Appendix C, ``ANFB Critical Power Correlation
Application for Coresident Fuel'', August 1997, and NRC SER,
``Acceptance for Referencing of Licensing Topical Report EMF-
1125(P), Supplement 1 Appendix C, ``ANFB Critical Power Correlation
Application for Co-Resident Fuel'', J.E. Lyons to R.A. Copeland, May
9, 1997] to Section 6.9.A.6.b of the Quad Cities Technical
Specifications and Bases Section 2.1.2 and Section 6.6.A.6.b of the
LaSalle Technical Specifications. Reference 3 determines the
additive constants and the associated uncertainty for application of
the ANFB correlation to the coresident GE [General Electric] fuel.
Therefore, it provides data that is used in the determination of the
MCPR Safety Limit. This approved methodology for applying the ANFB
critical power correlation to the GE fuel will protect the fuel from
boiling transition. Operational MCPR limits will also be applied to
ensure that the MCPR Safety Limit is protected during all modes of
operation and anticipated operational occurrences. Because Reference
3 contains conservative methods and calculations and because the
operability of plant systems designed to mitigate any consequences
of accidents have not changed, the probability or consequences of an
accident previously evaluated will not increase.
Addition of SPC Topical for Revised ANFB Correlation Uncertainty
(Quad Cities Units 1 and 2, Dresden Units 2 and 3, and LaSalle Units
1 and 2)
The probability or consequences of a previously evaluated
accident is not increased by adding Reference 7 [ANF-1125(P),
Supplement 1, Appendix D, ``ANFB Critical Power Correlation
Uncertainty For Limited Data Sets''. Submitted to the NRC by SPC
letter, ``Request for Review of ANFB Critical Power Correlation
Uncertainty for Limited Data Sets, ANF-1125(P), Supplement 1,
Appendix D'', HDC:97:032, H.D. Curet to Document Control Desk, April
18, 1997] to Section 6.9.A.6.b of the Quad Cities and Dresden
Technical Specifications and Bases Section 2.1.2 and Section
6.6.A.6.b of the LaSalle Technical Specifications.
[[Page 2275]]
Reference 7 documents the additive constant uncertainty for SPC
ATRIUM-9B fuel design with an internal water channel. This
methodology is used to determine an input to the MCPR Safety Limit
calculations, which ensures that more than 99.9% of the fuel rods
avoid transition boiling during normal operation as well as
anticipated operational occurrences. This change does not require
any physical plant modifications, physically affect any plant
components, or entail changes in plant operation. This methodology
for determining the ATRIUM-9B additive constant uncertainty for the
MCPR Safety Limit calculation will continue to support protecting
the fuel from boiling transition. Operational MCPR limits will be
applied to ensure the MCPR Safety Limit is not violated during all
modes of operation and anticipated operational occurrences.
Therefore, no individual precursors of an accident are affected and
the operability of plant systems designed to mitigate the
probability of consequences of an accident previously evaluated are
not affected by these changes.
Change to Minimum Critical Power Ratio Safety Limit (Quad Cities
Units 1 and 2 and Dresden Units 2 and 3)
Changing the MCPR Safety Limit at Quad Cities Units 1 and 2 and
Dresden Units 2 and 3 will not increase the probability of an
accident previously evaluated. This change implements the MCPR
Safety Limits resulting from the SPC ANFB critical power correlation
methodology using a revised additive constant uncertainty from
Reference 7. The MCPR Safety Limit of 1.09 that is proposed for Quad
Cities Units 1 and 2 and Dresden Units 2 and 3 is anticipated to be
conservative and acceptable for future cycles. Cycle specific MCPR
Safety Limit calculations will be performed, consistent with SPC's
approved methodology, to confirm the appropriateness of the MCPR
Safety Limit. Additionally, operational MCPR limits will be applied
that will ensure the MCPR Safety Limit is not violated during all
modes of operation and anticipated operational occurrences. Changing
the MCPR Safety Limit will not alter any physical systems or
operating procedures. The MCPR Safety Limit is set to 1.09, which is
the CPR value where less than 0.1% of the rods in the core are
expected to experience boiling transition. This safety limit is
expected to be applicable for future cycles of ATRIUM-9B at Dresden
and Quad Cities. Therefore the probability or consequences of an
accident will not increase.
Removal of Footnotes Limiting Operation with ATRIUM-9B Fuel
Reloads (Quad Cities Unit 2 and Dresden Unit 3)
The removal of footnotes from the Quad Cities and Dresden
Technical Specifications does not involve any significant increase
in the probability or consequences of an accident previously
evaluated. The footnotes were added to clarify that cycle specific
methods were used until the generic methodology was approved by the
NRC. Since the NRC has approved SPC's generic methodology for
application of the ANFB correlation to the coresident GE fuel
(Reference 3) and SPC has addressed the concerns regarding the
database used to calculate the ATRIUM-9B additive constant
uncertainties (Reference 7), the footnotes are no longer necessary.
The removal of the Unit 2 specific ``a'' pages, 2-1a and B2-3a, in
the Quad Cities Technical Specifications is justified by the removal
of the footnotes. Therefore, removing these footnotes and ``a''
pages does not require any physical plant modifications, nor does it
physically affect any plant components or entail changes in plant
operation. Therefore, the probability or consequences of an accident
previously evaluated is not expected to increase.
Revision to Thermal Limit Descriptions (Quad Cities Units 1 and
2, Dresden Units 2 and 3, and LaSalle Units 1 and 2)
The revision to the Section 3 Technical Specification
description of the APLHGR limits has no implications on accident
analysis or plant operations. The purpose of the revision is to
allow flexibility for the MAPLHGR [maximum average planar linear
heat generation rate] limits and their exposure basis to be
specified in the COLR [core operating limits report] and to
establish consistency with approved methodologies currently utilized
by Siemens Power Corporation, which calculates MAPLHGR limits based
on bundle or planar average exposures. This revision also provides
for consistency in the APLHGR limit Technical Specification wording
between the ComEd BWRs [boiling water reactor]. The revision to the
3.11.D SLHGR [steady state linear heat generation rate] Technical
Specification for Dresden also has no implications on accident
analysis or plant operations. The purpose of this revision is to
allow flexibility for the LHGR [linear heat generation rate] limits
and their exposure basis to be specified in the COLR. This revision
makes the Dresden LHGR definition consistent with NUREG 1433/1434
wording. The definition of the Average Planar Exposure is deleted,
because the exposure basis of the APLHGR is being removed.
Therefore, no plant equipment or processes are affected by this
change. Thus, there is no alteration in the probability or
consequences of an accident previously evaluated.
2. Create the possibility of a new or different kind of accident
from any accident previously evaluated:
Creation of the possibility of a new or different kind of
accident would require the creation of one or more new precursors of
that accident. New accident precursors may be created by
modifications to the plant configuration, including changes in
allowable modes of operation. This Technical Specification submittal
does not involve any modifications to the plant configuration or
allowable modes of operation. No new precursors of an accident are
created and no new or different kinds of accidents are created.
Therefore, the proposed changes do not create the possibility of a
new or different kind of accident from any accident previously
evaluated.
Addition of SPC Revised Jet Pump Methodology (LaSalle Units 1
and 2)
The revised jet pump model methodology will be used to analyze
the LOCA for LaSalle Units 1 and 2, and does not introduce any
physical changes to the plant or the processes used to operate the
plant. This change only affects the methods used to analyze the LOCA
event and determine the MAPLHGR limits. Therefore, the possibility
of a new or different kind of accident is not created.
Addition of SPC Generic Methodology for Application of ANFB
Critical Power Correlation to Non-SPC Fuel (Quad Cities Units 1 and
2 and LaSalle Units 1 and 2)
Addition of the generic methodology for the application of the
ANFB critical power correlation to GE fuel in Section 6.9.A.6.b of
the Quad Cities Technical Specifications and Bases Section 2.1.2 and
Section 6.6.A.6.b of the LaSalle Technical Specifications does not
introduce any physical changes to the plant, the processes used to
operate the plant, or allowable modes of operation. This change only
involves adding an NRC approved methodology, which is used to
determine the additive constants and additive constant uncertainty
for GE fuel, to Section 6 of the Technical Specifications.
Therefore, no new precursors of an accident are created and no new
or different kinds of accidents are created.
Addition of SPC Topical for Revised ANFB Correlation Uncertainty
(Quad Cities Units 1 and 2, Dresden Units 2 and 3, and LaSalle Units
1 and 2)
Addition of the Reference 7 methodology to Section 6.9.A.6.b of
the Quad Cities and Dresden Technical Specifications and Bases
Section 2.1.2 and Section 6.6.A.6.b of the LaSalle Technical
Specifications will not create the possibility of a new or different
kind of accident from any accident previously evaluated. This
methodology describes the calculation of an input to the MCPR Safety
Limit--the ATRIUM-9B additive constant uncertainty. Therefore, no
new precursors of an accident are created and no new or different
kinds of accidents are created.
Change to Minimum Critical Power Ratio Safety Limit (Quad Cities
Units 1 and 2 and Dresden Units 2 and 3)
Changing the MCPR Safety Limit will not create the possibility
of a new accident from an accident previously evaluated. This change
will not alter or add any new equipment or change modes of
operation. The MCPR Safety Limit is established to ensure that 99.9%
of the rods avoid boiling transition.
The MCPR Safety Limit is changing for Quad Cities Unit 1 due to
the transition to SPC ATRIUM-9B fuel and SPC methodologies. The MCPR
Safety Limit is changing for Quad Cities Unit 2 due to the Reference
7 methodology, which documents a 0.0195 ATRIUM-9B additive constant
uncertainty and supports a 1.09 MCPR Safety Limit. This MCPR Safety
Limit is lower than the current MCPR Safety Limit for Quad Cities
Unit 2, 1.10, which is based on a higher interim conservative
additive constant uncertainty of 0.029. The lower ATRIUM-9B additive
constant uncertainty results in the lower MCPR Safety Limit for Quad
Cities Unit 2. The new MCPR Safety Limit for Dresden Units 2 and 3,
1.09, is greater than the current value at Dresden Units 2 and 3 and
is being increased now in anticipation of bounding future reloads of
ATRIUM-9B. Therefore, no new accidents are created that are
different from any accident previously evaluated.
[[Page 2276]]
Removal of Footnotes Limiting Operation with ATRIUM-9B Fuel
Reloads (Quad Cities Unit 2 and Dresden Unit 3)
The removal of the footnotes from the Quad Cities and Dresden
Technical Specifications does not create a new or different kind of
accident from any accident previously evaluated. The removal of the
footnotes does not affect plant systems or operation. The footnotes
were temporarily established to implement a conservative cycle
specific MCPR Safety Limit until the SPC generic methodology was
approved. With the approval of the generic Reference 3 methodology
and the anticipated approval of the Reference 7 additive constant
uncertainty methodology, these footnotes are no longer applicable.
The removal of the Unit 2 specific ``a'' pages, 2-1a and B2-3a, in
the Quad Cities Technical Specifications which is justified by the
removal of the footnotes, also does not create a new or different
kind of accident from any accident previously evaluated.
Revision to Thermal Limit Descriptions (Quad Cities Units 1 and
2, Dresden Units 2 and 3, and LaSalle 1 and 2)
The revision of the APLHGR and LHGR limit descriptions will not
create the possibility of a new or different kind of accident from
any accident previously evaluated. This revision will not alter any
plant systems, equipment, or physical conditions of the site. This
revision allows the flexibility of the APLHGR and the LHGR limits to
be specified in the COLR and to maintain consistency with the
calculated results of methodologies currently used to determine the
APLHGR. The definition of the Average Planar Exposure is deleted,
because it is being removed from LHGR and APLHGR Technical
Specifications.
3. Involve a significant reduction in the margin of safety for
the following reasons:
Addition of SPC Revised Jet Pump Methodology (LaSalle Units 1
and 2)
The revised jet pump model methodology, and the MAPLHGRs,
resulting from the revised jet pump methodology, will continue to
ensure fuel design criteria and 10 CFR 50.46 compliance. The results
of LOCA analyses performed with this methodology must continue to
comply with the requirements of 10 CFR 50.46. Therefore, there is no
significant reduction in the margin of safety.
Addition of SPC Generic Methodology for Application of ANFB
Critical Power Correlation to Non-SPC Fuel (Quad Cities Units 1 and
2 and LaSalle Units 1 and 2)
The margin of safety is not decreased by adding this reference
to Section 6.9.A.6.b of the Quad Cities Technical Specifications and
Bases Section 2.1.2 and Section 6.6.A.6.b of the LaSalle Technical
Specifications. Siemens Power Corporation methodology for
application of the ANFB Critical Power Correlation to coresident GE
fuel is approved by the NRC and is the same methodology used in the
cycle specific topical for coresident fuel (References 4 [EMF-96-
021(P), Revision 1, ``Application of the ANFB Critical Power
Correlation to Coresident GE fuel for LaSalle Unit 2 Cycle 8'',
February 1996, and NRC SER, ``Safety Evaluation for Topical Report
EMF-95-021(P), Revision 1, ``Application of the ANFB Critical Power
Correlation to Coresident GE Fuel for LaSalle Unit 2 Cycle 8' (TAC
NO. M94964'', D.M. Skay to I. Johnson, September 26, 1996] and 5
[EMF-96-051(P), ``Application of the ANFB Critical Power Correlation
to Coresident GE Fuel for Quad Cities Unit 2 Cycle 15'', May, 1996,
and NRC SER, ``Approval of Topical Report EMF-96-051(P)--Quad
Cities, Unit 2 (TAC NO. M96213)'', R. Pulsifer to I. Johnson, May
16, 1997] that greater than 99.9% of the rods in the core avoid
boiling transition. Additionally, operating limits will be
established to ensure the MCPR Safety Limit is not violated during
all modes of operation.
Addition of SPC Topical for Revised ANFB Correlation Uncertainty
(Quad Cities Units 1 and 2, Dresden Units 2 and 3, and LaSalle Units
1 and 2)
The MCPR Safety Limit provides a margin of safety by ensuring
that less than 0.1% of the rods are expected to be in boiling
transition if the MCPR Safety Limit is not violated. This Technical
Specification amendment proposes to insert the topical report that
describes SPC's calculation of the ATRIUM-9B additive constant
uncertainty. The new ATRIUM-9B additive constant uncertainty
calculation is conservative and is based on a larger database than
previous calculations. Because a conservative method is used to
calculate the ATRIUM-9B additive constant uncertainty, a decrease in
the margin to safety will not occur due to adding this methodology
to the Technical Specifications. In addition, operational limits
will be established to ensure the MCPR Safety Limit is protected for
all modes of operation. This revised methodology will only ensure
that the appropriate level of fuel protection is being employed.
Change to Minimum Critical Power Ratio Safety Limit (Quad Cities
Unit 1 and 2 and Dresden Units 2 and 3)
Changing the MCPR Safety Limit for Quad Cities and Dresden will
not involve any reduction in margin of safety. The MCPR Safety Limit
provides a margin of safety by ensuring that less than 0.1% of the
rods are expected to be in boiling transition if the MCPR Safety
Limit is not violated. The proposed Technical Specification
amendment reflects the MCPR Safety Limit results from conservative
evaluations by SPC using the ANFB critical power correlation with
the new 0.0195 ATRIUM-9B additive constant uncertainty documented in
Reference 7.
Because a conservative method is used to apply the ATRIUM-9B
additive constant uncertainty in the MCPR Safety Limit calculation,
a decrease in the margin to safety will not occur due to changing
the MCPR Safety Limit. The revised MCPR Safety Limit will ensure the
appropriate level of fuel protection. Additionally, operational
limits will be established based on the proposed MCPR Safety Limit
to ensure that the MCPR Safety Limit is not violated during all
modes of operation including anticipated operation occurrences. This
will ensure that the fuel design safety criterion of more than 99.9%
of the fuel rods avoiding transition boiling during normal operation
as well as during an anticipated operational occurrence is met.
Removal of Footnotes Limiting Operation with ATRIUM-9B Fuel
Reloads (Quad Cities Unit 2 and Dresden Unit 3)
The removal of the cycle specific footnotes in Quad Cities and
Dresden Technical Specifications does not impose a change in the
margin of safety. These footnotes were added due to concerns
regarding the calculation of the additive constant uncertainty for
the ATRIUM-9B fuel and the cycle specific application of the ANFB
critical power correlation to coresident GE fuel in Quad Cities Unit
2 Cycle 15. Because the generic ANFB application to coresident GE
fuel MCPR methodology (Reference 3) has received NRC approval and
the topical report describing the increased database used to
calculate the additive constant uncertainties for ATRIUM-9B
(Reference 7) have been submitted to the NRC and both are proposed
to be added to the Technical Specifications in this amendment, there
is no reason for the footnotes to remain. Removal of the Unit 2
specific ``a'' pages, 2-1a and B2-3a, in the Quad Cities Technical
Specifications is justified by the removal of the footnotes.
Therefore, the removal of the ``a'' pages, 2-1a and B2-3a, also does
not impose a change in the margin of safety.
Revision to Thermal Limit Descriptions (Quad Cities Units 1 and
2, Dresden Units 2 and 3, and LaSalle Units 1 and 2)
The revision to the APLHGR and LHGR limit descriptions will not
involve a reduction in the margin of safety. The methodology used to
calculate the APLHGR must comply with the guidelines of Appendix K
of 10 CFR Part 50, and the APLHGR and LHGR will still be required to
be maintained within the limits specified in the COLR. The
surveillance requirements for these two thermal limits remain
unchanged. Thus, there will be no 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: for Dresden, Morris Area
Public Library District, 604 Liberty Street, Morris, Illinois 60450;
for La Salle, Jacobs Memorial Library, Illinois Valley Community
College, Oglesby, Illinois 61348; for Quad Cities, Dixon Public
Library, 221 Hennepin Avenue, Dixon, Illinois 61021.
Attorney for licensee: Michael I. Miller, Esquire; Sidley and
Austin, One First National Plaza, Chicago, Illinois 60603.
NRC Project Director: Robert A. Capra.
[[Page 2277]]
Commonwealth Edison Company, Docket Nos. 50-237 and 50-249, Dresden
Nuclear Power Station, Units 2 and 3, Grundy County, Illinois
Docket Nos. 50-254 and 50-265, Quad Cities Nuclear Power Station, Units
1 and 2, Rock Island County, Illinois
Date of amendment request: October 27, 1997.
Description of amendment request: The proposed amendments would
clarify the applicability, action and surveillance requirements for the
Standby Liquid Control System.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below.
(1) Involve a significant increase in the probability or
consequences of an accident previously evaluated because of the
following:
The proposed changes represent the conversion of current
requirements which are based on generic guidance or previously
approved provisions for other stations. The proposed changes are
consistent with NUREG-1433 and do not significantly increase the
probability or consequences of any previously evaluated accidents
for Dresden or Quad Cities Stations. The proposed amendment is
consistent with the current safety analyses and represents
sufficient requirements for the assurance and reliability of
equipment assumed to operate in the safety analysis, or provide
continued assurance that specified parameters remain within their
acceptance limits. The proposed TS continue to ensure sufficient
requirements are in place for the SLCS during plant operation. The
proposed changes that eliminate Applicability and Actions during
refueling operations for the SLCS do not affect the probability of
any previously evaluated accident because only one control rod can
be withdrawn during refueling operations and Shutdown Margin
requirements are maintained in the Technical Specifications.
Therefore, the probability of an inadvertent criticality is not
increased as reactivity controls are maintained. Because the SLCS is
manually initiated and not assumed to mitigate any accident scenario
during refueling operations, the proposed changes do not affect the
consequences of any previously evaluated accident. As such, these
changes will not significantly increase the probability or
consequences of a previously evaluated accident.
The associated systems related to this proposed amendment are
not assumed in any safety analysis to initiate any accident sequence
for Dresden or Quad Cities Stations. In addition, the revisions
proposed to the surveillance requirements are administrative in
nature and either relocate procedural details to administrative
controls or allow provisions for manual alignment of a manual system
to the proper orientation. As such, because there is no effect on
any accident scenario, the probability of any accident previously
evaluated is not increased by the proposed amendment. Because the
proposed changes are administrative in nature, the consequences of
any previously evaluated accident are not increased.
(2) Create the possibility of a new or different kind of
accident from any accident previously evaluated because:
The proposed amendment for Dresden and Quad Cities Station's
Technical Specification is based on generic guidance or NRC accepted
changes for later operating BWR plants. The proposed amendment has
been reviewed for acceptability at the Dresden and Quad Cities
Nuclear Power Stations considering similarity of system or component
design versus the generic guidance. The proposed changes do not
create the possibility of a new or different kind of accident
previously evaluated for Dresden or Quad Cities Stations. No new
modes of operation are introduced by the proposed changes. SLCS
requirements are adequately retained to ensure sufficient controls
remain during plant operations. The proposed changes to the
Applicability and Actions during refueling operations for the SLCS
do not create a new or different kind of previously evaluated
accident. Because the SLCS is manually initiated to mitigate
accident concerns during power operations, the proposed deletion of
Applicability and Actions during refueling operations does not
affect the probability of a new or different kind of accident from
being created. The changes proposed to the surveillance requirements
are administrative in nature and do not affect the system operation;
as such, the proposed changes do not affect the probability of a new
or different kind of accident being created. Therefore, the proposed
changes do not create the possibility of a new or different kind of
accident from any previously evaluated.
The associated systems related to this proposed amendment are
not assumed in any safety analysis to initiate any accident sequence
for Dresden or Quad Cities Stations; therefore, the proposed changes
do not create the possibility of a new or different kind of accident
from any previously evaluated.
(3) Involve a significant reduction in the margin of safety
because:
The proposed amendment represents the conversion of current
requirements which are based on generic guidance or previously
approved provisions for other stations. The proposed changes are
consistent with NUREG-1433 and do not adversely affect existing
plant safety margins or the reliability of the equipment assumed to
operate in the safety analysis. The proposed changes have been
evaluated and found to be acceptable for use at Dresden or Quad
Cities based on system design, safety analysis requirements and
operational performance. SLCS provisions continue to be adequately
maintained during plant operation. The proposed changes to the
Applicability and Actions during refueling operations for the SLCS
do not significantly reduce existing plant safety margins. Because
the SLCS is manually initiated to mitigate accident concerns during
power operations, the proposed deletion of Applicability and Actions
during refueling operations has no effect on existing plant safety
margins as this system is not required during this mode of
operation. The changes proposed to the surveillance requirements are
administrative in nature and do not affect the system operation; as
such, the proposed changes do not adversely affect existing plant
safety margins as adequate system surveillance requirements are
maintained. Since the proposed changes are based on NRC accepted
provisions at other operating plants that are applicable at Dresden
or Quad Cities and maintain necessary levels of system or component
reliability, the proposed changes do not involve a significant
reduction in the margin of safety.
The proposed amendment for Dresden and Quad Cities Stations will
not reduce the availability of systems required to mitigate accident
conditions; therefore, the proposed changes do not involve a
significant reduction in the margin of safety.
Guidance has been provided in ``Final Procedures and Standards
on No Significant Hazards Considerations,'' Final Rule, 51 FR 7744,
for the application of standards to license change requests for
determination of the existence of significant hazards
considerations. This document provides examples of amendments which
are and are not considered likely to involve significant hazards
considerations.
This proposed amendment does not involve a significant
relaxation of the criteria used to establish safety limits, a
significant relaxation of the bases for the limiting safety system
settings or a significant relaxation of the bases for the limiting
conditions for operations. Therefore, based on the guidance provided
in the Federal Register and the criteria established in 10 CFR
50.92(c), the proposed change does not constitute 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
requested amendments involve no significant hazards consideration.
Local Public Document Room location: for Dresden, Morris Area
Public Library District, 604 Liberty Street, Morris, Illinois 60450;
for Quad Cities, Dixon Public Library, 221 Hennepin Avenue, Dixon,
Illinois 61021.
Attorney for licensee: Michael I. Miller, Esquire; Sidley and
Austin, One First National Plaza, Chicago, Illinois 60603.
NRC Project Director: Robert A. Capra.
Commonwealth Edison Company, Docket Nos. 50-237 and 50-249, Dresden
Nuclear Power Station, Units 2 and 3, Grundy County, Illinois
Date of amendment request: November 7, 1997.
Description of amendment request: The proposed amendments would
[[Page 2278]]
relocate the Unit 2 24/48 Vdc batteries, chargers, and distribution
systems operability and surveillances requirements from the Technical
Specifications to licensee administratively controlled documents.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below.
(1) Involve a significant increase in the probability or
consequences of an accident previously evaluated because of the
following:
Removal of the Unit 2 24/48 Vdc battery, charger, and
distribution panel requirements from the Technical Specification
requirements of 3/4.9.C, 3/4.9.D, 3/4.9.E, and 3/4.9.F and the
subsequent relocation of those requirements to licensee
administrative controls is an administrative change that will
continue [to] ensure the availability of the Unit 2 24/48 Vdc system
and will not increase the probability of accidents previously
evaluated. Relocation of the Unit 2 24/48 Vdc requirements to
administrative controls will have no effect on the control
instrumentation and cannot act as an initiator for any of the
accidents evaluated in the UFSAR [Updated Final Safety Analysis
Report].
Similarly, relocation of the Unit 2 24/48 Vdc system
requirements to licensee administrative controls will have no effect
on the availability of the loads which are supplied by the Unit 2
24/48 Vdc batteries nor on any of the consequences of accidents
previously evaluated in the UFSAR. Control of the Unit 2 24/48 Vdc
requirements by licensee administrative controls under 10 CFR 50.59
will not affect any of the protection or mitigation functions which
may be provided by any of the loads supplied by the batteries.
Operation under the proposed amendment will not significantly
increase the probability or consequences of any accidents previously
evaluated.
Because of the above evaluation, removal of the Unit 2 24/48 Vdc
system from the Technical Specifications will 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 accident previously evaluated because:
The Unit 2 24/48 Vdc batteries, chargers, and other components
will retain the separation, and redundancy under which they are
presently installed. No new failure modes are introduced by this
administrative relocation of requirements, for the Unit 2 24/48 Vdc
system, from the Technical Specifications to licensee administrative
control. Since the batteries are not being operated differently and
transferring ATS [Analog Trip System] loads to the 125 Vdc safety-
related battery system does not affect the function or mode of
operation of these loads, the possibility of a new or different
accident from any accident previously evaluated is not increased or
created by this administrative change.
(3) Involve a significant reduction in the margin of safety
because:
Relocation of the TS requirements for the Unit 2 24/48 Vdc
system does not affect the operating points or setpoints of any
systems or components. Plant operating points or parameters are not
changed by the proposed relocation of requirements in this amendment
request. The safety-related equipment that is supported by the Unit
2 24/48 Vdc system will continue to be required in the existing
modes of applicability as determined by the individual equipment
Technical Specifications. Thus operation under the proposed license
amendment removes some redundancy and constraints during refueling
but does not significantly reduce the margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
requested amendments involve no significant hazards consideration.
Local Public Document Room location: Morris Area Public Library
District, 604 Liberty Street, Morris, Illinois 60450.
Attorney for licensee: Michael I. Miller, Esquire; Sidley and
Austin, One First National Plaza, Chicago, Illinois 60603.
NRC Project Director: Robert A. Capra.
Commonwealth Edison Company, Docket No. 50-373, LaSalle County Station,
Unit 1, LaSalle County, Illinois
Date of amendment request: November 24, 1997.
Description of amendment request: The proposed amendment would
revise Technical Specification (TS) 3/4.3.2, ``Isolation Actuation
Instrumentation,'' to add instrumentation for the reactor water cleanup
(RWCU) pump rooms and valve room as a result of modifications to the
RWCU system. Also, additional instrumentation will be added in the RWCU
holdup pipe area, the filter/demineralizer valve rooms, and RWCU pump
suction high flow switch as a result of a high energy line break re-
evaluation. The setpoints for the RWCU heat exchanger room
instrumentation will be revised as a result of new design basis
calculations. The proposed amendment will also delete instrumentation
related to the residual heat removal (RHR) steam condensing mode which
is no longer utilized and will eliminate the alarm and isolation
functions for the RHR shutdown cooling mode.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
(1) Involve a significant increase in the probability or
consequences of an accident previously evaluated because:
(a) There is no effect on accident initiators so there is no
change in [the] probability of an accident. A line break in the
subject areas would consist of an instantaneous circumferential
break downstream of the outermost isolation valve of one of these
systems. The leak detection isolation is only a precursor of a
break, and thus does not affect the probability of a break.
(b) There is no or minimal effect on the consequences of
analyzed accidents, due to changing the leak detection ambient T or
Delta T setpoint and allowable values to detect 25 gpm equivalent
leakage. The addition of more ambient T and Delta T leak detection
monitoring, along with the addition of the high flow break detection
will actually decrease the consequences of the associated accidents.
The worst case accident outside the primary containment boundary is
a main steam line break which bounds the dose consequences of all
line breaks and therefore bounds any size of leak.
The deletion of the RHR steam condensing mode isolation
actuation instrumentation trip functions from the LaSalle TS does
not increase the probability or consequences of an accident
previously evaluated, because this mode of operation of the RHR
system has been deleted from the LaSalle design basis and the lines
that were previously high energy line are isolated during unit
operation, including Operational Condition 1 (Run mode), Operational
Condition 2 (Startup mode), and Operational Condition [3] (Hot
Shutdown).
The deletion of the RHR shutdown cooling mode leak detection T
and Delta T isolation actuation instrumentation trip functions from
the LaSalle TS does not increase the probability or consequences of
an accident previously evaluated, because the leak detection is only
a precursor of a break, and thus does not affect the probability of
a break. Also, there are two remaining different methods of
detecting abnormal leakage and isolating the system in technical
specification trip functions A.6.a, Reactor Vessel Water Level--Low,
Level 3 and A.6.c, RHR Pump Suction Flow--High. In addition, other
means to detect leakage from the RHR system, such as sump monitoring
and area radiation monitoring, are also available. In accordance
with TS Administrative Requirement 6.2.F.1, LaSalle has a leakage
reduction program to reduce leakage from those portions of systems
outside primary containment that contain radioactive fluids. RHR,
including piping and components associated with the shutdown cooling
mode, is part of this program, which includes periodic visual
inspection for system leakage. The sump monitoring, radiation
monitoring and periodic inspections for system leakage makes the
probability of a leak of 5 gpm going undetected for more than a day
very low.
Also, due to the low reactor pressures (less than 135 psig) at
which RHR shutdown
[[Page 2279]]
cooling mode is able to operate, reactor coolant makeup and outflow
is very low compared to normal plant operation. A change in flow
balance due to a leak is thus more readily detectable with reactor
coolant water level changes and makeup flow rate, and thus precludes
a significant leak going undetected before break detection
instrumentation would cause automatic isolation.
Therefore, there is not 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 accident previously evaluated because:
The purpose of the leak detection system, as it applies to the
RWCU and RHR system areas, is to provide the capability for leak
detection and automatic isolation as necessary of the system in the
event of leakage in these areas. This change maintains this
capability with at least two different methods of detection of
abnormal leakage for protection from the flooding concerns of a
significant leak or line break when the RHR system is operating in
the shutdown cooling mode, so that redundant systems will not be
affected.
This change also maintains or adds primary containment isolation
logic for the leak detection isolation based on temperature
monitoring in RWCU areas and break detection based on RWCU pump
suction flow--high. The additional instrumentation and the
associated isolation logic is the same or similar to existing
instrumentation and logic for containment isolation actuation
instrumentation, so no new failure modes are created in this way.
Therefore, the possibility of a new or different kind of
accident from any previously evaluated is not created.
(3) Involve a significant reduction in the margin of safety
because:
The change to the automatic isolation setpoint for high Delta T
leak detection in the heat exchanger rooms is based on current
configuration calculated/analyzed response to a small leak compared
to a circumferential break. The increased leakage rate in the RWCU
heat exchanger rooms that is necessary to actuate isolation on high
temperature during winter conditions, does not adversely affect the
margin of safety. This increased leakage rate is below the critical
crack leakage rate as represented in [Updated Final Safety Analysis
Report] UFSAR Figure 5.2-11. Additionally, differential temperature
leak detection is conservative under these same conditions, and will
actuate isolation at a leakage rate less than the established limit.
The leak detection isolation logic is unchanged and thus remains
single failure proof.
The addition of automatic primary containment isolation on
Ambient and Differential Temperature (Delta T)-High for the Reactor
Water Cleanup System (RWCU) Pump, Pump Valve, Holdup Pipe, and
Filter/Demineralizer (F/D) Valve Rooms and the addition of the RWCU
Pump Suction Flow High line break isolation add to the margin of
safety with respect to leak detection and line breaks in the RWCU
system, because the system isolation diversity is increased and the
amount of system piping monitored for leakage is increased.
The setpoints for the ambient temperature and differential
temperature leak detection isolations being changed or added and the
RWCU pump suction flow--high are set sufficiently high enough so as
not to increase the possibility of spurious actuation. In the event
that a spurious actuation does occur, little safety significance is
presented since the RWCU system performs no safety function. The
setpoints and allowable values for the proposed changes also assure
sufficient margin to the analytical values and [are] high enough to
prevent spurious actuations based on calculations consistent with
Regulatory Guide 1.105.
The deletion of the RHR steam condensing mode isolation
actuation instrumentation does not effect the margin of safety,
because this mode is no longer utilized by LaSalle in Operational
Conditions 1, 2, or 3 (Run mode, Startup Mode, or Hot Shutdown).
The elimination of the temperature based trip functions for the
RHR shutdown cooling mode area is based on the determination that
temperature is not the appropriate parameter as it does not provide
meaningful indication and will not provide setpoints that would be
sufficiently above the normal range of ambient conditions to avoid
spurious isolations.
There are two remaining different methods of detecting abnormal
leakage and isolating the system in technical specification trip
function A.6, namely A.6.a, Reactor Vessel Water Level--Low, Level 3
and A.6.c, RHR Pump Suction Flow--High. In addition, other means to
detect leakage from the RHR system, such as sump monitoring and area
radiation monitoring, are also available. Also, in accordance with
TS Administrative Requirement 6.2.F.1, LaSalle has a leakage
reduction program to reduce leakage from those portions of systems
outside primary containment that contain radioactive fluids. RHR,
including piping and components associated with the shutdown cooling
mode, is part of this program, which includes periodic visual
inspection of system for leakage.
The previous evaluation of diversity of isolation parameters, as
presented in Table 5.2-8 of the UFSAR remains unchanged. Adequate
diversity of isolation parameters is maintained because there are at
least two different methods available to detect and allow isolation
of the system for a line break, as necessary.
Therefore, this requested Technical Specification amendment 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
requested amendment involves no significant hazards consideration.
Local Public Document Room location: Jacobs Memorial Library,
Illinois Valley Community College, Oglesby, Illinois 61348.
Attorney for licensee: Michael I. Miller, Esquire; Sidley and
Austin, One First National Plaza, Chicago, Illinois 60603.
NRC Project Director: Robert A. Capra.
Detroit Edison Company, Docket No. 50-341, Fermi 2, Monroe County,
Michigan
Date of amendment request: December 10, 1997 (NRC-97-0105).
Description of amendment request: The proposed amendment would
modify the technical specifications (TS) and the bases to accommodate
the installation of an improved power range neutron monitoring system.
The modification and the TS changes represent part of the licensee's
actions in response to Generic Letter 94-02, ``Long-Term Solutions and
Upgrade of Interim Operating Recommendations for Thermal-Hydraulic
Instabilities in Boiling Water Reactors,'' dated July 11, 1994. The TS
revisions include changes to Action Statements and Surveillance
Requirements which are generally consistent with licensing topical
report NEDC-32410P-A, ``Nuclear Measurement Analysis and Control Power
Range Neutron Monitor (NUMAC PRNM) Retrofit Plus Option III Stability
Trip Function,'' and NEDC-32410P Supplement 1, ``NUMAC PRNM Retrofit
Plus Option III Stability Trip Function,'' which were reviewed by the
NRC as documented in a letter dated September 5, 1995, and a safety
evaluation dated August 15, 1997. The proposed amendment also includes
two unrelated changes. Surveillance Requirement 4.3.1.3 and its
associated bases are modified to clarify the applicability of response
time testing requirements. In addition, the first page of Table 3.3.6-2
is modified to correct a typographical error in the title.
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 consequences of an accident previously
evaluated.
This proposed TS change is associated with the NUMAC-PRNM
retrofit design. The proposed TS change involves modification of the
Limiting Conditions for Operation (LCOs) and Surveillance
Requirements (SRs) for equipment designed to mitigate events that
result in power increase transients. The APRM [average power range
monitor] system mitigative action is to block control rod withdrawal
or initiate a reactor scram, which terminates the power increase
when setpoints are exceeded. The Rod Block
[[Page 2280]]
Monitor (RBM) system mitigative action is to block continuous
control rod withdrawal prior to exceeding the fuel design limits
during a postulated Rod Withdrawal Error. The functional capability
of the previous Reactor Coolant System Recirculation Flow control
rod block trip functions have been incorporated into the modified
APRM control rod block trip functions. The worst case failure of
either the APRM or the RBM systems is failure to initiate mitigative
action (failure to scram or block rod withdrawal). Failure to
initiate mitigative action will not increase the probability of an
accident. Thus, the proposed change does not increase the
probability of an accident previously evaluated.
For the APRM and the RBM systems, the NUMAC PRNM design,
together with revised operability requirements (LCOs) and revised
testing requirements (SRs), continues to perform the same mitigation
functions under identical conditions with availability comparable to
the types of equipment that it replaces. Because there is no change
in mitigation functions and because availability of the functions is
maintained, the proposed change does not involve a significant
increase in the 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 involve modification and replacement of the
existing power range neutron monitoring equipment, and modification
of the setpoints and operational requirements for the APRM and RBM
systems. These proposed changes do not modify the basic functional
requirements of the affected equipment, create any new system
interfaces or interactions, nor create any new system failure modes
or sequence of events that could lead to an accident. The worst case
failure of the affected equipment is failure to perform a mitigation
action, and failure of this mitigative equipment does not create the
possibility of a new or different kind of accident. The proposed
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 proposed TS change is associated with the NUMAC PRNM
retrofit design. The NUMAC PRNM change does not impact reactor
operating parameters nor the functional requirements of the power
range neutron monitoring system. The replacement equipment continues
to provide information, enforce control rod blocks and initiate
reactor scrams under appropriate specified conditions. The proposed
change does not revise any safety margin requirements. The
replacement APRM/RBM equipment has improved channel trip accuracy
compared to the current system and meets or exceeds system
requirements previously assumed in setpoint analysis. Thus, the
ability of the new equipment to enforce compliance with margins of
safety equals or exceeds the ability of the equipment which it
replaces. The proposed change does 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. The editorial change in Table 3.3.6-2 and the clarification
in Surveillance Requirement 4.3.1.3 also satisfy the three standards 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: Monroe County Library System,
3700 South Custer Road, Monroe, Michigan 48161.
Attorney for licensee: John Flynn, Esq., Detroit Edison Company,
2000 Second Avenue, Detroit, Michigan 48226.
NRC Project Director: John N. Hannon.
Northeast Nuclear Energy Company, et al., Docket No. 50-336, Millstone
Nuclear Power Station, Unit No. 2, New London County, Connecticut
Date of amendment request: December 1, 1997.
Description of amendment request: The proposed amendment would
change the Technical Specifications (TSs) to add a time delay,
including allowance, to a portion of the Engineered Safety Feature
Actuation System undervoltage (UV) trip TSs. The proposed changes would
result in the TSs being consistent with the current design, as detailed
in the Final Safety Analysis Report, and the current surveillance
procedures.
Specifically, TS Table 3.3-4, Loss of Power, would be changed by
adding a 2.0 [plus or minus] 0.1 second time delay for the 4.16 kV
Emergency Bus UV (UV Relays) level 1--Trip Setpoint and the Allowable
values.
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 concludes that these proposed additions to Technical
Specification Table 3.3-4 do not involve a significant hazards
consideration (SHC) and do not involve a significant impact on
public health and safety. The basis for this conclusion is that the
three criteria of 10 CFR 50.92(c) are not compromised. That is, the
proposed changes do not:
1. Involve a significant increase in the probability or
consequences of an accident previously evaluated.
The proposed changes will add a time delay, including allowance,
to a portion of the Engineered Safety Feature Actuation System
(ESFAS) Undervoltage (UV) Trip Technical Specification Table 3.3-4.
These changes will align the Technical Specifications to the
existing plant design, as described in the Final Safety Analysis
Report (FSAR) system description and the existing surveillance
procedure. No new plant modifications are associated with this
addition to the Technical Specifications.
The addition of the Level One UV trip time delay setpoint does
not impact any system or component whose failure results in
initiation of the accidents described in the FSAR. Therefore, the
changes do not affect the probability of occurrence of the
previously evaluated accidents. The Level One UV trip time delay
potentially affects the Emergency Diesel Generator (EDG) response
time to accident conditions that occur coincident with a loss of
normal power (LPN). However, previous analysis of the increase in
the time delay (0.5 seconds to 2.0 [plus or minus] 0.1 second)
concluded that the ESFAS response times for those events considered
to occur coincident with an LNP, are not challenged by the time
delay. This conclusion is based upon a comparison between the EDG
start time and the maximum time required to complete those LNP trip
functions necessary to support EDG availability for worst case
accident conditions (Loss of Coolant Accident which results in a
Safety Injection Actuation Signal (SIAS) coincident with LNP). The
calculated EDG start time considered the ESFAS response time (0.5
seconds) in addition to the maximum EDG start time of 15 seconds
after receipt of an SIAS, as specified in Technical Specification
Surveillance Requirement 4.8.1.1.2.a.2. Since the calculated LNP
trip time delay of 15.14 seconds is less than the calculated SIAS
initiated EDG start time of 15.5 seconds, the proposed changes do
not increase the likelihood of an EDG malfunction during an accident
condition. Consequently, the proposed additions do not adversely
affect the ability of either the ESFAS or the EDGs to perform their
intended safety function. The proposed additions to Table 3.3-4 do
not modify the Limiting Condition for Operation or the specific
surveillance procedure acceptance criterion, nor do they change the
frequency of the surveillance. The proposed changes do not involve
any physical changes to the plant and do not alter the way any
structure, system, or component functions. The proposed changes do
not have any adverse impact on the design basis accidents previously
analyzed. The proposed changes do not result in an increase in
radiation exposure to either members of the public or site personnel
because accident mitigation systems will be available consistent
with the assumptions used in the accident analysis. Therefore, the
proposed additions to Technical Specification Table 3.3-4 do not
affect the consequences of the previously evaluated accidents.
Based on the above, the proposed changes do not involve a
significant increase in the probability of consequences of an
accident previously evaluated.
2. Create the possibility of a new or different kind of accident
from any accident previously evaluated.
The function availability and failure modes of equipment
important to safety are
[[Page 2281]]
unaffected by the addition of the 2.0 [plus or minus] 0.1 second
Level One UV trip time delay to Technical Specification Table 3.3-4.
The additions do not introduce any new, credible accidents, or any
new failure modes.
Therefore, the proposed changes do not create the possibility of
a new or different kind of accident from any accident previously
evaluated.
3. Involve a significant reduction in a margin of safety.
The proposed additions to Technical Specification Table 3.3-4 do
not have any adverse impact on the accident analyses. Actuation of
the required safety systems is not delayed because the proposed
additions do not delay the time at which the EDGs are required, by
the plant Technical Specifications, to be available to power the
required loads.
Therefore, based on the above, there is no 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, Connecticut, and the Waterford Library, ATTN: Vince Juliano,
49 Rope Ferry Road, Waterford, Connecticut.
Attorney for licensee: Lillian M. Cuoco, Esq., Senior Nuclear
Counsel, Northeast Utilities Service Company, P.O. Box 270, Hartford,
Connecticut.
NRC Deputy Director: Phillip F. McKee.
Public Service Electric & Gas Company, Docket Nos. 50-272 and 50-311,
Salem Nuclear Generating Station, Unit Nos. 1 and 2, Salem County, New
Jersey
Date of amendment request: December 15, 1997.
Description of amendment request: The proposed amendments would
revise the Technical Specifications (TSs) to adopt Option B, of 10 CFR
Part 50, Appendix J, to implement a performance-based approach for Type
B and C testing. Additionally, the wording in the TSs would be modified
for the previous adoption of Option B on Type A testing and a section
added on the primary containment leakage rate testing program.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. The proposed changes do not involve a significant increase in
the probability or consequences of an accident previously evaluated.
Containment leak rate testing is not an initiator of any
accident. The proposed changes do not make any physical changes to
the containment and do not affect reactor operations or the accident
analyses. Therefore, the proposed changes do not involve a
significant increase in the probability of any previously evaluated
accident.
Since the allowable leakage rate is not being changed and since
the analysis documented in NUREG-1493, ``Performance-Based
Containment Leak-Rate Program'' concludes that the impact on public
health and safety due to extended intervals is negligible, the
proposed changes will not involve a significant increase in the
consequences of any previously evaluated accident.
Therefore, adoption of a performance-based leakage testing
requirements will provide an equivalent level of safety and does not
involve a significant increase in the probability or consequences of
an accident previously evaluated.
2. The proposed change does not create the possibility of a new
or different kind of accident from any accident previously
evaluated.
No physical changes are being made to the plant, nor are there
any changes being made to the operation of the plant as a result of
the proposed changes. In addition, no new failure modes of plant
equipment previously evaluated are being introduced.
Therefore, the proposed amendment 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 proposed changes are based on NRC-approved provisions and
maintain adequate levels of reliability of containment integrity.
The performance-based approach to leakage rate testing recognizes
that historically good results of containment testing provide
appropriate assurance of future containment integrity. This supports
the conclusion that the impact on the health and safety of the
public as a result of extended test intervals is negligible. Since
the analysis documented in NUREG-1493 confirms that the performance
based schedule continues to maintain a minimal impact on public
risk, it can be concluded that the margin of safety is not
significantly affected by the proposed changes.
Therefore, the proposed amendment will not involve a significant
reduction in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: Salem Free Public Library, 112
West Broadway, Salem, NJ 08079.
Attorney for licensee: Jeffrie J. Keenan, Esquire, Nuclear Business
Unit--N21, P.O. Box 236, Hancocks Bridge, NJ 08038.
NRC Project Director: John F. Stolz.
Southern Nuclear Operating Company, Inc, Docket Nos. 50-348 and 50-364,
Joseph M. Farley Nuclear Plant, Units 1 and 2, Houston County, Alabama
Date of amendments request: July 23, 1997, as supplemented
September 30 and December 18, 1997. The July 23, 1997, application was
previously noticed in the Federal Register on September 10, 1997 (62 FR
47699). The December 18, 1997, supplement provided additional
information that revised the licensee's evaluation of the significant
hazards consideration. Therefore, renotification of the Commission's
proposed determination of no significant hazards consideration is
necessary.
Description of amendments request: The proposed amendments would
revise the Technical Specifications (TSs) by relocating the reactor
coolant system pressure and temperature limits from the TSs to the
proposed Pressure Temperature Limits Report in accordance with the
guidance provided by Generic Letter 96-03, ``Relocation of the Pressure
Temperature Limit Curves and Low Temperature Overpressure Protection
System Limits.''
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. The proposed changes do not involve a significant increase in
the probability or consequences of an accident previously evaluated.
The proposed removal of the Reactor Coolant System (RCS)
pressure temperature (P-T) limits from the Technical Specifications
(TSs) and relocation to the proposed Pressure Temperature Limits
Report (PTLR) in accordance with the guidance provided by Generic
Letter (GL) 96-03 is administrative in that the requirements for the
P-T limits are unchanged. The P-T limits proposed for inclusion in
the PTLR are based on the fluence associated with 2775 MW thermal
power and operation through 21.9 effective full power years (EFPY)
for Unit 1 and 33.8 EFPY for Unit 2. GL 96-03 requires that the P-T
limits be generated in accordance with the requirements of 10 CFR
[Part] 50, Appendices G and H, documented in an NRC-approved
methodology incorporated by reference in the TSs.
[[Page 2282]]
Accordingly, the proposed curves have been generated using the NRC-
approved methods described in WCAP-14040-NP-A, Revision 2, as
modified at the direction of the NRC Staff, and meet the
requirements of 10 CFR [Part] 50, Appendices G and H. TS 3.4.10.1
will continue to require that the RCS pressure and temperature be
limited in accordance with the limits specified in the PTLR. The
NRC-approval document will be specified in TS 6.9.1.15 and NRC
approval will be required in the form of a TS Amendment prior to
changing the methodology. Use of P-T limit curves generated using
the NRC-approved methods will provide additional protection for the
integrity of the reactor vessel, thereby assuring that the reactor
vessel is capable of providing its function as a radiological
barrier.
TS 3.4.10.3 for Farley Nuclear Plant (FNP) Unit 1 and Unit 2
provides the operability requirements for RCS low temperature
overpressure protection (LTOP). Specifically, TS 3.4.10.3 requires
that two residual heat removal (RHR) system suction relief valves
(RHRRVs) be operable or that the RCS be vented at RCS cold leg
temperatures less than or equal to 310[ deg.]F. Consistent with GL
96-03, the Farley Unit 1 and Unit 2 requirements for LTOP will be
retained in TS 3.4.10.3 and will be evaluated in accordance with the
proposed methodology.
Based on the above evaluation, the proposed changes are
administrative in nature and do not involve a significant increase
in the probability or consequences of an accident previously
evaluated.
2. The proposed change does not create the possibility of a new
or different kind of accident from any accident previously
evaluated.
As stated above, the proposed changes to remove the RCS P-T
limits from the TSs and relocate them to the proposed PTLR is an
administrative change. Consistent with the guidance provided by GL
96-03, the proposed P-T limits contained in the proposed PTLR meet
the requirements of 10 CFR [Part] 50, Appendices G and H, and were
generated using the NRC-approved methods described in WCAP-14040-NP-
A, Revision 2, as modified at the direction of the NRC Staff. The
proposed changes do not result in a physical change to the plant or
add any new or different operating requirements on plant systems,
structures, or components with the exception of limiting the number
of operating RCPs at RCS temperatures below 110[ deg.]F. Limiting
the number of operating RCPs below 110[ deg.]F results in a
reduction in the [delta]P between the reactor vessel beltline and
the RHRRVs, thereby providing additional margin to limits of
Appendix G. Provisions are made to allow the start of a second RCP
at temperatures below 110[ deg.]F in order to secure the pump that
was originally operating without interrupting RCS flow. The LTOP
enable temperature exceeds the minimum LTOP enable temperature
determined as described in WCAP-14040-NP-A, Rev. 2, thereby
providing additional assurance that the LTOP system will be
available to protect the RCS in the event of an overpressure
transient at RCS temperatures at or below 310[ deg.]F. Based on the
methods contained in WCAP-14040-NP-A, Rev. 2, the minimum boltup
temperature for the reactor vessel flange region is conservatively
established as 70[ deg.]F.
As stated in the above response, implementation of the proposed
changes do not result in a significant increase in the probability
of a new or different accident (i.e., loss of reactor vessel
integrity). The RCS P-T limits will continue to meet the
requirements of 10 CFR [Part] 50, Appendices G and H, and will be
generated in accordance with the NRC approved methodology described
in WCAP-14040-NP-A, Revision 2, as modified at the direction of the
NRC Staff. Therefore, the proposed changes do not result in a
significant increase in the possibility of a new or different
accident from any previously evaluated.
3. The proposed change does not involve a significant reduction
in a margin of safety.
The margin of safety is not affected by the removal of the RCS
P-T limits from the TSs and relocating them to the proposed PTLR.
The RCS P-T limits will continue to meet the requirements of 10 CFR
50, Appendices G and H. To provide additional assurance that the P-T
limits continue to meet the requirements of Appendices G and H, TS
6.9.1.15 will require the use of the NRC-approved methodology to
generate P-T limits. The RCS LTOP requirements will be retained in
TS 3.4.10.3 due to use of the RHRRVs for LTOP, consistent with the
guidance provided by GL 96-03, and will be verified to provide
adequate protection of the reactor coolant system against the limits
of Appendix G. The LTOP enable temperature exceeds the LTOP enable
temperature determined in accordance with the NRC-approved
methodology, thus protecting the RCS in the event of a low
temperature overpressure transient over a broader range of
temperatures than required by WCAP-14040-NP-A, Rev. 2.
Administrative procedures will preclude operation of the RCS at
temperatures below the minimum boltup temperature for the reactor
vessel head, thus precluding the possibility of tensioning the
reactor vessel head at RCS temperatures below the minimum boltup
temperature. Operation of the plant in accordance with the RCS P-T
limits specified in the PTLR and continued operation of the LTOP
system in accordance with TS 3.4.10.3 will continue to meet the
requirements of 10 CFR [Part] 50, Appendices G and H, and will
therefore, assure that a margin of safety is not significantly
decreased as the result of the proposed changes.
Based on the preceding analysis, SNC [Southern Nuclear Operating
Company] has determined that removal of the RCS P-T limits from the
TS and relocation to the proposed PTLR will not significantly
increase the probability or consequences of an accident previously
evaluated, create the possibility of a new or different kind of
accident from any accident previously evaluated, or involve a
significant reduction in a margin of safety. SNC therefore concludes
that the proposed change meets the requirements of 10 CFR 50.92(c)
and does 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: Houston-Love Memorial Library,
212 W. Burdeshaw Street, Post Office Box 1369, Dothan, Alabama.
Attorney for licensee: M. Stanford Blanton, Esq., Balch and
Bingham, Post Office Box 306, 1710 Sixth Avenue North, Birmingham,
Alabama.
NRC Project Director: Herbert N. Berkow.
Union Electric Company, Docket No. 50-483, Callaway Plant, Unit 1,
Callaway County, Missouri
Date of application request: October 17, 1997.
Description of amendment request: The proposed amendment would
modify the technical specifications (TS) for plant heatup and cooldown
curves and the maximum allowable power operated relief valve (PORV)
setpoint curve for cold overpressure protection, as found in TS Figures
3.4-2, 3.4-3, and 3.4-4. These changes are requested to incorporate
information gained from Surveillance Capsule V, which was removed
during Callaway Refuel 8 in the fall of 1996 after 9.85 effective full
power years (EFPY) of exposure. Capsule V is the third capsule to be
removed from the reactor vessel in the continuing surveillance program
that monitors the effects of neutron irradiation on the Callway reactor
vessel materials under actual plant operating conditions. The proposed
changes include:
(1) Figure 3.4-2, heatup limitation curve and Figure 3.4-3,
cooldown limitation curve, would be revised to reflect the
TRNDT calculated for 20 EFPY in the surveillance capsule
report.
(2) Figure 3.4-4 is the maximum allowable PORV setpoint curve for
cold overpressure protection. This curve would be (a) revised to
account for the changes made in the heatup and cooldown limitation
curves, (b) allow for the operation of the normal charging pump, and
(c) account for instrument accuracy and other uncertainties.
(3) TS Bases 3/4.4.9 and 3/4.5.2 through 3/4.5.4 would be revised
by correcting miscellaneous items and by adding discussion of the
normal charging pump.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the
[[Page 2283]]
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.
Pressure and temperature limits for the reactor pressure vessel
(RPV) are established to the requirements of 10 CFR 50, Appendix G
to ensure brittle fracture of the vessel does not occur. This
amendment revises the P/T curves in the TS to reflect the material
capsule surveillance results from the sample removed during the fall
outage of 1996.
The RPV surveillance capsule contained flux wires for neutron
flux monitoring and Charpy V notch impact and tensile test
specimens. The irradiated material properties were compared to
available unirradiated properties to determine the effect of
irradiation on material toughness for the base and weld materials
through Charpy testing. Irradiated tensile testing results are
compared with unirradiated data to determine the effect of
irradiation on the stress-strain relationship of the materials.
The P/T curves are modified to reflect the results of the above
examination. These curves and their operating limits were generated
using the NRC-approved methods described in WCAP-14040-NP-A,
Revision 2 and meet the requirements of 10 CFR 50, Appendices G and
H as modified by the provisions of ASME Code Case N-514. The new
curves therefore represent the latest information available on the
state of the reactor vessel materials. The P/T curves are generated
for reactor vessel protection against brittle fracture, they do not
affect the recirculation piping. Accordingly, the probability of
occurrence of a design basis Loss of Coolant Accident (LOCA) is not
increased. Likewise, no other previously evaluated accident and
transients, as defined in Chapter 15 of the Final Safety Analysis
Report are affected by this proposed change to the Callaway P/T
curves. Additionally, this proposed revision does not affect the
design, operation, or maintenance of any safety-related system
designed for the mitigation or prevention of previously analyzed
events.
Since no previously evaluated accidents or transients are
affected by this change, their probability of occurrence and
consequences is not increased.
2. The proposed change does not create the possibility of a new
or different kind of accident from any accident previously
evaluated.
Implementing the proposed P/T curves into the TS does not alter
the design or operation of any system or piece of equipment designed
for the prevention or mitigation of accidents and transients. As a
result, no new operating modes are introduced from which a new type
accident becomes possible. Existing systems will continue to be
operated per present design basis assumptions.
The proposed P/T limits were generated from the evaluation of
the material capsule removed during the fall outage of 1996 using
the NRC-approved methods described in WCAP-14040-NP-A, Revision 2.
As a result, these limits include the latest available information
on the reactor vessel materials. Furthermore, they will continue to
be monitored per the requirements of the TS and 10 CFR 50,
Appendices G and H. For the above reasons, the changes do not create
the possibility of a new type of accident.
3. The proposed change does not involve a significant reduction
in a margin of safety.
The purpose of the P/T limits is to avoid a brittle fracture of
the reactor vessel. As such, material capsules are removed
periodically to determine the effects of neutron irradiation on the
reactor vessel materials. This change to the Callaway curves is
proposed to incorporate the evaluation results of the latest capsule
removed during the fall outage of 1996. Accordingly, these curves
represent the latest information available on the reactor vessel
materials.
Also, the curves were generated using the approved methodologies
of 10 CFR 50, Appendix G.
The Cold Overpressure Mitigation System Curve (Figure 3.4-4) is
also revised to reflect exposure dependencies. This curve was
generated for 20 EFPY using approved methodologies and reflects the
results of this latest material capsule report. Utilizing the
methodology set forth in ASME Section XI, Appendix G, which includes
the provisions of Code Case N-514, and 10 CFR 50, Appendices G and H
ensures that proper limits and conservative safety factors are
maintained.
The proposed changes do not affect the evaluation of any FSAR
Chapter 15 transient and accident. Furthermore, the proposed change
does not affect the operation of systems or equipment important to
safety.
The Limiting Condition for Operation of Specification 3.4.9 will
not change. Also, no TS surveillance or surveillance frequencies are
revised as a result of this Technical Specification submittal,
besides the fact that the P/T surveillance will now refer to the
revised curves. Procedures regarding the monitoring of the P/T
limits during reactor startup, cooldown, and leakage testing will
not change as a result of this proposed Technical Specification
change with respect to frequency of the surveillance or the methods
used to perform the surveillance. Thus, the P/T limits will continue
to be surveilled as before per the same procedures and at the same
frequencies.
No other Technical Specifications are affected by the proposed
revision. The margin of safety to any Technical Specification safety
limit therefore is not reduced. For the above reasons the new curves
do not represent a significant reduction in the margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: Callaway County Public
Library, 710 Court Street, Fulton, Missouri 65251.
Attorney for licensee: Gerald Charnoff, Esq., Shaw, Pittman, Potts
& Trowbridge, 2300 N Street, N.W., Washington, D.C. 20037.
NRC Project Director: William H. Bateman.
Union Electric Company, Docket No. 50-483, Callaway Plant, Unit 1,
Callaway County, Missouri
Date of application request: October 31, 1997.
Description of amendment request: The proposed amendment would
revise the engineered safety features actuation system (ESFAS)
Functional Unit 6.f, Loss of Offsite Power-Start Turbine-Driven Pump,
in Technical Specification Tables 3.3-3, 3.3-4, and 4.3-2. The tables
would be revised to create separate functional units for the analog and
digital portions of the ESFAS function associated with starting the
turbine-driven auxiliary feedwater pump (TDAFP) upon a loss of offsite
power.
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.
Overall protection system performance will remain within the
bounds of the previously performed accident analyses since no
hardware changes are proposed. The recognition that different
operability and surveillance requirements apply to analog vs.
digital circuitry does not impact any previously analyzed accidents.
The proposed change will not affect any of the analysis assumptions
for any of the accidents previously evaluated. The proposed change
does not alter the current method or procedures for meeting the
surveillance requirements in Table 4.3-2. The proposed change will
not affect the probability of any event initiators nor will the
proposed change affect the ability of any safety-related equipment
to perform its intended function. There will be no degradation in
the performance of nor an increase in the number of challenges
imposed on safety-related equipment assumed to function during an
accident situation. Therefore, the proposed change does not involve
a significant increase in the probability or consequences of an
accident previously evaluated.
2. The proposed change does not create the possibility of a new
or different kind of accident from any accident previously
evaluated.
There are no hardware changes nor are there any changes in the
method by which any safety-related plant system performs its
[[Page 2284]]
safety function. The separation of analog and digital portions of
Functional Unit 6.f will not impact the normal method of plant
operation.
The operability requirements, ACTION Statement, and surveillance
requirements for the analog portion, new Functional Unit 6.f.1), are
identical to those of Functional Unit 8.a, while the requirements
for the digital portion, new Functional Unit 6.f.2), are consistent
with the current Technical Specifications, other than the new ACTION
Statement 39 provisions that defer to the TDAFP Specification
3.7.1.2 requirements and the performance of a TADOT during
appropriate plant conditions. These changes do not change any ESFAS
design standards and are appropriate for digital functions such as
this. No new accident scenarios, transient precursors, failure
mechanisms, or limiting single failures are introduced as a result
of this change. Therefore, the proposed change does 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 proposed change does not affect the acceptance criteria for
any analyzed event. There will be no effect on the manner in which
safety limits or limiting safety system settings are determined nor
will there be any effect on those plant systems necessary to assure
the accomplishment of protection functions. There will be no impact
on any margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: Callaway County Public
Library, 710 Court Street, Fulton, Missouri 65251.
Attorney for licensee: Gerald Charnoff, Esq., Shaw, Pittman, Potts
& Trowbridge, 2300 N Street, N.W., Washington, D.C. 20037.
NRC Project Director: William H. Bateman.
Union Electric Company, Docket No. 50-483, Callaway Plant, Unit 1,
Callaway County, Missouri
Date of application request: October 31, 1997.
Description of amendment request: The proposed amendment would
change Technical Specification Tables 2.2-1, 4.3-1, and 3.3-4, as well
as their associated Bases, in order to reduce repeated alarms, rod
blocks, and partial reactor trips that continue to manifest themselves,
especially during beginning of cycle operation following refueling
outages. Besides the potential for distracting operator attention away
from more safety significant evolutions, these occurrences have also
led to power reductions during surveillance testing in order to avoid
reactor trips, since the channel being tested is placed in the tripped
condition. These changes to various setpoint terms associated with the
overtemperature delta T, overpower delta T, and steam generator (SG)
water level low-low vessel delta T (Power-1 and Power-2) reactor trip
and auxiliary feedwater (AFW) start engineered safety feature actuation
system (ESFAS) functions will improve plant operations and reduce the
potential for unnecessary reactor trips, with no detrimental effect on
the plant's safety analysis or licensing basis.
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.
Overall protection system performance will remain within the
bounds of the previously performed accident analyses since no
hardware changes are proposed. The protection systems will continue
to function in a manner consistent with the plant design basis. The
proposed changes will not affect any of the analysis assumptions for
any of the accidents previously evaluated. The proposed changes will
not affect the probability of any event initiators nor will the
proposed changes affect the ability of any safety-related equipment
to perform its intended function. There will be no degradation in
the performance of nor an increase in the number of challenges
imposed on safety-related equipment assumed to function during an
accident situation. There will be no change to normal plant
operating parameters or accident mitigation capabilities. Therefore,
the proposed changes do not involve a significant increase in the
probability or consequences of an accident previously evaluated.
2. The proposed change does not create the possibility of a new
or different kind of accident from any accident previously
evaluated. There are no hardware changes associated with this
license amendment 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 unchanged. The Overtemperature
delta T Allowable Value increase is justified by the use of existing
setpoint margin and elimination of conservatisms not required by the
safety analysis and licensing basis. There will be a reduction in
the incidence of alarms, rod stops, and partial reactor trips. There
will also be less of a need to reduce power during on-line
surveillance testing. These changes represent substantial plant
operational improvements.
No new accident scenarios, transient precursors, failure
mechanisms, or limiting single failures are introduced as a result
of these changes. There will be no adverse effect or challenges
imposed on any safety-related system as a result of these changes.
Therefore, the proposed changes do 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 proposed changes do not affect the acceptance criteria for
any analyzed event nor is there a change to any Safety Analysis
Limit (SAL). Maintaining the SAL preserves the margin of safety.
There will be no effect on the manner in which safety limits or
limiting safety system settings are determined nor will there be any
effect on those plant systems necessary to assure the accomplishment
of protection functions. There will be no impact on the overpower
limit, DNBR limits, FQ, F(delta)H, LOCA PCT, peak local
power density, or any other 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: Callaway County Public
Library, 710 Court Street, Fulton, Missouri 65251.
Attorney for licensee: Gerald Charnoff, Esq., Shaw, Pittman, Potts
& Trowbridge, 2300 N Street, N.W., Washington, D.C. 20037.
NRC Project Director: William H. Bateman.
Washington Public Power Supply System, Docket No. 50-397, Nuclear
Project No. 2, Benton County, Washington
Date of amendment request: December 4, 1997.
Description of amendment request: This amendment as reflected in
Section 2.1.1.2 of the Technical Specifications would continue the use
of the existing Siemens Power Corporation minimum critical power ratio
(MCPR) safety limits for Cycle 14 and would change the Asea Brown
Boveri (ABB) MCPR safety limit for single loop operation from 1.08 for
Cycle 13 to 1.09 for Cycle 14.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. Does the proposed change involve a significant increase in
the probability or
[[Page 2285]]
consequences of an accident previously evaluated?
The probability of an evaluated accident is derived from the
probabilities of the individual precursors to that accident. The
consequences of an evaluated accident are determined by the
operability of plant systems designed to mitigate those
consequences. Limits have been established consistent with NRC
approved methods to ensure that fuel performance during normal,
transient, and accident conditions is acceptable. The proposed
Technical Specifications amendment continues the use of
conservatively established ATRIUM-9X MCPR safety limits for WNP-2
such that the fuel is protected during normal operation as well as
during plant transients or anticipated operational occurrences.
The probability of an evaluated accident is not increased by the
continued use of the interim ATRIUM-9X MCPR safety limit of 1.13
(two loop operation) or 1.14 (single loop operation) or from
changing the ABB single loop MCPR safety limit of 1.08 (Cycle 13) to
1.09 (Cycle 14). The changes do not require any physical plant
modifications, physically affect any plant component, or entail
changes in plant operation. The increase in single loop MCPR safety
limit is attributed to a slightly more conservative assembly power
distribution used in the Cycle 14 calculations following ABB
standard methodology. While the Cycle 13 result is also
conservative, the increase in Cycle 14 is intended to accommodate
small cycle to cycle variability. Therefore, no individual
precursors of an accident are affected.
This Technical Specification amendment proposes to continue
using the interim MCPR safety limits for ATRIUM-9X fuel to protect
the fuel during normal operation as well as during plant transients
or anticipated operational occurrences. The method that is used to
determine the ATRIUM-9X additive constant uncertainty is
conservative, such that the resulting interim ATRIUM-9X MCPR safety
limits are high enough to ensure that less than 0.1% of the fuel
rods are expected to experience boiling transition if the limit is
not violated. Using NRC approved methodology, ABB has utilized these
interim values as the basis for the Cycle 14 safety limit for the
co-resident ATRIUM-9X. Operational limits have been established
based on the interim ATRIUM-9X MCPR safety limits to ensure that the
safety limits are not violated. This will ensure that the fuel
design safety criteria (more than 99.9% of the fuel rods avoid
transition boiling during normal operation as well as anticipated
operational occurrences) is met. In addition, since the operability
of plant systems designed to mitigate any consequences of accidents
have not changed, the consequences of an accident previously
evaluated are not expected to increase.
2. Does the proposed change create the possibility of a new or
different kind of accident from any accident previously evaluated?
Creation of the possibility of a new or different kind of
accident would require the creation of one or more new precursors of
that accident. New accident precursors may be created by
modifications of the plant configuration, including changes in
allowable modes of operation. This Technical Specification submittal
does not involve any modifications of the plant configuration or
allowable modes of operation. This Technical Specification change
continues the use of added conservatism in the ATRIUM-9X MCPR safety
limits which resulted from analytical changes and use of an expended
database. Also, ABB has calculated single loop MCPR safety limit
[which is] about 0.006 greater in Cycle 14 than was used in Cycle
13. The increase in single loop MCPR safety limit is attributed to a
slightly more conservative assembly power distribution used in the
Cycle 14 calculations following ABB standard methodology. While the
Cycle 13 result is also conservative, the increase in Cycle 14 is
intended to accommodate small cycle to cycle variability. Therefore,
no new precursors of an accident are created and no new or different
kinds of accidents are created.
3. Does the proposed change involve a significant reduction in a
margin of safety?
The continued use of interim MCPR safety limits provides a
margin of safety by ensuring that less than 0.1% of the rods are
expected to be in boiling transition if the MCPR limit is not
violated. These interim limits are based on calculations by SPC
using the revised ATRIUM-9X additive constant uncertainty. These
calculations are based on a larger pool of data than previous
calculations (527 data points versus 82 data points). Additionally,
the revised additive constant uncertainty has been conservatively
applied in the calculation of the interim ATRIUM-9X MCPR safety
limits resulting in more restrictive limits.
The calculated single loop MCPR safety limit results are about
0.006 greater for Cycle 14 than they were for Cycle 13. The increase
in single loop MCPR safety limits is attributed to a slightly more
conservative assembly power distribution used in the Cycle 14
calculations following ABB standard methodology. Because the fuel
design safety criteria of more than 99.9% of the fuel rods avoiding
transition boiling during normal operation as well as anticipated
operational occurrences is met, there is 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 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: Richland Public Library, 955
Northgate Street, Richland, Washington 99352.
Attorney for licensee: Perry D. Robinson, Esq., Winston & Strawn,
1400 L Street, N.W., Washington, D.C. 20005-3502.
NRC Project Director: William H. Bateman.
Previously Published Notices of Consideration of Issuance of Amendments
to Facility Operating Licenses, Proposed No Significant Hazards
Consideration Determination, and Opportunity for a Hearing
The following notices were previously published as separate
individual notices. The notice content was the same as above. They were
published as individual notices either because time did not allow the
Commission to wait for this biweekly notice or because the action
involved exigent circumstances. They are repeated here because the
biweekly notice lists all amendments issued or proposed to be issued
involving no significant hazards consideration.
For details, see the individual notice in the Federal Register on
the day and page cited. This notice does not extend the notice period
of the original notice.
Public Service Electric & Gas Company, Docket No. 50-272, Salem Nuclear
Generating Station, Unit No. 1, Salem County, New Jersey
Date of amendment request: December 11, 1997.
Brief description of amendment request: The proposed amendment
would provide a one-time change to the Technical Specifications to
allow purging of the containment during Modes 3 (Hot Standby) and 4
(Hot Shutdown) upon return to power from the current refueling outage
(1R13).
Date of publication of individual notice in Federal Register:
December 18, 1997 (62 FR 66397).
Expiration date of individual notice: January 20, 1998.
Local Public Document Room location: Salem Free Public Library, 112
West Broadway, Salem, NJ 08079.
Notice of Issuance of Amendments to Facility Operating Licenses
During the period since publication of the last biweekly notice,
the Commission has issued the following amendments. The Commission has
determined for each of these amendments that the application complies
with the standards and requirements of the Atomic Energy Act of 1954,
as amended (the Act), and the Commission's rules and regulations. The
Commission has made appropriate findings as required by the Act and the
Commission's rules and regulations in 10 CFR Chapter I, which are set
forth in the license amendment.
Notice of Consideration of Issuance of Amendment to Facility
Operating License, Proposed No Significant Hazards Consideration
Determination,
[[Page 2286]]
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.
Commonwealth Edison Company, Docket Nos. STN 50-454 and STN 50-455,
Byron Station, Unit Nos. 1 and 2, Ogle County, Illinois
Date of application for amendments: May 6, 1997, as supplemented on
July 30, 1997.
Brief description of amendments: The amendments will change
Technical Specification 3/4.7.5, ``Ultimate Heat Sink'' and the
associated Bases to support steam generator replacement and incorporate
recent Ultimate Heat Sink (UHS) design evaluations.
Date of issuance: December 12, 1997.
Effective date: Immediately, to be implemented within 30 days.
Amendment Nos.: 95 and 95.
Facility Operating License Nos. NPF-37 and NPF-66: The amendments
revised the Technical Specifications.
Date of initial notice in Federal Register: July 2, 1997 (62 FR
35847). The July 30, 1997, submittal provided additional clarifying
information that did not change the initial proposed no significant
hazards consideration determination.
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated December 12, 1997.
No significant hazards consideration comments received: No.
Local Public Document Room location: Byron Public Library District,
109 N. Franklin, P.O. Box 434, Byron, Illinois 61010.
Commonwealth Edison Company, Docket Nos. 50-237 and 50-249, Dresden
Nuclear Power Station, Units 2 and 3, Grundy County, Illinois
Docket Nos. 50-254 and 50-265, Quad Cities Nuclear Power Station, Units
1 and 2, Rock Island County, Illinois
Date of application for amendments: September 30, 1997.
Brief description of amendments: The amendments would add a new
Technical Specification (TS) Section 3/4.12.C and associated bases to
allow certain reactor coolant pressure tests to be performed in MODE 4
when the reactor pressure vessel requires testing at temperatures
greater than 212 degrees Fahrenheit. This temperature normally
corresponds with MODE 3.
Date of issuance: January 5, 1998.
Effective date: Immediately, to be implemented within 60 days.
Amendment Nos.: 164, 159, 179 and 177.
Facility Operating License Nos. DPR-19, DPR-25, DPR-29 and DPR-30.
The amendments revised the Technical Specifications.
Date of initial notice in Federal Register: November 19, 1997 (62
FR 61839). The Commission's related evaluation of the amendments is
contained in a Safety Evaluation dated January 5, 1998.
No significant hazards consideration comments received: No.
Local Public Document Room location: for Dresden, Morris Area
Public Library District, 604 Liberty Street, Morris, Illinois 60450;
for Quad Cities, Dixon Public Library, 221 Hennepin Avenue, Dixon,
Illinois 61021.
Entergy Operations, Inc., Docket No. 50-368, Arkansas Nuclear One, Unit
No. 2, Pope County, Arkansas
Date of application for amendment: April 24, 1997.
Brief description of amendment: The amendment revises the inservice
inspection requirements associated with steam generator tube sleeves.
Date of issuance: December 23, 1997.
Effective date: December 23, 1997.
Amendment No.: 187.
Facility Operating License No. NPF-6: Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: July 16, 1997 (62 FR
38134).
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated December 23, 1997.
No significant hazards consideration comments received: No.
Local Public Document Room location: Tomlinson Library, Arkansas
Tech University, Russellville, AR 72801.
Entergy Gulf States, Inc., 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: October 15, 1997.
Brief description of amendment: The amendment revises the license
to reflect the transfer of the 30-percent undivided ownership interest
in the River Bend Station, Unit No. 1 from Cajun Electric Power
Cooperative, Inc. to Entergy Gulf States, Inc. The transfer was
approved by Order dated November 28, 1997, which was published in the
Federal Register on December 5, 1997 (62 FR 64404).
Date of issuance: December 23, 1997.
Effective date: December 23, 1997.
Amendment No.: 101.
Facility Operating License No. NPF-47: The amendment revised the
operating license.
Date of initial notice in Federal Register: October 24, 1997 (62 FR
55432).
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated December 23, 1997.
No significant hazards consideration comments received: No.
Local Public Document Room location: Government Documents
Department, Louisiana State University, Baton Rouge, LA 70803.
Florida Power Corporation, et al., Docket No. 50-302, Crystal River
Unit No. 3 Nuclear Generating Plant, Citrus County, Florida
Date of application for amendment: July 18, 1997, as supplemented
September 12 and October 25, 1997.
Brief description of amendment: Establish a new Low-Temperature
Overpressure Protection Technical Specification (TS).
Date of issuance: December 22, 1997.
Effective date: December 22, 1997.
Amendment No.: 161.
Facility Operating License No. DPR-31: Amendment revised the TSs.
Date of initial notice in Federal Register: August 13, 1997 (62 FR
43369) The supplemental letters dated September 12 and October 25, 1997
did not change the initial no significant hazards consideration
determination or expand the scope of the initial notice.
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated December 22, 1997.
[[Page 2287]]
No significant hazards consideration comments received: No.
Local Public Document Room location: Coastal Region Library, 8619
W. Crystal River, Florida 34428.
Florida Power Corporation, et al., Docket No. 50-302, Crystal River
Unit No. 3 Nuclear Generating Plant, Citrus County, Florida
Date of application for amendment: July 29, 1997, as supplemented
October 29, 1997.
Brief description of amendment: The amendment revises the Post-
Accident Monitoring Instrumentation Technical Specification (TS).
Date of issuance: December 22, 1997.
Effective date: December 22, 1997.
Amendment No.: 162.
Facility Operating License No. DPR-31: Amendment revised the TSs.
Date of initial notice in Federal Register: August 13, 1997 (62 FR
43369).
The supplemental letter October 29, 1997 did not change the initial
no significant hazards consideration determination.
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated December 22, 1997.
No significant hazards consideration comments received: No.
Local Public Document Room location: Coastal Region Library, 8619
W. Crystal River, Florida 34428.
Florida Power and Light Company, Docket Nos. 50-250 and 50-251, Turkey
Point Plant Units 3 and 4, Dade County, Florida
Date of application for amendments: February 24, 1997, as
supplemented on April 24 and December 4, 1997.
Brief description of amendments: The admendments change technical
specification section 6.9.1.7, Core Operating Limits Report, to reflect
use of the Westinghouse Best Estimate Large Break Loss-of-Coolant
Accident (LOCA) methodology for large break LOCA analysis, including
supporting documents.
Date of issuance: December 20, 1997.
Effective date: December 20, 1997.
Amendment Nos.: 195 and 189.
Facility Operating Licenses Nos. DPR-31 and DPR-41: Amendments
revised the TS.
Date of initial notice in Federal Register: June 4, 1997 (62 FR
30631). By letter dated December 4, 1997, the licensee provided
additional information which did not affect the original no significant
hazards determination.
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated December 20, 1997.
No significant hazards consideration comments received: No.
Local Public Document Room location: Florida International
University, University Park, Miami, Florida 33199.
Northeast Nuclear Energy Company, et al., Docket No. 50-423, Millstone
Nuclear Power Station, Unit No. 3, New London County, Connecticut
Date of application for amendment: October 7, 1997, as supplemented
December 17, 1997.
Brief description of amendment: Technical Specifications 4.6.1.1,
3/4.6.1.2, and 3/4.6.1.3 require the testing of the containment to
verify leakage limits at a specified test pressure. The amendment (1)
modifies the list of valves that can be opened in Modes 1 through 4,
(2) removes a footnote on Type A testing, and (3) makes editorial
changes to the Technical Specifications and makes changes to the
associated Bases sections.
Date of issuance: December 18, 1997.
Effective date: As of the date of issuance, to be implemented
within 60 days.
Amendment No.: 154.
Facility Operating License No. NPF-49: Amendment revised the
Technical Specifications. Date of initial notice in Federal Register:
November 5, 1997 (62 FR 59917) The December 17, 1997, letter
provide clarifying information that did not change the October 7, 1997,
application and the initial proposed no significant hazards
consideration determination.
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated December 18, 1997.
No significant hazards consideration comments received: No.
Local Public Document Room location: Learning Resources Center,
Three Rivers Community-Technical College, 574 New London Turnpike,
Norwich, Connecticut 06360, and the Waterford Library, ATTN: Vince
Juliano, 49 Rope Ferry Road, Waterford, Connecticut.
Public Service Electric & Gas Company, Docket No. 50-272, Salem Nuclear
Generating Station, Unit No. 1, Salem County, New Jersey
Date of application for amendment: October 6, 1997.
Brief description of amendment: The requested changes would
increase the allowable band for control and shutdown rod demanded
position versus indicated position from plus or minus 12 steps to plus
or minus 18 steps when the power level is not greater than 85% rated
thermal power. The changes have already been approved for Salem Unit 2
in Amendment No. 183, issued September 10, 1997, as an exigent
amendment.
Date of issuance: December 22, 1997.
Effective date: December 22, 1997.
Amendment No.: 202.
Facility Operating License No. DPR-70: This amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: November 19, 1997 (62
FR 61845).
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated December 22, 1997.
No significant hazards consideration comments received: No.
Local Public Document Room location: Salem Free Public Library, 112
West Broadway, Salem, NJ 08079.
Tennessee Valley Authority, Docket Nos. 50-260 and 50-296 Browns Ferry
Nuclear Plant, Units 2, and 3, Limestone County, Alabama
Date of application for amendment: June 19, 1997, with additional
information provided on August 15, 1997 (TS 391T).
Brief Description of amendment: The amendments revise the Technical
Specifications (TS) to temporarily extend the allowed outage time for
the emergency diesel generators from 7 to 14 days to permit completion
of preventive maintenance.
Date of issuance: December 22, 1997.
Effective Date: December 22, 1997.
Amendment Nos.: 250 and 209.
Facility Operating License Nos. DPR-52 and DPR-68: Amendment
revised the TS.
Date of initial notice in Federal Register: July 30, 1997 (62 FR
40858).
The additional information provided on August 15, 1997 does not
affect the staff's proposed finding of no significant hazards
consideration.
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated December 22, 1997.
No significant hazards consideration comments received: None.
Local Public Document Room Location: Athens Public library, 405 E.
South Street, Athens, Alabama 35611.
TU Electric Company, Docket Nos. 50-445 and 50-446, Comanche Peak Steam
Electric Station, Unit Nos. 1 and 2, Somervell County, Texas
Date of amendment request: July 9, 1996 (TXXX-96393), as
supplemented
[[Page 2288]]
on December 12, 1997 (TXXX-97268). (The supplement contains clarifying
information and does not change the staff's original proposed no
significant hazards determination.)
Brief description of amendments: The amendments change Technical
Specification 3.3-3, ``Engineered Safety Features Actuation System
Instrumentation Trip Setpoints.'' The proposed changes would increase
the minimum allowable value of the Unit 1 Steam Line Pressure--Low
Safety Injection and Steam Line Isolation functions. These changes are
needed to ensure that the instrumentation error is properly accounted
for in the TSs.
Date of issuance: December 30, 1997.
Effective date: December 30, 1997, to be implemented within 30
days.
Amendment Nos.: Unit 1--Amendment No. 56; Unit 2--Amendment No. 42.
Facility Operating License Nos. NPF-87 and NPF-89: The amendments
revised the Technical Specifications.
Date of initial notice in Federal Register: February 12, 1997 (62
FR 6579) The Commission's related evaluation of the amendments is
contained in a Safety Evaluation dated December 30, 1997.
No significant hazards consideration comments received: No.
Local Public Document Room location: University of Texas at
Arlington Library, Government Publications/Maps, 702 College, P.O. Box
19497, Arlington, TX 76019.
Dated at Rockville, Maryland, this 7th day of January 1998.
For the Nuclear Regulatory Commission.
Elinor G. Adensam,
Acting Director, Division of Reactor Projects--III/IV, Office of
Nuclear Reactor Regulation.
[FR Doc. 98-753 Filed 1-13-98; 8:45 am]
BILLING CODE 7590-13-P
