[Federal Register Volume 63, Number 37 (Wednesday, February 25, 1998)]
[Notices]
[Pages 9589-9618]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: 98-4620]
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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 February 2, 1998, through February 12, 1998.
The last biweekly notice was published on February 11, 1998 (63 FR
6968).
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
[[Page 9590]]
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 Administration Services, Office of
Administration, U.S. Nuclear Regulatory Commission, Washington, DC
20555-0001, and should cite the publication date and page number of
this Federal Register notice. Written comments may also be delivered to
Room 6D22, Two White Flint North, 11545 Rockville Pike, Rockville,
Maryland from 7:30 a.m. to 4:15 p.m. Federal workdays. Copies of
written comments received may be examined at the NRC Public Document
Room, the Gelman Building, 2120 L Street, NW., Washington, DC. The
filing of requests for a hearing and petitions for leave to intervene
is discussed below.
By March 27, 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.
[[Page 9591]]
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 amendment request: October 4, 1996, as supplemented by
letters dated June 6, September 19, November 7, and December 16, 1997.
Description of amendment request: The proposed amendment for each
unit identified above would change the distance criterion in Action b
to Limiting Condition for Operation (LCO) 3/4.1.3, ``Movable Control
Assemblies,'' by which more than one full-length or part-length control
element assembly (CEA) is misaligned from any other CEA in its group.
Action b states, in part, that if the misalignment is greater than the
specified distance criterion, the reactor core is to be placed in at
least hot standby within 6 hours. The proposed amendment would reduce
the distance criterion from 19 inches to 9.9 inches, and replace hot
standby in 6 hours by ``open the reactor trip breakers.''
This proposed amendment is included as a ``more restrictive''
change in the conversion of the current Technical Specifications (CTS)
to the Improved Technical Specifications, which was noticed in the
Federal Register (62 FR 18153) on April 14, 1997. The proposed
amendment would be included in Action F to LCO 3.1.5, ``Movable Control
Assemblies,'' of the Improved Technical Specifications. This proposed
amendment is a change to the current Technical Specifications and is in
addition to the six proposed changes to the CTS or proposed deviations
to the Improved Standard Technical Specifications (NUREG-1432) which
were identified in the notice of April 14, 1997.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. The proposed change does not involve a significant increase in
the probability or consequences of an accident previously evaluated.
The proposed changes provide more stringent requirements than
previously existed in the CTS. The more stringent requirements will not
result in operation that will increase the probability of initiating an
analyzed event. If anything, the new requirements may decrease the
probability or consequences of an analyzed event by incorporating the
more restrictive changes discussed in the specific Discussion of
Changes [for specification 3.1.5]. These changes will not alter
assumptions relative to mitigation of an accident or transient event.
The more restrictive requirements will not alter the operation and will
continue to ensure process variables, structures, systems, or
components are maintained consistent with safety analyses and licensing
basis [for the plant]. These changes have been reviewed to ensure that
no previously evaluated accident has been adversely affected.
Therefore, these changes will not involve a significant increase in the
probability or consequences of an accident evaluated.
2. The proposed change does not create the possibility of a new or
different kind of accident from any accident previously evaluated.
Making existing requirements more restrictive and adding more
restrictive requirements to the CTS will not alter the plant
configuration (no new or different type of equipment will be installed)
or change the methods governing normal plant operation. These changes
do impose different requirements. However, they are consistent with the
assumptions made in the safety analyses, licensing basis, and NUREG-
1432 [for the plant]. Therefore, these changes will not create the
possibility of a new or different kind of accident from any accident
previously evaluated.
3. The proposed change does not involve a significant reduction in
a margin of safety.
The proposed changes provide more stringent requirements than
previously existed in the CTS. An evaluation of these changes concluded
that adding these more restrictive requirements either increases or has
no impact on the margin of safety. The changes provide additional
restrictions which may enhance plant safety. These changes maintain
requirements of the safety analysis, licensing basis, and NUREG-1432
[for the plant]. As such, no question of safety is involved. Therefore,
these changes 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: 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.
Boston Edison Company, Docket No. 50-293, Pilgrim Nuclear Power
Station, Plymouth County, Massachusetts
Date of amendment request: September 19, 1997.
Description of amendment request: The proposed amendment would
relocate the Radioactive Effluent Technical Specifications (RETS) and
the Radiological Environmental Monitoring Program to the Offsite Dose
Calculation Manual (ODCM), in accordance with the recommendations of
Generic Letter 89-01 and NUREG-1433. In addition, changes to other
sections of the TSs are being proposed to align the current TSs with
NUREG-1433.
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 change involve a significant increase in the
probability or consequences of an accident previously evaluated?
Operation of PNPS in accordance with the proposed change will not
involve a significant increase in the probability or consequences of an
accident previously evaluated because of the following:
Definitions
Definitions perform a supporting function for other sections of the
TS. The proposed change to incorporate the definition for the Offsite
Dose Calculations Manual (ODCM) into Section 5.0, ``Programs and
Manuals'', subsection 5.5.1 of the proposed TS will carry forward the
requirements contained in the DEFINITION, with minor editorial
rewording to be consistent with NUREG 1433, and result in no technical
changes. Since the requirements will remain, the impact on initiators
of analyzed events or the assumptions assumed in the mitigation of
accidents or transient events will not change. Editorial rewording
(either adding or deleting) and reformatting is proposed to provide
clarity and does not change any technical requirements.
The definitions being proposed for relocation do not impact reactor
operation, identify a parameter which is an initial condition
assumption for a DBA or transient, identify a significant abnormal
degradation of the reactor coolant pressure boundary, and do not
[[Page 9592]]
provide any mitigation of a design basis event.
RAD Effluents
All editorial rewording (either adding or deleting) and renumbering
is made to restructure the section accounting for the requirements
relocated in accordance with Generic Letter 89-01. During the editorial
rewording and renumbering of the Improved Technical Specifications, no
technical changes (either actual or interpretational) to the TS were
made unless they were identified and justified.
Adding a note to clearly indicate that the first sample for noble
gas activity is not required for 31 days after SJAE is placed in
operation has always been considered the intent of this surveillance
requirement. This allowance is consistent with the frequency for the
required surveillance and allows time for concentrations of longer
lived isotopes to reach equilibrium. In addition, other instrumentation
continuously monitors the offgas to alert operators of significant
increases in radioactivity.
The proposed change provides more stringent requirements than
previously existed in the Technical Specifications. The more stringent
requirements will not result in operation that will increase the
probability of initiating an analyzed event. If anything, the new
requirements may decrease the probability or consequences of an
analyzed event by incorporating the more restrictive changes discussed
above. The change will not alter assumptions relative to mitigation of
an accident or transient event. The more restrictive requirements will
not alter the operation of process variables, structures, systems, or
components as described in the safety analyses.
These proposed changes relocate requirements from the Technical
Specifications to the T. S. BASES, FSAR, or ODCM. The licensee
controlled document containing the relocated requirements will be
maintained using the provisions of 10 CFR 50.59 or a change control
process in the Administrative Controls Section of the Technical
Specifications. Since any changes to these licensee controlled
documents will be evaluated per an NRC approved change control process,
no increase in the probability or consequences of an accident
previously evaluated will be allowed.
Basing the potential fission product release rate on gross gamma
activity rate is more representative of the whole body dose that would
be received by an individual at the site boundary should a release
occur. Therefore, reasonable assurance that the potential whole body
accident dose to an individual at the exclusion area boundary will not
exceed a small fraction of the limits specified in 10 CFR Part 100 is
maintained.
Allowing the sample to be taken from either pretreatment monitor
station will have no effect on the objective of assuring that the
potential whole body accident dose to an individual at the exclusion
area boundary will not exceed a small fraction of the limits specified
in 10 CFR Part 100, because both monitor stations are prior to
treatment, adsorption, or delay of the noble gases.
RAD Material Source
The requirements for miscellaneous radioactive materials do not
impact reactor operation, identify a parameter which is an initial
condition assumption for a DBA or transient, identify a significant
abnormal degradation of the reactor coolant pressure boundary, and do
not provide any mitigation of a design basis event.
Major Design Features
The reformatting, renumbering, and rewording along with the other
changes listed involve no technical changes to existing Technical
Specifications. The proposed changes are administrative in nature and
do not impact initiators or assumptions of analyzed accidents or
transient events.
The proposed change provides more stringent requirements than
previously existed in the Technical Specifications. The more stringent
requirements will not result in operation that will increase the
probability of initiating an analyzed event. If anything, the new
requirements may decrease the probability or consequences of an
analyzed event by incorporating the more restrictive changes discussed
above. The change will not alter assumptions relative to mitigation of
an accident or transient event. The more restrictive requirements will
not alter the operation of process variables, structures, systems, or
components as described in the safety analyses.
These proposed changes relocate requirements from the Technical
Specifications to the FSAR. Since any changes to the FSAR must be
evaluated per 10 CFR 50.59, no increase (significant or insignificant)
in the probability or consequences of an accident previously evaluated
will be allowed.
Administrative Controls
The reformatting, renumbering, and rewording along with the other
changes listed involves no technical changes to existing Technical
Specifications. The change to the existing Technical Specifications was
done in order to be consistent with the NUREG-1433. During development
of NUREG-1433, certain wording preferences or English language
conventions were adopted. The proposed change to this section is
administrative in nature and does not impact initiators of analyzed
events. It also does not impact the assumed mitigation of accidents or
transient events.
The proposed change provides more stringent requirements than
previously existed in the Technical Specifications. These more
stringent requirements are administrative in nature (e.g., specifying
additional responsibilities for plant personnel, ensuring overtime
control, incorporating program and manual requirements already in
place, and adding details to reports). These additional requirements
will not alter the plant configuration (no new or different type of
equipment will be installed) or changes in methods governing normal
plant operation, not alter assumptions relative to the mitigation of an
accident or transient event, or alter the operation of process
variables, structures, systems, or components as described in the
safety analyses.
This proposed change relocates requirements from the Technical
Specifications to licensee controlled documents. The licensee
controlled documents containing the relocated requirements are required
to meet the applicable regulation and any change process invoked by the
regulation. Since any changes to the licensee controlled document must
continue to meet the regulation, no increase (significant or
insignificant) in the probability or consequences of an accident
previously evaluated will be allowed.
This change proposes to provide flexibility in meeting the minimum
shift staffing for up to two hours in order to provide for unexpected
absence. The proposed change does not affect the probability of an
accident. The actions of an individual are not assumed to be an
initiator of any analyzed event. Also, the change does not negate the
requirement to have licensed individuals in the control room. This
proposed change does not impact the assumptions of any design basis
accident. This change will not alter assumptions relative to the
mitigation of an accident or transient event.
This change proposes to relax the requirement to have an individual
qualified in radiation protection procedures to be onsite when fuel is
in the reactor. The proposed change will allow the position to be
vacant for up
[[Page 9593]]
to two hours in order to provide for unexpected absence.
The proposed change does not affect the probability of an accident.
The actions of an individual qualified in radiation protection
procedures are not assumed to be an initiator of any analyzed event.
Also, the consequences of an accident are not affected by the presence
of an individual qualified in radiation protection. This proposed
change does not impact the assumptions of any design basis accident.
This change will not alter assumptions relative to the mitigation of an
accident or transient event. This change will not have any impact on
the plant safety because the presence of a person qualified in
radiation protection is not required for the mitigation of any
accident.
This change proposes to relax the requirement for submitting the
Radioactive Effluent Release Report and to relocate the report details
outside the TS. The current TS require the report to be submitted semi-
annually. This proposed change will allow the report to be submitted
annually as required by 10 CFR 50.36a. The proposed change does not
affect the probability of an accident. Neither the submittal
requirements nor the contents of the Radioactive Effluent Release
Report is assumed to be an initiator of any analyzed event. Also, the
consequences of an accident are not affected by submittal requirements
nor the contents of the Radioactive Effluent Release Report. This
proposed change does not impact the assumptions of any design basis
accident. This change will not alter assumptions relative to the
mitigation of an accident or transient event. This change has no impact
on the safe operation of the plant. The report will still be required
to be submitted and does not affect any plant equipment or requirements
for maintaining plant equipment. The submittal of this report is not
required for the mitigation of any accident.
The proposed alternatives for control of access to high radiation
areas are consistent with the intent of 10 CFR 20.1601(a) and (b). The
proposed changes do not affect the probability of an accident. The
controls used for access to high radiation areas are not assumed in the
initiation of any analyzed event. Also, the consequences of an accident
are not affected by these changes. These changes are both consistent
with good radiological safety practice and will provide an adequate
level of radiation protection. These proposed changes do not impact the
assumptions of any design basis accident. These changes will not alter
assumptions relative to the mitigation of an accident or transient
event. These changes have no impact on safe operation of the plant.
Radiological Environmental Monitoring
The proposed changes only alter the format and location of
procedural details and administrative controls of the radioactive
effluents, radiological environmental monitoring, and solid radioactive
waste programs. The changes are administrative in nature and do not
involve any change to the configuration or operation of plant
equipment. The Radiological Effluent Technical Specifications (RETS)
procedural details are being moved to the Offsite Dose Calculation
manual (ODCM). In addition, new administrative controls have been added
to the Technical Specifications which will provide an equivalent level
of assurance that activities involving radioactive effluents, solid
radioactive waste, and radiological environmental monitoring are
conducted in full compliance with regulatory requirements. Since any
changes to these requirements will require NRC approval, no increase in
the probability or consequences of an accident previously evaluated
will be allowed.
2. Does the change create the possibility of a new or different
kind of accident from any accident previously evaluated?
Operation of PNPS in accordance with the proposed change will not
create the possibility of a new or different kind of accident from any
accident previously evaluated because of the following:
Definitions
These proposed changes do not involve a physical alteration of the
plant (no new or different type of equipment will be installed) or
changes in methods governing normal plant operation. The proposed
change will not impose any new or different requirements or eliminate
any existing requirements.
Relocating these definitions will not alter the plant configuration
(no new or different type of equipment will be installed) or change
methods governing normal plant operation. Relocating requirements will
not impose different requirements and adequate control of information
will be maintained. Relocating these definitions will not alter
assumptions made in the safety analysis and licensing basis.
RAD Effluents
The proposed change does not involve a physical alteration of the
plant (no new or different type of equipment will be installed) or
changes in methods governing normal plant operation. The proposed
change will not impose any new or different requirements or eliminate
any existing requirements.
Making existing requirements more restrictive and adding more
restrictive requirements to the Technical Specifications will not alter
the plant configuration (no new or different type of equipment will be
installed) or change methods governing normal plant operation. These
changes are consistent with current design bases, licensing bases or
assumptions made in the safety analysis.
These changes do not alter the plant configuration (no new or
different type of equipment will be installed) or methods governing
normal plant operation. These changes will not impose different
requirements and adequate control of information will be maintained.
These changes do not alter assumptions made in the safety analysis and
licensing basis.
The proposed change will not involve a physical alteration of the
plant (no new or different type of equipment will be installed) or
changes in methods governing normal plant operation. Operation of the
plant will not be altered by this change. This change will not place
the plant in any new condition or introduce any mode of operation not
previously analyzed.
The proposed change will not involve a physical alteration of the
plant (no new or different type of equipment will be installed) or
changes in methods governing normal plant operation. Operation of the
plant will not be altered by this change. This change will not place
the plant in any new condition or introduce any mode of operation not
previously analyzed.
RAD Material Source
Relocating these requirements will not alter the plant
configuration (no new or different type of equipment will be installed)
or change methods governing normal plant operation. Relocating
requirements will not impose different requirements and adequate
control of information will be maintained. Relocating requirements does
not alter assumptions made in the safety analysis and licensing basis.
Major Design Features
The proposed change does not involve a physical alteration of the
plant (no new or different type of equipment will be installed) or
changes in methods governing normal plant operation. The proposed
change will not impose any new or different requirements or eliminate
any existing requirements.
[[Page 9594]]
Making existing requirements more restrictive and adding more
restrictive requirements to the Technical Specifications will not alter
the plant configuration (no new or different type of equipment will be
installed) or changes in methods governing normal plant operation. The
change does impose different requirements. However, the change is
consistent with assumptions made in the safety analyses.
These changes relocate requirements to the FSAR. These changes do
not alter the plant configuration (no new or different type of
equipment will be installed) or the methods governing normal plant
operation. These changes do not impose different requirements and
adequate control of information will be maintained. This change will
not alter assumptions made in the safety analysis and licensing basis.
Administrative Controls
The proposed change does not involve a physical alteration of the
plant (no new or different type of equipment will be installed) or
changes in methods governing normal plant operation. The proposed
change will not impose any new or different requirements or eliminate
any existing requirements.
Making existing requirements more restrictive and adding new
requirements to the Technical Specifications will not alter the plant
configuration (no new or different type of equipment will be installed)
or changes in the methods governing normal plant operation.
This change relocates requirements to a licensee controlled
document. This change will not alter the plant configuration (no new or
different type of equipment will be installed) or changes in methods
governing normal plant operation. This change will not impose different
requirements and adequate control of information will be maintained.
This change will not alter assumptions made in the safety analysis and
licensing basis.
This change proposes to provide flexibility in meeting the minimum
shift staffing for up to two hours in order to provide for an
unexpected absence. The proposed change will not create the possibility
of an accident. This change will not physically alter the plant (no new
or different type of equipment will be installed).
This change proposes to relax the requirement to have an individual
qualified in radiation protection procedures to be onsite when fuel is
in the reactor. The proposed change will allow the position to be
vacant for up to two hours in order to provide for unexpected absence.
The proposed change will not create the possibility of an accident.
This change will not physically alter the plant (no new or different
type of equipment will be installed) or the methods of operation.
This change will not physically alter the plant (no new or
different type of equipment will be installed). The changes in methods
governing normal plant operation are consistent with the current safety
analysis assumptions.
The proposed change will not create the possibility of an accident.
This change will not physically alter the plant (no new or different
type of equipment will be installed). The changes in methods governing
normal plant operation are consistent with the current safety analysis
assumptions.
Radiological Environmental Monitoring
The procedural requirements of the RETS will be maintained in the
ODCM. Operation of the plant will not be altered by the changes
proposed to the administration of the RETS. This change will not place
the plant in any new condition or introduce any mode of operation not
previously analyzed.
3. Does this change involve a significant reduction in a margin of
safety?
Operation of PNPS in accordance with the proposed change will not
involve a significant reduction in a margin of safety because of the
following:
Definitions
Definitions perform a supporting function for other sections of the
TS and the proposed editing, omission or relocation of definitions
associated with this change will not, by itself, reduce existing
restrictions on plant operations.
The definitions to be transposed from the Technical Specifications
to the ODCM are the same as the existing Technical Specifications.
Future changes to the ODCM will be controlled in accordance with
proposed technical specification 5.5.1 ``Offsite Dose Calculation
Manual (ODCM)''.
RAD Effluents
The change is administrative in nature and does not involve any
technical changes. The proposed change will not reduce a margin of
safety because it has no impact on any safety analysis assumptions.
Also, because the change is administrative in nature, no question of
safety is involved.
Adding these new requirements and making existing ones more
restrictive does not affect any safety analysis assumptions. As such,
no question of safety is involved.
The requirements to be relocated from the Technical Specifications
to the FSAR T.S. BASES, or ODCM are the same as the existing Technical
Specifications and any future changes to this licensee controlled
document will be evaluated per an NRC approved change control process.
Specifying a release rate based only on gamma activity is more
representative of the whole body dose that would be received by an
individual at the site boundary should a release occur. The actual
margin of safety could be increased because potential errors in
converting beta activity to whole body exposures are eliminated
The sample used to determine the gaseous activity rate will
continue to be taken prior to treatment, adsorption, or delay of the
noble gases.
RAD Material Source
This change relocates requirements from the Technical
Specifications to a licensee controlled document. This change will not
reduce a margin of safety since it has no impact on any safety analysis
assumptions. In addition, the requirements to be transposed from the
Technical Specifications to the licensee controlled documents are the
same as the existing Technical Specifications. Since any future changes
to these licensee controlled documents must be evaluated per the cited
regulations or requirements of 10 CFR 50.59, no reduction (significant
or insignificant) in a margin of safety will be allowed.
Major Design Features
The changes are administrative in nature and do not involve any
technical changes. The proposed changes do not impact initiators or
assumptions of analyzed accidents or transient events.
These new or more restrictive requirements are consistent with the
current design and licensing bases; therefore, a margin of safety is
not affected.
These changes relocate requirements from the Technical
Specifications to the FSAR. The requirements to be are the same as the
existing Technical Specifications. Since any future changes to the FSAR
must be evaluated per the requirements of 10 CFR 50.59, no reduction
(significant or insignificant) in a margin of safety will be allowed.
Administrative Controls
The change is administrative in nature and will not involve any
technical changes. The proposed change will not reduce a margin of
safety because it has no impact on any safety analysis assumptions.
[[Page 9595]]
Adding these new requirements and making existing ones more
restrictive does not introduce any new tests or changes in methods
governing normal plant operation. Therefore, the changes do not impact
any safety analysis assumptions.
This change relocates requirements from the Technical
Specifications to a licensee controlled document. The licensee
controlled documents containing the relocated requirements are required
to meet the applicable regulation and any change process invoked by the
regulation. Since any changes to a licensee controlled document must
continue to meet the regulation, no increase (significant or
insignificant) in the probability or consequences of an accident
previously evaluated will be allowed.
This change proposes to provide flexibility in meeting the minimum
shift staffing for up to two hours in order to provide for unexpected
absence. This proposed change has no effect on the assumptions of a
design basis accident. The safety analysis assumptions will still be
maintained; thus, no question of safety exists.
This change proposes to relax the requirement to have an individual
qualified in radiation protection procedures to be onsite when fuel is
in the reactor. The proposed change will allow the position to be
vacant for up to two hours in order to provide for unexpected absence.
The margin of safety is not affected by the presence or absence on site
of an individual qualified in radiation protection procedures. This
proposed change has no effect on the assumptions of the design basis
accident. This change will not have any impact on the plant safety
because the presence of a person qualified in radiation protection is
not required for the mitigation of any accident. The safety analysis
assumptions will still be maintained; thus, no question of safety
exists.
This proposed change has no effect on the assumptions of the design
basis accident. This change has no impact on the safe operation of the
plant. The report will still be required to be submitted and does not
affect any plant equipment or requirements for maintaining plant
equipment. The safety analysis assumptions will still be maintained;
thus, no question of safety exists.
The proposed alternatives for control of access to high radiation
areas are consistent with the intent of 10 CFR 20.1601(a) and (b). The
margin of safety is not reduced due to these proposed changes. These
changes are both consistent with good radiological safety practices and
have been found to provide an adequate level of radiation protection.
In addition, these changes provide the benefit of ensuring radiation
dose to all workers is minimized by providing the flexibility to select
the best means of providing a barrier and access control to a high
radiation area given the plant location and radiological conditions.
These proposed changes have no impact on the safe operation of the
plant. The safety analysis assumptions will still be maintained; thus,
no question of safety exists.
Radiological Environmental Monitoring
The proposed changes relocate the procedural details and Bases for
RETS from the Technical Specifications to the ODCM. The RETS procedural
details and Bases will be maintained by these programs. In addition,
new administrative controls have been added to the Technical
Specifications which assure the proper control and maintenance of these
documents and provide an equivalent level of assurance that activities
involving radioactive effluents, solid radioactive waste, and
radiological environmental monitoring are conducted in full compliance
with regulatory requirements.
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: Plymouth Public Library, 11
North Street, Plymouth, Massachusetts 0236.
Attorney for licensee: W.S. Stowe, Esquire, Boston Edison Company,
800 Boylston Street, 36th Floor, Boston, Massachusetts 02199.
NRC Project Director: Cecil O. Thomas.
Commonwealth Edison Company, Docket Nos. STN 50-454 and STN 50-455,
Byron Station, Unit Nos. 1 and 2, Ogle County, Illinois, Docket Nos.
STN 50-456 and STN 50-457, Braidwood Station, Unit Nos. 1 and 2, Will
County, Illinois
Date of amendment request: November 7, 1997.
Description of amendment request: The proposed amendment would
revise the technical specifications and associated bases to allow the
licensee to perform 10 CFR Part 50, Appendix J, Type A testing on
Byron, Unit 2, and Braidwood, Unit 2, containments at least once per 10
years based on a single successful Type A test, rather than two
successful Type A tests.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. The proposed change does not involve a significant increase in
the probability or consequences of an accident previously evaluated.
Performance of Type A tests at a different interval does not
involve a change to any structures, systems, or components, does not
affect reactor operations, is not an accident initiator, and does not
change any existing safety analysis previously evaluated in the UFSAR
[Updated Final Safety Analysis Report]. Therefore, there is no
significant increase in the probability of an accident previously
evaluated.
Several tables of UFSAR Chapter 15, ``Accident Analyses,'' provide
containment leak rate values used in assessing the consequences of
accidents discussed in this chapter. Although decreasing the test
frequency can increase the probability that an increase in containment
leakage could go undetected for an extended period of time, the risk
resulting from this proposed change is inconsequential as documented in
NUREG-1493, ``Performance-Based Containment Leakage Test Program''.
This document indicated that given the insensitivity of reactor risk to
containment leakage rate and a small fraction of leakage paths are
detected solely by Type A testing, increasing the interval between
integrated leak rate tests is possible with minimal impact on public
risk. Further, industry experience presented in this document indicated
that Type A testing has had insignificant impact on uncertainties
involved with containment leak rates.
Based on risk information presented in NUREG-1493, the proposed
change does not increase 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.
The proposed change does not alter the plant design, systems,
components, or reactor operations, only the frequency of test
performance. New conditions or parameters that contribute to the
initiation of accidents would not be created as a result of this
proposed change. The change does not involve new equipment and existing
equipment does not have to be operated in a
[[Page 9596]]
different manner, therefore there are no new failure modes to consider.
Changing test intervals as shown in NUREG-1493 has no impact on,
nor contributes to the possibility of a new or different kind of
accident as evaluated in the UFSAR. Therefore, the proposed change does
not create the possibility of a new or different kind of accident from
any accident previously evaluated.
3. The proposed change does not involve a significant reduction in
a margin of safety.
With the exception of the test frequency, the actual tests will not
change. Quantitative risk studies documented in NUREG-1493 regarding
extended testing intervals demonstrated that there was minimal impact
on the public health and safety. Reducing the frequency, as stated in
the NUREG resulted in an ``imperceptible'' increase in risk to public
safety. Further, a table in this NUREG regarding risk impacts due to a
reduction in testing frequency suggested that there was also minimal
difference in risk to the public safety when the test frequency was
relaxed.
The proposed change will not reduce the availability of systems and
components associated with containment integrity that would be required
to mitigate accident conditions nor are any containment leakage rates,
parameters or accident assumptions affected by the proposed change.
The proposed change does not involve a significant reduction in a
margin of safety, based on the above information.
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 Byron, the Byron Public
Library District, 109 N. Franklin, P.O. Box 434, Byron, Illinois 61010;
for Braidwood, the Wilmington Public Library, 201 S. Kankakee Street,
Wilmington, Illinois 60481.
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. STN 50-454 and STN 50-455,
Byron Station, Unit Nos. 1 and 2, Ogle County, Illinois, Docket Nos.
STN 50-456 and STN 50-457, Braidwood Station, Unit Nos. 1 and 2, Will
County, Illinois
Date of amendment request: December 30, 1997.
Description of amendment request: The proposed amendment would
revise the Technical Specification (TS) 3.7.1.3, ``Condensate Storage
Tank,'' (CST) and its associated Bases for Byron and Braidwood to raise
the minimum allowable CST level to ensure that a sufficient volume of
water is available to meet the design basis requirements for the
auxiliary feedwater (AFW) system supply. The proposed amendment would
also revise the AFW system transfer to essential service water (SX)
trip setpoint and allowable value in Table 3.3-4 to ensure that the
design basis requirements for the AFW system are accurately reflected
in the TS.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. The change does not involve a significant increase in the
probability or consequences of an accident previously evaluated.
The amount of water in the CST [Condensate Storage Tank] at the
beginning of an accident and the setpoint for AF [auxiliary feedwater]
pump suction pressure-low trip have no impact on the probability of
occurrence of any accident analyzed in the UFSAR [Updated Final Safety
Analysis Report]. This is due to the availability of the safety-related
SX [essential service water] water supply as a backup system.
Therefore, the probability of an accident previously evaluated is
unchanged.
The loss of the Safety Category II CST under accident conditions
has already been evaluated in the UFSAR. The SX system is the emergency
source of water supply to the AF system under accident conditions. The
design basis analysis for the essential service water (SX) system and
the Limiting Condition for Operation requirements for the ultimate heat
sink ensure that a sufficient supply of water is available to plant
operators to mitigate the consequences of all analyzed accidents. None
of the proposed changes to the CST minimum level or the setpoints
documented in TS Table 3.3-4, functional unit 6.g. has any negative
impact on the assumptions or results of these analyzed accidents. To
the contrary, the proposed changes will ensure that the CST remains
available as the primary supply of water to the AF system and that
automatic suction transfer will occur for circumstances where the
Safety Category II CST becomes unavailable (e.g., seismic event or
tornado).
The level in the CST and the associated instrumentation and
setpoints help ensure that sufficient water is available to plant
operators to mitigate the consequences of accidents that are analyzed
in the UFSAR. The SX system is the emergency source of water credited
in the UFSAR. However, the proposed Technical Specification Bases
require that sufficient water be maintained in the CST to respond to
postulated events where the CST remains available (e.g., non-seismic
related events and events with no tornado assumed). The proposed CST
levels ensure that this requirement is met. The water level requirement
for the CST provides additional assurance that plant operators remain
capable of responding to postulated events as described in the UFSAR.
Therefore, the proposed changes do not increase the consequences of an
accident previously evaluated.
Therefore this proposed amendment does not involve a significant
increase in the probability or consequences of an accident previously
evaluated.
2. The change does not create the possibility of a new or different
kind of accident from any accident previously evaluated.
The proposed changes are being implemented to account for
instrument accuracy and AF system suction requirements that affect the
volume of useable water in the CST. The amendment request incorporates
the full design requirements of the AF System and components to ensure
that sufficient water is maintained in the CST. The changes reduce the
probability of an undesirable introduction of lower quality essential
service (SX) system water into the steam generators unless required due
to the unavailability of the CST during emergency conditions (e.g.,
seismic event or tornado). Although the SX system is the safety-related
water supply to AF, the water contains high levels of impurities and
sediment that could eventually degrade the steam generators. The CST
contains demineralized water. Therefore, the long term reliability and
availability of the steam generators is enhanced by precluding
introduction of SX water into the steam generators unless required
under emergency conditions. The proposed CST levels account for the
incremental increase in CST water
[[Page 9597]]
volume required due to the larger metal mass and primary volume of the
replacement steam generators for Byron Unit 1 and Braidwood Unit 1.
Finally, the trip setpoint and allowable values in Table 3.3-4 of the
TS are being updated to reflect the current design basis of the AF
system. The required CST level changes when plant modifications are
completed. Each configuration has been evaluated and the associated CST
level maintains a sufficient water volume to perform its design
function.
The modification to the suction pressure circuitry involves the
addition of an electronic ``lead-lag'' circuit card for the motor-
driven AF pump, which experiences the most severe startup suction
pressure transients. This circuit card will be set up for ``lag'' only
operation and will filter the suction pressure signal during transients
associated with pump startup or other sudden changes in flow or
pressure. This will prevent an inadvertent trip during transient
conditions when the CST is available. In situations where the CST is
unavailable, the suction pressure will decrease with no recovery until
switchover. Under this condition, the output of the lead-lag card will
continue to decrease as well until the switchover setpoint is reached.
The time constant of the lead-lag card was selected such that the
resulting time delays in actuating SX switchover and pump trip are
consistent with pump protection requirements.
Therefore, the proposed changes do not create the possibility of a
new or different kind of accident from any previously evaluated. This
conclusion is also valid when considering the planned modifications to
the AF suction pressure transient circuitry.
3. The change does not involve a significant reduction in a margin
of safety.
The proposed change is made in the conservative direction with
respect to the current TS requirements for minimum CST level and AF
pump CST to SX switchover setpoints. Increasing the volume of water
contained in the CST level provides redundancy to the safety-related
source of water to the AF supply, which is the SX system. In
combination, the CST and the SX system ensure that sufficient water is
available to feed the steam generators under all anticipated normal and
emergency conditions to cool a unit from full power conditions down to
350 degrees Fahrenheit, when the residual heat removal system can be
placed into service. The proposed changes ensure the CST will have
sufficient water to meet all normal operating conditions and mitigate
the consequences of all analyzed accidents except those that result in
CST unavailability. In addition, automatic switchover of the AF water
supply from the CSTs to SX will occur as assumed in the current safety
analyses for events where the CST becomes unavailable. The SX system
remains capable of supplying the emergency source of water to the AF
supply.
Therefore, the proposed changes do not involve a significant
reduction in the margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
requested amendments involve no significant hazards consideration.
Local Public Document Room location: For Byron, the Byron Public
Library District, 109 N. Franklin, P.O. Box 434, Byron, Illinois 61010;
for Braidwood, the Wilmington Public Library, 201 S. Kankakee Street,
Wilmington, Illinois 60481.
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: January 28, 1998 (NRC-98-0002).
Description of amendment request: The proposed amendment would
revise technical specification (TS) surveillance requirements
4.8.2.1.a.2, 4.8.2.1.b, and 4.8.2.1.c.4 to accommodate differences in
the monitored parameters between the existing batteries and the
batteries that will be installed for Division II during the sixth
refueling outage.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. The changes do not involve a significant increase in the
probability or consequences of an accident previously evaluated.
The proposed changes do not involve a change in the manner in which
the plant is operated. TS Section 4.8.2.1 is being revised to reflect
the new Division II battery cell/system characteristics and associated
requirements. The new battery will have an increased capacity over the
present battery, while maintaining the existing battery system voltage
requirements. This is possible because the present and new battery
specific gravity (1.215) and type (lead calcium) are the same. Also,
the end of battery system discharge voltage remains the same as 210
VDC. The Division II batteries will continue to furnish power to
redundant essential loads as required and as designed. The new
surveillance requirement voltages are based on the same volts/cell
criteria used for the existing batteries. Furthermore, failure or
malfunction of the station batteries does not initiate any of the
analyzed accidents previously evaluated in the UFSAR [updated final
safety analysis report]. The changes described will therefore not
involve an increase in the probability or consequences of an accident
previously evaluated.
2. The changes do not create the possibility of a new or different
kind of accident from any accident previously evaluated.
The new battery is Class 1E qualified equipment and is being
maintained within the same overall design parameters as the existing
battery. That is, the battery terminal voltage on float voltage
conditions (2.167 volt[s]/cell), overvoltage conditions (2.5 volts/
cell) and charger capability (2.15 volts/cell) are the same as the
original design. Furthermore, the end of system discharge voltage of
the battery system is maintained the same; therefore, there is no
negative impact to plant loads supplied by the batteries. Failures of
the batteries and chargers have been considered in both the existing
and modified configurations. The proposed changes will not change
performance or reliability nor introduce any new or different failure
modes or common mode failure and will therefore not create the
possibility of a new or different kind of accident from any accident
previously evaluated.
3. The changes do not involve a significant reduction in the margin
of safety.
The changes act to increase overall battery capacity from 560
ampere-hours to 1200 ampere-hours with the minimum battery discharge
voltage remaining at 210 VDC (or 105 VDC per battery). The battery
terminal voltage on float voltage conditions (2.167 volt[s]/cell),
overvoltage conditions (2.5 volts/cell) and charger capability (2.15
volts/cell) are the same as the original design. The new surveillance
requirement voltages are based on the same volts/cell criteria used for
the existing batteries. The batteries' ability to satisfy the design
requirements (battery duty cycle) of the dc system will not be reduced
from original plant design and will therefore not have any negative
impact to plant loads the battery supplies. The
[[Page 9598]]
proposed changes therefore do not involve a 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: 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: Cynthia A. Carpenter.
Detroit Edison Company, Docket No. 50-341, Fermi 2, Monroe County,
Michigan
Date of amendment request: January 28, 1998 (NRC-98-0003).
Description of amendment request: The proposed amendment would
revise technical specification (TS) 3.4.10, TS Figure 3.4.10-1 and the
associated bases by changing the prohibited and restricted operating
regions associated with core thermal-hydraulic stability. TS 3.4.1.4,
TS Figure 3.4.1.4-1, and the associated bases would also be revised to
reflect stability-related improvements in operating restrictions for
idle recirculation loop startup. Finally, in an unrelated change, TS
Tables 3.3.7.5-1 and 4.3.7.5-1 would be revised to delete neutron flux
from the parameters the licensee is required to monitor by TS 3.3.7.5,
Accident Monitoring Instrumentation.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
Thermal Hydraulic Stability and Idle Recirculation Loop Startup
1. The proposed TS changes do not involve a significant increase in
the probability or consequences of an accident previously evaluated.
These changes act to prohibit operations which have been found to
carry a significant potential for the formation of core thermal-
hydraulic instabilities and eliminates inappropriate technical
specifications for maintaining <50% recirculation="" loop="" flow="" before="" starting="" the="" idle="" recirculation="" pump.="" as="" such,="" operation="" in="" compliance="" with="" the="" proposed="" provisions="" does="" not="" affect="" any="" initiating="" mechanism="" for="" previously="" evaluated="" accidents="" or="" the="" response="" of="" the="" plant="" to="" a="" previously="" evaluated="" accident.="" the="" actions="" taken="" lead="" to="" placing="" the="" plant="" in="" a="" safe="" condition="" and="" are="" not="" themselves="" associated="" with="" an="" initiator="" for="" a="" previously="" evaluated="" accident.="" therefore,="" the="" change="" does="" not="" represent="" a="" significant="" increase="" in="" the="" probability="" or="" consequences="" of="" any="" previously="" evaluated="" accident.="" 2.="" the="" proposed="" ts="" changes="" do="" not="" create="" the="" possibility="" of="" a="" new="" or="" different="" kind="" of="" accident="" from="" any="" accident="" previously="" evaluated.="" as="" discussed="" above,="" the="" change="" acts="" to="" restrict="" operations="" previously="" allowed.="" the="" change="" also="" provides="" remedial="" actions="" that="" act="" to="" place="" the="" plant="" in="" a="" safe="" condition.="" the="" actions="" specified="" are="" within="" the="" analyzed="" domain="" of="" plant="" operations.="" unless="" an="" instability="" event="" is="" in="" progress,="" the="" new="" allowance="" to="" use="" a="" core="" flow="" increase="" to="" leave="" the="" exit="" region="" is="" no="" different="" than="" normal="" plant="" maneuvering.="" if="" an="" instability="" event="" is="" in="" progress,="" the="" new="" action="" 3.4.10.c="" to="" scram="" the="" reactor="" takes="" precedence.="" the="" allowance="" to="" start="" an="" idle="" loop="" with="" the="" active="" loop="" flow=""><50% of="" rated="" flow="" has="" been="" shown="" to="" have="" no="" adverse="" [e]ffect="" on="" scram="" avoidance="" or="" jet="" pump="" riser="" brace="" vibration.="" therefore,="" the="" proposed="" changes="" do="" not="" create="" a="" new="" or="" different="" type="" of="" accident.="" 3.="" the="" proposed="" ts="" changes="" do="" not="" involve="" a="" significant="" reduction="" in="" a="" margin="" of="" safety.="" consistent="" with="" the="" latest="" bwrog="" [boiling="" water="" reactor="" owners="" group]="" guidance,="" the="" changes="" act="" to="" expand="" the="" exit="" region="" compared="" to="" the="" current="" ts="" for="" core="" thermal-hydraulic="" instability="" and="" provide="" improved="" remedial="" actions="" which="" promptly="" terminate="" the="" potential="" for="" instability.="" these="" changes="" therefore="" do="" not="" involve="" a="" significant="" reduction="" in="" a="" margin="" of="" safety.="" post-accident="" monitoring="" 1.="" the="" proposed="" ts="" changes="" do="" not="" involve="" a="" significant="" increase="" in="" the="" probability="" or="" consequences="" of="" an="" accident="" previously="" evaluated.="" the="" proposed="" change="" does="" not="" involve="" a="" change="" in="" plant="" design="" or="" a="" change="" in="" the="" manner="" in="" which="" the="" plant="" is="" operated.="" the="" long="" term="" post-accident="" design="" requirements="" of="" the="" neutron="" monitoring="" system="" (nms)="" are="" not="" based="" on="" operator="" use="" for="" transients="" with="" scram,="" accidents="" with="" scram,="" and="" other="" occurrences="" without="" scram="" (reference="" 6="" [of="" january="" 28,="" 1998,="" application]).="" for="" lesser="" events="" such="" as="" transients="" without="" scram,="" the="" nms="" enhances="" the="" operator="" actions,="" since="" successful="" verification="" that="" power="" is="" below="" approximately="" 3%="" power="" can="" avoid="" non-routine="" operator="" actions="" (reference="" 6).="" these="" lesser="" events="" establish="" design="" requirements="" for="" the="" nms.="" the="" failure="" of="" this="" instrumentation="" during="" post-accident="" conditions="" will="" not="" prevent="" the="" operator="" from="" determining="" reactor="" power="" levels.="" alternate="" parameter="" status="" will="" be="" available="" from="" which="" reactor="" power="" may="" be="" inferred.="" based="" on="" the="" multiple="" inputs="" available="" to="" the="" operator,="" sufficient="" information="" will="" be="" available="" upon="" which="" to="" base="" operational="" decisions="" and="" to="" conclude="" that="" reactivity="" control="" has="" been="" accomplished.="" this="" change="" will="" therefore="" not="" represent="" a="" significant="" increase="" in="" the="" probability="" or="" consequences="" of="" an="" accident="" previously="" evaluated.="" 2.="" the="" proposed="" ts="" changes="" do="" not="" create="" the="" possibility="" of="" a="" new="" or="" different="" kind="" of="" accident="" from="" any="" accident="" previously="" evaluated.="" the="" proposed="" change="" does="" not="" introduce="" a="" new="" mode="" of="" plant="" operation="" and="" does="" not="" involve="" the="" installation="" of="" any="" new="" equipment="" or="" modifications="" to="" the="" plant.="" therefore,="" it="" 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="" change="" eliminates="" a="" ts="" listing="" of="" a="" function="" to="" reflect="" the="" actual="" safety="" significance.="" as="" such="" it="" has="" no="" effect="" on="" actual="" plant="" operation="" and="" thus="" no="" impact="" on="" any="" margin="" of="" safety.="" based="" on="" the="" above,="" detroit="" edison="" has="" determined="" that="" the="" proposed="" amendment="" 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:="" 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:="" cynthia="" a.="" carpenter.="" detroit="" edison="" company,="" docket="" no.="" 50-341,="" fermi="" 2,="" monroe="" county,="" michigan="" date="" of="" amendment="" request:="" january="" 28,="" 1998="" (nrc-98-0006).="" [[page="" 9599]]="" description="" of="" amendment="" request:="" the="" proposed="" amendment="" would="" revise="" technical="" specification="" (ts)="" surveillance="" requirement="" 4.4.3.2.2.a="" to="" extend="" the="" interval="" for="" leak="" rate="" testing="" of="" pressure="" isolation="" valves="" from="" 18="" months="" to="" 24="" months.="" basis="" for="" proposed="" no="" significant="" hazards="" consideration="" determination:="" as="" required="" by="" 10="" cfr="" 50.91(a),="" the="" licensee="" has="" provided="" its="" analysis="" of="" the="" issue="" of="" no="" significant="" hazards="" consideration,="" which="" is="" presented="" below:="" 1.="" the="" proposed="" ts="" change="" does="" not="" involve="" a="" significant="" increase="" in="" the="" probability="" or="" consequences="" of="" an="" accident="" previously="" evaluated.="" the="" proposed="" change="" revises="" the="" periodicity="" of="" ts="" surveillance="" requirement="" (sr)="" 4.4.3.2.2.a="" from="" ``at="" least="" once="" per="" 18="" months''="" to="" ``at="" least="" once="" per="" 24="" months.''="" this="" change="" revises="" the="" testing="" periodicity="" only;="" no="" other="" testing="" methodology="" is="" being="" affected.="" the="" testing="" periodicity="" is="" being="" revised="" to="" be="" consistent="" with="" other="" category="" ``a''="" valves="" since="" the="" pressure="" isolation="" valves="" (pivs)="" are="" classified="" as="" category="" ``a''="" valves.="" both="" asme="" [american="" society="" of="" mechanical="" engineers]="" [code]="" section="" xi="" and="" nureg-1482="" require="" category="" ``a''="" valves="" to="" be="" leak="" tested="" on="" a="" periodicity="" of="" at="" least="" once="" every="" 2="" years.="" the="" function="" of="" the="" pivs="" is="" to="" protect="" the="" low="" pressure="" portions="" of="" safety="" systems="" from="" the="" rcs="" [reactor="" coolant="" system]="" pressure.="" periodic="" valve="" leak="" rate="" testing="" is="" performed="" on="" the="" pivs="" to="" assure="" system="" integrity="" is="" maintained="" and="" to="" prevent="" the="" design="" pressure="" of="" the="" low="" pressure="" systems="" from="" being="" exceeded.="" the="" frequency="" of="" the="" inservice="" test="" could="" increase="" the="" probability="" that="" an="" increase="" in="" piv="" seat="" leakage="" may="" occur.="" if="" this="" were="" to="" occur="" and="" the="" leakage="" was="" significant="" (assuming="" leakage="" through="" both="" the="" inboard="" and="" outboard="" valves="" of="" the="" same="" penetration),="" the="" excess="" leakage="" would="" be="" detected="" by="" the="" system="" leakage="" detection="" instrumentation="" which="" would="" require="" corrective="" actions="" to="" be="" taken="" to="" assure="" that="" leakage="" remained="" within="" allowable="" limits.="" considering="" that="" past="" test="" results="" show="" very="" minimal="" seat="" leakage="" changes="" over="" years="" of="" service,="" the="" consequences="" and="" probabilities="" resulting="" from="" the="" proposed="" change="" is="" considered="" minimal.="" the="" proposed="" change="" does="" not="" impose="" or="" eliminate="" any="" testing="" requirements.="" this="" change="" is="" only="" a="" change="" to="" the="" frequency="" (testing="" interval)="" for="" measuring="" the="" seat="" leakage="" through="" the="" pivs.="" the="" pivs="" will="" continue="" to="" be="" tested="" in="" accordance="" with="" asme="" code="" section="" xi.="" this="" change="" does="" not="" affect="" any="" of="" the="" parameters="" or="" conditions="" that="" could="" contribute="" to="" the="" initiation="" of="" any="" accidents="" previously="" evaluated="" and="" therefore="" cannot="" increase="" the="" consequences="" or="" probabilities="" of="" any="" accident="" previously="" evaluated.="" 2.="" the="" proposed="" ts="" change="" does="" not="" create="" the="" possibility="" of="" a="" new="" or="" different="" kind="" of="" accident="" from="" any="" accident="" previously="" evaluated.="" the="" proposed="" change="" does="" not="" involve="" a="" change="" to="" the="" plant="" design="" or="" operation.="" as="" a="" result,="" the="" proposed="" change="" does="" not="" affect="" any="" of="" the="" parameters="" or="" conditions="" that="" could="" contribute="" to="" the="" initiation="" of="" any="" accidents.="" this="" change="" only="" involves="" the="" lengthening="" of="" the="" pivs'="" testing="" frequency="" from="" 18="" months="" to="" 24="" months.="" the="" method="" for="" performing="" the="" actual="" tests="" are="" not="" changed.="" no="" new="" accident="" scenarios="" are="" created="" by="" extending="" the="" testing="" intervals.="" no="" safety-related="" equipment="" or="" safety="" functions="" are="" altered="" as="" a="" result="" of="" this="" change.="" therefore,="" extending="" the="" test="" frequency="" does="" not="" create="" the="" possibility="" of="" a="" new="" or="" different="" kind="" of="" accident="" or="" malfunction="" from="" those="" previously="" analyzed.="" 3.="" the="" proposed="" ts="" change="" does="" not="" involve="" a="" significant="" reduction="" in="" a="" margin="" of="" safety.="" the="" proposed="" change="" only="" affects="" the="" frequency="" of="" the="" pivs'="" seat="" leakage="" tests.="" the="" frequency="" is="" proposed="" to="" be="" extended="" to="" reflect="" the="" asme="" section="" xi,="" 1980="" edition,="" winter="" 1980="" addenda,="" section="" iwv-3422="" seat="" leakage="" testing="" periodicity="" requirement="" of="" 24="" months.="" no="" other="" testing="" methodology="" is="" being="" changed.="" the="" allowable="" leakage="" limits="" will="" not="" be="" affected="" by="" this="" change.="" the="" margin="" of="" safety="" as="" defined="" in="" the="" bases="" of="" any="" technical="" specification="" will,="" therefore,="" not="" be="" reduced="" by="" extending="" the="" testing="" periodicity="" of="" the="" subject="" valves.="" 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:="" 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:="" cynthia="" a.="" carpenter.="" detroit="" edison="" company,="" docket="" no.="" 50-341,="" fermi="" 2,="" monroe="" county,="" michigan="" date="" of="" amendment="" request:="" january="" 28,="" 1998="" (nrc-98-0008).="" description="" of="" amendment="" request:="" the="" proposed="" amendment="" would="" revise="" the="" technical="" specifications="" (tss)="" by="" modifying="" the="" ``#''="" footnote="" to="" table="" 1.2="" and="" the="" ``*''="" footnote="" to="" surveillance="" requirements="" 4.9.1.2="" and="" 4.9.1.3="" to="" permit="" the="" reactor="" mode="" switch="" to="" be="" placed="" in="" the="" run="" or="" startup/hot="" standby="" positions="" to="" test="" switch="" interlock="" functions="" provided="" that="" all="" control="" rods="" are="" verified="" to="" remain="" fully="" inserted="" in="" core="" cells="" containing="" one="" or="" more="" fuel="" assemblies.="" basis="" for="" proposed="" no="" significant="" hazards="" consideration="" determination:="" as="" required="" by="" 10="" cfr="" 50.91(a),="" the="" licensee="" has="" provided="" its="" analysis="" of="" the="" issue="" of="" no="" significant="" hazards="" consideration,="" which="" is="" presented="" below:="" 1.="" the="" proposed="" change="" does="" not="" involve="" a="" significant="" increase="" in="" the="" probability="" or="" consequences="" of="" an="" accident="" previously="" evaluated.="" the="" proposed="" change="" would="" permit="" the="" reactor="" mode="" switch="" to="" be="" placed="" in="" the="" run="" or="" startup/hot="" standby="" positions="" to="" test="" the="" switch="" interlock="" functions="" provided="" that="" all="" control="" rods="" are="" verified="" to="" remain="" inserted="" in="" core="" cells="" containing="" one="" or="" more="" fuel="" assemblies.="" the="" existing="" ts="" requires="" that="" all="" control="" rods="" be="" verified="" to="" remain="" inserted="" regardless="" of="" whether="" core="" cells="" are="" defueled.="" the="" reactor="" mode="" switch="" refuel="" position="" interlocks="" restrict="" the="" operation="" of="" refueling="" equipment="" or="" withdrawal="" of="" control="" rods="" to="" reinforce="" unit="" procedures="" that="" prevent="" the="" reactor="" from="" achieving="" criticality="" during="" refueling="" operations.="" as="" such,="" the="" refueling="" equipment="" interlocks="" preserve="" the="" assumptions="" for="" the="" analyses="" of="" a="" control="" rod="" withdrawal="" event="" or="" loading="" of="" a="" fuel="" assembly="" into="" an="" uncontrolled="" cell="" during="" refueling="" operations.="" the="" reactor="" mode="" switch="" refuel="" position="" interlocks="" are="" not="" initiators="" of="" any="" previously="" evaluated="" accident.="" the="" revised="" footnote="" requires="" that="" all="" control="" rods="" remain="" fully="" inserted="" in="" core="" cells="" containing="" one="" or="" more="" fuel="" assemblies="" while="" the="" mode="" switch="" is="" moved="" to="" support="" interlock="" testing.="" additionally,="" when="" the="" reactor="" mode="" switch="" is="" unlocked="" to="" support="" interlock="" testing,="" ts="" 3.9.1="" prohibits="" core="" alterations.="" with="" all="" control="" rods="" fully="" inserted="" in="" core="" cells="" containing="" one="" or="" more="" fuel="" assemblies="" and="" no="" core="" alterations="" in="" progress,="" there="" are="" no="" credible="" mechanisms="" to="" initiate="" a="" reactivity="" excursion="" during="" the="" interlock="" [[page="" 9600]]="" testing.="" therefore,="" the="" proposed="" change="" does="" not="" involve="" a="" significant="" increase="" in="" the="" probability="" of="" a="" previously="" evaluated="" accident.="" the="" proposed="" change="" accommodates="" reactor="" mode="" switch="" refuel="" position="" interlock="" testing="" with="" one="" or="" more="" control="" rods="" removed="" as="" permitted="" by="" ts="" 3.9.10.1="" and="" 3.9.10.2.="" in="" addition="" to="" requiring="" all="" fuel="" assemblies="" to="" be="" removed="" from="" core="" cells="" associated="" with="" removed="" control="" rods,="" ts="" 3.9.10.1="" and="" 3.9.10.2="" require="" minimum="" shutdown="" margin="" to="" be="" maintained="" in="" accordance="" with="" ts="" 3/4.1.1.="" under="" these="" conditions,="" it="" is="" not="" possible="" for="" criticality="" to="" occur="" in="" the="" event="" of="" a="" withdrawal="" of="" a="" single="" control="" rod="" or="" loading="" of="" fuel="" assemblies="" into="" a="" single="" core="" cell="" with="" no="" control="" rod="" inserted.="" therefore,="" the="" proposed="" change="" does="" not="" involve="" a="" significant="" increase="" in="" the="" consequences="" of="" a="" previously="" evaluated="" accident.="" 2.="" the="" proposed="" change="" does="" not="" create="" the="" possibility="" of="" a="" new="" or="" different="" kind="" of="" accident="" from="" any="" accident="" previously="" evaluated.="" repositioning="" of="" the="" reactor="" mode="" switch="" to="" test="" refueling="" position="" interlocks="" is="" permitted="" by="" both="" the="" existing="" and="" proposed="" ts.="" the="" proposed="" change="" affects="" only="" the="" conditions="" under="" which="" the="" mode="" switch="" can="" be="" repositioned.="" the="" proposed="" changes="" do="" not="" change="" underlying="" principles="" affecting="" the="" way="" in="" which="" the="" plant="" is="" operated="" and="" no="" new="" or="" different="" failure="" modes="" are="" introduced="" by="" the="" proposed="" change="" for="" any="" plant="" system="" or="" component.="" no="" new="" limiting="" single="" failure="" has="" been="" identified="" as="" a="" result="" of="" the="" proposed="" changes.="" therefore,="" no="" new="" or="" different="" types="" of="" failures="" or="" accident="" initiators="" are="" introduced="" by="" the="" proposed="" changes.="" 3.="" the="" change="" does="" not="" involve="" a="" significant="" reduction="" in="" the="" margin="" of="" safety.="" the="" proposed="" change="" described="" above="" affects="" the="" conditions="" under="" which="" the="" reactor="" mode="" switch="" can="" be="" repositioned="" to="" accommodate="" refuel="" position="" interlock="" testing.="" the="" proposed="" change="" in="" combination="" with="" existing="" restrictions="" within="" the="" ts="" provide="" assurance="" that="" there="" is="" no="" credible="" mechanism="" to="" initiate="" a="" reactivity="" excursion="" during="" interlock="" testing.="" therefore,="" the="" proposed="" change="" does="" not="" involve="" a="" significant="" reduction="" in="" a="" margin="" of="" safety.="" the="" nrc="" staff="" has="" reviewed="" the="" licensee's="" analysis="" and,="" based="" on="" this="" review,="" it="" appears="" that="" the="" three="" standards="" of="" 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:="" 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:="" cynthia="" a.="" carpenter.="" detroit="" edison="" company,="" docket="" no.="" 50-341,="" fermi="" 2,="" monroe="" county,="" michigan="" date="" of="" amendment="" request:="" january="" 28,="" 1998="" (nrc-98-0011).="" description="" of="" amendment="" request:="" the="" proposed="" amendment="" would="" revise="" technical="" specification="" (ts)="" 3.4.2.1="" by="" changing="" the="" tolerance="" for="" the="" as-found="" setpoints="" of="" the="" safety/relief="" valves="" (srvs)="" from="" [plus="" or="" minus]="" 1="" percent="" to="" [plus="" or="" minus]="" 3="" percent="" of="" the="" nominal="" setpoint.="" the="" revised="" tolerance="" would="" be="" used="" when="" evaluating="" whether="" setpoint="" test="" results="" were="" acceptable.="" however,="" after="" initial="" testing,="" the="" as-left="" setpoints="" of="" the="" srvs="" would="" be="" adjusted="" to="" within="" [plus="" or="" minus]="" 1="" percent="" of="" the="" nominal="" setpoint.="" 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="" this="" change="" involve="" a="" significant="" increase="" in="" the="" probability="" or="" consequences="" of="" an="" accident="" previously="" evaluated?="" the="" proposed="" change="" allows="" an="" increase="" in="" the="" srv="" setpoint="" tolerance,="" determined="" by="" test="" after="" the="" valves="" have="" been="" removed="" from="" service,="" from="" [plus="" or="" minus]="" 1%="" to="" [plus="" or="" minus]="" 3%.="" the="" proposed="" change="" does="" not="" alter="" the="" srv="" lift="" setpoints,="" the="" srv="" lift="" setpoint="" test="" frequency,="" or="" the="" number="" of="" srvs="" required="" to="" be="" operable.="" this="" change="" does="" not="" involve="" physical="" changes="" to="" the="" srvs,="" nor="" does="" it="" change="" the="" operating="" characteristics="" or="" safety="" function="" of="" the="" srvs.="" this="" change="" requires="" that="" the="" srvs="" be="" adjusted="" to="" within="" [plus="" or="" minus]="" 1%="" of="" their="" nominal="" lift="" setpoints="" following="" testing="" and="" prior="" to="" installation="" in="" the="" plant.="" the="" only="" change,="" other="" than="" the="" change="" in="" setpoint="" tolerance,="" will="" be="" to="" increase="" the="" maximum="" rated="" speed="" of="" the="" rcic="" [reactor="" core="" isolation="" cooling]="" turbine="" and="" pump.="" the="" increased="" speed="" is="" within="" the="" design="" limits="" of="" the="" system="" and="" the="" overspeed="" trip="" function="" retains="" adequate="" margin;="" therefore,="" rcic="" operability="" is="" not="" affected="" by="" this="" change.="" additionally,="" srv="" actuation="" is="" not="" a="" precursor="" to="" any="" design="" basis="" accident="" analyzed="" for="" the="" fermi="" 2="" plant.="" therefore,="" this="" change="" will="" not="" significantly="" increase="" the="" probability="" of="" an="" accident="" previously="" evaluated.="" generic="" considerations="" related="" to="" the="" change="" in="" setpoint="" tolerance="" were="" addressed="" in="" nedc-31753p,="" ``bwrog="" in-service="" pressure="" relief="" technical="" specification="" revision="" licensing="" topical="" report,''="" and="" were="" reviewed="" and="" approved="" by="" the="" nrc.="" the="" plant="" specific="" evaluations="" identified="" in="" the="" nrc[']s="" safety="" evaluation="" for="" nedc-31753p="" were="" performed="" in="" order="" to="" support="" the="" proposed="" change="" (cycle="" 6="" reload="" licensing="" report,="" power="" uprate="" safety="" analysis,="" and="" nedc-32788p,="" ``safety="" review="" for="" enrico="" fermi="" energy="" center="" unit="" 2="" safety/relief="" valve="" setpoint="" tolerance="" relaxation="" analyses'').="" these="" evaluations="" included="" transient="" analysis="" of="" the="" anticipated="" operational="" occurrences="" (aoos);="" analysis="" of="" the="" design="" basis="" overpressurization="" event;="" evaluation="" of="" the="" performance="" of="" high="" pressure="" systems,="" motor="" operated="" valves,="" and="" vessel="" instrumentation="" and="" associated="" piping;="" and="" evaluation="" of="" the="" containment="" response="" during="" loca="" [loss="" of="" coolant="" accident]="" and="" the="" hydrodynamic="" loads="" on="" the="" srv="" discharge="" lines="" and="" containment.="" although="" not="" specified="" in="" the="" generic="" topical="" report="" nedc-="" 31753p,="" an="" analysis="" of="" the="" short="" term="" pressurization="" phase="" of="" an="" atws="" [anticipated="" transient="" without="" scram]="" event="" was="" also="" performed.="" these="" analyses="" show="" that="" there="" is="" adequate="" margin="" to="" the="" design="" core="" thermal="" limits="" and="" to="" the="" reactor="" vessel="" pressure="" limits="" using="" a="" [plus="" or="" minus]="" 3%="" srv="" setpoint="" tolerance.="" they="" also="" show="" that="" operation="" of="" the="" high="" pressure="" injection="" systems="" will="" not="" be="" adversely="" affected;="" and="" the="" containment="" response="" during="" loca="" will="" be="" acceptable.="" therefore,="" this="" change="" will="" not="" involve="" a="" significant="" increase="" in="" the="" probability="" or="" consequences="" of="" any="" accident="" previously="" evaluated.="" 2.="" does="" this="" change="" create="" the="" possibility="" of="" a="" new="" or="" different="" kind="" of="" accident="" from="" any="" accident="" previously="" evaluated?="" the="" proposed="" change="" to="" allow="" an="" increase="" in="" the="" srv="" setpoint="" tolerance="" from="" [plus="" or="" minus]="" 1%="" to="" [plus="" or="" minus]="" 3%="" does="" not="" alter="" the="" srv="" lift="" setpoints,="" the="" minimum="" srv="" lift="" setpoint="" test="" frequency,="" or="" the="" number="" of="" srvs="" required="" to="" be="" operable.="" this="" change="" does="" not="" involve="" physical="" changes="" to="" the="" srvs,="" nor="" does="" it="" change="" the="" operating="" characteristics="" or="" the="" safety="" function="" of="" the="" srvs.="" the="" only="" change="" to="" plant="" equipment="" will="" be="" to="" increase="" the="" rcic="" turbine/pump="" maximum="" rated="" speed="" from="" 4550="" rpm="" to="" 4600="" rpm.="" the="" rcic="" pump="" and="" turbine="" [[page="" 9601]]="" have="" been="" verified="" to="" be="" capable="" of="" operating="" at="" the="" increased="" speed,="" pressure="" and="" temperature="" associated="" with="" this="" increase="" in="" maximum="" rated="" speed.="" these="" changes="" do="" not="" result="" in="" any="" changed="" component="" interactions.="" the="" srvs="" and="" the="" rcic="" system="" will="" continue="" to="" function="" as="" designed.="" therefore,="" this="" change="" will="" not="" create="" the="" possibility="" of="" a="" new="" or="" different="" kind="" of="" accident="" from="" any="" accident="" previously="" evaluated.="" 3.="" does="" this="" change="" involve="" a="" significant="" reduction="" in="" a="" margin="" of="" safety?="" while="" the="" calculated="" peak="" vessel="" pressures="" for="" the="" asme="" [american="" society="" of="" mechanical="" engineers]="" overpressure="" event="" and="" the="" atws="" msivc="" [main="" steam="" isolation="" valve="" closure]="" event="" are="" higher="" than="" those="" calculated="" without="" the="" setpoint="" tolerance="" relaxation,="" both="" are="" still="" within="" the="" respective="" licensing="" acceptance="" limits="" associated="" with="" these="" events.="" similarly,="" although="" the="" loads="" associated="" with="" srv="" blowdown="" could="" increase="" slightly,="" containment="" loadings="" have="" been="" determined="" to="" remain="" within="" acceptance="" limits.="" these="" licensing="" acceptance="" limits="" have="" been="" determined="" by="" the="" nrc="" to="" provide="" a="" sufficient="" margin="" of="" safety.="" additionally,="" the="" increased="" setpoint="" tolerances="" have="" been="" determined="" to="" have="" a="" negligible="" effect="" on="" the="" other="" accidents="" and="" transients="" analyzed.="" therefore,="" the="" proposed="" change="" will="" not="" involve="" a="" significant="" reduction="" in="" a="" margin="" of="" safety.="" the="" nrc="" staff="" has="" reviewed="" the="" licensee's="" analysis="" and,="" based="" on="" this="" review,="" it="" appears="" that="" the="" three="" standards="" of="" 10="" cfr="" 50.92(c)="" are="" satisfied.="" therefore,="" the="" nrc="" staff="" proposes="" to="" determine="" that="" the="" amendment="" request="" involves="" no="" significant="" hazards="" consideration.="" local="" public="" document="" room="" location:="" 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:="" cynthia="" a.="" carpenter.="" duquesne="" light="" company,="" et="" al.,="" docket="" no.="" 50-334,="" beaver="" valley="" power="" station,="" unit="" no.="" 1,="" shippingport,="" pennsylvania="" date="" of="" amendment="" request:="" january="" 17,="" 1998.="" description="" of="" amendment="" request:="" the="" proposed="" amendment="" would="" revise="" the="" waste="" gas="" system="" line="" break="" accident="" analysis.="" the="" proposed="" changes="" would="" affect="" beaver="" valley="" power="" station,="" unit="" no.="" 1="" updated="" final="" safety="" analysis="" report="" (ufsar)="" tables="" 11.3-7,="" ``postulated="" control="" room="" accident="" dose,''="" and="" 14.2-8,="" ``parameters="" used="" in="" control="" room="" habitability="" analysis="" of="" the="" waste="" gas="" system="" failure="" analysis.''="" the="" analysis="" references="" on="" tables="" 11.3-7="" and="" 14.2-8="" would="" be="" revised="" due="" to="" the="" reanalysis="" of="" the="" waste="" gas="" system="" line="" break="" accident.="" in="" table="" 11.3-7,="" the="" waste="" gas="" system="" line="" break="" accident="" gamma="" dose="" value="" would="" be="" revised="" from="" 0.0031="" rem="" to="" less="" than="" 0.01="" rem="" and="" the="" beta="" dose="" value="" would="" be="" revised="" from="" 0.013="" rem="" to="" less="" than="" 1.0="" rem.="" 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="" change="" involve="" a="" significant="" increase="" in="" the="" probability="" or="" consequences="" of="" an="" accident="" previously="" evaluated?="" the="" proposed="" change="" has="" no="" effect="" on="" the="" probability="" of="" an="" accident="" previously="" evaluated.="" the="" proposed="" change="" results="" from="" the="" correction="" of="" values="" and="" change="" to="" assumptions="" utilized="" in="" the="" original="" calculation="" to="" address="" resultant="" dose="" to="" control="" room="" operators="" in="" the="" event="" of="" the="" postulated="" waste="" gas="" system="" line="" break.="" the="" proposed="" change="" also="" corrects="" an="" error="" in="" ufsar="" table="" 14.2-8="" whereby="" the="" fraction="" of="" fuel="" with="" defects="" was="" assumed="" to="" be="" one="" percent,="" not="" 0.0026.="" this="" correction="" reflects="" the="" value="" used="" in="" the="" calculation="" and="" does="" not="" alter="" the="" results.="" the="" proposed="" change="" does="" not="" significantly="" increase="" the="" consequences="" of="" an="" accident="" previously="" analyzed.="" although="" the="" correction="" to="" the="" calculation="" and="" revision="" to="" the="" assumptions="" used="" result="" in="" an="" insignificant="" increase="" to="" the="" postulated="" dose="" to="" the="" control="" room="" operators,="" the="" results="" remain="" below="" the="" acceptance="" limit="" of="" other="" postulated="" accidents="" presented="" in="" the="" ufsar="" (table="" 11.3-7)="" and="" the="" acceptance="" approved="" by="" the="" nrc="" in="" the="" nrc="" safety="" evaluation="" report,="" section="" 15.1,="" dated="" october="" 1974.="" the="" proposed="" change="" does="" not="" alter="" the="" currently="" approved="" technical="" specification.="" the="" proposed="" change="" does="" not="" affect="" the="" dose="" to="" the="" public.="" 2.="" does="" the="" change="" create="" the="" possibility="" of="" a="" new="" or="" different="" kind="" of="" accident="" from="" any="" accident="" previously="" evaluated?="" the="" proposed="" change="" does="" not="" alter="" the="" physical="" plant="" or="" modify="" the="" modes="" of="" operation.="" the="" proposed="" change="" does="" not="" involve="" modifications="" to="" plant="" equipment="" nor="" does="" it="" alter="" operation="" of="" plant="" systems.="" therefore="" operation="" of="" the="" facility="" with="" the="" proposed="" change="" does="" not="" create="" the="" possibility="" of="" a="" new="" or="" different="" kind="" of="" accident="" from="" any="" accident="" previously="" evaluated.="" 3.="" does="" the="" change="" involve="" a="" significant="" reduction="" in="" a="" margin="" of="" safety?="" the="" proposed="" change="" does="" not="" reduce="" the="" margin="" of="" safety.="" the="" proposed="" change="" does="" not="" affect="" any="" plant="" systems="" or="" equipment.="" therefore,="" the="" response="" of="" the="" plant="" to="" any="" actual="" events="" will="" not="" be="" affected,="" and="" the="" 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.="" therefore,="" the="" nrc="" staff="" proposes="" to="" determine="" that="" the="" amendment="" request="" involves="" no="" significant="" hazards="" consideration.="" local="" public="" document="" room="" location:="" b.="" f.="" jones="" memorial="" library,="" 663="" franklin="" avenue,="" aliquippa,="" pa="" 15001.="" attorney="" for="" licensee:="" jay="" e.="" silberg,="" esquire,="" shaw,="" pittman,="" potts="" &="" trowbridge,="" 2300="" n="" street,="" nw.,="" washington,="" dc="" 20037.="" nrc="" project="" director:="" john="" f.="" stolz.="" entergy="" operations="" inc.,="" docket="" no.="" 50-382,="" waterford="" steam="" electric="" station,="" unit="" 3="" (waterford="" 3),="" st.="" charles="" parish,="" louisiana="" date="" of="" amendment="" request:="" november="" 13,="" 1997.="" description="" of="" amendment="" request:="" the="" proposed="" change="" will="" modify="" technical="" specification="" (ts)="" 6.8.4.a,="" ``primary="" coolant="" sources="" outside="" containment,''="" to="" add="" portions="" of="" the="" containment="" vacuum="" relief="" (cvr)="" system="" and="" the="" primary="" sampling="" system="" to="" the="" program="" at="" waterford="" 3.="" basis="" for="" proposed="" no="" significant="" hazards="" consideration="" determination:="" as="" required="" by="" 10="" cfr="" 50.91(a),="" the="" licensee="" has="" provided="" its="" analysis="" of="" the="" issue="" of="" no="" significant="" hazards="" consideration,="" which="" is="" presented="" below:="" 1.="" will="" operation="" of="" the="" facility="" in="" accordance="" with="" this="" proposed="" change="" involve="" a="" significant="" increase="" in="" the="" probability="" or="" consequences="" of="" an="" accident="" previously="" evaluated?="" response:="" no.="" the="" proposed="" change="" adds="" the="" containment="" vacuum="" relief="" (cvr)="" system="" and="" the="" primary="" sampling="" system="" to="" the="" primary="" coolant="" sources="" outside="" containment="" program="" in="" the="" technical="" specifications.="" the="" program="" will="" require="" preventative="" maintenance="" [[page="" 9602]]="" and="" periodic="" visual="" inspection,="" and="" leak="" rate="" testing="" on="" appropriate="" portions="" of="" these="" systems="" to="" ensure="" leakage="" of="" radioactive="" fluids="" are="" as="" low="" as="" practicable.="" the="" addition="" of="" these="" two="" systems="" to="" the="" program="" will="" not="" affect="" the="" probability="" of="" an="" accident.="" neither="" the="" cvr="" system="" nor="" the="" primary="" sampling="" system="" are="" initiators="" of="" any="" analyzed="" event.="" the="" consequences="" of="" an="" accident="" are="" not="" affected="" by="" this="" change.="" the="" maximum="" allowed="" leakage="" limits="" are="" not="" being="" increased="" due="" to="" the="" addition="" of="" these="" two="" systems.="" any="" leakage="" from="" the="" cvr="" system="" will="" be="" factored="" into="" the="" overall="" leakage="" limits="" and="" any="" leakage="" from="" the="" primary="" sampling="" system="" will="" be="" kept="" to="" a="" minimum="" by="" performing="" required="" maintenance.="" this="" change="" does="" not="" affect="" the="" mitigation="" capabilities="" of="" any="" component="" or="" system="" nor="" does="" it="" affect="" the="" assumptions="" relative="" to="" the="" mitigation="" of="" accidents="" or="" transients.="" the="" addition="" of="" these="" systems="" to="" the="" program="" also="" helps="" ensure="" that="" the="" systems="" will="" perform="" their="" intended="" function.="" therefore,="" the="" proposed="" change="" will="" not="" involve="" a="" significant="" increase="" in="" the="" probability="" or="" consequences="" of="" any="" accident="" previously="" evaluated.="" 2.="" will="" operation="" of="" the="" facility="" in="" accordance="" with="" this="" proposed="" change="" create="" the="" possibility="" of="" a="" new="" or="" different="" type="" of="" accident="" from="" any="" accident="" previously="" evaluated?="" response:="" no.="" the="" proposed="" change="" adds="" the="" cvr="" system="" and="" the="" primary="" sampling="" system="" to="" the="" primary="" coolant="" sources="" outside="" containment="" program="" in="" the="" technical="" specifications.="" the="" program="" will="" require="" preventative="" maintenance="" and="" periodic="" visual="" inspection,="" and="" leak="" rate="" testing="" on="" appropriate="" portions="" of="" these="" systems="" to="" ensure="" leakage="" of="" radioactive="" fluids="" are="" as="" low="" as="" practical.="" neither="" the="" design="" nor="" configuration="" of="" the="" plant="" is="" being="" changed="" due="" to="" the="" addition="" of="" the="" cvr="" system="" to="" the="" program.="" also,="" as="" a="" result="" of="" the="" cvr="" system="" being="" added="" to="" the="" program,="" there="" has="" been="" no="" physical="" change="" to="" plant="" systems,="" structures="" or="" components="" nor="" will="" the="" addition="" of="" the="" cvr="" system="" reduce="" the="" ability="" of="" any="" of="" the="" safety-related="" equipment="" required="" to="" mitigate="" anticipated="" operational="" occurrences="" (aoos)="" or="" accidents.="" although="" the="" addition="" of="" the="" primary="" sampling="" system="" to="" the="" program="" was="" a="" result="" of="" a="" change="" to="" the="" configuration="" of="" the="" plant,="" it="" does="" not="" reduce="" the="" ability="" of="" any="" safety-related="" equipment="" required="" to="" mitigate="" aoos="" or="" accidents.="" any="" leakage="" from="" the="" primary="" sampling="" system="" will="" be="" kept="" to="" a="" minimum="" by="" performing="" required="" maintenance.="" therefore,="" the="" proposed="" change="" will="" not="" create="" the="" possibility="" of="" a="" new="" or="" different="" kind="" of="" accident="" from="" any="" accident="" previously="" evaluated.="" 3.="" will="" operation="" of="" the="" facility="" in="" accordance="" with="" this="" proposed="" change="" involve="" a="" significant="" reduction="" in="" a="" margin="" of="" safety?="" response:="" no.="" the="" proposed="" change="" adds="" the="" cvr="" system="" and="" the="" primary="" sampling="" system="" to="" the="" primary="" coolant="" sources="" outside="" containment="" program="" in="" the="" technical="" specifications.="" the="" program="" will="" require="" preventative="" maintenance="" and="" periodic="" visual="" inspection,="" and="" leak="" rate="" testing="" on="" appropriate="" portions="" of="" these="" systems="" to="" ensure="" leakage="" of="" radioactive="" fluids="" are="" as="" low="" as="" practical.="" this="" change="" will="" not="" affect="" the="" maximum="" containment="" leakage="" allowed="" in="" the="" technical="" specifications.="" the="" leakage="" from="" the="" cvr="" system="" will="" be="" added="" to="" the="" overall="" containment="" leakage="" rate.="" any="" leakage="" from="" the="" primary="" sampling="" system="" will="" be="" kept="" to="" a="" minimum="" by="" performing="" required="" maintenance.="" the="" overall="" containment="" leakage="" requirement="" is="" required="" to="" be="" met="" and="" therefore,="" this="" change="" will="" not="" result="" in="" an="" increase="" in="" the="" analyzed="" dose="" consequences="" assumed="" in="" the="" waterford="" 3="" safety="" analysis.="" therefore,="" the="" proposed="" change="" will="" not="" involve="" a="" significant="" reduction="" in="" a="" margin="" of="" safety.="" the="" nrc="" staff="" has="" reviewed="" the="" licensee's="" analysis="" and,="" based="" on="" this="" review,="" it="" appears="" that="" the="" three="" standards="" of="" 10="" cfr="" 50.92(c)="" are="" satisfied.="" therefore,="" the="" nrc="" staff="" proposes="" to="" determine="" that="" the="" amendment="" request="" involves="" no="" significant="" hazards="" consideration.="" local="" public="" document="" room="" location:="" university="" of="" new="" orleans="" library,="" louisiana="" collection,="" lakefront,="" new="" orleans,="" la="" 70122.="" attorney="" for="" licensee:="" n.s.="" reynolds,="" esq.,="" winston="" &="" strawn="" 1400="" l="" street="" n.w.,="" washington,="" d.c.="" 20005-3502.="" nrc="" project="" director:="" john="" n.="" hannon.="" florida="" power="" and="" light="" company,="" et="" al.,="" docket="" no.="" 50-389,="" st.="" lucie="" plant,="" unit="" no.="" 2,="" st.="" lucie="" county,="" florida="" date="" of="" amendment="" request:="" december="" 31,="" 1997.="" description="" of="" amendment="" request:="" the="" proposed="" amendment="" will="" revise="" technical="" specification="" 5.6.1="" and="" associated="" figure="" 5.6-1,="" and="" specification="" 5.6.3,="" to="" permit="" an="" increase="" in="" the="" allowed="" spent="" fuel="" pool="" (sfp)="" storage="" capacity.="" the="" analyses="" supporting="" this="" request,="" in="" part,="" assume="" credit="" for="" up="" to="" 1266="" ppm="" boron="" concentration="" existing="" in="" the="" sfp.="" basis="" for="" proposed="" no="" significant="" hazards="" consideration="" determination:="" as="" required="" by="" 10="" cfr="" 50.91(a),="" the="" licensee="" has="" provided="" its="" analysis="" of="" the="" issue="" of="" no="" significant="" hazards="" consideration,="" which="" is="" presented="" below:="" 1.="" the="" proposed="" amendment="" will="" not="" involve="" a="" significant="" increase="" in="" the="" probability="" or="" consequences="" of="" an="" accident="" previously="" evaluated.="" analyses="" to="" support="" the="" proposed="" fuel="" pool="" capacity="" increase="" have="" been="" developed="" using="" conservative="" methodology.="" the="" analysis="" of="" the="" potential="" accidents="" summarized="" below="" has="" shown="" that="" there="" is="" no="" significant="" increase="" in="" the="" consequences="" of="" any="" accident="" previously="" analyzed.="" a="" review="" of="" relevant="" plant="" operations="" has="" also="" demonstrated="" that="" there="" is="" no="" significant="" increase="" in="" the="" probability="" of="" occurrence="" of="" any="" accident="" previously="" analyzed.="" this="" conclusion="" is="" also="" discussed="" below.="" previously="" evaluated="" accidents="" that="" were="" examined="" for="" this="" proposed="" license="" amendment="" include:="" fuel="" handling="" accident,="" spent="" fuel="" cask="" drop="" accident,="" and="" loss="" of="" all="" fuel="" pool="" cooling.="" there="" will="" be="" no="" change="" in="" the="" mode="" of="" plant="" operation="" or="" in="" the="" availability="" of="" plant="" systems="" as="" a="" result="" of="" this="" proposed="" change;="" the="" systems="" interfacing="" with="" the="" spent="" fuel="" pool="" have="" previously="" encountered="" borated="" pool="" water="" and="" are="" designed="" to="" interact="" with="" irradiated="" spent="" fuel="" and="" remove="" the="" residual="" heat="" load="" generated="" by="" isotopic="" decay.="" the="" proposed="" amendment="" does="" not="" require="" a="" change="" in="" the="" maintenance="" interval="" or="" maintenance="" scope="" for="" the="" fuel="" pool="" cooling="" system="" or="" for="" the="" spent="" fuel="" cask="" crane.="" the="" frequency="" of="" cask="" handling="" operations="" and="" the="" maximum="" weight="" carried="" by="" the="" crane="" is="" not="" increased="" as="" a="" result="" of="" the="" proposed="" license="" amendment.="" thus,="" there="" will="" be="" no="" increase="" in="" the="" probability="" of="" a="" loss="" of="" fuel="" pool="" cooling="" or="" in="" the="" probability="" of="" a="" failure="" of="" the="" cask="" crane="" as="" a="" result="" of="" the="" proposed="" amendment.="" there="" will="" not="" be="" a="" significant="" increase="" in="" the="" frequency="" of="" handling="" discharged="" assemblies="" in="" the="" fuel="" pool="" as="" a="" result="" of="" this="" change;="" any="" handling="" of="" fuel="" in="" the="" spent="" fuel="" pool="" will="" continue="" to="" be="" performed="" in="" borated="" water.="" if="" the="" license="" amendment="" is="" approved,="" there="" will="" be="" a="" one-time="" repositioning="" of="" certain="" discharged="" assemblies="" stored="" in="" the="" fuel="" pool="" to="" comply="" with="" the="" revised="" positioning="" requirements,="" but="" the="" increased="" pool="" storage="" capacity="" will="" permit="" the="" deferral="" of="" spent="" fuel="" handling="" associated="" with="" cask="" loading="" operations.="" fuel="" manipulation="" during="" the="" repositioning="" activity="" will="" be="" performed="" in="" the="" same="" [[page="" 9603]]="" manner="" as="" for="" fuel="" placed="" in="" the="" spent="" fuel="" pool="" during="" refueling="" outages.="" there="" will="" be="" no="" changes="" in="" the="" manner="" of="" handling="" fuel="" discharged="" from="" the="" core="" as="" a="" result="" of="" refueling;="" administrative="" controls="" will="" continue="" to="" be="" used="" to="" specify="" fuel="" assembly="" placement="" requirements.="" the="" relative="" positions="" of="" region="" i="" and="" region="" ii="" storage="" locations="" will="" remain="" the="" same="" within="" the="" fuel="" pool.="" therefore,="" the="" probability="" of="" a="" fuel="" handling="" accident="" has="" not="" been="" significantly="" increased.="" the="" consequences="" of="" a="" fuel="" handling="" accident="" have="" been="" evaluated.="" the="" radioactive="" release="" consequences="" of="" a="" dropped="" fuel="" assembly="" are="" not="" affected="" by="" the="" proposed="" increase="" in="" fuel="" pool="" storage="" capacity.="" they="" remain="" bounded="" by="" the="" results="" of="" calculations="" performed="" to="" justify="" the="" existing="" st.="" lucie="" unit="" 2="" fuel="" storage="" racks="" and="" burnup="" limits.="" at="" the="" limiting="" fuel="" assembly="" burnup,="" radioactive="" releases="" from="" a="" dropped="" assembly="" would="" be="" only="" a="" small="" fraction="" of="" nrc="" guidelines.="" the="" input="" parameters="" employed="" in="" analyzing="" this="" event="" are="" consistent="" with="" the="" current="" values="" of="" fuel="" enrichment,="" discharge="" burnup="" and="" uranium="" content="" used="" at="" st.="" lucie="" unit="" 2="" and="" with="" future="" use="" of="" the="" ``value-added''="" fuel="" pellet="" design.="" thus,="" the="" consequences="" of="" the="" fuel="" assembly="" drop="" accident="" would="" not="" be="" significantly="" increased="" from="" those="" previously="" evaluated.="" the="" capability="" of="" the="" fuel="" pool="" cooling="" system="" to="" handle="" the="" increased="" number="" of="" discharged="" assemblies="" has="" been="" examined.="" the="" impact="" of="" a="" total="" loss="" of="" spent="" fuel="" pool="" cooling="" flow="" on="" available="" equipment="" recovery="" time="" and="" on="" fuel="" cladding="" integrity="" has="" also="" been="" evaluated.="" for="" the="" limiting="" full="" core="" discharge,="" sufficient="" time="" remains="" available="" to="" restore="" cooling="" flow="" or="" to="" provide="" an="" alternate="" makeup="" source="" before="" boiloff="" results="" in="" a="" fuel="" pool="" water="" level="" less="" than="" that="" needed="" to="" maintain="" acceptable="" radiation="" dose="" levels.="" analysis="" has="" shown="" that="" in="" the="" event="" of="" a="" total="" loss="" of="" fuel="" pool="" cooling="" fuel="" cladding="" integrity="" is="" maintained.="" therefore,="" the="" consequences="" of="" a="" loss="" of="" fuel="" pool="" cooling="" event,="" including="" the="" effect="" of="" the="" proposed="" increase="" in="" fuel="" pool="" storage="" capacity,="" have="" not="" been="" significantly="" increased="" from="" previously="" analyzed="" results="" for="" this="" type="" of="" accident.="" the="" analysis="" of="" record="" pertaining="" to="" the="" radiological="" consequences="" of="" the="" hypothetical="" drop="" of="" a="" loaded="" spent="" fuel="" cask="" just="" outside="" the="" fuel="" handling="" building="" was="" examined="" to="" determine="" the="" impact="" of="" the="" increased="" fuel="" storage="" capacity="" on="" this="" accident's="" results.="" the="" results="" of="" the="" previously="" performed="" analysis="" were="" determined="" to="" bound="" the="" conditions="" described="" by="" the="" proposed="" license="" amendment,="" thus="" the="" consequences="" of="" the="" cask="" drop="" accident="" would="" not="" be="" significantly="" increased="" as="" a="" result="" of="" this="" change.="" it="" is="" concluded="" that="" the="" proposed="" amendment="" to="" increase="" the="" storage="" capacity="" of="" the="" st.="" lucie="" unit="" 2="" spent="" fuel="" pool="" will="" not="" involve="" a="" significant="" increase="" in="" the="" probability="" or="" consequences="" of="" any="" accident="" previously="" evaluated.="" 2.="" the="" proposed="" amendment="" will="" not="" create="" the="" possibility="" of="" a="" new="" or="" different="" type="" of="" accident="" from="" any="" accident="" previously="" evaluated.="" in="" this="" license="" amendment="" fpl="" proposes="" to="" credit="" the="" negative="" reactivity="" associated="" with="" a="" portion="" of="" the="" soluble="" boron="" present="" in="" the="" spent="" fuel="" pool.="" soluble="" boron="" has="" always="" been="" present="" in="" the="" st.="" lucie="" unit="" 2="" spent="" fuel="" pool;="" as="" such="" the="" possibility="" of="" an="" inadvertent="" fuel="" pool="" dilution="" has="" always="" existed.="" however,="" the="" spent="" fuel="" pool="" dilution="" analysis="" demonstrates="" that="" a="" dilution="" of="" the="" unit="" 2="" spent="" fuel="" pool="" which="" could="" increase="" the="" pool="">eff to greater than 0.95
is not a credible event. Neither implementation of credit for the
reactivity of fuel pool soluble boron nor the proposed increase in the
fuel pool storage capacity will create the possibility of a new or
different type of accident at St. Lucie Unit 2.
An examination of the limiting fuel assembly misload has determined
that this would not represent a new or different type of accident. None
of the other accidents examined as a part of this license submittal
represent a new or different type of accident; each of these situations
has been previously analyzed and determined to produce acceptable
results.
The proposed license amendment will not result in any other changes
in the mode of spent fuel pool operation at St. Lucie Unit 2 or in the
method of handling irradiated nuclear fuel. The spatial relationship
between the fuel storage racks and the cask crane range of motion is
not affected by the proposed change.
As a result of the evaluation and supporting analyses, FPL has
determined that the proposed fuel pool capacity increase does not
create the possibility of a new or different type of accident from any
accident previously evaluated.
3. The proposed amendment will not involve a significant reduction
in the margin of safety.
FPL has determined, based on the nature of the proposed license
amendment that the issue of margin of safety, when applied to this fuel
pool capacity increase, should address the following areas:
(1) Fuel Pool reactivity considerations
(2) Fuel Pool boron dilution considerations
(3) Thermal-Hydraulic considerations
(4) Structural loading and seismic considerations
The Technical Specification changes proposed by this license
amendment, the proposed spent fuel pool storage configuration and the
existing Technical Specification limits on fuel pool soluble boron
concentration provide sufficient safety margin to ensure that the array
of fuel assemblies stored in the spent fuel pool will always remain
subcritical. The revised spent fuel storage configuration is based on a
Unit 2 specific criticality analysis performed using methodology
consistent with that approved by the NRC. Additionally, the soluble
boron concentration required by current Technical Specifications
ensures that the fuel pool keff will always be maintained
substantially less than 0.95.
The Unit 2 criticality analysis established that the
keff of the spent fuel pool storage racks will be less than
1.0 with no soluble boron in the fuel pool water, including the effect
of all uncertainties and tolerances. Credit for the soluble boron
actually present is used to offset uncertainties, tolerances, off-
normal conditions and to provide margin such that the spent fuel pool
keff is maintained less than or equal to 0.95. FPL has also
demonstrated that a decrease in the fuel pool boron concentration such
that keff exceeds 0.95 is not a credible event.
Current Technical Specifications require that the fuel pool boron
concentration be maintained greater than or equal to 1720 ppm. This
boron value is substantially in excess of the 520 ppm required by the
uncertainty and reactivity equivalencing analyses discussed in this
evaluation and the 1266 ppm value required to maintain keff
less than or equal to 0.95 in the presence of the most adverse
mispositioned fuel assembly.
The St. Lucie Unit 2 fuel pool boron concentration will continue to
be maintained significantly in excess of 1266 ppm; the proposed license
amendment will not result in changes in the mode of operation of the
refueling water tank (RWT) or in its use for makeup to the fuel pool.
Thus, operation of the spent fuel pool following the proposed change,
combined with the existing fuel pool boron concentration Technical
Specification limit of 1720 ppm, will continue to ensure that
keff of the fuel pool will be substantially less than 0.95.
[[Page 9604]]
Even if this not-credible dilution event was to occur, no radiation
would be released; the only consequence would be a reduction of
shutdown margin in the fuel pool. The volume of unborated water
required to dilute the fuel pool to a keff of 0.95 is so
large (in excess of 358,900 gallons to dilute the fuel pool to 520 ppm
boron) that only a limited number of water sources could be considered
potential dilution sources. The likelihood that this level of water use
could remain undetected by plant personnel is extremely remote.
In meeting the acceptance criteria for fuel pool reactivity, the
proposed amendment to increase the storage capacity of the existing
fuel pool racks does not involve a significant reduction in the margin
of safety for nuclear criticality.
Calculations of the spent fuel pool heat load with an increased
fuel pool inventory were performed using ANSI/ANS-5.1-1979 methodology.
This method was demonstrated to produce conservative results through
benchmarking to actual St. Lucie Unit 2 fuel pool conditions and by
comparison of its results to those generated by a calculation using
Auxiliary Systems Branch Technical Position 9-2 methodology.
Conservative methods were also used to demonstrate fuel cladding
integrity is maintained in the absence of cooling system forced flow.
The results of these calculations demonstrate that, for the limiting
case, the existing fuel pool cooling system can maintain fuel pool
conditions within acceptable limits with the increased inventory of
discharged assemblies. Therefore, the proposed change does not result
in a significant reduction in the margin of safety with respect to
thermal-hydraulic or spent fuel cooling considerations.
The primary safety function of the spent fuel pool and the fuel
storage racks is to maintain discharged fuel assemblies in a safe
configuration for all environments and abnormal loadings, such as an
earthquake, a loss of pool cooling or a drop of a spent fuel assembly
during routine spent fuel handling. The proposed increase in spent fuel
inventory on the fuel pool and the existing storage racks have been
evaluated and show that relevant criteria for fuel rack stresses and
floor loadings have been met and that there has been no significant
reduction in the margin of safety for these criteria.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: Indian River Junior College
Library, 3209 Virginia Avenue, Fort Pierce, Florida 34954-9003.
Attorney for licensee: M.S. Ross, Attorney, Florida Power & Light,
P.O. Box 14000, Juno Beach, Florida 33408-0420.
NRC Project Director: Frederick J. Hebdon.
Florida Power and Light Company, Docket Nos. 50-250 and 50-251, Turkey
Point Plant Units 3 and 4, Dade County, Florida
Date of amendment request: November 22, 1996, as revised and
replaced on February 2, 1998.
Description of amendment request: The licensee proposed to change
the Technical Specifications (TS) to allow the use of a temporary fuel
oil storage system for up to 10 days in order to perform a surveillance
requirement on the Unit 3 fuel oil storage tank with Unit 3 in Modes 5,
6, or defueled.
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:
Question 1 Does the proposed license amendment involve a
significant increase in the probability or consequences of an accident
previously evaluated?
The proposed amendment will allow the installation of a temporary
fuel oil storage and transfer system for up to 10 days, once every 10
years. EDGs [emergency diesel generators] are designed as backup AC
power sources for essential safety systems in the event of a loss of
offsite power. Since the EDGs are not accident initiators, the
probability of occurrence of accidents previously analyzed has not been
increased.
The temporary fuel oil storage tanks will be located greater than
fifty (50) feet from safety related or safe shutdown components or
circuits. This does not produce any threat to fire protection or safe
shutdown capability and therefore represents a configuration that is
bounded by existing fire hazards analysis.
The proposed amendment will not change the condition or minimum
amount of operating equipment assumed in the plant safety analyses for
accident mitigation. The temporary fuel storage and transfer system
provides a reliable means of performing the required delivery support
function for the Unit 3 EDGs.
An insignificant increase in the consequences of an accident
previously evaluated is possible since the temporary storage and
transfer system will not meet requirements for Seismic Category I or
Class 1E. However, the probability of a seismic event will be very low
due to the limited time that the temporary storage system will be in
use.
The increase in the consequences of an accident previously
evaluated is insignificant due to the following:
Manual actions required to provide a 7 day supply of fuel to the
EDGs can easily be accomplished in the 17 hours of EDG operation
provided by the 3880 gallon capacity of a single EDG day and skid tank.
The location of the temporary fuel oil supply inside the protected area
security fence by the Central Receiving Facility provides multiple
access routes to transfer fuel to the Unit 3 EDGs and is in close
proximity to a severe weather shelter for the mobile tanker.
Additionally, more than 17 hours will be available to manually
transfer fuel from the temporary fuel storage tanks located inside the
protected area, by filling the Unit 4 EDG storage tanks with
approximately 8600 gallons of fuel oil above that required for Unit 4
EDG operability. This extra capacity will be available to the Unit 3
EDGs prior to taking the permanent Unit 3 storage tank out of service.
This will be done by filling the Unit 4 fuel tanks to 39,000 gallons,
which is just below the high level alarm. This gives a capacity of 4300
gallons in each tank above the Unit 4 Technical Specification minimum
required volume of 34,700 gallons. The Unit 4 tanks are contained
within a Seismic Class 1 structure and protected by installed fire
protection equipment.
Combining the excess available fuel from the Unit 4 storage tanks
and the nominal volume of the Unit 3 day and skid tanks gives a total
of 12,480 gallons (4300 x 2+3880) of available fuel to either of the
Unit 3 EDGs. This allows a run time for a Unit 3 EDG of 55 hours
(assuming fuel oil transfer from Unit 4) prior to reaching the
Technical Specification minimum volume for the Unit 4 fuel oil storage
tanks. Manual actions to replenish the Unit 4 or Unit 3 fuel oil
storage tanks from the temporary storage tanks, via the mobile tanker,
can easily be accomplished within the 55 hours. Procedures currently
exist for the transfer of fuel from (1) the mobile tanker to the
auxiliary fill station at the Unit 3 EDGs, and (2) from the Unit 4 EDG
storage tanks to the Unit 3 day tanks by using either of the Unit 4
transfer pumps. The
[[Page 9605]]
Unit 4 transfer pumps are powered from redundant Class 1E power
supplies.
The temporary storage tanks will be located inside the protected
area in the vicinity of the Nuclear Plant Central Receiving Facility.
The temporary tanks will be located greater than fifty (50) feet from
safety related or safe shutdown components or circuits. This does not
produce any threat to fire protection or safe shutdown capability and
therefore represents a configuration that is bounded by existing fire
hazards analysis.
A dedicated mobile tanker staged inside the protected area to
transfer fuel from the temporary storage tanks to the permanent day/
skid tank system. The mobile tanker will have an integral transfer pump
to facilitate movement of fuel to either of the two truck fills at the
Unit 4 EDG building or day tank truck fills (auxiliary fill station) at
the Unit 3 EDGs. One truck fill at the Unit 4 EDG building supplies
fuel to the 4A and 4B storage tanks, the other truck fill at the Unit 4
EDG building can provide fuel directly to the Unit 3 day tanks. This
fuel supply will provide continued operation for 7 days. The temporary
storage and transfer system will not meet requirements for Seismic
Category I or Class 1E.
The capability to operate an Unit 3 EDG for 7 days during the tank
cleaning evolution will be assured by an approved plant procedure that
controls the following:
A minimum fuel supply of 3880 gallons from the Unit 3 day and
skid tank. This provides 17 hours of operation.
The extra fuel supply of 8600 gallons in the Unit 4 EDG tanks
which will be transferred by using one of the installed Unit 4 transfer
pumps. This provides an additional 38 hours of operation.
Three temporary tanks containing a minimum fuel supply of
38,000 gallons. This fuel supply will provide continued operation for 7
days.
Consequently, operation of the facility in accordance with the proposed
amendment would not involve a significant increase in the probability
or consequences of an accident previously evaluated.
Question 2 Does the proposed license amendment create the
possibility of a new or different kind of accident from any accident
previously evaluated?
The proposed amendment will not change the physical plant or modes
of plant operation defined in the Turkey Point Units 3 and 4 operating
license. The change will not involve addition or modification of
equipment for Unit 3 EDG fuel storage and transfer. The temporary fuel
supply system provides a reliable means of performing the required fuel
delivery support function for the Unit 3 EDGs.
Consequently, operation of either unit in accordance with the
proposed amendment would not create the possibility of a new or
different kind of accident from any accident previously evaluated.
Question 3 Does the proposed amendment involve a significant
reduction in the margin of safety?
The proposed amendment is designed to provide flexibility to
schedule and perform required surveillance activities. Surveillance
intervals or operating requirements are not changed by the proposal;
only the method of fuel oil storage on a temporary basis for a single
operable EDG is addressed. The proposed change will not alter the basis
for any Technical Specification that is related to the establishment
of, or maintenance of, a nuclear safety margin.
Consequently, operation of Turkey Point Units 3 and 4 in accordance
with this proposed amendment would not involve a significant reduction
in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: Florida International
University, University Park, Miami, Florida 33199.
Attorney for licensee: M.S. Ross, Attorney, Florida Power & Light,
P.O. Box 14000, Juno Beach, Florida 33408-0420.
NRC Project Director: Frederick J. Hebdon.
Florida Power and Light Company, Docket Nos. 50-250 and 50-251, Turkey
Point Plant Units 3 and 4, Dade County, Florida
Date of amendment request: January 9, 1998.
Description of amendment request: The licensee proposed to change
the Technical Specifications (TS) to allow the use of
ZIRLOtm fuel rod clad material.
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:
Question 1 Does the proposed license amendment involve a
significant increase in the probability or consequences of an accident
previously evaluated?
Implementation of ZIRLOtm fuel rod cladding will have no
impact on the probability or consequences of any Design Basis Event
occurrences which were previously evaluated. The determination that
fuel design limits are met will continue to be performed using NRC
approved fuel performance analysis methodology. Changing to
ZIRLOtm fuel rod cladding poses no significant increase in
the probability or consequences of any accident previously evaluated.
No new performance requirements are being imposed on any system or
component in order to support implementation of ZIRLOtm fuel
rod cladding. Since the LOCA and Non-LOCA analysis results will remain
within design limits, the inputs to the radiation dose analysis do not
change. Therefore, the consequences to the public resulting from any
accident previously evaluated in the Updated Final Safety Analysis
Report (UFSAR) is not increased.
Fuel rod design criteria will be evaluated every cycle to ensure
proper compliance with fuel rod design limits and therefore the UFSAR.
The evaluation of the fuel design against fuel design limits will be
performed in accordance with 10 CFR 50.59, which ensures that the
reload will not involve an increase in the probability or consequence
of an accident previously evaluated.
Question 2 Does the proposed license amendment create the
possibility of a new or different kind of accident from any accident
previously evaluated?
Implementation of ZIRLOtm fuel rod cladding will have no
impact, nor does it contribute in any way to the probability or
consequences of an accident.
No new accident scenarios, failure mechanisms or limiting single
failures are introduced as a result of using ZIRLOtm fuel
rod cladding. The institution of ZIRLOtm fuel rod cladding
will have no adverse effect on, and does not challenge the performance
of, any safety related system.
The determination that the fuel rod design limits are met will be
performed using NRC approved methodology. Therefore, the proposed
amendment does not in any way create the possibility of a new or
different kind of accident from any accident previously evaluated.
[[Page 9606]]
Question 3 Does the proposed amendment involve a significant
reduction in the margin of safety?
The margin of safety is not affected by the implementation of
ZIRLOtm fuel rod cladding. Use of ZIRLOtm fuel
rod cladding has been approved by the NRC and does not constitute a
significant reduction in the margin of safety.
The margin of safety provided in the fuel design limits is
acceptable and will be maintained and not reduced.
In addition, each future reload will involve a 10 CFR 50.59 review
to assure that operation of the units within the cycle specific limits
will not involve a reduction in the margin of safety. Therefore, the
proposed amendment 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 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: Florida International
University, University Park, Miami, Florida 33199.
Attorney for licensee: M.S. Ross, Attorney, Florida Power & Light,
P.O. Box 14000, Juno Beach, Florida 33408-0420.
NRC Project Director: Frederick J. Hebdon.
Northeast Nuclear Energy Company (NNECO), et al., Docket No. 50-423,
Millstone Nuclear Power Station, Unit No. 3, New London County,
Connecticut
Date of amendment request: January 22, 1998.
Description of amendment request: The amendment would incorporate
the proposed revision into Chapter 9 of the Millstone Unit 3 Final
Safety Analysis Report. The proposed revision to the Millstone Unit 3
licensing basis would accept the existing use of epoxy coatings on
safety-related components.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
NNECO has reviewed the proposed revision in accordance with
10CFR50.92 and has concluded that the revision does not involve a
significant hazards consideration (SHC). The basis for this conclusion
is that the three criteria of 10CFR50.92(c) are not satisfied. The
proposed revision does not involve [an] SHC because the revision would
not:
1. Involve a significant increase in the probability or consequence
of an accident previously evaluated.
Past experience indicates that failure of previous ARCOR
applications may have degraded the performance of SWS [service water
system] heat exchangers within one train, but there is no indication
that failure of multiple heat exchangers on both trains is feasible.
Furthermore, the likelihood of ARCOR material being released has been
reduced by improving the application procedure and performing
destructive testing to detect disbondment. In addition, the completion
of normal heat exchanger performance surveillance's and periodic visual
inspections minimizes the potential for disbonded ARCOR to degrade SWS
components.
Therefore, the presence of ARCOR coating material within the SWS
does not involve a significant increase in the probability or
consequence of an accident previously evaluated.
2. Create the possibility of a new or different kind of accident
from any accident previously evaluated.
The application of ARCOR material may lead to the degradation of
SWS heat exchangers. However, multiple ARCOR application failures
occurring simultaneously either instantaneously or gradually resulting
in failure of all SWS heat exchangers in both trains is not considered
feasible. An instantaneous failure is discounted by analysis which
concludes that normal system operations are more likely to cause the
release of degraded ARCOR than what might be expected following a
seismic event. Gradual degradation is not expected since normal SWS
heat exchanger performance surveillance's will identify heat exchanger
tubesheet fouling and thus, provide early detection of coating
failures. Therefore, the use of ARCOR coating material within the SWS
does 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.
Although the gradual release of ARCOR material creates the
potential to simultaneously degrade the performance of mitigating
equipment in both trains of safety systems, it is determined to be
unrealistic due to normal heat exchanger performance surveillance's.
These surveillance's are expected to identify heat exchanger tubesheet
fouling and provide early detection and mitigation of a problem with
the pipe coatings. Therefore, the application of ARCOR coating within
the SWS does not involve a significant reduction in the margin of
safety.
In conclusion, based on the information provided, it is determined
that the proposed revision does not involve an SHC.
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.
Northern States Power Company, Docket No. 50-263, Monticello Nuclear
Generating Plant, Wright County, Minnesota
Date of amendment request: July 26, 1996, as supplemented September
5 and December 4, 1997.
Description of amendment request: The proposed amendment would, as
part of the licensee's power rerate program, increase the maximum power
level to 1775 megawatts thermal (MWt). This change is approximately 6.3
percent above the current maximum power level of 1670 MWt.
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:
A. The proposed amendment will not involve a significant increase
In the probability or consequences of an accident previously evaluated.
The probability of occurrence and consequences of an [accident]
previously evaluated have been evaluated for MNGP [Monticello Nuclear
Generating Plant] Power Rerate. This evaluation has concluded that MNGP
Power Rerate will not involve a significant increase in the probability
of occurrence or consequences of previously evaluated accidents.
1. Evaluation of Accident Consequences
(a) ECCS-LOCA Analysis
The Emergency Core Cooling System Loss of Coolant Accident (ECCS-
LOCA)
[[Page 9607]]
performance analysis has been evaluated for MNGP Power Rerate using
methodology which has been approved by the NRC for LOCA 10CFR50.46
analyses [requirements]. The current ECCS performance requirements were
used in the power rerate analysis; no further parameter relaxations
were included in the analysis. The ECCS-LOCA analysis was performed for
MNGP Power Rerate for the existing licensed rated thermal power and at
a bounding thermal power level of 1880 MWt that is approximately 6%
greater than the proposed power rerate to 1775 MWt [megawatts thermal].
In addition, the bounding thermal power level was increased by an
additional 2% in accordance with regulatory guidance. The licensing
peak clad temperature for the bounding analyzed thermal power level
remains below the 10CFR50.46 required limit of 2,200'F. Therefore the
analysis demonstrates that MNGP will continue to comply with 10CFR50.46
and 10CFR50, Appendix K at rerated conditions thus the consequences of
a LOCA is not significantly increased for the proposed power rerate.
(b) Abnormal Operating Transient Analysis
An evaluation of the Updated Safety Analysis Report (USAR) and
reload transients has been performed for MNGP Power Rerate to
demonstrate that the proposed power rerate has no adverse effect on
plant safety. This evaluation was performed for a power level of 1775
MWt, with the exception that certain event evaluations were performed
at 102% of the rerate power level. The transient analysis performed to
demonstrate the acceptability of MNGP Power Rerate used the NRC
approved methods identified in the MNGP Technical Specifications.
The limiting transient events at the power rerate conditions have
been analyzed. This includes all events that establish the core thermal
operating limits and the events that bound other transient acceptance
criteria. These limiting transients were benchmarked against the
existing rated thermal power level by performance of the event analysis
at both the proposed rerate power level and the existing rated power
level. In addition, an expanded group of transient events was evaluated
to confirm that these events were less severe with the power rerate
than the most limiting transients. The events included in the expanded
group of transient events were chosen based on those events which have
been demonstrated to be sensitive to initial power level. This
evaluation confirmed that the existing set of limiting transient events
remains valid for MNGP Power Rerate. The evaluation was performed for a
representative core and demonstrated the overall capability to meet all
transient safety criteria for the power rerate. Cycle specific analysis
will continue to be performed for each fuel reload to demonstrate
compliance with the applicable transient criteria and to establish
cycle specific operating limits.
The results of the evaluation of transients demonstrate that the
power rerate can be accomplished without a significant increase in the
consequences of the transients evaluated. The fuel thermal-mechanical
limits at the power rerate conditions are within the specific design
criteria for the GE [General Electric] fuels currently loaded in the
MNGP core. Also, the power-dependent and flow-dependent MCPR [minimum
critical power ratio] and Maximum Average Planar Linear Heat Generation
Rate (MAPLHGR) methods developed as part of the core performance
improvement program remain applicable to rerate conditions. The
transient event evaluation confirmed that MNGP Power Rerate has no
significant effect on the power-dependent and flow-dependent MCPR and
MAPLHGR limits. The peak reactor pressure vessel bottom head pressure
remains within the ASME [American Society of Mechanical Engineers]
requirement for reactor pressure vessel overpressure protection.
The effects of plant transients were evaluated by assessing a
number of disturbances of process variables and malfunctions or
failures of equipment consistent with USAR. The transient events were
evaluated against the Safety Limit Minimum Critical Power Ratio,
(SLMCPR). The SLMCPR is determined using NRC-approved methods. The
limiting transient events are slightly more severe when initiated from
the rerate power level. The power rerate transient evaluation results
show a slightly more limiting event initial CPR [critical power ratio]
(less than or equal to 0.02) than that initiated from the present rated
power level for the near limiting transients. However, for the most
limiting transient, the evaluation of a representative core showed that
no change is required to the Operating Limit MCPR for the power rerate
and that the integrity of the SLMCPR is maintained. The margin of
safety established by the SLMCPR is not affected and the event
consequences are not significantly affected by the proposed power
rerate to 1775 MWt. Cycle specific analysis will continue to be
performed for each fuel reload to demonstrate compliance with the
applicable transient criteria and to establish cycle specific operating
limits.
The results demonstrate that the MNGP core thermal power output can
be safely increased to the power rerate level without significant
effect on the consequences of previously evaluated postulated transient
events. The results of the rerate transient analysis are summarized as
follows.
(1) Events Resulting in a Nuclear System Pressure Increase
(a) Main Generator Load Rejection with No Steam Bypass
At rerated conditions, the fuel transient thermal and mechanical
overpower results remain below the NRC accepted design criteria.
(b) Main Turbine Trip with No Steam Bypass
At rerate conditions, the fuel transient thermal and mechanical
overpower results remain below the NRC accepted design criteria.
(c) Main Steam Isolation Valve Closure, Flux Scram
The peak reactor pressure vessel bottom head pressure for rerate
conditions is slightly higher than the reactor pressure vessel bottom
head pressure at current conditions. However, the resultant pressure is
still below the ASME overpressure limit of 1,375 psig [pounds per
square inch].
(d) Slow Closure of a Single Turbine Control Valve
The results of this transient for the power rerate remain non-
limiting as compared with other more severe pressurization events.
(2) Event Resulting in a Reactor Vessel Water Temperature Decrease
(a) Feedwater Controller Failure-Maximum Demand
The delta CPR calculated for this event at rerate conditions is
about 0.01 higher than the corresponding value for the current rated
power when the impact of the new condensate pumps is factored in. The
trend for the Feedwater Controller Failure-Maximum Demand event is
consistent with the analysis for the current rated power. The fuel
thermal margin results are within the acceptable limits for the fuel
types analyzed.
(b) Loss of Feedwater Heating
This event at the rerate conditions remains significantly less than
the cycle operating MCPR limit. The results at low core flow conditions
are actually slightly higher than for the high core
[[Page 9608]]
flow condition because of increased inlet coolant subcooling into the
reactor core. The calculated thermal and mechanical overpower limits at
the power rerate conditions for this event also meet the fuel design
criteria.
(c) Inadvertent HPCI [high-pressure coolant injection] Actuation
For the limiting condition analyzed, both the high water level
setpoint and the high reactor pressure vessel steam dome pressure scram
setpoints are not reached. Based on the peak average fuel surface heat
flux results, the HPCI actuation event will be bounded by the limiting
pressurization event with respect to delta Critical Power Ratio
([delta] CPR) considerations. In addition, the fuel transient thermal
and mechanical overpower limits remain within the NRC accepted design
values.
(3) Event Resulting in a Positive Reactivity Insertion
(a) Rod Withdrawal Error (RWE)
The current Rod Block Monitor (RBM) system for MNGP with power
dependent setpoints was analyzed for the rod withdrawal error event at
the power rerate conditions using a statistical approach consistent
with NRC approved methods. The analysis concluded that the transient is
slightly more severe with a greater delta Critical Power Ratio ([delta]
CPR) from the initial most limiting CPR. However, the fuel and
mechanical overpower results remain within the NRC accepted design
criteria.
(4) Event Resulting in a Reactor Vessel Coolant Inventory Decrease
(a) Pressure Regulator Failure to Full Open
The results of this transient for the power rerate remain non-
limiting as compared with other more severe pressurization events.
(b) Loss of Feedwater Flow
This transient event does not pose any direct threat to the fuel in
terms of a power increase from the initial conditions. Water level
declines rapidly and a low level causes a reactor scram. The closure of
the main steam isolation valves and the actuation of High Pressure
Coolant Injection and Reactor Core Isolation Cooling terminate the
event. This event was included in the power rerate evaluation to
provide assurance that sufficient water makeup capability is available
to keep the core covered when all normal feedwater is lost. The generic
analysis performed in support of the extended power uprate program
shows that at the power rerate conditions a large amount of water
remains above the top of the active fuel. These sequences of events do
not require any new operator actions or shorter operator response
times. Therefore, the operator actions for the event do not
significantly change for the power rerate.
(5) Event Resulting in a Core Coolant Flow Decrease
(a) Recirculation Pump Seizure
The recirculation pump seizure assumes instantaneous stoppage of
the pump motor shaft of one recirculation pump. As a result, the core
flow decreases rapidly. The heat flux decline lags core power and flow
and could result in a degradation of core heat transfer. At the power
rerate conditions, the transient results confirmed that the
consequences of the pump seizure event remain non-limiting.
(6) Event Resulting in a Core Coolant Flow Increase
(a) Recirculation Flow Controller Failure Increasing Flow
The results of this transient for the power rerate remain non-
limiting as compared with other more severe pressurization events.
(c) Design Basis Accident Challenges to the Containment
The primary containment response to the limiting design basis
accident was evaluated for a bounding reactor power level approximately
6% greater than the proposed power rerate to 1775 MWt. In addition, the
bounding reactor power level was increased by an additional 2% in
accordance with regulatory guidance. The effect of the power rerate on
the short term containment response (peak values) as well as the long
term containment response for containment pressure and temperature
confirms the suitability of the plant for operation at the bounding
power level, thus the proposed power rerate to 1775 MWt is acceptable.
Factors of safety provided in the ASME Code are maintained and safety
margin is not affected for the power rerate to 1775 MWt.
Short-term containment response analyses were performed for the
limiting design basis LOCA consisting of a double-ended guillotine
break of a recirculation suction line, to demonstrate that operation at
a bounding reactor power will not result in exceeding the containment
design limits. This limiting design basis LOCA event results in the
highest short-term containment pressures and dynamic loads. The
analysis determined that for a bounding reactor power the maximum
drywell pressure values are bounded by the current USAR analysis value
and by the containment design pressure. The power rerate to 1775 MWt
has no adverse effect on the containment structural design pressure.
Because there will be more residual heat with increased thermal
power, the containment long term response will have slightly higher
temperatures. Long term suppression chamber temperatures remain within
the design temperature of the structure, thus factors of safety
provided in the ASME code are maintained and safety margin is not
affected. Analysis confirmed that ECCS pump NPSH is adequate for this
temperature response. It was confirmed that the long term response does
not adversely affect the containment structure or the environmental
qualification (EQ) of equipment located in the drywell or suppression
chamber room. The drywell long term temperature response is not
adversely affected for a bounding reactor power. An analytical power
level of 1880 MWt bounds the decay heat associated with the 1775 MWt
power level with a one sided confidence interval of 95%. The
containment long term response is therefore acceptable for the power
rerate to 1775 MWt.
The impact of a reactor power increase on the containment dynamic
loads have been determined, evaluated and found to have no adverse
effects for conditions which well bound the proposed power rerate. Thus
the containment dynamic loads were found to be acceptable for the power
rerate to 1775 MWt.
The MNGP Power Rerate evaluation of the primary containment
response to the design basis accident confirmed that the power rerate
does not result in a significant increase in consequences for a
bounding reactor power approximately 6% greater than the proposed power
rerate to 1775 MWt.
(d) Radiological Consequences of Design Basis Accidents
For MNGP Power Rerate, the radiological consequences of the
limiting design basis accidents were re-evaluated. These evaluations
included the effect of the power rerate on the radiological
consequences of accidents presented in USAR Section 14.7.
This evaluation was performed using inputs and evaluation
techniques consistent with the current regulatory guidance, the current
GE analysis methods, and the appropriate plant design basis. The inputs
and analysis methods used for MNGP Power Rerate differ from those
utilized in the current licensing basis evaluation presented in
[[Page 9609]]
the USAR and the AEC [Atomic Energy Commission] safety evaluation
supporting plant initial licensing. The MNGP Power Rerate evaluations
used the more contemporary staff approved methods. The inputs used in
the MNGP Power Rerate evaluation provide a conservative assessment of
the potential radiological consequences. The conclusions of these
evaluations are consistent with the original licensing basis
evaluations. The radiological consequences of the limiting design basis
accidents remain well within 10CFR100 guidelines for a bounding thermal
power approximately 6% greater than the proposed power rerate of 1775
MWt. In addition the bounding thermal power level was increased by an
additional 2% in accordance with regulatory guidance.
To conservatively analyze the change in consequences, the
evaluation of radiological consequences using the analysis inputs and
methods was performed for the existing licensed rated thermal power and
a thermal power bounding the proposed power rerate. This provides a
conservative bounding change in consequences for the requested power
rerate to 1775 MWt.
The MNGP Power Rerate evaluation of the radiological consequences
of design basis accidents confirmed that the power rerate does not
result in a significant increase in consequences for a bounding power
level approximately 6% greater than the proposed power rerate. The
results remain below the 10CFR100 guideline values as well as the
licensing basis established in the March 18, 1970 AEC safety
evaluation. Therefore, the postulated radiological consequences do not
represent a significant change in accident consequences and are clearly
within the regulatory guidelines for the proposed power rerate to 1775
MWt.
(e) Other Evaluations
(1) Performance Improvements
The MNGP Power Rerate safety analysis has been performed taking
into account the implementation of the following previously approved
special operational features.
(a) Maximum Extended Load Line Limit/Increase Core Flow (MELLL/ICF)
The safety analysis for rerate conditions shows that the extended
operating domain as analyzed by MELLL/ICF remains valid for the power
rerate conditions.
(b) Average Power Range Monitor/Rod Block Monitor Technical
Specification (ARTS) Improvements
The safety analysis for rerate conditions shows that the ARTS
improvements remain valid for the power rerate conditions.
(c) Single Loop Operation (SLO)
The safety analysis for rerate conditions shows that the single
loop operating mode remains valid for the power rerate conditions. The
MELLLA trip setpoints determined for two-loop operation were confirmed
to be acceptable for single loop operation with a correction applied to
account for the actual effective drive flow applied when operating in
single loop. The single loop settings have been conservatively
established to be consistent with the two loop settings while ensuring
the appropriate corrections are applied to the MAPLHGR and the
operating limit MCPR to account for single loop operation.
(2) Effect of Power Rerate on Support Systems
An evaluation was performed to address the effect of MNGP Power
Rerate on accident mitigation features, structures, systems, and
components within the balance of plant. The results are as follows:
Auxiliary systems such as, building heating, Ventilation and Air
Conditioning (HVAC) systems, reactor building closed cooling water,
service water and emergency service water, spent fuel pool cooling,
process auxiliaries such as instrument air and makeup water and the
post-accident sampling system were confirmed to operate acceptably
under normal and accident conditions at rerate conditions.
The secondary containment and standby gas treatment system were
confirmed to be able to adequately contain, process, and control the
release of normal and post-accident levels of radioactivity at rerate
conditions.
Instrumentation was reviewed and confirmed to be capable of
performing its control and monitoring functions under rerate
conditions. As required, analyses were performed to determine the need
for setpoint changes for various functions (e.g., APRM [average power
range monitor] neutron flux scram setpoints). In general, setpoints are
to be changed only to maintain adequate difference between plant
operating parameters and trip setpoints, while ensuring safety
performance is demonstrated. The revised setpoints have been
established using the NRC reviewed methodology as guidance.
Electric power systems including the turbine generator and
switchgear components were verified as being capable of providing the
electrical load as a result of the rerate power levels. An evaluation
of the auxiliary power system for the power rerate conditions confirmed
that the system has sufficient capacity with the changes identified in
Exhibit I [of the 12/4/97 submittal] to support all required loads for
safe shutdown, to maintain a safe shutdown condition, and to operate
the required engineered safeguards equipment following postulated
accidents. No safety-related electrical loads were affected which would
adversely impact the emergency diesel generators.
Piping systems were evaluated for the effect of operation at higher
power levels, including transient loading. The evaluation confirmed
that, with few exceptions, piping and supports are adequate to
accommodate the increased loading resulting from operation at rerate
power conditions. In a few cases, piping supports will be modified to
accept higher forces due to rerate conditions.
The effect of rerate conditions on high energy line break (HELB)
was evaluated. The evaluation confirmed structures, systems, and
components important to safety are capable of accommodating the effects
of jet impingement and blowdown forces and the environmental effects
resulting from HELB events at rerate conditions.
Control room habitability was evaluated. With the implementation of
minor hardware and non-hardware changes to the control room ventilation
system, Post-accident Control Room and Technical Support Center doses
at rerate conditions were confirmed to be within the guidelines of
General Design Criterion 19 of 10CFR50, Appendix A.
The environmental qualification of equipment important to safety
was evaluated for the effect on normal and accident operating
conditions at rerate power levels. The equipment remains qualified for
the new conditions. Minor adjustments will reflect some changes to
maintenance frequencies. The preventative maintenance program will
continue to provide for equipment maintenance or replacement to ensure
equipment environmental qualification at rerate power conditions.
(3) Effect on Special Events
The consequences of special events (i.e., ATWS [anticipated
transient without scram], 10CFR50, Appendix R, and Station Blackout)
remain within NRC accepted criteria for rerate conditions. Concurrent
malfunctions assumed to occur during accidents have
[[Page 9610]]
been accounted for in the safety analyses for rerate conditions. The
consequences of these equipment malfunctions does not change with
implementation of the MNGP Power Rerate program. The generic ATWS
analysis for operation at rerate conditions is being revised. The
revision is not expected to affect MNGP compliance with NRC acceptance
criteria.
(f) Conclusion
The evaluation of the Emergency Core Cooling System performance has
demonstrated the criteria of 10CFR50.46 are satisfied, thus the margin
of safety established by the criteria is maintained. The analysis
demonstrated that the ECCS will function with the most limiting single
failure to mitigate the consequences of the accidents and maintain fuel
integrity. The system will continue to perform as required under rerate
conditions to mitigate the consequences of accidents and thus the power
rerate does not adversely affect ECCS performance in a manner to
increase the severity of consequences. Challenges to the containment
have been evaluated and the integrity of the fission product barrier
has been confirmed. The radiological consequences of design basis
accidents have been evaluated and it was found that the effect of the
proposed power rerate on postulated radiological consequences does not
result in a significant increase in accident consequences. These
evaluations have been performed for a bounding reactor power
approximately 6% greater than the proposed power rerate. In addition
the bounding reactor power level was increased by an additional 2% in
accordance with regulatory guidance. Thus the evaluations provide
conservative bounding results for the proposed power rerate to 1775 MWt
and demonstrate that the proposed power rerate does not result in
significant increase in accident consequences.
The abnormal transients have been analyzed under the power rerate
conditions, and the analysis has confirmed that the power rerate to
1775 MWt has only a minor effect on the minimum critical power ratio
and that no change to the safety limit critical power ratio results,
thus the margin of safety as assured by the safety limit critical power
ratio is maintained. The effect of the power rerate on the consequences
of abnormal transients which result from potential component
malfunctions has been shown to be acceptable, thus the power rerate
does not result in a significant increase in transient event
consequences.
The spectrum of analyzed postulated accidents and transients has
been investigated, and has been determined to meet the current
regulatory criteria for the MNGP at rerate conditions. In the area of
core design, the fuel operating limits will still be met at the rerate
power level, and fuel reload analyses will show plant transients meet
the criteria accepted by the NRC as specified in the plant Technical
Specifications. The evaluation of transient and accident consequences
was performed consistent with the proposed changes to the plant
Technical Specifications. Therefore, the proposed Operating License and
Technical Specification changes will not cause a significant increase
in the consequences of an accident previously evaluated for the
Monticello plant.
2. Evaluation of the Probability of Previously Evaluated Accidents
The proposed power rerate imposes only minor increases in the plant
operating conditions. No changes are required to the rated core flow,
rated reactor pressure, or turbine throttle pressure. The power rerate
will result in moderate flow increases in those system[s] associated
with the turbine cycle (i.e., condensate, feedwater, main steam, etc.).
For MNGP Power Rerate, the small increase in operating temperatures for
balance of plant support systems has no significant effect on LOCA or
other accident probabilities.
The increase in flow rates in balance of plant systems is
addressed by compliance with NRC Generic Letter 89-08, ``Erosion/
Corrosion in Piping.'' The MNGP Power Rerate evaluations have confirmed
that the power rerate has no significant effect on flow induced
erosion/corrosion. The worst case limiting feedwater and main steam
piping flow increases were evaluated to be approximately proportional
to the power increase. The affected systems are currently monitored by
the MNGP Erosion/Corrosion program. Continued monitoring of the systems
provides a high level of confidence in the integrity of potentially
susceptible high energy piping systems.
The occurrence frequency of accident precursors and transients
[has] been addressed when required by applying the guidance of NRC
reviewed setpoint methodology to insure that acceptable trip avoidance
is provided during operational transients subsequent to implementation
of rerate. The setpoint evaluation has confirmed that MNGP Power Rerate
does not result in any increase in challenges to the plant protective
instrumentation.
Plant systems, components, and structures have been verified to be
capable of performing their intended functions under rerate conditions
with a few minor exceptions. Where necessary, some components will be
modified prior to implementation of the MNGP Power Rerate Program to
accommodate the revised operating conditions (e.g., a limited number of
pipe supports changes, instrumentation setpoint changes, control room
habitability improvements). MNGP Power Rerate does not significantly
affect the reliability of plant equipment. Where reliability effects
have been identified, modifications and administrative controls will be
implemented prior to the power rerate to adequately compensate. No new
components or system interactions that could lead to an increase in
accident probability are created due to the power rerate.
The probability (i.e., frequency of occurrence) of design basis
accidents occurring is not affected by the increased power level, as
the applicable criteria established for plant equipment (e.g., ANSI
Standard B31.1, ASME Code,) will still be followed as the plant is
operated at the rerate power level. The MNGP Power Rerate analysis
basis assures that the power dependent margin prescribed by the Code of
Federal Regulations (CFR) will be maintained by meeting the appropriate
regulatory criteria. Similarly, factors of safety specified by
application of the Code design rules have been demonstrated to be
maintained, as have other margin-assuring acceptance criteria used to
judge the acceptability of the plant. Reactor scram setpoints as
established are such that there is no significant increase in scram
frequency due to rerate conditions. No new challenges to safety-related
equipment will result from the power rerate. Therefore, the proposed
Operating License and Technical Specifications changes do not involve a
significant increase in the probability of an accident previously
evaluated.
B. The proposed Operating License changes do not create the
possibility of a new or different kind of accident from any accident
previously evaluated.
The basic Boiling Water Reactor configuration, operation and event
response is unchanged by the power rerate. Analysis of transient events
has confirmed that the same transients remain limiting and that no
transient events result in a new sequence of events which could lead to
a new accident scenario. The MNGP Power Rerate analyses confirmed that
the accident progression is basically unchanged by the power rerate.
[[Page 9611]]
An increase in power level will not create a new fission product
release path, or result in a new fission product barrier failure mode.
The same fission product barriers such as the fuel cladding, the
reactor coolant pressure boundary and the reactor containment, remain
in place. Fuel rod cladding integrity is ensured by operating within
thermal, mechanical, and exposure design limits and is demonstrated by
the MNGP Power Rerate transient analysis and accident analysis.
Similarly, analysis of the reactor coolant pressure boundary and
primary containment have demonstrated that the power rerate has no
adverse effect on these fission product barriers. The proposed changes
to the plant Technical Specifications to support the power rerate
implementation are consistent with the MNGP Power Rerate analyses and
assure transient and accident mitigation capability in compliance with
regulatory requirements.
The effect of MNGP Power Rerate on plant equipment has been
evaluated. No new operating mode, safety-related equipment lineup,
accident scenario, or equipment failure mode resulting from the power
rerate was identified. The full spectrum of accident considerations
defined in the USAR have been evaluated and no new or different kind of
accident resulting from the power rerate has been identified. MNGP
Power Rerate uses already developed technology and applies it within
the capabilities of already existing plant equipment in accordance with
presently existing regulatory criteria which includes accepted codes,
standards, and methods. GE has designed BWRs of higher power levels
than the rerate power of any of the currently operating BWR fleet and
no new power dependent accidents have been identified. In addition,
MNGP Power Rerate does not create any new sequence of events or failure
modes that lead to a new type of accident.
All actions to ensure that safety-related structures, systems, and
components will remain within their design allowable values and ensure
they can perform their intended functions under rerate conditions will
be taken prior to implementation of the power rerate. MNGP Power Rerate
does not increase challenges to or create any new challenge to safety-
related equipment or other equipment whose failure could cause an
accident. Plant modifications required to support implementation of
MNGP Power Rerate will be made to existing systems (e.g., a limited
number of pipe supports, instrumentation setpoints, control room
habitability improvements), rather than by adding new systems of a
different design which might introduce new failure modes or accident
sequences. The Technical Specification changes required to implement
the power rerate require little change to the plant's configuration,
and all changes have been evaluated and are acceptable.
Therefore, the proposed Operating License and Technical
Specification changes do not create the possibility of a new or
different kind of accident from any previously evaluated.
C. The proposed Operating License changes do not involve a
significant reduction in a margin of safety.
The accident analysis, as well as a majority of the plant specific
evaluations performed in support of MNGP Power Rerate have been
performed assuming a bounding steady state power level 112.6% of the
existing licensed limit of 1670 MWt, and approximately 6% above the
licensed maximum thermal power level of 1775 MWt proposed by MNGP Power
Rerate. In addition, the bounding reactor power level was increased by
an additional 2% in accordance with regulatory guidance when applicable
for the evaluation of accidents and transients. For plant conditions
associated with a bounding analysis power level, the analyses
demonstrated operating margin to criteria establishing margins of
safety, thus additional operating margin is demonstrated and assured
for the proposed power rerate to 1775 MWt and added confidence is
established in the integrity of criteria establishing margin to safety.
The cycle specific transient analysis, as well as the analysis to
establish plant instrumentation set points have been performed assuming
a plant steady state power level of 1775 MWt. This analysis approach
was taken in order to demonstrate safety and equipment margins while
ensuring appropriate cycle specific operating limits. The evaluation of
transient events and instrument setpoints demonstrated operating margin
to criteria establishing margins of safety for the proposed power
rerate conditions.
The MNGP Power Rerate analysis basis assures that the power
dependent safety margin assuring criteria prescribed by the Code of
Federal Regulations (CFR) will be maintained by meeting the appropriate
regulatory criteria. Similarly, factors of safety specified by
application of the code design rules have been maintained, as have
other margin-assuring acceptance criteria used to judge the
acceptability of the plant.
1. Fuel Thermal Limits
No change is required in the basic fuel design to achieve the
rerate power levels or to maintain the margins as discussed above. No
increase in the allowable peak bundle power is requested for the power
rerate. The abnormal transients have been evaluated under the power
rerate conditions for a representative core configuration. The analysis
has confirmed that the power rerate has no adverse effect on the
operating limit Minimum Critical Power Ratio (MCPR) and that no change
to the safety limit MCPR results, thus the margin of safety as assured
by the safety limit MCPR is maintained. The fuel operating limits such
as Maximum Average Planar Linear Heat Generation Rate (MAPLHGR) and the
operating limit MCPR will still be met at the rerate power level. The
MNGP Power Rerate analyses have confirmed the acceptability of these
operating limits for the power rerate without an adverse effect on
margins to safety. Cycle specific analysis will continue to be
performed for each fuel reload to demonstrate compliance with the
applicable transient criteria and to establish cycle specific operating
limits.
2. Design Basis Accidents Challenges to Fuel
The evaluation of the Emergency Core Cooling System performance has
demonstrated the criteria of 10CFR50.46 are satisfied, thus the margin
of safety established by the criteria is maintained. This evaluation
was performed for a bounding reactor power level approximately 6%
greater than the proposed power rerate. In addition the bounding
reactor power level was increased by an additional 2% in accordance
with regulatory guidance. The analysis demonstrates that MNGP will
continue to comply [with] the 10 CFR 50.46 at the rerate conditions and
that the margin of safety established by the regulation is maintained
for the proposed power rerate.
3. Design Basis Accident Challenges to Containment
The primary containment response to the limiting design basis
accident was evaluated for a bounding reactor power level approximately
6% greater than the proposed power rerate to 1775 MWt. In addition, the
bounding reactor power level was increased by an additional 2% in
accordance with regulatory guidance. The effect of the power rerate on
the short term containment response (peak values) as well as the long
term containment response for containment pressure and temperature
confirms the
[[Page 9612]]
suitability of the plant for operation at the bounding power level,
thus the proposed power rerate to 1775 MWt is acceptable. Factors of
safety provided in the ASME Code are maintained and safety margin is
not affected for the power rerate to 1775 MWt.
Short-term containment response analyses were performed for the
limiting design basis LOCA consisting of a double-ended guillotine
break of a recirculation suction line, to demonstrate that operation at
a bounding reactor power will not result in exceeding the containment
design limits. The analysis determined that for a bounding reactor
power the maximum drywell pressure values are bounded by the current
USAR analysis value and by the containment design pressure. The power
rerate to 1775 MWt has no adverse effect on the containment structural
design pressure.
Long term suppression chamber temperatures remain within the design
temperature of the structure, thus factors of safety provided in the
ASME code are maintained and safety margin is not affected. An
analytical power level of 1880 MWt bounds the decay heat associated
with the 1775 MWt power level with a one sided confidence interval of
95%. Analysis confirmed that ECCS pump NPSH is not adversely affected
with this temperature response. It was confirmed that the long term
response does not significantly affect the containment structure or the
environmental qualification (EQ) of equipment located in the drywell or
suppression chamber room.
The impact of a reactor power increase on the containment dynamic
loads [has] been determined, evaluated and found to have no adverse
effects for conditions which well bound the proposed power rerate. Thus
the containment dynamic loads were found to be acceptable for the power
rerate to 1775 MWt.
The MNGP Power Rerate evaluation of the primary containment
response to the design basis accident confirmed that the power rerate
does not result in a reduction in margins of safety for a bounding
reactor power approximately 6% greater than the proposed power rerate
to 1775 MWt.
4. Design Basis Accident Radiological Consequences
The Updated Safety Analysis Report (USAR) provides the radiological
consequences for each of the design basis accidents. The magnitude of
the potential consequences is dependent upon the quantity of fission
products released to the environment, the atmospheric dispersion
factors and the dose exposure pathways. For power rerate, the
atmospheric dispersion factors and the dose exposure pathways do not
change. Therefore, the only factor which will influence the magnitude
of the consequences is the quantity of activity released to the
environment. This quantity is a product of the activity released from
the core and the transport mechanisms between the core and the effluent
release point.
The radiological consequences of design basis accidents have been
evaluated, and it was found that the consequences did not result in a
significant increase in consequences for a bounding reactor power level
approximately 6% greater than the proposed power rerate. In addition,
the bounding reactor power level was increased by an additional 2% in
accordance with regulatory guidance. The results remain below the
10CFR100 guideline values as well as the licensing basis established in
the March 18, 1970 AEC safety evaluation. Therefore, the postulated
radiological consequences are clearly within the regulatory guidelines
and all radiological safety margins are maintained for the power rerate
to 1775 MWt.
5. Transient Evaluations
The effects of plant transients were evaluated by assessing a
number of disturbances of process variables and malfunctions or
failures of equipment consistent with USAR. The transient events were
evaluated against the Safety Limit Minimum Critical Power Ratio,
(SLMCPR). The SLMCPR is determined using NRC-approved methods. The
Power Rerate transient analyses were performed using the approved
methodology specified in the plant Technical Specifications. The
limiting transient events are slightly more severe when initiated from
the rerate power level. The power rerate transient evaluation results
show a slightly more limiting transient initial CPR (less than or equal
to 0.02) than that initiated from the present rated power level for the
near limiting transients. However, for the most limiting transient, the
evaluation of a representative core showed that no change is required
to the Operating Limit MCPR for the power rerate and that the integrity
of the SLMCPR is maintained. Cycle specific analysis will continue to
be performed for each fuel reload to demonstrate compliance with the
applicable transient criteria and to establish cycle specific operating
limits.
The fuel thermal-mechanical limits at the power rerate conditions
are within the specific design criteria for the GE fuels currently
loaded in the MNGP core. Also, the power-dependent and flow-dependent
MCPR and Maximum Average Planar Linear Heat Generation Rate (MAPLHGR)
methods developed as part of the core performance improvement program
remain applicable to rerate conditions. The transient event evaluation
confirmed that MNGP Power Rerate has no significant effect on the
power-dependent and flow-dependent MCPR and MAPLHGR limits. The peak
reactor pressure vessel bottom head pressure remains within the ASME
requirement for reactor pressure vessel over pressure protection.
The margin of safety established by the SLMCPR is not affected by
the proposed power rerate to 1775 MWt.
6. Technical Specification Changes
The Technical Specifications ensure that the plant and system
performance parameters are maintained at the values assumed in the
safety analysis. The Technical Specification (setpoints, trip settings,
etc.) are selected such that the actual equipment is maintained equal
to or conservative with respect to the inputs used in the safety
analysis. Proper account is taken of inaccuracies introduced by
instrument drift, instrument accuracy, and calibration accuracy. The
Technical Specifications address equipment availability and limit
equipment out-of-service to assure that the plant can be expected to
have at least the complement of equipment available to deal with plant
transients as that assumed in the safety analysis. The evaluations and
analyses performed to demonstrate the acceptability of MNGP Power
Rerate were performed using inputs consistent with the proposed changes
to the plant Technical Specifications.
The events that form the Technical Specification Bases were
evaluated for the power rerate conditions using inputs and initial
conditions consistent with the proposed Technical Specification
changes. Although some changes to the Technical Specifications are
required for the power rerate, no NRC acceptance limit will be
exceeded. Therefore, the margins of safety assured by safety limits and
other Technical Specification limits will be maintained. The changes to
the Technical Specification Bases proposed by this submittal are
consistent with the evaluations which demonstrated acceptability of the
power rerate.
7. Conclusion
The spectrum of postulated accidents, transients, and special
events has been investigated and [has] been determined to meet the
current regulatory criteria
[[Page 9613]]
for the MNGP at the power rerate conditions. In the area of core
design, the fuel operating limits will still be met at the rerate power
level, and fuel reload analyses will show plant transients meet the
criteria accepted by the NRC as specified in the plant Technical
Specifications. Challenges to fuel or ECCS performance were evaluated
and shown to meet the criteria of 10 CFR 50.46 and 10 CFR 50, Appendix
K. Challenges to the containment have been evaluated and the integrity
of the fission product barrier has been confirmed. Radiological release
events have been evaluated and shown to meet the guidelines of 10 CFR
100. The proposed Operating License and Technical Specification changes
are consistent with the MNGP Power Rerate evaluation performed. The
evaluations demonstrated compliance with the margin assuring acceptance
criteria contained in applicable codes and regulations. Therefore, the
proposed Operating License and Technical Specifications changes will
not involve a significant reduction in the margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: Minneapolis Public Library,
Technology and Science Department, 300 Nicollet Mall, Minneapolis,
Minnesota 55401
Attorney for licensee: Gerald Charnoff, Esq., Shaw, Pittman, Potts
and Trowbridge, 2300 N Street, NW, Washington, DC 20037
NRC Project Director: Cynthia A. Carpenter
Philadelphia Electric Company, Docket No. 50-352, Limerick Generating
Station (LGS), Unit 1, Montgomery County, Pennsylvania
Date of amendment request: February 9, 1998.
Description of amendment request: The amendment request proposes to
revise the LGS, Unit 1 Technical Specifications (TS) Section 2.1 and
its associated TS Basis to reflect the change in the minimum critical
power ratio (MCPR) safety limit due to the plant-specific evaluation
performed by General Electric Company (GE) for LGS, Unit 1, Cycle 8.
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 change does not involve a significant increase
in the probability or consequences of an accident previously evaluated.
The revised MCPR Safety Limits for LGS Unit 1 Technical
Specifications, and their use to determine cycle-specific thermal
limits, have been calculated using NRC-approved methods (i.e., GESTAR-
II, Rev. 13) and are based on LGS Unit 1 Cycle 8 specific inputs. The
use of these methods assures that the [safety limit for minimum
critical power ratio] SLMCPR value is within the existing design and
licensing basis, and cannot increase the probability or severity of an
accident.
The basis of the MCPR Safety Limit calculation is to ensure that
greater than 99.9% of all fuel rods in the core avoid transition
boiling if the limit is not violated. The MCPR Safety limit preserves
the existing margin to transition boiling and fuel damage in the event
of a postulated accident. The probability of fuel damage is not
increased.
Therefore, the proposed TS change does not involve an increase in
the probability or consequences of an accident previously evaluated.
2. The proposed TS change does not create the possibility of a new
or different kind of accident from any accident previously evaluated.
The MCPR Safety Limit is a Technical Specification numerical value
designed to ensure that fuel damage from transition boiling does not
occur as a result of the limiting postulated accident. The MCPR Safety
Limit is not an accident initiator; therefore, it cannot create the
possibility of any new type of accident. The new MCPR Safety Limits are
calculated using NRC-approved methods (i.e., GESTAR-II, Rev. 13) and
are based on LGS Unit 1, Cycle 8 specific inputs.
Therefore, the proposed TS change does not create the possibility
of a new or different kind of accident from any accident previously
evaluated.
3. The proposed TS change does not involve a significant reduction
in the margin of safety.
The margin of safety as defined in the TS Bases will remain the
same. The new MCPR Safety Limits are calculated using NRC-approved
methods (i.e., GESTAR-II, Rev. 13), which are in accordance with the
current fuel design and licensing criteria, and are based on LGS Unit 1
Cycle 8 specific inputs. The MCPR Safety Limit remains high enough to
ensure that greater than 99.9% of all fuel rods in the core will avoid
transition boiling if the limit is not violated, thereby preserving the
fuel cladding integrity.
Therefore, the proposed TS change does not involve a 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: Pottstown Public Library, 500
High Street, Pottstown, PA 19464.
Attorney for licensee: J. W. Durham, Sr., Esquire, Sr. V.P. and
General Counsel, Philadelphia Electric Company, 2301 Market Street,
Philadelphia, PA 19101.
NRC Project Director: John F. Stolz.
Power Authority of the State of New York, Docket No. 50-333, James A.
FitzPatrick Nuclear Power Plant, Oswego County, New York
Date of amendment request: October 14, 1997.
Description of amendment request: The proposed changes would
correct the maximum exposure dependent, infinite lattice multiplication
factor for fuel bundles and provide for installation of additional
storage racks to increase spent fuel capacity.
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:
Operation of the FitzPatrick plant in accordance with the proposed
Amendment would not involve a significant hazards consideration as
defined in 10 CFR 50.92, since it would not:
1. Involve a significant increase in the probability or
consequences of an accident previously evaluated because:
A change in the infinite lattice neutron multiplication factor for
a fuel bundle in the reactor core geometry which ensures the
criticality limit for fuel in the spent fuel pool [SFP] geometry is met
does not affect initiation of any accident.
Operation in accordance with the revised limit ensures the
consequences of previously analyzed accidents are not changed. Storage
of additional fuel assemblies in the pool does not affect the
probability or consequences of dropping a fuel assembly, since this
accident is localized to a small area of the storage array. Likewise,
addition of
[[Page 9614]]
specifications containing details presently in plant design documents
and editorial changes do not change the probability or consequences of
a previously analyzed accident.
2. Create the possibility of a new or different kind of accident
for any accident previously evaluated because:
A change in the infinite lattice neutron multiplication factor for
a fuel bundle in the reactor core geometry which ensures the
criticality limit for fuel in the spent fuel pool geometry is met does
not affect the types of reactivity accidents which may occur. Therefore
changing the limit will not [create the possibility of] a new or
different type of accident. Maintenance of available decay heat removal
systems ensures that no new type of loss of cooling accident associated
with the SFP will occur as a result of storing additional irradiated
fuel assemblies. Likewise, addition of specifications containing
details presently in plant design documents and editorial changes do
not create the possibility of a new or different type of accident.
3. Involve a significant reduction in a margin of safety because:
The revised limit on infinite lattice neutron multiplication factor
for a fuel bundle in the reactor core geometry ensures maintenance of
the same margin of safety with respect to criticality as presently
exists for storage of fuel in the SFP. Storing additional irradiated
fuel assemblies in the pool does not affect the margin of safety with
regard to pool cooling since sufficient heat removal systems will be
maintained available to ensure maintenance of acceptable pool
temperatures. Addition of specifications containing details presently
in other design documents and editorial changes have no effect on the
margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 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: Reference and Documents
Department, Penfield Library, State University of New York, Oswego, New
York 13126.
Attorney for licensee: Mr. David E. Blabey, 1633 Broadway, New
York, New York 10019.
NRC Project Director: S. Singh Bajwa.
South Carolina Electric & Gas Company (SCE&G), South Carolina Public
Service Authority, Docket No. 50-395, Virgil C. Summer Nuclear Station,
Unit No. 1, Fairfield County, South Carolina
Date of amendment request: February 9, 1998.
Description of amendment request: The proposed amendment would
revise the Virgil C. Summer Nuclear Station Technical Specifications
(TS) to remove emergency diesel generator (1) accelerated testing
requirements (TS 3/4.8.1, Table 4.8-1), and (2) special reporting
requirements (TS Surveillance Requirement 4.8.1.1.3) in accordance with
NRC Generic Letter (GL) 94-01, ``Removal of Accelerated Testing and
Special Reporting Requirements for Emergency Diesel Generators.''
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. This request does not involve a significant increase in the
probability or consequences of an accident previously evaluated.
This change will provide flexibility to structure the emergency
diesel generator maintenance program based on the risk significance of
the structures, systems, and components that are within the scope of
the maintenance rule. The removal of the diesel generator accelerated
testing is acceptable as the maintenance rule applies system and train
specific performance criteria to monitor diesel generator performance.
These criteria include a running availability and reliability measure.
The performance criteria for the diesel generator reliability and
unavailability established by the maintenance rule, and the causal
determinations and corrective actions required for functional failures
and/or exceeding performance criteria, is considered to be an
acceptable method for monitoring diesel generator performance.
As the diesel generator performance will [continue] to be assured
by the maintenance rule, the proposed changes do not affect any of the
initiators for an accident previously evaluated. The changes do not
impact the diesel's design sources, operating characteristics, system
functions, or system interrelationships. The failure mechanisms for the
accidents previously analyzed are not affected, and no additional
failure modes are created that could cause an accident previously
evaluated. Since the changes are administrative in nature, and the
diesel generator performance and reliability will continue to be
assured by the maintenance rule, the proposed changes cannot involve a
significant increase in the probability or consequences of an accident
previously evaluated.
2. This request does not create the possibility of a new or
different kind of accident from any accident previously evaluated.
This proposed change does not involve a change to the plant design
or operation. As a result, the proposed change does not affect any of
the parameters or conditions that could contribute to the initiation of
any accidents. The proposed changes only affect the methods used to
monitor and assure diesel generator performance. The performance
criteria for both the diesel generator reliability and unavailability
established by the maintenance rule, and the causal determinations and
corrective actions required for functional failures and/or exceeding
performance criteria, is considered by GL 94-01 to be an acceptable
method for monitoring diesel generator performance.
No SSC [structure, system, or component], method of operating, or
system interface is altered by this change. The changes do not impact
the diesel's design sources, operating characteristics, system
functions, or system interrelationships. The failure mechanisms for the
accidents are not affected, and no additional failure modes are
created. Because the proposed changes are administrative in nature, and
the diesel generator performance and reliability will continue to be
assured by the maintenance rule, the proposed changes cannot create the
possibility of a new or different kind of accident from any accident
previously evaluated.
3. This request does not involve a significant reduction in a
margin [of] safety.
The proposed changes only affect the methods used to monitor and
assure diesel generator performance. The performance criteria for both
the diesel generator reliability and unavailability established by the
maintenance rule, and the causal determinations and corrective actions
required for functional failures and/or exceeding performance criteria,
is considered by GL 94-01 to be an acceptable method for monitoring
diesel generator performance. No margin [of] safety as defined in the
basis for any technical specification is impacted by these changes.
This change does not impact any uncertainty in the design,
construction, or operation of any SSC.
[[Page 9615]]
Diesel generator response to accident initiators is unchanged. No SSC,
method of operating, or system interface is altered by this change. The
changes do not impact the diesel's design sources, operating
characteristics, system functions, or system interrelationships.
Because the proposed changes are administrative in nature, and the
diesel generator performance and reliability will continue to be
assured by the maintenance rule, the proposed changes cannot involve a
significant reduction in the margin [of] safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: Fairfield County Library, 300
Washington Street, Winnsboro, SC 29180.
Attorney for licensee: Randolph R. Mahan, South Carolina Electric &
Gas Company, Post Office Box 764, Columbia, South Carolina 29218.
NRC Project Director: William M. Dean.
Wolf Creek Nuclear Operating Corporation, Docket No. 50-482, Wolf Creek
Generating Station, Coffey County, Kansas
Date of amendment request: January 28, 1998.
Description of amendment request: The proposed amendment would
revise Technical Specification (TS) Sections 6.3 and 6.12 to reflect
the merger of the positions of Superintendent Radiation Protection and
Superintendent Chemistry into one new position, Manager Chemistry/
Radiation Protection.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. The proposed change does not involve a significant increase in
the probability or consequences of an accident previously evaluated.
The proposed change does not involve a significant increase in the
probability of consequences of an accident previously evaluated. These
changes involve administrative changes to the WCNOC organization.
2. The proposed change does not create the possibility of a new or
different kind of accident from any accident previously evaluated.
The proposed change does not create the possibility of a new or
different kind of accident from any accident previously evaluated. This
change is administrative in nature and does not involve a change to the
installed plant systems or the overall operating philosophy of Wolf
Creek Generating Station.
3. The proposed change does not involve a significant reduction in
a margin of safety.
The proposed change does not involve a significant reduction in a
margin of safety. This change does not involve any changes in overall
organizational commitments and will not affect qualification
requirements of any unit staff personnel. A position and title change
alone does not 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
amendment request involves no significant hazards consideration.
Local Public Document Room locations: Emporia State University,
William Allen White Library, 1200 Commercial Street, Emporia, Kansas
66801 and Washburn University School of Law Library, Topeka, Kansas
66621.
Attorney for licensee: Jay Silberg, Esq., Shaw, Pittman, Potts and
Trowbridge, 2300 N Street, N.W., Washington, D.C. 20037.
NRC Project Director: William H. Bateman.
Previously Published Notices of 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.
Northern States Power Company, Docket No. 50-282, Prairie Island
Nuclear Generating Plant, Unit 1, Goodhue County, Minnesota
Date of amendment request: January 15, 1998.
Description of amendment request: The proposed amendment would
initiate a one-time only change for Prairie Island Unit 1 Cycle 19 that
would allow the use of the moveable incore detector system for
measurement of the core peaking factors with less than 75% and greater
than or equal to 50% of the detector thimbles available.
Date of individual notice in the Federal Register: January 30, 1998
(63 FR 4676).
Expiration date of individual notice: March 2, 1998.
Local Public Document Room location: Minneapolis Public Library,
Technology and Science Department, 300 Nicollet Mall, Minneapolis,
Minnesota 55401.
Attorney for licensee: Jay Silberg, Esq., Shaw, Pittman, Potts, and
Trowbridge, 2300 N Street, NW, Washington, DC 20037.
NRC Project Director: Cynthia A. Carpenter.
Notice of Issuance of Amendments to Facility Operating Licenses
During the period since publication of the last biweekly notice,
the Commission has issued the following amendments. The Commission has
determined for each of these amendments that the application complies
with the standards and requirements of the Atomic Energy Act of 1954,
as amended (the Act), and the Commission's rules and regulations. The
Commission has made appropriate findings as required by the Act and the
Commission's rules and regulations in 10 CFR Chapter I, which are set
forth in the license amendment.
Notice of Consideration of Issuance of Amendment to Facility
Operating License, Proposed No Significant Hazards Consideration
Determination, and Opportunity for A Hearing in connection with these
actions was published in the Federal Register as indicated.
Unless otherwise indicated, the Commission has determined that
these amendments satisfy the criteria for categorical exclusion in
accordance with 10 CFR 51.22. Therefore, pursuant to 10 CFR 51.22(b),
no environmental impact statement or environmental assessment need be
prepared for these amendments. If the Commission has prepared an
environmental assessment under the special circumstances
[[Page 9616]]
provision in 10 CFR 51.12(b) and has made a determination based on that
assessment, it is so indicated.
For further details with respect to the action see (1) the
applications for amendment, (2) the amendment, and (3) the Commission's
related letter, Safety Evaluation and/or Environmental Assessment as
indicated. All of these items are available for public inspection at
the Commission's Public Document Room, the Gelman Building, 2120 L
Street, NW., Washington, DC, and at the local public document rooms for
the particular facilities involved.
Baltimore Gas and Electric Company, Docket No. 50-317, Calvert Cliffs
Nuclear Power Plant, Unit No. 1, Calvert County, Maryland
Date of application for amendment: May 16, 1997, as supplemented
November 14, 1997.
Brief description of amendment: The amendment involves replacing
the service water (SRW) heater exchangers with new plate and frame heat
exchangers (PHEs), having increased thermal performance capability. The
Saltwater (SW) and SRW piping configuration will be modified as
necessary to allow proper fit-up to the new components. A flow control
scheme to throttle saltwater flow to the heat exchangers and the
associated bypass lines will be added. Saltwater strainers with an
automatic flushing arrangement will be added upstream of each heat
exchanger. The majority of the physical work associated with this
modification is restricted to the SRW pump room. The amendment is
partially denied to the extent that the licensee is not authorized to
operate with one PHE secured, and removing one containment air cooler
from service to enable the affected subsystem to remain operable while
the one PHE is secured.
Date of issuance: February 10, 1998.
Effective date: As of the date of issuance to be implemented within
30 days.
Amendment No.: 225.
Facility Operating License No. DPR-53: Amendment revised the
Updated Final Safety Analysis Report.
Date of initial notice in Federal Register: June 18, 1997 (62 FR
33118).
The November 14, 1997, letter provided clarifying information that
did not change the initial proposed no significant hazards
consideration.
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated February 10, 1998.
No significant hazards consideration comments received: No.
Local Public Document Room location: Calvert County Library, Prince
Frederick, Maryland 20678.
Carolina Power & Light Company, et al., Docket Nos. 50-325 and 50-324,
Brunswick Steam Electric Plant, Units 1 and 2, Brunswick County, North
Carolina
Date of application for amendments: November 6, 1997, as
supplemented by letter dated January 28, 1998.
Brief Description of amendments: The amendments to Technical
Specification (TS) Limiting Conditions for Operation (LCO) 3.3.5.5,
Instrumentation for Control Room Emergency Ventilation System (CREVS)
and 3.7.2, Control Room Emergency Ventilation System, and associated
Bases for the Brunswick Steam Electric Plant (BSEP) Units 1 and 2 will
be limited in duration (approximately 3 months) and will allow
operation of both BSEP units to continue while upgrades to the control
building ventilation system, including new air conditioning (AC) units
and improved ductwork supports, are being installed. Part of the
planned work requires opening the ductwork at the evaporative (i.e.
cooling) coils. Temporary barriers will be constructed to preserve the
leakage integrity of the control room pressure boundary; however, the
temporary barriers will not be seismically qualified. While the
permanent AC units are out of service, temporary AC units will be
utilized. During the upgrade installation, the AC for the control room
will not be protected from certain external events (e.g., seismic
events, environmental hazards such as tornadoes and hurricanes,
radiological sabotage, and missile hazards), as required by the system
design and licensing basis, and will not fully meet single failure
criteria.
Date of issuance: February 6, 1998.
Effective date: February 6, 1998.
Amendment Nos.: 191 and 222.
Facility Operating License Nos. DPR-71 and DPR-62: Amendments
authorize changes to the facility's Technical Specifications.
Date of initial notice in Federal Register: December 3, 1997 (62 FR
63973).
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated February 6, 1998.
No significant hazards consideration comments received: No.
Local Public Document Room location: University of North Carolina
at Wilmington, William Madison Randall Library, 601 S. College Road,
Wilmington, North Carolina 28403-3297.
Commonwealth Edison Company, Docket Nos. STN 50-454 and STN 50-455,
Byron Station, Unit Nos. 1 and 2, Ogle County, Illinois, Docket Nos.
STN 50-456 and STN 50-457, Braidwood Station, Unit Nos. 1 and 2, Will
County, Illinois
Date of application for amendments: February 28, 1997. Information
related to the proposed restoration of the primary coolant dose
equivalent iodine-131 (DEI) to their original licensing basis had been
previously submitted in Commonwealth Edison Company's (ComEd) letter
dated November 13, 1996, which was supplemented in subsequent letters
dated March 20, June 24, August 19 and November 3, 1997.
Brief description of amendments: The amendments revise the
technical specifications (TS) to reflect the forthcoming replacement of
the original steam generators (OSG) in Byron, Unit 1, and Braidwood,
Unit 1, which are Westinghouse Model D4 steam generators (SG), with the
replacement steam generators (RSG) which are Babcock and Wilcox,
International (BWI) SG. The present revisions to the TS remove the
interim plugging criteria (IPC) related to outer diameter stress
corrosion cracking (ODSCC) in the OSG as well as the F* alternative
repair criteria and two separate SG tube sleeving methodologies which
are not needed for the RSG.
Date of issuance: February 3, 1998
Effective date: This license amendment is effective as of the date
of its issuance and shall be implemented in the first operating cycle
after installation of the BWI replacement steam generators
Amendment Nos.: 101, 101, 92 and 92.
Facility Operating License Nos. NPF-37, NPF-66, NPF-72 and NPF-77:
The amendments revised the Technical Specifications.
Date of initial notice in Federal Register: December 17, 1997 (62
FR 66134). The November 13, 1996, and March 20, June 24, August 19 and
November 3, 1997, submittals provided clarifying information that did
not change the initial proposed no significant hazards consideration
determination.
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated February 3, 1998.
No significant hazards consideration comments received: No
Local Public Document Room location: For Byron, the Byron Public
Library District, 109 N. Franklin, P.O. Box 434, Byron, Illinois 61010;
for
[[Page 9617]]
Braidwood, the Wilmington Public Library, 201 S. Kankakee Street,
Wilmington, Illinois 60481.
Duquesne Light Company, et al., Docket Nos. 50-334 and 50-412, Beaver
Valley Power Station, Unit Nos. 1 and 2, Shippingport, Pennsylvania
Date of application for amendments: November 6, 1995, and March 11,
1996, as supplemented June 5, 1997. The June 5, 1997, letter provided
clarifying information that did not change the initial proposed no
significant hazards consideration determination or expand the amendment
request beyond the scope of the December 20, 1995, and April 10, 1996,
Federal Register notices.
Brief description of amendments: These amendments revise the alarm
setpoints for the effluent radiation and in-containment area radiation
monitors listed in Technical Specification (TS) Table 3.3-6. These
revisions make these alarm setpoints consistent with criteria for the
Emergency Action Levels (EALs) approved by the Nuclear Regulatory
Commission in August 1994. The EALs use these monitors as an indication
of fission product barrier challenges or failures. These amendments
also revise Action Statement 36 of TS Table 3.3-6 to reflect a
previously approved change (License Amendment Nos. 188 and 70) in
reporting frequency (change from semi-annual to annual) for effluent
releases. The revision to Action Statement 36 makes it consistent with
the previously approved change. These amendments include several
editorial changes to the TSs which do not change the intent of the TSs.
Date of issuance: February 9, 1998.
Effective date: Both units, as of date of issuance, to be
implemented within 60 days.
Amendment Nos.: 211 and 89.
Facility Operating License Nos. DPR-66 and NPF-73: Amendments
revised the Technical Specifications.
Dates of initial notice in Federal Register: December 20, 1995 (60
FR 65677) and April 10, 1996 (61 FR 15988). The Commission's related
evaluation of the amendments is contained in a Safety Evaluation dated
February 9, 1998.
No significant hazards consideration comments received: No.
Local Public Document Room location: B. F. Jones Memorial Library,
663 Franklin Avenue, Aliquippa, PA 15001.
Duquesne Light Company, et al., Docket No. 50-334, Beaver Valley Power
Station, Unit No. 1, Shippingport, Pennsylvania
Date of application for amendment: November 4, 1997.
Brief description of amendment: The amendment revises Item 6.a.2,
``4.16kV Emergency Bus (Start Diesel),'' of Table 3.3-4 of Technical
Specification 3.3.2.1. The change reduces the trip setpoint for
starting the emergency diesel generators on emergency bus undervoltage
from a trip setpoint of greater than or equal to 83 percent with a 12-
cycle delay time to a setpoint of greater than or equal to 75 percent
of nominal bus voltage with a time delay of less than 0.9 seconds
including auxiliary relay times. The amendment also revises the
allowable value from greater than or equal to 81 percent of nominal bus
voltage to greater than or equal to 74 percent of nominal bus voltage
with a time delay of less than 0.9 seconds including auxiliary relay
times.
Date of issuance: February 11, 1998.
Effective date: As of date of issuance, to be implemented within 60
days.
Amendment No: 212.
Facility Operating License No. DPR-66. Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: December 3, 1997 (62 FR
63976).
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated February 11, 1998.
No significant hazards consideration comments received: No.
Local Public Document Room location: B. F. Jones Memorial Library,
663 Franklin Avenue, Aliquippa, PA 15001.
Entergy Operations, Inc., System Energy Resources, Inc., South
Mississippi Electric Power Association, and Entergy Mississippi, Inc.,
Docket No. 50-416, Grand Gulf Nuclear Station, Unit 1, Claiborne
County, Mississippi
Date of application for amendment: August 6, 1997.
Brief description of amendment: The amendment eliminated the
provisions in Technical Specification 3.8.1, ``AC Sources--Operating,''
for accelerated testing of the emergency diesel generators (DG). The
changes are the following: (1) the frequency of verifying DG starts and
operation in Surveillance Requirements (SRs) 3.8.1.2 and 3.8.1.3,
respectively, are changed to 31 days, from the present reference to
Table 3.8.1-1, and (2) Table 3.8.1-1, ``Diesel Generator Test
Schedule,'' is deleted. The emergency diesel generators provide
emergency AC power to the site with the loss of offsite AC power.
Date of issuance: February 9, 1998.
Effective date: February 9, 1998.
Amendment No: 134.
Facility Operating License No. NPF-29: Amendment revises the
Technical Specifications.
Date of initial notice in Federal Register: September 24, 1997 (62
FR 50003).
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated February 9, 1998.
No significant hazards consideration comments received: No.
Local Public Document Room location: Judge George W. Armstrong
Library, 220 S. Commerce Street, Natchez, MS 39120.
Northeast Nuclear Energy Company, et al., Docket No. 50-336, Millstone
Nuclear Power Station, Unit 2, New London County, Connecticut
Date of application for amendment: September 3, 1997.
Brief description of amendment: The amendment authorizes Northeast
Nuclear Energy Company, through a license condition, to incorporate
changes to the description of the facility in the Updated Final Safety
Analysis Report (UFSAR). This change revises the UFSAR by modifying the
operation of the onsite emergency diesel generators and their
associated fuel oil supplies.
Date of issuance: January 23, 1998.
Effective date: As of the date of issuance to be implemented within
30 days.
Amendment No.: 212.
Facility Operating License No. DPR-65: Amendment revised the
Updated Final Safety Analysis Report.
Date of initial notice in Federal Register: September 24, 1997 (62
FR 50009).
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated January 23, 1998.
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, and the Waterford Library, ATTN: Vince Juliano,
49 Rope Ferry Road, Waterford, Connecticut.
[[Page 9618]]
Public Service Electric & Gas Company, Docket Nos. 50-272 and 50-311.
Salem Nuclear Generating Station, Unit Nos. 1 and 2, Salem County, New
Jersey
Date of application for amendments: October 24, 1997.
Brief description of amendments: The amendments revise the
containment hydrogen analyzer Technical Specifications (TSs)
surveillance requirements of TS 4.6.4.1 to increase the calibration
frequency from once per refueling outage to quarterly.
Date of issuance: January 29, 1998.
Effective date: As of the date of issuance.
Amendment Nos. 204 and 186.
Facility Operating License Nos. DPR-70 and DPR-75. The amendments
revised the Technical Specifications.
Date of initial notice in Federal Register: December 17, 1997 (62
FR 66140).
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated January 29, 1998.
No significant hazards consideration comments received: No.
Local Public Document Room location: Salem Free Public Library, 112
West Broadway, Salem, NJ 08079.
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: June 13, 1997, as supplemented by
letter dated January 7, 1998.
Brief Description of amendments: The amendments change Technical
Specification (TS) 3.9.13 by adding a footnote to clarify the required
electrical power sources for the penetration room filtration system
when it is aligned to the spent fuel pool room during refueling
operations. In addition, the associated Bases section of the TS will be
modified to provide additional details concerning the proposed TS
change.
Date of issuance: February 5, 1998.
Effective date: As of the date of issuance to be implemented within
30 days.
Amendment Nos.: Unit 1-134; Unit 2-126.
Facility Operating License Nos. NPF-2 and NPF-8: Amendments revise
the Technical Specifications.
Date of initial notice in Federal Register: July 16, 1997 (62 FR
38138).
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated February 5, 1998.
No significant hazards consideration comments received: No.
Local Public Document Room location: Houston-Love Memorial Library,
212 W. Burdeshaw Street, Post Office Box 1369, Dothan, Alabama.
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: October 16, 1997.
Brief Description of amendments: The amendments change the Farley
Units 1 and 2 TS by revising the number of allowable charging pumps
capable of injecting into the reactor coolant system (RCS) when the
temperature of one or more of the RCS cold legs is equal to or less
than 180 deg. F.
Date of issuance: February 5, 1998.
Effective date: As of the date of issuance to be implemented within
30 days.
Amendment Nos.: Unit 1-135; Unit 2-127.
Facility Operating License Nos. NPF-2 and NPF-8: Amendments revise
the Technical Specifications.
Date of initial notice in Federal Register: December 3, 1997 (62 FR
63983).
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated February 5, 1998.
No significant hazards consideration comments received: No.
Local Public Document Room location: Houston-Love Memorial Library,
212 W. Burdeshaw Street, Post Office Box 1369, Dothan, Alabama.
Toledo Edison Company, Centerior Service Company, and The Cleveland
Electric Illuminating Company, Docket No. 50-346, Davis-Besse Nuclear
Power Station, Unit 1, Ottawa County, Ohio
Date of application for amendment: October 28, 1996, as
supplemented by letters dated August 19, 1997, and October 16, 1997.
Brief description of amendment: This amendment revises TS Section
3/4.8.1, ``A.C. Sources,'' TS Section 3/4.8.2, ``Onsite Power
Distribution Systems,'' TS Table 4.8.1, ``Battery Surveillance
Requirements,'' and the associated bases. Surveillance requirements
have been modified to account for the increase in the fuel cycle.
Administrative changes were also made.
Date of issuance: February 3, 1998.
Effective date: February 3, 1998.
Amendment No.: 219.
Facility Operating License No. NPF-3: Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: January 2, 1997 (62 FR
132).
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated February 3, 1998.
No significant hazards consideration comments received: No. The
supplemental information provided by the Licensees did not affect the
proposed no significant hazards consideration determination.
Local Public Document Room location: University of Toledo, William
Carlson Library, Government Documents Collection, 2801 West Bancroft
Avenue, Toledo, OH 43606.
Dated at Rockville, Maryland, this 18th day of February 1998.
For the Nuclear Regulatory Commission.
Elinor G. Adensam,
Acting Director, Division of Reactor Projects--III/IV, Office of
Nuclear Reactor Regulation.
[FR Doc. 98-4620 Filed 2-24-98; 8:45 am]
BILLING CODE 7590-01-P
