[Federal Register Volume 64, Number 164 (Wednesday, August 25, 1999)]
[Notices]
[Pages 46424-46455]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: 99-21914]
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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 July 31, 1999, through August 13, 1999. The
last biweekly notice was published on August 11, 1999 (64 FR 43764).
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
[[Page 46425]]
determination. Any comments received within 30 days after the date of
publication of this notice will be considered in making any final
determination.
Normally, the Commission will not issue the amendment until the
expiration of the 30-day notice period. However, should circumstances
change during the notice period such that failure to act in a timely
way would result, for example, in derating or shutdown of the facility,
the Commission may issue the license amendment before the expiration of
the 30-day notice period, provided that its final determination is that
the amendment involves no significant hazards consideration. The final
determination will consider all public and State comments received
before action is taken. Should the Commission take this action, it will
publish in the Federal Register a notice of issuance and provide for
opportunity for a hearing after issuance. The Commission expects that
the need to take this action will occur very infrequently.
Written comments may be submitted by mail to the Chief, Rules and
Directives Branch, Division of 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 September 24, 1999, 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
[[Page 46426]]
document room for the particular facility involved.
Carolina Power & Light Company, et al., Docket No. 50-400, Shearon
Harris Nuclear Power Plant, Unit 1, Wake and Chatham Counties, North
Carolina
Date of amendment request: August 2, 1999.
Description of amendment request: The proposed amendment would
revise Technical Specification 6.2.2.e to require either the Operations
Manager or an off-shift Operations superintendent to hold a senior
reactor operator (SRO) license. This revision would delete the option
which allows the Manager-Operations to have at one time held a Senior
Reactor Operator License for a similar unit and replaces it with the
requirement for an off-shift Operations superintendent who holds an SRO
license to supervise shift work and licensed activities.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. The proposed amendment does not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
The change to Technical Specification 6.2.2.e to require the
Manager-Operations or an off-shift Operations superintendent to hold
an SRO license is administrative in nature and does not directly
affect plant operations. The change does not physically alter the
facility in any manner and, as such, does not affect the means in
which any safety-related system performs its intended safety
function.
Therefore, there would be no increase in the probability or
consequences of an accident previously evaluated.
2. The proposed amendment does not create the possibility of a
new or different kind of accident from any accident previously
evaluated.
As stated above, the proposed change is administrative in
nature. There is no physical alteration to any plant system, nor is
there a change in the method in which any safety related system
performs its function.
Therefore, the proposed change does not create the possibility
of a new or different kind of accident from any accident previously
evaluated.
3. The proposed amendment does not involve a significant
reduction in the margin of safety.
The proposed amendment does not reduce the margin of safety as
defined in the Safety Analysis Report or the bases contained in the
Technical Specifications. The requirement to have a licensed SRO
management position responsible for plant operations is maintained
within the proposed amendment. The proposed amendment is consistent
with (1) 10 CFR 50.54(l), which requires individuals responsible for
directing the licensed activities of licensed operators to hold an
SRO license, (2) Revision 1 of NUREG-1431, ``Standard Technical
Specifications Westinghouse Plants,'' and Technical Specification
Traveler Form (TSTF) 65, Revision 1, and (3) the intent of ANSI/ANS-
3.1, ``Standard for Selection and Training of Personnel for Nuclear
Power Plants,'' (September 1979 Draft).
Therefore, the proposed change does not involve a significant
reduction in the margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: Cameron Village Regional
Library, 1930 Clark Avenue, Raleigh, North Carolina 27605
Attorney for licensee: William D. Johnson, Vice President and
Corporate Secretary, Carolina Power & Light Company, Post Office Box
1551, Raleigh, North Carolina 27602
NRC Section Chief: Sheri R. Peterson.
Carolina Power & Light Company, et al., Docket No. 50-400, Shearon
Harris Nuclear Power Plant, Unit 1, Wake and Chatham Counties, North
Carolina
Date of amendment request: August 4, 1999.
Description of amendment request: The proposed amendment would
revise Technical Specification 6.9.1.6.2 to incorporate analytical
methodology references which are used to determine core operating
limits. The analytical methodologies to be referenced are documented in
topical reports which have been accepted by the Nuclear Regulatory
Commission for referencing in licensing applications.
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 license amendment does not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
The proposed changes incorporate additional references to
methodologies used to evaluate core operating limits. These
methodologies have been approved for use by the NRC. Plant
structures, systems, and components will not be operated in a
different manner as a result of these proposed changes and no
physical modifications to equipment are involved. Adding these
references to the Core Operating Limits Report section of Technical
Specifications does not increase the probability or consequences of
an accident previously evaluated.
2. The proposed amendment does not create the possibility of a
new or different kind of accident from any accident previously
evaluated.
The proposed changes incorporate additional references to
methodologies used to evaluate core operating limits. These
methodologies have been approved for use by the NRC. Plant
structures, systems, and components will not be operated in a
different manner as a result of these proposed changes and no
physical modifications to equipment are involved. Adding these
references to the Core Operating Limits Report section of Technical
Specifications does not create the possibility of a new or different
type of accident from any previously evaluated.
3. The proposed amendment does not involve a significant
reduction in the margin of safety.
The proposed changes incorporate additional references to
methodologies used to evaluate core operating limits. These
methodologies have been approved for use by the NRC. Plant
structures, systems, and components will not be operated in a
different manner as a result of these proposed changes and no
physical modifications to equipment are involved. Adding these
references to the Core Operating Limits Report section of Technical
Specifications 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: Cameron Village Regional
Library, 1930 Clark Avenue, Raleigh, North Carolina 27605.
Attorney for licensee: William D. Johnson, Vice President and
Corporate Secretary, Carolina Power & Light Company, Post Office Box
1551, Raleigh, North Carolina 27602.
NRC Section Chief: Sheri R. Peterson.
Commonwealth Edison Company, Docket Nos. 50-237 and 50-249, Dresden
Nuclear Power Station, Units 2 and 3, Grundy County, Illinois
Date of amendment request: May 20, 1999.
Description of amendment request: The proposed amendments would
revise Technical Specification (TS) 3.8.A to identify the specific
Containment Cooling Service Water (CCSW) equipment required to support
operation of the Control Room Emergency Ventilation System (CREVS). The
proposed amendment would also
[[Page 46427]]
revise TS 3/4.5.C.2 to ensure that the suppression pool water level is
adequate to prevent vortexing in the Low Pressure Coolant Injection and
Core Spray pump suctions.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. Involve a significant increase in the probability or
consequences of an accident previously evaluated because of the
following:
The proposed changes to the technical specifications provide
clarity in the support system relationship and requirements for the
CCSW system support of the CREVS operation. [Neither] [t]he CCSW
system nor the CREVS system are assumed to be accident precursors
for previously evaluated accident[s]. Therefore, the proposed
changes have no effect on the probability or consequences of
accidents previously evaluated.
The proposed change to the allowable suppression chamber level
does not involve a significant increase in the probability or
consequences of an accident previously evaluated. The proposed
change revises a Technical Specification acceptance value to [a]
more conservative value and serves to ensure operability of
equipment important to safety. By ensuring equipment availability,
the probability or consequences of an accident previously evaluated
are not increased. In addition, the proposed changes have no impact
on any initial condition assumptions for accident scenarios. Onsite
or offsite dose consequences resulting from an event previously
evaluated are not affected by this proposed amendment request.
2. Create the possibility of a new or different kind of accident
from any accident previously evaluated because:
The proposed changes do not create the possibility of a new or
different kind of accident from that previously evaluated. The
changes to the CCSW specifications more appropriate[ly] reflect the
design requirements and clarify the support role of the CCSW system
as it relates the CREVS. Neither the CCSW system nor the CREVS will
be operated differently with the proposed change. Therefore new or
different failure modes will not be created. Therefore, the
possibility of new and different accidents has not been created with
the proposed change. The proposed change to the suppression pool
allowable level restores margin to the Technical Specifications and
ensures equipment operability. The proposed change is conservative
with respect to current requirements. The proposed amendment does
not involve any plant physical changes that would create the
possibility of a new or different kind of accident from any accident
previously evaluated.
Therefore, the proposed amendment does not create the
possibility of a new or different kind of accident from any accident
previously evaluated.
3. Involve a significant reduction in the margin of safety
because:
The proposed change to the CCSW technical specification will not
result in a significant reduction in the margin of safety. The
proposed change has greater consistency with the current design
requirements for CSSW support of CREVS operation. Therefore, the
margin of safety has been not been altered. [Therefore, the margin
of safety has not been altered. SIC]
The proposed changes for suppression pool level does not involve
a significant reduction in a margin of safety. In fact, the proposed
changes restore margin and ensure equipment operability. Since the
changes maintain the necessary level of system reliability, they do
not involve a significant reduction in the margin of safety.
The proposed amendment for Dresden will not reduce the
availability of systems required to mitigate accident conditions;
therefore, the proposed changes do not involve a significant
reduction in the margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
requested amendments involve no significant hazards consideration.
Local Public Document Room location: Morris Area Public Library
District, 604 Liberty Street, Morris, Illinois 60450.
Attorney for licensee: Ms. Pamela B. Stroebel, Senior Vice
President and General Counsel, Commonwealth Edison Company, P.O. Box
767, Chicago, Illinois 60690-0767.
NRC Section Chief: Anthony J. Mendiola.
Commonwealth Edison Company, Docket Nos. 50-373 and 50-374, LaSalle
County Station, Units 1 and 2, LaSalle County, Illinois
Date of amendment request: July 14, 1999.
Description of amendment request: The proposed amendments would
allow the units to operate at an uprated power level of 3489 MWt, an
increase of 5 percent rated core thermal power.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
Does the change involve a significant increase in the
probability or consequences of an accident previously evaluated?
A. Evaluation of the Probability of Previously Evaluated Accidents
The proposed power uprate imposes only minor increases in plant
operating conditions. No change is made to the reactor operating
pressure. Operation at uprated conditions will result in moderate
flow increases in those systems associated with the turbine cycle in
that steam flow increases by approximately six (6)% and feed flow
increases by approximately six (6)%. The increase in flow in the
carbon steel piping systems was evaluated for the effect on flow
induced erosion and corrosion rates and it was confirmed that power
uprate has no significant effect on flow induced erosion or
corrosion. The affected systems are currently monitored by the Flow
Accelerated Corrosion (FAC) program that addresses erosion and
corrosion concerns. Continued monitoring of the systems provides a
high level of confidence in the integrity of potentially susceptible
high energy piping systems.
Plant systems and components have been verified to be capable of
performing their intended design functions at uprated power
conditions. Where necessary, some components will be modified prior
to implementation of uprated power conditions to accommodate the
revised operating conditions. The review has concluded that
operation at power uprate conditions will not affect the reliability
of plant equipment, and that current Technical Specifications (TS)
surveillance requirements ensure adequate monitoring of system
operability. Systems continue to be operated in accordance with
current design requirements under uprated conditions, therefore no
new components or system interactions were identified that could
lead to an increase in accident probability. Changes to reactor
scram setpoints are such that no significant increase in scram
frequency due to operation at uprated conditions will occur.
B. Evaluation of the Consequences of Previously Evaluated Accidents
The radiological consequences due to the Loss of Coolant
Accident (LOCA) were calculated and are found to be below the
applicable regulatory limits. The results are presented in Table 9-3
of Attachment E [of the July 14, 1999 submittal].
The LOCA radiological consequences have not significantly
increased due to power uprate, and radiological consequences
continue to meet established regulatory limits.
The radiological evaluations for other non-LOCA Design Basis
Accidents (DBAs) were also performed and the dose consequences for
these events did not significantly increase. These changes are
outlined in Section 9.2 of Attachment E and they demonstrate that
LaSalle County Station (LCS), Units 1 and 2 still meets the
applicable regulatory limits.
Non-DBA Radiological Doses
All of the other radiological releases discussed in Updated
Final Safety Analysis Report (UFSAR) are either unchanged because
they are not power-dependent, or increase approximately in linear
proportion to the amount of the uprate. The dose consequences for
all of the non-LOCA radiological release accident events did not
significantly increase, and are bounded by the ``LOCA Radiological
Consequences''
[[Page 46428]]
events discussed above and were shown to meet the current dose
acceptance limits. These events are discussed in Section 9.2 of
Attachment E.
Therefore, the proposed change does not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
Does the change create the possibility of a new or different
kind of accident from any accident previously evaluated?
The configuration, operation and event response of the LCS,
Units 1 and 2 systems, structures or components are [unchanged] by
operation at uprated power conditions. Analysis of transient events
has confirmed that the same transients remain limiting and that no
transient event results in a new sequence of events that could lead
to a new accident scenario.
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 current fission product barriers consisting of the reactor
fuel rod cladding, the reactor coolant pressure boundary, and the
containment structure remain in place. Fuel rod cladding integrity
is ensured by operating within thermal, mechanical, and exposure
design limits, and was confirmed for a representative core by
performance of transient and accident analysis. Cycle specific
analysis will continue to be performed for each fuel reload to
demonstrate the compliance with the applicable transient analysis
criteria and to establish the cycle specific Minimum Critical Power
Ratio (MCPR) safety limit and fuel operating limits. The integrity
of the reactor coolant pressure boundary was confirmed by evaluation
of the bounding overpressurization event and ensuring that the
corresponding pressure remained below the American Society of
Mechanical Engineers (AMSE) Boiler and Pressure Vessel (B&PV) Code,
Section III, ``Rules for Construction of Nuclear Power Plant
Components,'' overpressure protection requirements. Similarly,
analysis of the primary containment structure has demonstrated under
worst case design basis accident conditions that the containment
structure remains below the containment design pressure.
The effect of operation at uprated conditions on plant equipment
has been evaluated. No new operating mode, safety-related equipment
lineup, accident scenario, or equipment failure mode was identified
as a result of operating at uprated conditions. In addition,
operation at power uprated conditions does not create any new
sequence of events or failure modes that lead to a new type of
accident. Plant modifications required to support implementation of
power uprated conditions will be made to existing systems rather
than by adding new systems of a different design, which might
introduce new failure modes or accident sequences.
Therefore, the proposed change does not create the possibility
of a new or different kind of accident from any previously
evaluated.
Does the change involve a significant reduction in a margin of
safety?
The power uprate analysis for LCS, Units 1 and 2 assures that
the power dependent safety margin will be maintained by meeting the
appropriate regulatory criteria as prescribed by the applicable
regulations. Similarly, factors of safety specified by application
of the regulatory required design rules have been maintained, as
have other acceptance criteria used to judge the acceptability of
current plant operation.
No change is required in the basic duel deign to achieve the
uprated power levels, or to maintain current operating and safety
margins. No increase in the allowable peak bundle power is requested
as a result of operation at uprated conditions. The abnormal
transients have been evaluated for a representative core
configuration and confirmed that operation at uprated conditions
does not have an adverse effect on the operating limit MCPR. 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 related to heat generation rate would still be met
at uprated conditions. Cycle specific analysis will continue to be
performed for each fuel reload to demonstrate the compliance with
the applicable transient analysis criteria and to establish the
cycle specific safety limit and fuel operating limits.
The Emergency Core Cooling System (ECCS)-LOCA performance has
been evaluated at power uprated conditions using methodologies that
have been approved by the NRC for 10CFR50.46, ``Acceptance Criteria
for Emergency Core Cooling Systems for Light-Water Nuclear Power
Reactors,'' analysis. The current ECCS performance requirements were
used in the power uprate analysis. The ECCS-LOCA analysis was
conducted at 102% of the proposed uprated thermal power in
accordance with regulatory guidance. The necessary analysis for
operation of General Electric (GE) fuel under uprated conditions and
the determination that the peak cladding temperature (PCT) remains
below the 10CFR50.46 limit of 2200 deg.F have been performed.
However, LCS Unit 2 currently contains a mixed core of GE and
Siemens Power Corporation (SPC) fuel. LCS obtained [a] TS amendment
that allows operation with SPC fuel, and approved the use of the SPC
analytical methodology. The ECCS-LOCA analysis performed to support
use of the SPC fuel was conducted at a power level that bounds 102%
of the proposed uprated power level and determined that the PCT, for
SPC fuel, remains below the 10CFR50.46 limit of 2200 deg.F. The
analysis for both GE and SPC fuel types demonstrate all 10CFR50.46
criteria are met. Therefore, there is no reduction in margin with
respect to maintaining ECCS performance.
The margin of safety of the reactor coolant pressure boundary is
maintained under power uprated conditions. The design pressure of
the RPV and reactor pressure coolant pressure boundary remains at
1250 psig. The ASME B&PV Code allowable peak pressure is 1375 psig
(i.e., 110% of design value), which is the acceptance limit for
pressurization events. The limiting pressurization event is a Main
Steam Isolation Valve (MSIV) closure with a failure of valve
position scram and this event results in a calculated peak RPV
pressure of 1332 psig at the bottom of the RPV. The peak pressure
remains below the 1375 psig ASME limit. Therefore, there is no
decrease in margin of safety in the reactor coolant pressure
boundary.
The margin of safety of the containment structure is maintained
under power uprated conditions. The analyses were conducted using a
newer NRC-reviewed methodology. The pre-uprated cases were run using
the new methodology and the re-baselined cases were compared to the
uprated cases. The short-term containment peak pressure analysis re-
baseline result was 39.3 psig compared to the original analysis of
39.6 psig. At uprated conditions the peak containment drywell
pressure would be 39.9 psig, and is below the design value of 45
psig. The long-term containment suppression pool temperature
analysis re-baseline result was 190 deg.F compared to the original
analysis result of 200 deg.F. At uprated conditions the analysis
concluded that in the event of a LOCA, the calculated peak bulk
suppression pool temperature would be 193 deg.F. This is less than
the design temperature of the suppression pool of 275 deg.F, and the
criteria used to ensure adequate Net Positive Suction Head (NPSH) to
the ECCS pumps which is 212 deg.F. Therefore, power uprate does not
challenge the structural integrity of the containment structure and
ECCS NPSH is assured.
Therefore, operation at power uprated conditions does not
involve a significant reduction in the margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
requested amendments involve no significant hazards consideration.
Local Public Document Room location: Jacobs Memorial Library, 815
North Orlando Smith Avenue, Illinois Valley Community College, Oglesby,
Illinois 61348-9692
Attorney for licensee: Ms. Pamela B. Stroebel, Senior Vice
President and General Counsel, Commonwealth Edison Company, P.O. Box
767, Chicago, Illinois 60690-0767
NRC Section Chief: Anthony J. Mendiola.
Commonwealth Edison Company, Docket Nos. 50-373 and 50-374, LaSalle
County Station, Units 1 and 2, LaSalle County, Illinois
Date of amendment request: August 6, 1999
Description of amendment request: The proposed amendments would
revise Technical Specification 3/4.6.4, ``Vacuum Relief'' to remove
specific operability requirements related to position indication for
the suppression chamber-drywell vacuum breakers. The amendments also
reformat the action statements for inoperable vacuum breakers, increase
the surveillance
[[Page 46429]]
interval for verifying that the vacuum breakers are closed, and delete
the requirement to verify that the manual isolation valves are closed
for an inoperable and open vacuum breaker.
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:
Do the proposed changes involve a significant increase in the
probability or consequences of an accident previously evaluated?
The proposed changes do not change the hardware configuration of
the suppression chamber-drywell vacuum breakers, and the vacuum
breakers are not considered an initiator in any accident scenario.
The removal of specific indication requirements and the extension of
the surveillance interval does not impact the ability of the vacuum
breakers to perform their safety function. The vacuum breakers
continue to meet their intended design function. The proposed
changes do not impact the assumed source term for any analyzed
accident. Therefore, no increases in the probability of an accident
or consequences will result due to this proposed change.
Do the proposed changes create the possibility of a new or
different kind of accident from any accident previously evaluated?
The proposed changes do not involve any physical alterations to
the suppression chamber-drywell vacuum breakers, or cause any
changes in the method by which the vacuum breakers or the
containment vacuum relief system performs their associated design
basis functions. Therefore, the proposed changes do not create the
possibility of a new or different kind of accident from any accident
previously evaluated.
Do the proposed changes involve a significant reduction in a
margin of safety?
The proposed changes do not impact the design function assumed
for the containment vacuum relief system. The proposed changes do
not require the vacuum breakers to operate in a condition not
previously assumed in the facility accident analysis. The
containment vacuum relief system will continue to operate and
provide the protection assumed in the accident analysis. In order to
limit bypass, the vacuum breakers are in a normally closed position.
These vacuum breakers cannot be permanently placed in the open
position. The proposed decrease in the surveillance frequency
verifying the closed vacuum breakers will not increase the risk of
the vacuum breakers being in the open position, since they will only
open in response to a pressure differential or manual cycling.
Therefore, the assurance of the operability of the containment
vacuum breakers would be the same as provided under current
Technical Specifications. The containment response analysis is
unchanged, in that the vacuum breakers protect the containment
structure, the peak containment pressure remains as calculated, and
the vacuum breakers continue to maintain bypass leakage rates as
assumed. Therefore this proposed change does not cause 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
requested amendments involve no significant hazards consideration.
Local Public Document Room location: Jacobs Memorial Library, 815
North Orlando Smith Avenue, Illinois Valley Community College, Oglesby,
Illinois 61348-9692.
Attorney for licensee: Ms. Pamela B. Stroebel, Senior Vice
President and General Counsel, Commonwealth Edison Company, P.O. Box
767, Chicago, Illinois 60690-0767.
NRC Section Chief: Anthony J. Mendiola.
Commonwealth Edison Company, Docket Nos. 50-254 and 50-265, Quad Cities
Nuclear Power Station, Units 1 and 2, Rock Island County, Illinois
Date of amendment request: July 16, 1999.
Description of amendment request: The proposed change to Technical
Specification Section 3/4.7.D is to eliminate the limit for any one
main steam line isolation valve (MSIV) leakage of less than or equal to
11.5 standard cubic feet per hour (scfh), and to replace that with an
aggregate value of less than or equal to 46 scfh for all four MSIVs.
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:
Does the change involve a significant increase in the
probability or consequences of an accident previously evaluated?
The proposed changes to the Technical Specifications, Appendix
A, modifies the allowed leakage limit to an aggregate value with no
change to the total allowed leakage rate. This change does not
affect either the automatic or manual features that would close the
MSIVs. There are no physical changes to the plant and plant
operations remain unchanged. Therefore, this proposed amendment does
not involve a significant increase in the probability or
consequences of an accident previously evaluated.
Does the change create the possibility of a new or different
kind of accident from any accident previously evaluated?
The safety function of the MSIVs is to provide a timely steam
line isolation to mitigate the release of radioactive steam and
limit reactor inventory loss under certain accident and transient
conditions. The MSIVs are designed to automatically close whenever
plant conditions warrant main steam line isolation. Changing the
leakage limits to include an aggregate value does not affect the
isolation function. No new equipment will be installed or utilized,
and no new operating conditions will be initiated as a result of
this change. Therefore, the proposed change does not create the
possibility of a new or different kind of accident from any
previously evaluated.
Does the change involve a significant reduction in a margin of
safety?
The total allowed leakage rate for all MSIVs remains unchanged
at 46 scfh. Therefore, there will be no change in the types or
significant increase in the amounts of any effluents released
offsite, and, thus, the radiological analyses remain unchanged and
within the guidelines of 10 CFR 100 and General Design Criteria 19.
Therefore, these changes do not involve a significant reduction in
the margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
requested amendments involve no significant hazards consideration.
Local Public Document Room location: Dixon Public Library, 221
Hennepin Avenue, Dixon, Illinois 61021.
Attorney for licensee: Ms. Pamela B. Stroebel, Senior Vice
President and General Counsel, Commonwealth Edison Company, P.O. Box
767, Chicago, Illinois 60690-0767.
NRC Section Chief: Anthony J. Mendiola.
Duke Energy Corporation, Docket Nos. 50-269, 50-270, and 50-287, Oconee
Nuclear Station, Units 1, 2, and 3, Oconee County, South Carolina
Date of amendment request: July 27, 1999.
Description of amendment request: The proposed amendments would add
a surveillance requirement to verify the Keowee out-of-tolerance logic
trips and blocks closure of the appropriate overhead or underground
power path breakers. This logic is being added as part of a
modification to provide voltage and frequency protection for the Keowee
Hydro Units to protect them from being exposed to out-of-tolerance
voltage and frequency.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
[[Page 46430]]
1. Involve a significant increase in the probability or
consequences of an accident previously evaluated:
This change does not create any conditions or events, which lead
to accidents previously, evaluated in the SAR. The Keowee Hydro
units are used for mitigation of loss of power scenarios. The
proposed changes do not change the current function of the Keowee
Hydro Units. Therefore, the proposed change does not involve a
significant increase in the probability of an accident previously
evaluated. The Keowee Hydro units and their role in the Oconee
emergency power system currently meet the design/licensing basis
requirements for the system. There is no adverse affect on
containment integrity and no new release paths are created. The
proposed changes do not cause any adverse effects to the Keowee
single failure design or adversely affect the Keowee start time of
23 seconds. Therefore, the proposed changes do not involve a
significant increase in the consequences of an accident previously
evaluated.
2. Create the possibility of a new or different kind of accident
from any accident previously evaluated:
The Keowee Hydro units are used for mitigation of loss of power
scenarios. No accidents new or different than already evaluated in
the SAR are postulated as a result of the proposed change. No
setpoints for parameters, which initiate protective or mitigative
action, are being changed. Therefore, this proposed amendment does
not create the possibility of any new or different kind of accident.
3. Involve a significant reduction in a margin of safety:
The proposed change does not adversely affect any plant safety
limits, set points, or design parameters. The change also does not
adversely affect the fuel, fuel cladding, Reactor Coolant System, or
containment integrity. Therefore, the proposed change does not
involve a significant reduction in a margin of safety.
Duke has concluded, based on the above, that there are no
significant hazards considerations involved in this amendment
request.
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: Oconee County Library, 501
West South Broad Street, Walhalla, South Carolina.
Attorney for licensee: Anne W. Cottington, Winston and Strawn, 1200
17th Street, NW., Washington, DC.
NRC Section Chief: Richard L. Emch, Jr.
Energy Northwest, (formerly known as the Washington Public Power Supply
System), Docket No. 50-397, WNP-2, Benton County, Washington
Date of amendment request: July 29, 1999.
Description of amendment request: The proposed amendment would
change the applicability of Section 3.4.9 of the Technical
Specifications (TS) from ``Mode 3 with steam drum pressure less than
the RHR [residual heat removal] cut in permissive'' to ``Mode 3 with
steam drum pressure less than 48 psig.'' Notes associated with TS
Surveillance Requirements 3.4.9.1 and 3.5.1.2 would be changed to
reflect the proposed 48 psig limit.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. The proposed change does not involve a significant increase
in the probability or consequences of an accident previously
evaluated.
This change involves further restrictions on the use of RHR in
the shutdown cooling mode of operation during hot shutdown
conditions. Chapter 15 of the FSAR [Final Safety Analysis Report]
defines the start of hot shutdown as the point when generated power
is below one percent rated power. During entry into hot shutdown
conditions the RHR system will be aligned in the Low Pressure
Coolant Injection (LPCI) mode of operation. Thus, it will be aligned
to provide water to the Reactor Pressure Vessel in the event the
high pressure systems (HPCS and RCIC) are not able to perform this
function. The change being proposed here has no impact on loss of
coolant accidents (LOCAs) requiring mitigation using RHR aligned in
the LPCI mode of operation.
During the high pressure portion of the hot shutdown condition,
intersystem (LOCAs) are a concern. The purpose of the RHR SDC
Isolation Reactor Pressure--High (cut-in permissive) at 135 psig is
to prevent over-pressurization of portions of the RHR system. This
protection is not being modified by this change. The instrumentation
that provides this protection will continue to function as designed.
This change only impacts the applicability of Technical
Specification 3.4.9 and when RHR SDC is required to be operable.
During hot shutdown the reactor is normally cooled down through
use of the main steam system and the condenser. Other means of
cooling are also available using the reactor water cleanup system or
a combination of emergency core cooling system (ECCS) pumps and
safety relief valves (SRVs). The RHR system aligned in the SDC mode
is used at the end of this cooling process to reach cold shutdown
conditions of less than or equal to 200 deg.F. The change being
proposed results in the RHR SDC being manually initiated at a lower
pressure and temperature. This change will have no significant
impact on the capability to cool the reactor.
FSAR Chapter 15, ``Accident Analysis,'' describes two events
associated with the RHR system. FSAR section 15.1.6, ``Inadvertent
Residual Heat Removal Shutdown Cooling Operation,'' describes the
impact of system operation during startup or cool-down when the
reactor is near critical. The proposed change involves the point at
which RHR is started in the SDC mode with the reactor sub-critical
with control rods inserted. Therefore, there will be no change in
the probability or consequences of this accident.
FSAR section 15.2.9, ``Failure of Residual Heat Removal Shutdown
Cooling,'' describes the failure of the RHR system to function in
SDC mode. This evaluation assumes a failure of the SDC mode of
operation but does not disable the remaining modes of RHR operation.
The alternate shutdown cooling paths involve the use of the SRVs
[safety relief valves] to establish a cooling path to the
containment suppression pool. This evaluated accident does not
result in any fuel failure. The proposed change will not result in
any fuel failures. The evaluated accident does result in normal
coolant activity being released to the suppression pool through the
safety relief valves. The proposed activity will not result in a
significant change in the release of this coolant activity.
The proposed change will not cause a significant increase in the
probability of a loss of SDC accident. This change proposes a delay
in the use of SDC because of temperature limitations. During this
time other means of decay heat removal would be used. This will
result in a decrease in use of RHR in SDC mode and a decrease in the
probability of failure of the system by restricting operation to be
within analyzed temperature limits. The proposed change will not
involve a significant increase in the consequences of the loss of
shutdown cooling accident. The accident evaluated in the FSAR
assumes SDC does not operate at any time and alternate means of
cooling are evaluated. Section 15.2.9.6 states there is no fuel
failure and release is limited to normal primary coolant activity to
the suppression pool. The proposed change results in a short delay
in the use of SDC because of temperature limitations. The accident
described in FSAR section 15.2.9 bounds this condition and, as a
result, there will be no increase in accident consequences.
With multiple means of reactor water makeup and heat removal
available the restriction in the use of RHR caused by this change
will not result in a significant increase in the probability or
consequences of an accident previously evaluated.
2. The proposed change does not create the possibility of a new
or different kind of accident from any accident previously
evaluated.
The proposed change will not cause any new inadvertent shutdown
cooling startup, loss of water inventory or loss of cooling
accidents. New or different inadvertent RHR SDC startup accidents
are not possible because this change is only a further restriction
on when the system is operated. The LOCA accidents during Mode 3 are
[[Page 46431]]
bounded by the LOCAs defined for Modes 1 and 2. No new primary sytem
LOCAs can be initiated because of this change. The purpose of the
RHR cut-in permissive at 135 psig is to prevent overpressurization
of portions of the RHR system that could cause an intersystem LOCA.
This change will not result in a new or different kind of
intersystem LOCA because this is only a further restriction on RHR
SDC operation. The use of RHR in the SDC mode is restricted to
operation at a lower pressure and temperature but other systems are
available to remove the decay heat. No new or different accidents
are created because of this change.
The FSAR section 15.2.9 accident, ``Failure of Resident Heat
Removal Shutdown Cooling,'' is bounding for all other accidents
which postulate failure of the capability to remove decay heat. No
additional accidents resulting in the loss of decay heat removal
capability will be caused by this change.
Therefore, the operation of WNP-2 in accordance with the
proposed amendment 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 amendment will increase the reliability of the RHR
system when operated in shutdown cooling mode by providing assurance
that the temperature limits of the piping and pipe supports will not
be exceeded. The ability to protect against an intersystem LOCA is
unchanged. The ability to remove decay heat from the reactor is not
changed by this modification as alternate means of heat removal are
available. Therefore, operation of WNP-2 in accordance with the
proposed amendment will 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 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: Richland Public Library, 955
Northgate Street, Richland, Washington 99352.
Attorney for licensee: Perry D. Robinson, Esq., Winston & Strawn,
1400 L Street, N.W., Washington, D.C. 20005-3502.
NRC Section Chief: Stephen Dembek.
Energy Northwest (formerly known as the Washington Public Power Supply
System), Docket No. 50-397, WNP-2, Benton County, Washington
Date of amendment request: July 29, 1999.
Description of amendment request: The proposed amendment would
revise Technical Specification Table 3.3.5.1-1, ``Emergency Core
Cooling System (ECCS) Instrumentation Items 1.a, 2.a, 4.a and 5.a,'' to
change the Reactor Vessel Water Level--Low Low Low, Level 1 allowable
value from the current value of -148 inches to a new value of -142.3
inches.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. The proposed change does not involve a significant increase
in the probability or consequences of an accident previously
evaluated.
This change involves the measurement of water level in the
Reactor Pressure Vessel (RPV) used to initiate the ECCS. The
accident evaluated for this condition is the spectrum of loss of
coolant accidents (LOCA) severe enough to decrease the RPV water
inventory by a significant amount.
The additional uncertainty introduced because of harsh
environmental effects could not be accommodated between the existing
Technical Specification allowable value and the analytical limit.
This uncertainty results in a requirement that the ECCS be initiated
at a slightly higher water level than previously calculated.
Therefore, operation of WNP-2 in accordance with the proposed
amendment will not involve a significant increase in the probability
or consequences of an accident previously evaluated.
2. The proposed change does not create the possibility of a new
or different kind of accident from any accident previously
evaluated.
The proposed change will not create a new or different kind of
accident since it only makes a small change in the RPV water level
at which the ECCS is initiated. This change is in the conservative
direction requiring a greater volume of water in the RPV to
accommodate the uncertainty associated with the harsh environment of
the water level sensors.
The level indicating switches are located on instrument racks in
the Reactor Building. The harsh environment in this building would
have no impact on the initial trip needed to initiate the ECCS on
loss of RPV level since conditions in the Reactor Building would be
benign at the initial stages of the accident. Only if the Level 1
trip was reset and initiated after a significant period of time
would the harsh environmental conditions have an impact on the
accuracy of the level indicating switches. However, increasing the
water level at which the ECCS is initiated results in a more
conservative value that adequately includes post-accident harsh
environment uncertainties and ensures that the associated analytical
limit is met.
Therefore, the operation of WNP-2 in accordance with the
proposed amendment 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 amendment increases the allowable value for water
level in the RPV. This small increase will result in an increase in
the margin of safety. A review of the plant settings for the Level 1
trip indicated that previous settings were within the new allowable
value.
Therefore, operation of WNP-2 in accordance with the proposed
amendment will not involve a significant reduction in the margin of
safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: Richland Public Library, 955
Northgate Street, Richland, Washington 99352.
Attorney for licensee: Perry D. Robinson, Esq., Winston & Strawn,
1400 L Street, N.W., Washington, D.C. 20005-3502.
NRC Section Chief: Stephen Dembek.
Energy Northwest (formerly known as the Washington Public Power Supply
System), Docket No. 50-397, WNP-2, Benton County, Washington
Date of amendment request: July 29, 1999.
Description of amendment request: The proposed amendment would
revise Technical Specification Surveillance Requirement (SR) 3.5.2.2.
This requirement verifies the adequacy of the water supply in the
condensate storage tanks (CSTs) which support operation of the high
pressure core spray (HPCS) system during Modes 4 and 5. Current
Technical Specification SR 3.5.2.2 requires that CST water level be
maintained above 13.25 feet in a single tank or above 7.6 feet in each
tank if the suppression pool level is below its minimum level. It is
proposed that the CST water level be maintained above 14.8 feet in a
single tank or above 9.1 feet in each tank.
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.
During Modes 4 and 5 HPCS may be required to provide water to
the reactor vessel if the water level decreases. The revised
condensate storage tank allowable levels increase the operating
margins by providing an increased water inventory. The previously
evaluated accident involving the loss of decay heat cooling
inventory will not have an increase in probability because the
inventory of water will be increased with the change being proposed.
[[Page 46432]]
The consequences of any accident involving the loss of decay
heat cooling inventory will not change as the consequences are
unaffected by the increased water inventory.
Therefore, operation of WNP-2 in accordance with the proposed
amendment will not involve a significant increase in the probability
or consequences of an accident previously evaluated.
2. The proposed change does not create the possibility of a new
or different kind of accident from any accident previously
evaluated.
The proposed change will not create a new or different kind of
accident since it only increases the amount of water held in reserve
to support reactor vessel inventory loss. The proposed change does
not introduce any credible mechanisms for unacceptable radiation
release nor does it require physical modification to the plant. The
inventory of water in the CSTs will increase to support any loss of
water inventory in the reactor vessel during shutdown.
The proposed change modifies the monitored values for CST level.
The plant has operated well within the existing allowable values.
The increased margin provided by the increased level will assure no
new or different kinds of accidents result from the proposed change.
Therefore, the operation of WNP-2 in accordance with the
proposed amendment 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 amendment increases the allowable value for water
level in the CSTs. This results in an increase in the inventory of
water available for cooling and inventory control during reactor
shutdown. This will result in an increase in the margin of safety.
Therefore, operation of WNP-2 in accordance with the proposed
amendment will not involve a significant reduction in the margin of
safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: Richland Public Library, 955
Northgate Street, Richland, Washington 99352.
Attorney for licensee: Perry D. Robinson, Esq., Winston & Strawn,
1400 L Street, N.W., Washington, D.C. 20005-3502.
NRC Section Chief: Stephen Dembek.
Energy Northwest (formerly known as the Washington Public Power Supply
System), Docket No. 50-397, WNP-2, Benton County, Washington
Date of amendment request: July 29, 1999.
Description of amendment request: The proposed amendment request
would revise Technical Specification Surveillance Requirement (SR) SR
3.8.4.6 of Technical Specification 3.8.4, ``DC Sources--Operating,''
and SR 3.8.5.1 of Technical Specification 3.8.5, ``DC Sources--
Shutdown.'' The proposed change to SR 3.8.4.6 would prohibit
surveillance testing of Division 1, 2, and 3 125 and 250 volt DC,
battery charger capacity during Modes 1, 2, and 3. However, credit
could be taken for unplanned events that satisfied the surveillance
requirement. The proposed change to SR 3.8.5.1 would include SR 3.8.4.6
as one of the surveillance tests that are not required to be performed.
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 has no impact on previously analyzed
accidents or transients, and has no effect on operation, capacity or
surveillance test details of the DC system battery chargers. The
change only imposes a mode restriction on performance of specified
surveillance testing and allows taking credit for unplanned events
that satisfy the surveillance. Therefore, operation of WNP-2 in
accordance with the proposed amendment will not involve a
significant increase in the probability or consequences of an
accident previously evaluated.
2. The proposed change does not create the possibility of a new
or different kind of accident from any accident previously
evaluated.
The proposed change has no effect on operation, capacity, or
surveillance test details of the DC system battery chargers. The
change only prohibits performing specified battery charger capacity
surveillance testing from being implemented during Mode 1, 2, or 3
and allows taking credit for unplanned events that satisfy the
surveillance. The proposed change to SR 3.8.4.6 of Technical
Specification 3.8.4 and SR 3.8.5.1 of Technical Specification 3.8.5
are consistent with the wording previously evaluated and approved by
the NRC in NUREG-1434 Rev. 1.
Therefore, operation of WNP-2 in accordance with the proposed
amendment will not create the possibility of a new or different kind
of accident previously evaluated.
3. The proposed change does not involve a significant reduction
in a margin of safety.
The proposed change only imposes a mode restriction, prohibiting
battery charger capacity surveillance testing from being performed
during Modes 1, 2, and 3, allowing credit to be taken for unplanned
events that satisfy the surveillance, and allowing such testing to
be omitted under certain conditions during Modes 4 and 5 and during
movement of irradiated fuel in secondary containment. Performance of
this testing would remove a DC electrical power subsystem from
service and could present a safety risk were an event to occur if
the testing was performed in Modes 1, 2, and 3, or while DC service
is required in other operating conditions. Therefore, operation of
WNP-2 in accordance with the proposed amendment will not involve a
significant reduction in the margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: Richland Public Library, 955
Northgate Street, Richland, Washington 99352.
Attorney for licensee: Perry D. Robinson, Esq., Winston & Strawn,
1400 L Street, N.W., Washington, D.C. 20005-3502.
NRC Section Chief: Stephen Dembek.
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 amendment request: July 20, 1998, as supplemented June 29,
1999.
Description of amendment request: The amendment would incorporate
the Technical Specification changes necessary for implementation of the
Boiling Water Reactor Owners' Group Reactor Stability Long-Term
Solution, Enhanced Option 1-A (E1A). E1A consists of modifications to
the plant operating procedures and associated plant components that
provide a means for reliably detecting and avoiding reactor
instabilities. By letter dated February 25, 1998, the Nuclear
Regulatory Commission (NRC) staff recognized E1A as a technically
acceptable implementation of a long-term stability solution satisfying
the requirements of NRC IE Bulletin 88-07, Supplement 1, and Generic
Letter 94-02, ``Long Term Solutions and Upgrade of Interim Operating
Recommendations for Thermal-Hydraulic Instabilities in Boiling Water
Reactors.''
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
[[Page 46433]]
1. This request does not involve a significant increase in the
probability or consequences of an accident previously evaluated.
The proposed amendments allow the implementation of the Enhanced
Option I-A (E1A) long term solution to the neutronic/thermal-
hydraulic instability issue. Current Technical Specification (TS)
restrictions on power and flow conditions, number of operating
recirculation loops and operator actions implemented to reduce the
probability of neutronic/thermal-hydraulic instability are
eliminated and new stability requirements consistent with NEDO-
32339-A, Supplement 4, Revision 1, are imposed. These requirements
include restrictions on power and flow conditions and actions
associated with the modified Average Power Range Monitor (APRM) flow
biased scram and control rod block functions. Required actions
include adherence to the boiling boundary limit stability control
prior to entry and during operation in the region of the power and
flow operating domain which is potentially susceptible to neutronic/
thermal-hydraulic instability in the absence of the stability
control. In addition, the proposed amendments require operator
actions based upon control room indications generated by a new
Period Based Detection System (PBDS). The PBDS is designed to
provide alarm indication that conditions consistent with a
significant degradation in the stability performance of the reactor
has occurred and the potential for imminent onset of neutronic/
thermal-hydraulic instability may exist. The PBDS also provides
analog indication of the highest and second highest successive
period confirmation count of all of the Local Power Range Monitors
(LPRMs) monitored. This provides the plant operators with continuous
indication of reactor stability operating conditions.
The proposed amendments will permit operation in regions of the
power and flow operating domain postulated to be susceptible to
neutronic/thermal-hydraulic instability. Operation in these regions
does not increase the probability of occurrence of initiators and
precursors of previously analyzed accidents when neutronic/thermal-
hydraulic instability is not possible. The proposed amendments
permit the implementation of the features of the E1A solution which
prevent neutronic/thermal-hydraulic instability including preemptive
reactor scram upon entry into the regions of the power and flow
operating domain most susceptible to neutronic/thermal-hydraulic
instability. The E1A solution also requires implementation of
stability control prior to entry into a region of the power and flow
operating domain which is potentially susceptible, in the absence of
stability control, to neutronic/thermal-hydraulic instability. The
E1A solution prevents neutronic/thermal-hydraulic instability during
operation in regions of the power and flow operating domain
previously excluded from operation and therefore does not
significantly increase the probability of a previously analyzed
accident.
Operation in the regions of the power and flow operating domain
excluded by current TS 3.4.1 and Figure 3.4.1-1 can occur as a
result of anticipated operational occurrences. The severity of these
transients may increase in the absence of operator actions due to
the potential occurrence of neutronic/thermal-hydraulic instability
as a result of operation in these regions. The proposed amendments
will permit the implementation of the E1A long term solution to the
stability issue. Required features of the E1A solution include
adherence to a boiling boundary limit stability control prior to
selection by the operator of APRM flow biased scram and control rod
block function ``Setup'' setpoints which allow operation in a region
of the power and flow operating domain potentially susceptible, in
the absence of the stability control, to neutronic/thermal-hydraulic
instability. Upon entry, as a result of an anticipated operational
occurrence, into the region most susceptible to neutronic/thermal-
hydraulic instability, the preemptive reactor scram prevents
neutronic/thermal-hydraulic instability. Therefore, the consequences
of an accident do not significantly increase while operating with
the stability control met.
After exiting the region requiring the stability control to be
met, the setpoints can be manually reset to their normal values.
Stability controls are required to be in place when setpoints are
``Setup''. As a backup E1A feature, the APRM flow biased setpoints
automatically reset to their normal values above a pre-determined
flow condition. This automatic reset to the more conservative
setpoints ensures that the preemptive reactor scram will prevent
operation as a result of an anticipated operational occurrence into
the region most susceptible to neutronic/thermal-hydraulic
instability should the operator not select the more conservative
setpoints appropriate for operation following exit from the region
requiring stability control.
Other required E1A features, including the PBDS, control rod
block alarms associated with entry into the region susceptible to
neutronic/thermal-hydraulic instabilities in the absence of
stability controls, and required operator actions, including manual
reactor scram, help ensure prevention of neutronic/thermal-hydraulic
instabilities. Therefore, the proposed amendments prevent the
occurrence of neutronic/thermal-hydraulic instability as a
consequence of an anticipated operational occurrence and do not
significantly increase the consequences of any previously analyzed
accident.
2. This request does not create the possibility of a new or
different kind of accident from any accident previously evaluated.
The proposed amendments replace current restrictions on power
and flow conditions with alternative restrictions which permit the
implementation of the E1A long term stability solution. The current
restrictions on the power and flow conditions and operating
recirculation loops in the RUN mode do not automatically prevent the
entry into regions of the power and flow operating domain most
susceptible to neutronic/thermal-hydraulic instability and therefore
the possibility of neutronic/thermal-hydraulic instability exists in
the absence of operator action. The required features of the E1A
solution implement a preemptive scram upon entry into the region
most susceptible to neutronic/thermal-hydraulic instability, without
operator action. The accessible operating domain allowed by the
proposed amendments is a subset of the power and flow operating
domain currently allowed. Current initiators and precursors of
accidents and anticipated operational occurrences [cannot] occur
with new or different initial conditions as a result of this change.
Additionally, there are no new event initiators or precursors of
accidents and anticipated operational occurrences created by this
change. Therefore, the proposed amendments do not create the
possibility of a new or different kind of accident from that
previously evaluated.
Concurrent with the implementation of the proposed amendments, a
modified Flow Control Trip Reference (FCTR) card, the E1A FCTR card,
and a new Period Based Detection System (PBDS) will be installed as
required by the E1A solution. The function of the E1A FCTR card is
to aid the operator in the identification of entry into regions of
the power and flow operating domain potentially susceptible to
neutronic/thermal-hydraulic instability in the absence of stability
controls and to initiate a preemptive scram upon entry into the
regions most susceptible to neutronic/thermal-hydraulic instability.
This is accomplished by altering the existing values of setpoints of
the APRM flow biased scram and the control rod block functions
generated by the E1A FCTR card. The E1A FCTR card design includes
components which may be susceptible to electromagnetic interference
or other environmental effects. The plant specific environmental
conditions (temperature, humidity, pressure, seismic, and
electromagnetic compatibility) have been confirmed to be enveloped
by the environmental qualification values for the E1A FCTR cards.
Therefore, the potential for spurious scrams or common mode failures
induced by environmental effects (e.g., electromagnetic
interference) is considered negligible. The installation of the E1A
FCTR card will therefore not create the possibility of a new or
different kind of accident from any accident previously evaluated.
The function of the PBDS is to provide the operator with an
indication that conditions consistent with a significant degradation
in the stability performance of the reactor has occurred and the
potential for imminent onset of neutronic/thermal-hydraulic
instability may exist. This is accomplished by the installation of a
new PBDS card in the Neutron Monitoring System. The PBDS card takes
inputs from individual local power range monitors and provides
analog indication of the highest and second highest successive
period confirmation count, provides a High Decay Ratio (Hi DR) and
High-High Decay Ratio (Hi-Hi DR) alarms, and INOP status indication
to the operator in the control room. These displays [cannot] create
the possibility of a new or different kind of accident from any
accident previously evaluated. The PBDS card design includes
components which may be susceptible to electromagnetic interference
or other environmental effects. However, the plant specific
environmental conditions (temperature, humidity, pressure, seismic,
[[Page 46434]]
and electromagnetic compatibility) have been confirmed to be
enveloped by the PBDS environmental qualification values. Therefore,
the installation of the PBDS card will not 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 to safety.
The proposed amendments permit the implementation of the E1A
long term solution to the stability issue. Under certain conditions,
existing BWR [boiling water reactor] designs are susceptible to
neutronic/thermal-hydraulic instability. General Design Criterion
(GDC) 12 of 10 CFR 50, Appendix A, requires thermal-hydraulic
instability to be prevented by design or be readily and reliably
detected and suppressed. When the design of the reactor system does
not prevent the occurrence of neutronic/thermal-hydraulic
instability, instability is an anticipated operational occurrence.
GDC 10 of 10 CFR 50, Appendix A, requires that specified acceptable
fuel design limits not be exceeded during anticipated operational
occurrences.
Analyses performed by the BWROG [Boiling Water Reactor Owners'
Group] indicate that neutronic/thermal-hydraulic instability induced
power oscillations could result in conditions exceeding the Minimum
Critical Power Ratio (MCPR) Safety Limit (SL) prior to detection and
suppression by the current design of the Neutron Monitoring System
and Reactor Protection System.
To ensure compliance with GDC 12 the BWROG developed Interim
Corrective Actions (ICAs) to enhance the capability of the operator
to readily and reliably detect and suppress neutronic/thermal-
hydraulic instability. The BWROG ICAs also provided additional
guidance for monitoring local power range monitors beyond the
requirements of current TS 3.4.1 to ensure adequate margin to the
onset of neutronic/thermal-hydraulic instability. Reliance on
operator actions to comply with GDC 12 was accepted on an interim
basis by the NRC pending final implementation of a long term
solution to the stability issue. Neutronic/thermal-hydraulic
instability is prevented by implementation of the E1A solution
through the modified design of the Reactor Protection System (APRM
[average power range monitor] flow biased scram) and the stability
control prior to entry into a region of the power and flow operating
domain which is potentially susceptible, in the absence of stability
control, to neutronic/thermal-hydraulic instability. In addition,
significant backup protection features, including the PBDS, control
rod block alarms associated with entry into the region susceptible
to neutronic/thermal-hydraulic instabilities in the absence of
stability controls, and specified operator actions, including manual
reactor scram, are required to be implemented. As a result, the
margin to the onset of neutronic/thermal-hydraulic instability
provided by the existing TS requirements and BWROG ICAs
recommendations is not significantly reduced by the implementation
of the E1A solution. The E1A solution assures compliance with GDC 12
by the prevention of neutronic/thermal-hydraulic instability and
therefore precludes neutronic/thermal-hydraulic instability from
becoming a credible consequence of an anticipated operational
occurrence. The consequences of anticipated operational occurrences
will not increase and the margin to the MCPR SL will not decrease
upon implementation of the E1A solution. Therefore, the proposed
amendments do not involve a significant reduction in a margin of
safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room Location: Judge George W. Armstrong
Library, 220 S. Commerce Street, Natchez, Mississippi 39120
Attorney for licensee: Nicholas S. Reynolds, Esquire, Winston and
Strawn, 1400 L Street, NW., 12th Floor, Washington, DC 20005-3502.
NRC Section Chief: Robert A. Gramm.
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 amendment request: May 6, 1999.
Description of amendment request: The proposed amendments would
change those Technical Specifications (TS) required to support Grand
Gulf Nuclear Station (GGNS), Cycle 11 operation. The changes would
include a change to the minimum critical power ratio safety limit
(SLMCPR) that would reflect a decrease of the two recirculation loop
SLMCPR limit from 1.11 to 1.09, and the single recirculation loop
SLMCPR limit from 1.12 to 1.10. These values were developed with
General Electric's cycle-specific SLMCPR methodology in GESTAR-II
Amendment 25, which was recently approved by the Nuclear Regulatory
Commission in a Safety Evaluation Report dated March 11, 1999.
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:
I. The proposed change does not significantly increase the
probability or consequences of an accident previously evaluated.
The Minimum Critical Power Ratio (MCPR) safety limit is defined
in the Bases to Technical Specification 2.1.1 as that limit which
``ensures that during normal operation and during Anticipated
Operational Occurrences (AOOs), at least 99.9% of the fuel rods in
the core do not experience transition boiling.'' The MCPR safety
limit is re-evaluated for each reload and, for GGNS Cycle 11, the
analyses have concluded that a two-loop MCPR safety limit of 1.09,
based on the application of GE's [General Electric's] NRC-approved
cycle-specific MCPR safety limit methodology demonstrates that this
acceptance criterion is satisfied. For single-loop operation, a MCPR
safety limit of 1.10, based on GE's [NRC-approved cycle-specific
MCPR safety limit methodology, also demonstrates that this
acceptance criterion is satisfied. Core MCPR operating limits are
developed to support the Technical Specification 3.2 requirements
and ensure these safety limits are maintained in the event of the
worst-case transient. Since the MCPR safety limit will be maintained
at all times, operation under the proposed changes will ensure at
least 99.9% of the fuel rods in the core do not experience
transition boiling. Therefore, these changes to the Minimum Critical
Power Ratio (MCPR) safety limit do not affect the probability or
consequences of an accident.
GE's NRC-approved GESTAR-II cycle-specific MCPR safety limit
methodology has been applied and has no effect on the probability or
consequences of any accidents previously evaluated. As previously
licensed, one exception to GESTAR is that the mis-oriented and mis-
located bundle events will continue to be analyzed as accidents
subject to the acceptance criteria in the current licensing basis.
The design of the GE11 fuel bundles is such that the bundles are not
likely to be mis-oriented or mis-located and the normal
administrative controls will be in effect for assuring proper
orientation and location. Therefore, the probability of a fuel
loading error is not increased. This analysis ensures that
postulated dose releases will not exceed a small fraction (10
percent) of 10CFR100 limits. Therefore, the probability or
consequences of accidents previously evaluated are unchanged.
II. The proposed change does not create the possibility of a new
or different kind of accident from any accident previously
evaluated.
The GE11 fuel to be used in Cycle 11 is of a design compatible
with fuel present in the core and used in the previous cycle.
Therefore, the GE11 fuel will not create the possibility of a new or
different kind of accident. The proposed changes do not involve any
new modes of operation, any changes to setpoints, or any plant
modifications. The proposed revised MCPR safety limits have been
shown to be acceptable for Cycle 11 operation. Compliance with the
applicable criterion for incipient boiling transition continues to
be ensured. The proposed MCPR safety limits do not result in the
creation of any new precursors to an accident.
Therefore, the proposed changes do not create the possibility of
a new or different type of accident from any accident previously
evaluated.
III. The proposed change does not involve a significant
reduction in a margin of safety.
[[Page 46435]]
The MCPR safety limits have been evaluated in accordance with
GE's NRC-approved cycle-specific methodology to ensure that during
normal operation and during AOOs, at least 99.9% of the fuel rods in
the core are not expected to experience transition boiling. One
exception to GESTAR is that the mis-oriented and mis-located bundle
events will continue to be analyzed as accidents subject to the
acceptance criteria in the current licensing basis. This analysis
ensures that postulated dose releases for the worst case mis-
oriented and mis-located bundle will not exceed a small fraction (10
percent) of 10CFR100 limits. On this basis, the implementation of
this GE methodology 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: Judge George W. Armstrong
Library, 220 S. Commerce Street, Natchez, Mississippi 39120.
Attorney for licensee: Nicholas S. Reynolds, Esquire, Winston and
Strawn, 1400 L Street, NW., 12th Floor, Washington, DC 20005-3502.
NRC Section Chief: Robert A. Gramm.
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 amendment request: June 23, 1999.
Description of amendment request: The requested Technical
Specification changes would revise those specifications associated with
various engineered safety feature systems, which need no longer be
credited following a design-basis fuel handling accident. The proposed
changes affect conditions where irradiated fuel is handled in the
primary or secondary containment, and also affect certain
specifications related to performing core alterations. These changes
are based on the revised analysis of the design-basis fuel handling
accident for the Grand Gulf Nuclear Station. This requested change is
consistent with the changes approved for the Perry Nuclear Power Plant
Operating License (Amendment 102), and the industry-proposed change to
the Technical Specification NUREGs, TSTF-51.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. The proposed changes do not significantly increase the
probability or consequences of an accident previously evaluated.
A new term to describe irradiated fuel is used to establish
operational conditions where specific activities represent
situations where significant radioactive releases can be postulated.
These operational conditions are consistent with the design basis
analysis. Because the equipment affected by the revised operational
conditions is not considered an initiator to any previously analyzed
accident, inoperability of the equipment cannot increase the
probability of any previously evaluated accident. The proposed
requirements bound the conditions of the current design basis fuel
handling accident analysis which concludes that the radiological
consequences are within the acceptance criteria of NUREG 0800,
Section 15.7.4 and General Design Criteria 19. Therefore, the
proposed changes do not significantly increase the probability or
consequences of any previously evaluated accident.
Removing a one time only allowance granted by Amendment 129 to
the Operating License that is no longer in affect is an
administrative change. Therefore, the proposed change does not
significantly increase the probability or consequences of any
previously evaluated accident.
Based on the above, neither the proposed changes to the
Technical Specifications nor that to the Operating License
significantly increase the probability or consequences of any
accident previously evaluated.
2. The proposed changes would not create the possibility of a
new or different kind of accident from any previous analyzed.
The new term to describe irradiated fuel is used to establish
operational conditions where specific activities represent
situations where significant radioactive releases can be postulated.
These operational conditions are consistent with the design basis
analysis. The proposed changes do not introduce any new modes of
plant operation and do not involve physical modifications to the
plant. Therefore, the proposed changes do not create the possibility
of a new or different kind of accident from any previous analyzed.
Removing a one time only allowance granted by Amendment 129 to
the Operating License that is no longer in affect is an
administrative change. Therefore, the proposed change does not
create the possibility of a new or different kind of accident from
any previous analyzed.
Based on the above, neither the proposed changes to the
Technical Specifications nor that to the Operating License create
the possibility of a new or different kind of accident from any
accident previously analyzed.
3. The proposed changes do not involve a significant reduction
in a margin of safety.
The new term to describe irradiated fuel is used to establish
operational conditions where specific activities represent
situations where significant radioactive releases can be postulated.
These operational conditions are consistent with the design basis
analysis and are established such that the radiological consequences
are at or below the current GGNS [Grand Gulf Nuclear Station]
licensing limit. Safety margins and analytical conservatisms have
been evaluated and are well understood. Substantial margins are
retained to ensure that the analysis adequately bounds all
postulated event scenarios. The proposed change only eliminates the
unnecessary margin from the analysis. The current margin of safety
is retained.
Specifically, the margin of safety for the fuel handling
accident is the difference between the 10CFR100 limits and the
licensing limit defined by NUREG 0800, Section 15.7.4. With respect
to the control room personnel doses, the margin of safety is the
difference between the 10CFR100 limits and the licensing limit
defined by 10CFR50, Appendix A, Criterion 19 (GDC 19). The
additional margin between the calculated doses for the postulated
events and the corresponding licensing limit provides no useful
purpose.
The proposed applicability continues to ensure that the whole-
body and thyroid doses at both the control room and the exclusion
area and low population zone boundaries are at or below the
corresponding licensing limit. The margin of safety is unchanged;
therefore, the proposed changes do not involve a significant
reduction in a margin of safety.
Removing a one time only allowance granted by Amendment 129 to
the Operating License that is no longer in affect is an
administrative change. Therefore, the proposed change does not
involve a significant reduction in a margin of safety.
Based on the above, neither the proposed changes to the
Technical Specifications nor that to the Operating License result in
a significant reduction in a margin of safety.
Based on the above evaluation, operation in accordance with the
proposed amendment involves no significant hazards considerations.
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: Judge George W. Armstrong
Library, 220 S. Commerce Street, Natchez, Mississippi 39120.
Attorney for licensee: Nicholas S. Reynolds, Esquire, Winston and
Strawn, 1400 L Street, NW., 12th Floor, Washington, DC 20005-3502.
NRC Section Chief: Robert A. Gramm.
Entergy Operations Inc., Docket No. 50-382, Waterford Steam Electric
Station, Unit 3, St. Charles Parish, Louisiana
Date of amendment request: October 6, 1998.
[[Page 46436]]
Description of amendment request: The proposed change modifies the
requirement to perform a Moderator Temperature Coefficient (MTC) test
near the end of each cycle. This request constitutes a lead-plant
submittal, submitted by Waterford 3 on behalf of the Combustion
Engineering Owners Group (CEOG). CE NPSD-911, Amendment 1, ``Analysis
of Moderator Temperature Coefficients in Support of a Change in the
Technical Specifications End of Cycle Negative MTC Limit'' dated
January, 1998 is provided as an Attachment to the application.
Specifically, the proposed change modifies Technical Specification (TS)
4.1.1.3.2c by adding a provision that eliminates the need to determine
the MTC upon reaching two-thirds of core burnup if the results of the
MTC tests required in TS 4.1.1.3.2a and 4.1.1.3.2b are within a
specified tolerance. In addition, some editorial changes are proposed
and the Bases change is included to support the changes 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. 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.
Under the proposed change, compliance with the TS Limiting
Condition for Operation is achieved through a surveillance program
consisting of beginning-of-cycle (BOC) measurements, plant parameter
monitoring, and end-of-cycle (EOC) MTC predictions. This change
eliminates the performance of the 2/3 Cycle MTC Surveillance when
the BOC MTC Surveillances are within a required tolerance of the
design value.
The probability and consequences of an accident previously
evaluated will not be increased because this change does not modify
any assumptions used in the input to the safety analyses. The
current safety calculations will remain valid because the allowed
range of MTC values will not change.
The Combustion Engineering analysis CE NPSD-911 and CE NPSD-911
Amendment 1, demonstrate that if the startup test program has
established that the core is operating as intended, and if the
isothermal temperature coefficients measured at zero power during
the cycle startup program, and at power prior to 40 EFPD [Effective
Full Power Days], fall within the design value of plus or minus
0.16 x 10-4 delta k/k/ deg.F, then the end-of-cycle best
estimate prediction will also be within plus or minus
0.16 x 10-4 delta k/k/ deg.F of true MTC.
Removing the footnote that was applicable during Cycle 7 and
providing a plus/minus for SR 4.1.1.3.2c is purely an administrative
change.
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.
Plant operation and plant parameter TS limits will remain
unchanged. There are no new changes in plant design nor are any new
failure modes introduced. CE NPSD-911 analysis determined that if
the MTC at the beginning-of-cycle is within plus or minus
0.16 x 10-4 delta k/k/ deg.F of the design value then the
MTC at the end-of-cycle will also be within plus or minus
0.16 x 10-4 delta k/k/ deg.F of the design value.
Removing the footnote that was applicable during Cycle 7 and
providing a plus/minus for SR 4.1.1.3.2c is purely an administrative
change.
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 margin of safety will not be reduced because the range of
allowed temperature coefficients will not be changed. The
surveillance program consisting of beginning-of-cycle measurements,
plant parameter monitoring, and end-of-cycle MTC predictions will
ensure that the MTC remains within the range of acceptable values.
Removing the footnote that was applicable during Cycle 7 and
providing a plus/minus for SR 4.1.1.3.2c is purely an administrative
change.
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, Louisiana 70122.
Attorney for licensee: N.S. Reynolds, Esquire, Winston & Strawn
1400 L Street NW., Washington, DC 20005-3502.
NRC Section Chief: Robert A. Gramm.
FirstEnergy Nuclear Operating Company, Docket No. 50-346, Davis-Besse
Nuclear Power Station, Unit 1, Ottawa County, Ohio
Date of amendment request: July 26, 1999.
Description of amendment request: The proposed amendment would make
the following line-item Technical Specification (TS) improvements:
(1) Relocate TS Section 3/4.3.3.2, Instrumentation--Incore
Detectors; TS 3/4.3.3.9, Instrumentation--Waste Gas System Oxygen
Monitor; and TS 3/4.4.7, Reactor Coolant System `` Chemistry, to the
Updated Safety Analysis Report (USAR) Technical Requirements Manual
(TRM);
(2) Change to TS 3/4.11.2, Radioactive Effluents--Explosive Gas
Mixture, and TS Bases 3/4.11.2, Explosive Gas Mixture, to reflect the
above proposed relocation of TS 3/4.3.3.9;
(3) Revise the requirements of TS 3/4.4.6.1, Reactor Coolant System
Leakage--Leakage Detection Systems, to require one monitor (gaseous or
particulate) of the containment atmosphere radioactivity monitoring
systems to be operable, rather than requiring both systems to be
operable simultaneously; and
(4) Revise the requirements of TS 3/4.3.3.1, Radiation Monitoring
Instrumentation, to be consistent with the above proposed revision to
TS 3/4.4.6.1.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensees have
provided their analysis of the issue of no significant hazards
consideration, which is presented below:
The Davis-Besse Nuclear Power Station (DBNPS) has reviewed the
proposed changes and determined that a significant hazards
consideration does not exist because operation of the Davis-Besse
Nuclear Power Station, Unit Number 1, in accordance with these
changes would:
1a. Not involve a significant increase in the probability of an
accident previously evaluated because no accident initiator,
conditions or assumptions are affected by the proposed revisions to
Technical Specification (TS) 3/4.3.3.1, Radiation Monitoring
Instrumentation, TS 3/4.3.3.2, Instrumentation--Incore Detectors; TS
3/4.3.3.9, Instrumentation--Waste Gas System Oxygen Monitor; TS 3/
4.4.7, Reactor Coolant System--Chemistry; TS 3/4.11.2, Radioactive
Effluents--Explosive Gas Mixture; and TS 3/4.4.6.1, Reactor Coolant
System Leakage--Leakage Detection Systems, and their associated TS
Bases.
The requirements of TS 3/4.3.3.2, TS 3/4.3.3.9, and TS 3/4.4.7
are proposed to be relocated from the TS to the DBNPS Updated Safety
Analysis Report (USAR) Technical
[[Page 46437]]
Requirements Manual (TRM). These requirements would be relocated
generally intact to the TRM whereby future changes would be subject
to the regulatory controls of 10 CFR 50.59. These relocations are
consistent with the NRC guidance provided in Generic Letter (GL) 95-
10, ``Relocation of Selected Technical Specifications Requirements
Related to Instrumentation,'' or NUREG-1430, Revision 1, ``Standard
Technical Specifications--Babcock and Wilcox Plants,'' dated April
1995.
The proposed revision to TS 3/4.11.2, Radioactive Effluents--
Explosive Gas Mixture, and its Bases is an administration change to
a reference necessitated by the proposed relocation of TS 3/4.3.3.9
to the USAR TRM.
The proposed revision to TS 3/4.3.3.1 and TS 3/4.4.6.1 regarding
the number of Reactor Coolant System (RCS) leakage detection
monitors required and their allowed outage times is based upon the
NRC's guidance of NUREG-1430, Revision 1. This proposed revision
affects the TS only and does not reduce the number, diversity, or
sensitivity of Reactor Coolant System leakage detection systems
inside the containment building or as committed to in the DBNPS
USAR.
1b. Not involve a significant increase in the consequences of an
accident previously evaluated because no accident condition or
assumption is affected by the proposed revisions. As described
above, the revisions are consistent with the guidance of NRC GL 95-
10 or NUREG-1430, Revision 1. The proposed revisions, as described
above, do not alter the source term, containment isolation, or
allowable releases. The proposed changes, therefore, will not
increase the radiological consequences of a previously evaluated
accident.
2. Not create the possibility of a new or different kind of
accident from any accident previously evaluated because no new
accident initiators or assumptions are introduced by the proposed TS
revisions. No new accident scenarios, transient precursors, failure
mechanisms, or limiting failures are introduced as a result of the
proposed changes.
3. Not involve a significant reduction in a margin of safety
because the proposed revisions do not reduce or adversely affect the
capabilities of any plant structures, systems or components. The
proposed relocation of TS 3/4.3.3.2, TS 3/4.3.3.9, and TS 3/4.4.7 to
the USAR TRM is essentially an administrative change to the location
and process by which these requirements are controlled and revised.
Future revisions to these requirements relocated to the USAR TRM
will be subject to the regulatory controls of 10 CFR 50.59.
Therefore, these revisions will not result in a significant
reduction in a margin of safety.
The proposed revision to TS 3/4.11.2 and its Bases is
administrative and reflects the relocation of TS 3/4.3.3.9 to the
USAR TRM. Therefore, this revision will not result in a significant
reduction in a margin of safety.
The proposed revisions to TS 3/4.3.3.1 and TS 3/4.4.6.1 affect
the number of containment atmosphere radioactivity monitors required
by TS to be operable simultaneously. However, redundancy and
diversity requirements are maintained in the TS for detecting
Reactor Coolant System leakage. Although TS-allowed outage times are
proposed to be increased consistent with NUREG-1430, Revision 1
guidance, related compensatory action requirements are also being
increased. Furthermore, the DBNPS commitments made for complying
with Regulatory Guide 1.45, May, 1973, ``Reactor Coolant Pressure
Boundary Leakage Detection Systems,'' are not changed by the
proposed revisions. Along with the applicable revised TS
requirements, 10 CFR 50, Appendix B, Criterion XVI will require
prompt corrective action for inoperable leakage detection systems.
Accordingly, these proposed revisions will not result in 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 Toledo, William
Carlson Library, Government Documents Collection, 2801 West Bancroft
Avenue, Toledo, OH 43606.
Attorney for licensee: Jay E. Silberg, Esquire, Shaw, Pittman,
Potts and Trowbridge, 2300 N Street, NW., Washington, DC 20037.
NRC Section Chief: Anthony J. Mendiola.
FirstEnergy Nuclear Operating Company, Docket No. 50-346, Davis-Besse
Nuclear Power Station, Unit 1, Ottawa County, Ohio
Date of amendment request: July 26, 1999.
Description of amendment request: The proposed amendment would
change the Technical Specifications to adopt the performance-based 10
CFR Part 50, Appendix J, Option B approach for Type B and C containment
leakage rate testing, and to relocate certain details of the tests into
a Containment Leakage Testing Program.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensees have
provided their analysis of the issue of no significant hazards
consideration, which is presented below:
The Davis-Besse Nuclear Power Station has reviewed the proposed
changes and determined that a significant hazards consideration does
not exist because operation of the Davis-Besse Nuclear Power
Station, Unit No. 1, in accordance with these changes would:
1a. Not involve a significant increase in the probability of an
accident previously evaluated because accident initiators,
conditions, or assumptions are not affected by the proposed changes.
The proposed changes to the Technical Specifications and Bases
implement 10 CFR [Part] 50 Appendix J Option B for Type B and C
Local Leak Rate Testing, based on the guidance of Regulatory Guide
1.163,
``Performance-Based Containment Leak-Test Program.'' Provided
that components have performed satisfactorily on a historical basis,
this guidance permits the use of extended testing frequencies. These
proposed changes do not affect accident initiators, conditions, or
assumptions.
1b. Not involve a significant increase in the consequences of an
accident previously evaluated because the proposed changes do not
change the source term or total allowable releases. With the
exception of the proposed increase in the containment air lock
leakage limits, the proposed changes do not affect the total
allowable containment leakage rates presently specified in the
Technical Specifications. Although the air lock leakage limits are
proposed to be increased, the accident analyses are based on the
current TS allowable maximum bypass leakage, which is not proposed
to be changed. Therefore, increases in leakage limits for individual
components, such as the air locks and their door seals, which are
constituents of bypass leakage, will have no effect on the
radiological consequences described in the accident analyses.
The proposed TS changes relating to implementation of 10 CFR
[Part] 50 Appendix J Option B may result in a small, but acceptable
increase in post-accident containment leakage, due to the increased
probability that due to generally increased intervals between tests,
an unacceptable leakage rate could go undetected for a longer length
of time. NUREG-1493, ``Performance-Based Containment Leak-Test
Program,'' September, 1995, which provided the technical basis for
the 10 CFR [Part] 50 Appendix J Option B rulemaking, provides a
detailed evaluation of the expected leakage and its consequences and
concludes that increased test frequencies are workable without
significant risk impacts.
2. Not create the possibility of a new or different kind of
accident from any accident previously evaluated because no new
accident initiators or assumptions are introduced by the proposed
changes. The proposed changes do not affect the methodology used in
conducting containment leak rate testing. The proposed changes do
not involve a change to the plant design or operation and,
therefore, will not introduce any new or different failure modes or
initiators.
3. Not involve a significant reduction in a margin of safety.
The proposed changes relating to implementation of 10 CFR [Part]
50, Appendix J, Option B do not significantly affect the allowable
containment leakage rates presently specified in the Technical
Specifications. The Technical Specifications, under the proposed
changes, will continue to ensure containment reliability by periodic
testing performed in full compliance with 10 CFR [Part] 50, Appendix
J.
[[Page 46438]]
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 Toledo, William
Carlson Library, Government Documents Collection, 2801 West Bancroft
Avenue, Toledo, OH 43606.
Attorney for licensee: Jay E. Silberg, Esquire, Shaw, Pittman,
Potts and Trowbridge, 2300 N Street, NW., Washington, DC 20037.
NRC Section Chief: Anthony J. Mendiola.
FirstEnergy Nuclear Operating Company, Docket No. 50-346, Davis-Besse
Nuclear Power Station, Unit 1, Ottawa County, Ohio
Date of amendment request: July 28, 1999.
Description of amendment request: The proposed amendment would
change Technical Specification (TS) Section 3/4.7.5.1, ``Ultimate Heat
Sink,'' to allow operation on Modes 1 through 4 with an Ultimate Heat
Sink water temperature of less than or equal to 90 deg.F, instead of
the current limit of less than or equal to 85 deg.F.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensees have
provided their analysis of the issue of no significant hazards
consideration, which is presented below:
The Davis-Besse Nuclear Power Station has reviewed the proposed
changes and determined that a significant hazards consideration does
not exist because operation of the Davis-Besse Nuclear Power
Station, Unit No. 1, in accordance with these changes would:
1a. Not involve a significant increase in the probability of an
accident previously evaluated because no accident initiators,
conditions, or assumptions are significantly affected by the
proposed change. The proposed change would increase the allowable
Ultimate Heat Sink (UHS) water temperature, as specified in TS LCO
3.7.5.1.b, from less than or equal to 85 deg.F to less than or equal
to 90 deg.F. This water is used by the Service Water System to
provide cooling to equipment that is used to mitigate accidents such
as a Large Break Loss of Coolant Accident. This increase in Service
Water temperature has been evaluated and the proposed change does
not result in the operation of equipment important to safety outside
their acceptable operating ranges.
1b. Not involve a significant increase in the consequences of an
accident previously evaluated because the proposed change does not
change the source term, containment isolation, or allowable
releases. The proposed increase in the Service Water System
temperature has been evaluated with respect to the containment and
equipment used to mitigate the consequences of accidents previously
evaluated. These evaluations have determined that there are no
significant increases in consequences.
2. Not create the possibility of a new or different kind of
accident from any accident previously evaluated because no new
accident initiators or assumptions are introduced by the proposed
5 deg.F increase in UHS temperature. The proposed change does not
result in installed equipment being operated outside their design
operating ranges. No new or different equipment failure modes or
mechanisms are introduced by the proposed change.
3. Not involve a significant reduction in a margin of safety
because the proposed 5 deg.F increase in UHS temperature does not
result in significant changes to the initial conditions contributing
to accident severity or consequences.
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 Toledo, William
Carlson Library, Government Documents Collection, 2801 West Bancroft
Avenue, Toledo, OH 43606.
Attorney for licensee: Jay E. Silberg, Esquire, Shaw, Pittman,
Potts and Trowbridge, 2300 N Street, NW., Washington, DC 20037 .
NRC Section Chief: Anthony J. Mendiola.
FirstEnergy Nuclear Operating Company, Docket No. 50-440, Perry Nuclear
Power Plant, Unit 1, Lake County, Ohio
Date of amendment request: June 17, 1999.
Description of amendment request: The proposed amendment modifies
multiple surveillance requirements to support implementation of a 24-
month operating cycle.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration which is presented below:
The proposed amendment does not involve a significant increase
in the probability or consequences of an accident previously
evaluated.
A. Frequency Extensions
The proposed Technical Specification (TS) changes involve a
change in the surveillance testing intervals to facilitate a change
in the Perry Nuclear Power Plant (PNPP) operating cycle from 18
months to 24 months. The proposed TS changes do not physically
impact the plant, nor do they impact any design or functional
requirements of the associated systems. That is, the proposed TS
changes do not degrade the performance of, or increase the
challenges to, any safety systems assumed to function in the
accident analysis. The proposed TS changes do not impact the TS
surveillance requirements themselves, or the way in which the
surveillances are performed. In addition, the proposed TS changes do
not introduce any accident initiators, since no accidents previously
evaluated have, as their initiators, anything related to the
frequency of surveillance testing. Also, evaluation of the proposed
TS changes demonstrated that the availability of equipment and
systems required to prevent or mitigate the radiological
consequences of an accident are not significantly affected because
of other, more frequent testing that is performed, the availability
of redundant systems and equipment, or the high reliability of the
equipment. Since the impact on the systems is minimal, it is
concluded that the overall impact on the plant accident analysis is
negligible. Furthermore, a historical review of surveillance test
results and associated maintenance records indicated that there was
no evidence of any failures that would invalidate the above
conclusions. Therefore, the proposed TS changes do not significantly
increase the probability or consequences of an accident previously
evaluated.
B. Allowable Value Changes
The proposed changes in Allowable Values for the instrumentation
include in Table 3.3.8.1-1 Items d and e of the Technical
Specifications are the result of application of the Perry Instrument
Setpoint Methodology (ISM) using plant specific drift values.
Application of this methodology results in Allowable Values which
more accurately reflect total instrumentation loop accuracy as well
as that of test equipment and calculated drift between
surveillances. The proposed changes will not result in any hardware
changes. The instrumentation is not assumed to be an initiator of
any analyzed event. Existing operating margin between plant
conditions and actual plant setpoints is not significantly reduced
due to these changes. The role of the instrumentation is in
mitigating and thereby limiting the consequences of accidents. The
Allowable Values have been developed to ensure that the design and
safety analysis limits will be satisfied. The methodology used for
the development of the Allowable Values ensures the affected
instrumentation remains capable of mitigating design basis events as
described in the safety analyses and that the results and
radiological consequences described in the safety analyses remain
bounding. Additionally, the proposed change does not alter the
plant's ability to detect and mitigate events. Therefore, this
change does not involve a significant increase in the probability or
consequences of an accident previously evaluated.
[[Page 46439]]
C. Frequency Reductions to Semiannual
The proposed Technical Specification (TS) changes involve a
change in the surveillance testing intervals from 18 months to
either 6 months or quarterly. The shorter frequencies are based on
PNPP specific results of setpoint drift evaluations. The proposed
more restrictive TS changes do not physically impact the plant, nor
do they impact any design or functional requirements of the
associated systems. That is, the proposed TS changes do not degrade
the performance of, or increase the challenges to, any safety
systems assumed to function in the accident analysis. The proposed
TS changes do not impact the TS surveillance requirements
themselves, or the way in which the surveillances are performed. In
addition, the proposed TS changes do not introduce any accident
initiators, since no accidents previously evaluated have, as their
initiators, anything related to the frequency of surveillance
testing. The proposed TS frequencies will demonstrate that the
equipment and systems required to prevent or mitigate the
radiological consequences of an accident are continuing to meet the
assumptions of the setpoint evaluation, on a more frequent basis.
Since the impact on the systems is minimal, and the assumptions of
the safety analyses will be maintained, it is concluded that the
overall impact on the plant accident analysis is negligible.
Furthermore, a historical review of surveillance test results and
associated maintenance records indicated that there was no evidence
of any failures that would invalidate the proposed test frequencies.
Therefore, the proposed TS changes do not significantly increase the
probability or consequences of an accident previously evaluated.
The proposed amendment would not create the possibility of a new
or different kind of accident from any accident previously
evaluated.
A. Frequency Extensions
The proposed TS changes involve a change in the surveillance
testing intervals to facilitate a change in the PNPP operating cycle
length. The proposed TS changes do not introduce any failure
mechanisms of a different type than those previously evaluated,
since there are no physical changes being made to the facility. No
new or different equipment is being installed. No installed
equipment is being operated in a different manner. As a result, no
new failure modes are being introduced. In addition, the
surveillance test requirements themselves, and the way surveillance
tests are performed, will remain unchanged. Furthermore, a
historical review of surveillance test results and associated
maintenance records indicated there was no evidence of any failures
that would invalidate the above conclusions. Therefore, the proposed
TS changes do not create the possibility of a new or different kind
of accident from any previously evaluated.
B. Allowable Value Changes
The proposed changes are the result of application of the ISM
using plant specific drift values and do not create the possibility
of a new or different kind of accident from any accident previously
evaluated. This is based on the fact that the method and manner of
plant operation is unchanged. The use of the proposed Allowable
Values does not impact safe operation of PNPP in that the safety
analysis limits will be maintained. The propose Allowable Values
involve no system additions or physical modifications to systems in
the station. These Allowable Values were revised to ensure the
affected instrumentation remains capable of mitigating accidents and
transients. Plant equipment will not be operated in a manner
different from previous operation, except that setpoints may be
changed. Since operational methods remain unchanged and the
operating parameters have been evaluated to maintain the station
within existing design basis criteria, no different type of failure
or accident is created.
C. Frequency Reductions to Semiannual or Quarterly
The proposed TS changes involve a change in the surveillance
testing interval due to the application of the ISM and plant
specific drift analysis results. Also, the quarterly tests reflect
current PNPP calibration practices, since the components are
normally calibrated during the Channel Functional Test. The proposed
TS changes do not introduce any failure mechanisms of a different
type than those previously evaluated, since there are no physical
changes being made to the facility. No new or different equipment is
being installed. No installed equipment is being operated in a
different manner. The proposed change does not impact core
reactivity, or the manipulation of fuel bundles. As a result, no new
failure modes are being introduced. In addition, the surveillance
test requirements themselves, and the way surveillance tests are
performed, will remain unchanged. Furthermore, a historical review
of surveillance test results and associated maintenance records
indicated there was no evidence of any failures that would
invalidate the above conclusions. Therefore, the proposed TS changes
do not create the possibility of a new or different kind of accident
from any previously evaluated.
The proposed amendment will not involve a significant reduction
in a margin of safety.
A. Frequency Extensions
Although the proposed TS changes will result in changes in the
interval between surveillance tests, the impact, if any, on system
availability is small, based on other, more frequent testing that is
performed, or the existence of redundant systems and equipment, or
overall system reliability. Evaluations have shown there is no
evidence of time dependent failures that would impact the
availability of the systems. The proposed change does not
significantly impact the condition or performance of structures,
systems, and components relied upon for accident mitigation. The
proposed change does not significantly impact any safety analysis
assumptions or results. Therefore, the proposed change does not
involve a significant reduction in a margin of safety.
B. Allowable Value Changes
The proposed change does not involve a reduction in a margin of
safety. The proposed changes have been developed using a methodology
to ensure safety analysis limits are not exceeded. As such, this
proposed change does not involve a significant reduction in a margin
of safety.
C. Frequency Reductions to Semiannual or Quarterly
The proposed TS changes will result in a shorter interval
between surveillance tests to ensure that the assumptions of the
safety analysis are maintained. The impact, if any, on system
availability is small, as a result of this more frequent testing
that is performed. The proposed change does not significantly impact
the condition or performance of structures, systems, and components
relied upon for accident mitigation. The proposed change does not
significantly impact any safety analysis assumptions or results.
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: Perry Public Library, 3753
Main Street, Perry, OH 44081.
Attorney for licensee: Jay Silberg, Esq., Shaw, Pittman, Potts &
Trowbridge, 2300 N Street, NW., Washington, DC 20037.
NRC Section Chief: Anthony J. Mendiola.
FirstEnergy Nuclear Operating Company, Docket No. 50-440, Perry Nuclear
Power Plant, Unit 1, Lake County, Ohio
Date of amendment request: August 4, 1999.
Description of amendment request: The amendment would incorporate
an additional option into the Required Actions for Technical
Specification 3.9.1, ``Refueling Equipment Interlocks.'' The change
would provide additional Required Actions when the refueling interlocks
are inoperable. The alternative would permit continued refueling
activities once control rod withdrawal is blocked and operators verify
that all appropriate controls rods are fully inserted.
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
[[Page 46440]]
consequences of an accident previously evaluated.
The refueling interlocks are explicitly assumed in the Perry
Nuclear Power Plant Updated Safety Analysis Report (USAR) analyses
of the control rod removal error and fuel loading error during
refueling. This analysis evaluates the probability and consequences
of control rod withdrawal during refueling. Criticality and,
therefore, subsequent prompt reactivity excursions are prevented
during the loading of fuel, provided all required control rods are
fully inserted. The refueling interlocks accomplish this by
preventing loading fuel into the core with any control rod
withdrawn, or by preventing withdrawal of a rod from the core during
fuel loading. When the refueling interlocks are inoperable, the
current method of preventing fuel loading when a control rod is
withdrawn, is to prevent fuel movement. This method is currently
required by the Technical Specifications. An alternate method to
ensure that fuel is not loaded into a cell with the control rod
withdrawn is to prevent control rods from being withdrawn and verify
that all control rods required to be inserted are fully inserted.
The proposed Technical Specification Required Actions will require
that a control rod block be placed in effect, thereby ensuring that
control rods are not subsequently inappropriately withdrawn.
Additionally, following placing the control rod withdrawal block in
effect, the proposed actions will require that all required control
rods be verified to be fully inserted. This verification is in
addition to the requirements to periodically verify control rod
position by other Technical Specification requirements. These
proposed actions will ensure that control rods are not withdrawn and
cannot be inappropriately withdrawn, because an electrical or
hydraulic block to control rod withdrawal is in place. Like the
current requirements, the proposed will ensure that unacceptable
operations are blocked (e.g., loading fuel into a cell with a
control rod withdrawn, except when following the requirements of LCO
3.10.6, ``Multiple Control Rod Removal--Refueling,'' which is
unaffected by this change). The proposed additional Required Actions
provide an equivalent level of assurance that fuel will not be
loaded into a core cell with a control rod withdrawn as do the
current Required Action or the Surveillance Requirement. Therefore,
the proposed change does not significantly increase the probability
or consequences of an accident previously evaluated.
2. The proposed change would not create the possibility of a new
or different kind of accident from any accident previously
evaluated.
The change in the Technical Specification requirements does not
involve a change in the plant design, or to the status of the
reactor core during refueling. The proposed actions will ensure that
control rods are not withdrawn and cannot be inappropriately
withdrawn, because an electrical or hydraulic block to control rod
withdrawal is in place. Although the exact method by which the
control rod withdrawal block is inserted is revised, the net effect
is equivalent. The requirements will continue to ensure that fuel is
not loaded into the core when a control rod is withdrawn, except
when following the requirements of LCO 3.10.6, ``Multiple Control
Rod Removal--Refueling,'' which is unaffected by this change.
Therefore, no new failure modes are introduced, and the proposed
changes do not create the possibility of a new or different kind of
accident from any accident previously evaluated.
3. The proposed change will not involve a significant reduction
in the margin of safety.
As discussed in the Bases for the affected Technical
Specification requirements, inadvertent criticality is prevented
during the loading of fuel provided all required control rods are
fully inserted during the fuel insertion. The refueling interlocks
function to support the refueling procedures by preventing control
rod withdrawal during fuel movement and the inadvertent loading of
fuel when a control rod is withdrawn. The proposed change will allow
the refueling interlocks to be inoperable and fuel movement to
continue only if a control rod withdrawal block is in effect and all
required control rods are verified to be fully inserted. These
proposed Required Actions provide an equivalent level of protection
as the refueling interlocks by preventing a configuration which
could lead to an inadvertent criticality event. The refueling
procedures will continue to be supported by the proposed Required
Actions because control rods cannot be withdrawn and as a result
fuel cannot be inadvertently loaded when a control rod is withdrawn,
except when following the requirements of LCO 3.10.6, ``Multiple
Control Rod Removal--Refueling,'' which is unaffected by this
change. Therefore, the proposed changes do not cause 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: Perry Public Library, 3753
Main Street, Perry, OH 44081.
Attorney for licensee: Mary E. O'Reilly, Attorney, FirstEnergy
Corporation, 76 South Main Street, Akron, OH 44308.
NRC Section Chief: Anthony J. Mendiola.
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: February 23, 1999.
Description of amendment request: The proposed license amendment
would remove redundant boron concentration monitoring requirements
specified for operating Modes 3 through 6 by deleting Technical
Specification 3/4.1.2.9, ``Reactivity Control Systems--Boron
Dilution.'' These requirements were interim measures intended to apply
until a permanent boron dilution alarm system was installed and
functional.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
(1) Operation of the facility in accordance with the proposed
amendment would not involve a significant increase in the
probability or consequences of an accident previously evaluated.
The proposed amendment does not involve changes to previously
evaluated accident initiators. The proposed deletion of the
redundant boron concentration verification requirements do not
impact the results of existing accident analyses, and will have no
adverse impact on any plant system performance. TS 3/4.1.2.9
provides mode and charging pump dependent monitoring requirements
for RCS boron concentration that are designed to detect an unplanned
boron dilution event in MODES 3 through 6 in the absence of an
automatic alarm system, and is based on the time requirements for
operator action specified in Section 15.4.6 of the Standard Review
Plan (SRP). This specification evolved from interim measures that
were proposed by FPL until the boron dilution alarm system (BDAS)
could be made completely functional following initial start-up of
St. Lucie Unit 2. The BDAS is completely functional and provides
redundant control room alarms to alert operators to the occurrence
of an unplanned boron dilution event in Modes 3 through 6. The alarm
setpoints are based on Chemical and Volume Control System (CVCS)
malfunction analyses, and satisfy the same SRP acceptance criteria
upon which the monitoring requirements of TS 3/4.1.2.9 were based.
Therefore, operation of the facility in accordance with the proposed
amendment will not involve a significant increase in the probability
or consequences of an accident previously evaluated.
(2) Operation of the facility in accordance with the proposed
amendment would not create the possibility of a new or different
kind of accident from any accident previously evaluated.
The proposed amendment will not change the physical plant or the
modes of operation defined in the facility license. The amendment
will remove requirements from the facility technical specifications
that were proposed by FPL as interim measures until the boron
dilution alarm system became completely functional. The amendment
will not alter the design of St. Lucie plant systems described in
the Updated Final Safety Analysis Report (UFSAR), and the plant
configuration will continue to remain consistent with assumptions
used in the existing accident analyses. Therefore, operation of the
facility in accordance with the proposed amendment would not create
the possibility of a new or different kind of
[[Page 46441]]
accident from any accident previously evaluated.
(3) Operation of the facility in accordance with the proposed
amendment would not involve a significant reduction in a margin of
safety. The proposed amendment has been evaluated with respect to
the applicable safety analyses. The BDAS provides a continuous,
early warning capability to detect a boron dilution event in Modes
3, 4, 5 and 6, and satisfies the same SRP time requirements for
operator action as the interim TS that is proposed for deletion.
BDAS setpoints are determined and/or validated for each fuel cycle
to ensure they remain consistent with the CVCS malfunction analyses
of record, and changes that may become necessary are controlled
pursuant to 10 CFR 50.59. The minimum required Shutdown Margin is
not changed by this proposal. Therefore, operation of the facility
in accordance with the proposed amendment would not involve a
significant reduction in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: Indian River Junior College
Library, 3209 Virginia Avenue, Fort Pierce, Florida 34954-9003.
Attorney for licensee: M.S. Ross, Attorney, Florida Power & Light,
P.O. Box 14000, Juno Beach, Florida 33408-0420.
NRC Section Chief: Sheri R. Peterson.
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: July 27, 1999.
Description of amendment request: The proposed amendments request
that Turkey Point Unit 3 Technical Specification (TS) 3/4.8.1, A.C.
SOURCES,TS 3/4.4.3, PRESSURIZER, and TS 3/4.5.2, ECCS SUBSYSTEMS--
Tavg GREATER THAN OR EQUAL TO 350 deg.F, be revised on a
one-time basis to extend the Allowed Outage Time (AOT) for an
inoperable Emergency Diesel Generator (EDG) from 72 hours to 7 days.
The proposed one-time AOT extension will be used to replace the Unit 3
EDG engine radiators prior to the Spring 2000 refueling outage.
However, replacement of the radiator is a very labor-intensive
evolution that cannot be performed within the existing 72 hour AOT. The
proposed AOT extension will allow the radiator replacement activity to
be completed successfully in a safe manner. The extended AOT will be
applied to one EDG at a time in a sequential manner. When the radiator
replacement activity is complete on one engine, it will be returned to
service so that work can proceed on the redundant EDG. It should be
noted that although the proposed changes apply only to Unit 3, the Unit
4 TSs are administratively affected since the TSs are combined for both
units.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
(1) Operation of the facility in accordance with the proposed
amendment would not involve a significant increase in the
probability or consequences of an accident previously evaluated.
The Emergency Diesel Generators (EDG) are part of the on-site
electrical power distribution system. They function as a standby
power source in the event that the preferred A.C. power supply,
i.e., offsite power, is interrupted. While certain failures in the
electrical distribution system can lead to a loss of offsite power
which is a design basis event for the plant, the EDGs are not
assumed to be an initiating condition of any accident evaluated in
the safety analysis report. Therefore, a one-time extension in the
EDG Allowed Outage Time (AOT) does not involve a significant
increase in the probability of an accident previously evaluated.
The purpose of the proposed license amendment is to permit on-
line replacement of the Unit 3 EDG radiators. The radiators are part
of the closed-loop diesel engine cooling water system and do not
interface with any system or component that contains radioactivity.
The EDGs do supply A.C. power to the emergency core cooling and
containment heat removal systems during accidents that involve loss
of offsite power. However, no changes are predicted for the
postulated post-accident releases since adequate EDG capacity will
be available under the conditions of the proposed license amendment
to accommodate any design basis accident condition. Accordingly, the
consequences of accidents previously evaluated in the safety
analysis report are not changed by an extended EDG outage.
Probabilistic Safety Assessment (PSA) techniques were used to
evaluate the impact of a one-time extension of the EDG AOT from 72
hours to 7 days. The results of these analyses indicate that
extending the AOT for the purpose of replacing the engine radiator
cores represents an acceptably small impact on Core Damage
Probability.
Based on the above, FPL concludes that the proposed amendment
does not involve a significant increase in the probability or
consequences of any accident previously evaluated.
(2) Operation of the facility in accordance with the proposed
amendment would not create the possibility of a new or different
kind of accident from any previously evaluated.
The proposed change does not alter the design, physical
configuration, or modes of operation of the plant. Plant
configurations that are prohibited by Technical Specifications will
not be created by the one-time EDG AOT extension. Therefore, the
proposed activity does not create the possibility of a new or
different kind of accident from any previously evaluated.
(3) Operation of the facility in accordance with the proposed
amendment would not involve a significant reduction in a margin of
safety.
The proposed license amendment will extend by 96 hours the
requirement to shutdown the plant when a Unit 3 EDG is removed from
service for maintenance. The one-time AOT extension will not alter
plant equipment, setpoints, or operating practices that provide the
existing margins of safety. Therefore, the change does not involve a
significant reduction in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment 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 Section Chief: Sheri R. Peterson.
Portland General Electric Company, et al., Docket No. 50-344, Trojan
Nuclear Plant, Columbia County, Oregon
Date of amendment request: August 27, 1998.
Description of amendment request: The amendment would delete the
requirements for an emergency plan from the 10 CFR Part 50 license and
technical specifications after the spent nuclear fuel is transferred to
a Part 72 licensed independent spent fuel storage installation (ISFSI).
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 elimination of the emergency plan requirements from
the 10 CFR 50 license is predicated on completion of transfer of the
spent nuclear fuel to the proposed 10 CFR 72 ISFSI licensed area and
removal of the reactor vessel and internals from the 10 CFR 50
licensed area of the site.
[[Page 46442]]
Removal of the potential radiological source terms for accidents
previously evaluated effectively eliminates the credibility of the
accidents, therefore, elimination of the emergency plan requirements
does not involve a significant increase in the probability or
consequences of an accident previously evaluated.
2. The proposed change does not create the possibility of a new
or different kind of accident from any accident previously
evaluated.
The proposed change is deletion of emergency plan requirements
and, as such, has no direct impact on plant equipment or the
procedures for operating plant equipment. Therefore, it 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.
Following the removal of the spent nuclear fuel and the reactor
vessel and internals from the 10 CFR 50 licensed area, the remaining
credible accidents are limited to decommissioning activities. The
potential accidents associated with decommissioning activities are
presented in the TNP [Trojan Nuclear Plant] Decommissioning Plan and
have been shown to have consequences less than the EPA PAGs
[Environmental Protection Agency Protective Action Guidelines].
Following the removal of the spent nuclear fuel and the reactor
vessel (including the internals) from the 10 CFR 50 site, no
credible accidents associated with the remaining decommissioning
activities would require pre-planned emergency measures to avoid
acute radiation doses. The deletion of the Trojan Nuclear Plant
Permanently Defueled Emergency Plan will not result in a reduction
in the margin of safety previously analyzed. Therefore, the proposed
10 CFR 50 license amendment 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: Branford Price Millar Library,
Portland State University, 934 S.W. Harrison Street, P.O. Box 1151,
Portland, Oregon 97207.
Attorney for licensee: Leonard A. Girard, Esq., Portland General
Electric Company, 121 S.W. Salmon Street, Portland, Oregon 97204.
NRC Section Chief: Michael T. Masnik.
Power Authority of The State of New York, Docket No. 50-286, Indian
Point Nuclear Generating Unit No. 3, Westchester County, New York
Date of amendment request: February 19, 1998, as supplemented July
28, 1999.
Description of amendment request: This application for amendment to
the Indian Point 3 Technical Specifications (TSs) proposes to revise
the Radioactive Effluents Technical Specifications (RETS) in accordance
with Generic Letter 89-01 (GL-89-01), to make changes to implement
revised 10 CFR Part 20 requirements, and to make administrative changes
under 10 CFR 50.36a.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. Does the proposed license amendment involve a significant
increase in the probability or consequences of any accident
previously evaluated?
A. The proposed changes involve (1) combining related LCO and
surveillance requirements from Sections 2.0 and 3.0, respectively,
of the Indian Point 3 (IP3) RETS and relocating this text to the new
Radiological Effluent Controls (REC) section of the ODCM, (2)
relocating the bases contained in Section 4.0 of the RETS to the
ODCM REC, (3) relocating the detailed reporting requirements
contained in Section 5.0 of the RETS to the ODCM REC, and (4)
updating references to 10 CFR Part 20. Additional changes include
formatting both the remaining RETS and the new REG to more closely
model Standard Technical Specifications (STS), revising the
frequency of the Radioactive Effluent Release Report in accordance
with 10 CFR 50.36a, relocating all definitions to Appendix A of the
Technical Specifications and adding/deleting definitions as
necessary, and adding a new Special Reports section to the ODCM.
Most of the changes are (1) consistent with the guidance provided in
the generic letter, NUREG-1301, or provisions of 10 CFR; or (2)
editorial. Editorial changes include the relocation of text,
correction of typographical and punctuation errors, renumbering,
reformatting, immaterial wording revisions/deletions/clarifications
which do not change intent, and updating references.
B. The proposed revisions to the liquid and gaseous release rate
limits, the relocation of the old 10 CFR 20.106 requirements to the
new 10 CFR 20.1302, and the revision to the TS bases for the Liquid
Holdup Tank activity will involve no change in the types or amounts
of effluents that will be released, nor will there be an increase in
individual or cumulative occupational radiation exposures.
The changes of definitions, terminology, paragraph references,
and report submittal frequency are necessary to keep IP3 TS
consistent with revised federal regulations (i.e., 10 CFR 20 and 10
CFR 50.36(a)). Record retention and reporting requirements will
continue to meet NRC regulations. These changes are administrative
in nature and do not affect plant hardware or operation.
The changes do not impact the operation, design, configuration,
or testing of plant structures, systems or components. As such, the
proposed changes do not involve a significant increase in the
probability or consequences of any accident previously evaluated.
2. Does the proposed license amendment create the possibility of
a new or different kind of accident from any previously evaluated?
A. The changes do not impact the operation, design,
configuration, or testing of plant structures, systems or
components. The changes do not result in a change in type or amount
of radiological effluents released. As such, the proposed changes do
not create the possibility of a new or different kind of accident
from any previously evaluated.
3. Does the proposed amendment involve a significant reduction
in a margin of safety?
A. The changes are being made in accordance with NRC guidance
and continue to assure compliance with the applicable regulatory
requirements including 10 CFR 20. The changes do not result in a
change in the types or amounts of effluents released. The current
level of radiological effluent control will be maintained. As such,
the proposed changes do not involve a significant reduction in a
margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: White Plains Public Library,
100 Martine Avenue, White Plains, New York 10601.
Attorney for licensee: Mr. David E. Blabey, 10 Columbus Circle, New
York, New York 10019.
NRC Section Chief: S. Singh Bajwa.
Sacramento Municipal Utility District (the District), Docket No. 50-
312, Rancho Seco Nuclear Station, Sacramento County, California
Date of amendment request: March 18, 1996 (PA-192).
Description of amendment request: The proposed amendment would
update the Rancho Seco cask drop analysis and establish the cask drop
event as the design-basis event for plant operation in the permanently
defueled mode. The proposed amendment would also make editorial changes
to the Permanently Defueled Technical Specifications and Bases by
adding the word ``heavy'' to specification D3.3 and eliminating
references to the MP-187 cask in specification D3.3 and D4.3.3.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
[[Page 46443]]
The District has reviewed the proposed changes against each of
the criteria in 10 CFR 50.92, and, based on the above safety
analysis, concludes:
Using the Gantry Crane to handle a fully loaded transfer cask in
the Fuel Storage Building will not create a significant increase in
the probability or consequences of an accident previously evaluated
in the SAR [safety analysis report], because the conservative dose
consequence calculated for the updated, design basis cask drop event
resulted in an exposure (224 mrem) that is:
1. A very small percentage ([approximately] 0.9%) of the 10 CFR
100 design basis accident dose limit of 25 rem total body;
2. A small percentage ([approximately] 3.6%) of the NUREG-0612
control of heavy loads accident dose limit of 6.25 rem total body;
3. Well within ([approximately] 4.5%) of the old EPA
[Environmental Protection Agency] NUREG-0654 plume exposure
Protective Action Guidelines of 500 mrem total body dose;
4. Well within the new EPA 1 to 5 rem Total Effective Dose
Equivalent (TEDE) Protective Action Guidelines (PAGs) specified in
document EPA-400-R-92-001, Table 2-1, May 1992;
5. Less than the maximum hypothetical Rancho Seco Independent
Spent Fuel Storage Installation design basis accident (375 mrem
total body dose);
6. Less than the original Rancho Seco operating design basis for
the Fuel Storage Building FHA [fuel-handling accident] exposure (399
mrem);
7. Less than the original Rancho Seco operating design basis for
the Reactor Building FHA exposure (477 mrem); and
8. Much less than the original Rancho Seco operating design
basis Maximum Hypothetical Accident exposure (3,600 mrem).
Therefore, the conservatively calculated 224 mrem cask drop
design basis accident exposure is (1) relatively small and (2) not
considered a significant hazard.
Also, the probability of occurrence of the FHA, which is the
current design basis accident, is similar to the probability of
occurrence of the updated cask drop event. The FHA is assumed to
occur because the fuel handling bridge is not single failure proof.
Likewise for the cask drop scenario, since the Gantry Crane is not
single failure proof, this Safety Analysis Report evaluates the
Gantry Crane dropping a loaded spent fuel cask.
This Safety Analysis Report analyzes the dropped cask accident
scenario even though the Gantry Crane and fuel handling bridge are:
1. Designed to safely handle their respective loads (i.e., a
loaded transfer cask and a spent fuel assembly, respectively; and
2. In compliance with the design and administrative requirements
addressed in NUREG-0612, ``Control of Heavy Loads at Nuclear Power
Plants.''
A loaded cask transfer drop is a very unlikely event because of
the numerous Gantry Crane safety features described in the above
safety Analysis Report. These features described above include:
1. Gantry Crane Administrative Safety Features;
2. Gantry Crane Design Safety Features;
3. General Gantry Crane Control System Design Safety Features;
4. Gantry Crane Radio Control System Design Safety Features;
5. Hoist Design Safety Features; and
6. Trolly and Bridge Design Safety Features.
The updated cask drop accident scenario will not create the
possibility of a new or different type of accident than previously
evaluated in the SAR, because the DSAR [defueled SAR] currently
evaluates a cask drop event. The cask drop scenario evaluated in the
above Safety Analysis Report just updates the existing cask drop
analysis. The updated cask drop analysis only:
1. Identifies the type of spent fuel cask that Rancho Seco will
use;
2. Results in a change to the calculated dose consequence
associated with the current, bounding, design basis accident (i.e.,
the FHA); and
3. Results in a change to the existing Rancho Seco cask drop
analysis.
The updated, design basis, cask drop event will not involve a
significant reduction in the margin of safety, because the
conservatively calculated dose consequence associated with the
postulated drop of a spent fuel transfer cask is:
1. Relatively small (i.e., 224 mrem) compared to the eight
accident limits and previously calculated accident doses summarized
above;
2. A very unlikely event;
3. Not a significant hazard; and
4. Not a public health and safety concern.
This conclusion is the same for the FHA, which is the current,
bounding, Rancho Seco design basis accident.
Also, the Emergency Planning Zone remains unchanged for this
updated, cask drop accident scenario. No significant changes to the
Rancho Seco Emergency Plan result from this proposed change to the
updated, design basis accident at Rancho Seco.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. The staff also reviewed the proposed editorial changes for
no significant hazards consideration. The proposed editorial changes do
not affect the design or operation of the facility and also satisfy the
three standards of 10 CFR 50.92(c). Therefore, the NRC staff proposes
to determine that the requested amendment involves no significant
hazards consideration.
Local Public Document Room location: Central Library, Government
Documents, 828 I Street, Sacramento, California 95814
Attorney for licensee: Dana Appling, Esq., Sacramento Municipal
Utility District, P.O. Box 15830, Sacramento, California 95852-1830
NRC Section Chief: Michael T. Masnik
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 amendment request: March 12, 1998, as supplemented April
24, August 20 and November 20, 1998, and February 3, 1999
Description of amendment request: The proposed amendments would
revise the Technical Specifications (TS) of each unit to conform with
NUREG-1431, Revision 1, ``Standard Technical Specifications--
Westinghouse Plants.'' The Commission had previously issued a Notice of
Consideration of Issuance of Amendments in the Federal Register on May
25, 1999, (64 FR 28218) covering all the proposed changes that were
within the scope of NUREG-1431. The following descriptions and no
significant hazard analyses cover only those items that are beyond the
scope of NUREG-1431. Associated with each change are administrative/
editorial changes which would make the new or revised requirements fit
into the format of NUREG-1431.
1. The Standard Technical Specification (STS) terms FQW(Z) and
FQC(Z) in Limiting Condition for Operation (LCO) 3.2.1 would be deleted
and the terms FQ(Z), ``steady state'' limit and ``transient'' limit
would be used. (Significant Hazards Evaluation A)
2. The STS wording in Required Action 3.2.4.A to ``reduce'' thermal
power would be revised to ``limit'' thermal power to allow entry into
the LCO applicability during startup when QPTR may be in excess of 1.02
due to transient core conditions which are usually self-correcting. (A)
3. The Applicability of LCO 3.2.4 would be revised to be consistent
with the Applicability for the AFD LCO to eliminate subtle differences
between the two LCO Applications which were previously the same. (M)
4. The Reactor Coolant System Loop Test specified in the TS LCO 3/
4.10.4 would not be included. (L-1)
5. A new Action would be added to the Emergency Core Cooling System
(ECCS)--Shutdown LCO 3.5.3. The new Action deals with the centrifugal
charging subsystem. (L-2)
6. The Reactor Coolant Pump (RCP) seal injection flow requirements
of 3.5.5 would be revised. The requirement to verify a single operating
point would be changed to require verification of a range of values on
an operating curve. (M)
7. The time allowed to reduce the power range neutron flux setpoint
in 3.7.1 to within the required limit would be extended and made
applicable in Mode 1 only. (L-3 and L-3a)
[[Page 46444]]
8. The Actions in 3.7.2 for an inoperable Main Steam Isolation
Valve (MSIV) would be revised to take credit for the redundant MSIVs in
each steam line. (L-4)
9. An Action would be added to the Service Water (SW) LCO 3.7.8
that accounts for the redundant automatic turbine building isolation
valves in each Farley SW train. (L-5)
10. The diesel generator accelerated Test Table 3.8.1-1 would be
deleted. (LA)
11. The AC Sources--Shutdown surveillance 3.8.2.1 would be revised
to more clearly state the required surveillances. (L-6 and L-6a)
12. The Actions 3.8.4 and 3.8.9 for an inoperable SW intake
structure Battery and Distribution System would be revised to more
accurately reflect the Farley design. (L-7)
13. The STS footnote to ESFAS Table 3.3.2-1 would be revised to be
consistent with the design of the Farley main steam system. (L-8)
14. A new Condition C would be added to LCO 3.3.4 to address
actions associated with the source range neutron flux monitor. (M)
15. LCO 3.3.5 would be revised to accommodate the addition of a
degraded grid alarm function. (M)
16. The specific title in 5.1.2 for the control room command
function would be replaced with a more general description. (L-9)
17. The specific title in 5.3.1 of Health Physics Supervisor would
be replaced with a more general description. (A)
18. The inspection frequency specified in 5.5.7 for the RCP
flywheel would be revised to be consistent with the NRC-approved WCAP-
14535A, ``Topical Report on RCP Flywheel Inspection Elimination,''
November 1996. (L-10)
19. The Health Physics Supervisor title in 5.7.1.c would be
replaced with a more general description. (L-11)
20. The Emergency Diesel General (DG) Failure Report in 5.6.7 would
be revised to be consistent with the latest Farley commitments for DG
failure tracking and reporting. (L-12)
21. A note would be added to Surveillance Requirement (SR) 3.4.1.4
that would not require this surveillance until 7 days after reaching
greater than 90% power. (M)
22. SR 3.4.5.2 would require verification that steam generator
secondary side water levels are 74% (wide range). (M)
23. LCO 3.4.15 would differ from the STS in several aspects. One
aspect would extend the Allowable Outage Time from 7 days to 30 days
for an inoperable leakage detection system. (L-13)
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. Each proposed out-of-scope item described above is
followed in parenthesis by either an A (for administrative changes), an
M (for changes which would be more restrictive), an LA (for
requirements that would be removed from the TS), or an L and a number
(for changes that would be less restrictive). Following are the no
significant hazards analyses corresponding to each of these
designations.
[A--Administrative Changes]
1. Does the change involve a significant increase in the
probability or consequences of an accident previously evaluated?
The proposed changes involve reformatting, renumbering, and
rewording of the CTS. These changes involve no technical revisions
to the CTS and were made to conform with the format and style of the
STS. As such, these changes are administrative in nature and do not
impact initiators of analyzed events or safety analyses assumptions
relative to the mitigation of accidents or transient events.
Therefore, these changes do not involve a significant increase in
the probability or consequences of an accident previously evaluated.
2. Does the change create the possibility of a new or different
kind of accident from any accident previously evaluated?
The proposed changes do not involve a physical alteration of the
plant (no new or different type of equipment will be installed) or
changes in the methods governing normal plant operation. The
proposed changes will not impose any new or different requirements
or eliminate any existing requirements. In addition, the change does
not alter assumptions made in the safety analyses and licensing
basis. Therefore, the changes do 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?
The proposed changes are administrative in nature and do not
involve any technical changes. As such, these changes do not impact
any safety analysis assumptions and no question of safety is
involved. Therefore, the changes do not involve a significant
reduction in a margin of safety.
[M--More Restrictive]
1. Does the change 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. These more stringent requirements are
not assumed to be initiators of analyzed events and will not alter
assumptions relative to mitigation of accident or transient events.
The changes are evaluated to ensure no previously analyzed accident
has been adversely affected. The more stringent requirements are
imposed to ensure process variables, structures, systems and
components are maintained consistent with the safety analyses and
licensing basis. These changes will not alter assumptions relative
to mitigation of an accident or transient event nor will they alter
the operation of process variables, structures, systems, or
components described in the safety analyses. Therefore, these
changes do not involve a significant increase in the probability or
consequences of an accident previously evaluated.
2. Does the change create the possibility of a new or different
kind of accident from any accident previously evaluated?
The proposed changes add more restrictive requirements to the TS
or make existing requirements more restrictive. The proposed changes
do not involve a physical alteration of the plant (no new or
different type of equipment will be installed) or a change in the
methods governing normal plant operation. The proposed changes do
impose new or different requirements. However, these changes are
consistent with assumptions made in the safety analysis and
licensing basis. Thus, these changes do 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?
The proposed changes add more restrictive requirements to the TS
or make existing requirements more restrictive and have been
evaluated to ensure consistency with the safety analysis and
licensing basis. As such, these changes do not impact any safety
analyses assumptions and no question of safety is involved.
Therefore, these changes do not involve a reduction in a margin of
safety.
[LA--Removal of Requirements]
1. Does the change involve a significant increase in the
probability or consequences of an accident previously evaluated?
The proposed changes relocate requirements from the CTS to a
licensee controlled document. The document containing the relocated
requirements will be maintained using the provisions of 10 CFR
50.59. Therefore, the proposed changes will only reduce the level of
regulatory control on these requirements. The level of regulatory
control has no impact on the probability or the consequences of an
accident previously evaluated. Thus, the proposed changes do not
involve a significant increase in the probability or consequences of
an accident previously evaluated.
2. Does the change create the possibility of a new or different
kind of accident from any accident previously evaluated?
The proposed changes relocate requirements from the CTS to a
licensee controlled document. The changes do not involve a physical
alteration of the plant (no new or different type of equipment will
be installed) or a change in the methods governing normal plant
operation. In addition, the changes do not impose any new or
different requirements or eliminate any
[[Page 46445]]
existing requirements. The changes do not alter assumptions made in
the safety analyses and licensing basis. Thus, the changes do 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?
The proposed changes relocate requirements from the CTS to a
licensee controlled document for which future changes will be
evaluated pursuant to the requirements of 10 CFR 50.59. The proposed
changes do not reduce a margin of safety because they have no impact
on any safety analysis assumptions. Therefore, these changes do not
involve a significant reduction in a margin of safety.
[L-1--Less Restrictive]
1. Does the change involve a significant increase in the
probability or consequences of an accident previously evaluated?
The proposed change involves deleting the CTS 3/4.10.4, Reactor
Coolant Loops Test Exception, requirements and does not result in
any hardware changes. The proposed change deletes a test exception
LCO that is no longer used or required at FNP. The natural
circulation test, for which this exception is designed, was only
required to be performed at FNP during the initial plant startup
test program. The proposed changes do not impact the capability of
the plant or any equipment to provide the required safety function
as described in the FSAR. In addition, the results of the analyses
described in the FSAR remain bounding. Also, the proposed changes do
not impose any new safety analyses limits or alter the plants
ability to detect and mitigate events. Therefore, this change does
not involve a significant increase in the probability or
consequences of an accident previously evaluated.
2. Does the change create the possibility of a new or different
kind of accident from any accident previously evaluated?
The proposed change involves changing the CTS requirements to
delete a test exception that is no longer used and does not
necessitate a physical alteration of the plant or changes in
parameters governing normal plant operation. Thus, this change does
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?
The proposed change, which deletes CTS 3/4.10.4 does not involve
a significant reduction in a margin of safety. The proposed change
does not impact any safety analysis assumptions and does not impose
any new safety analyses limits or alter the plants ability to detect
and mitigate events. Therefore, the proposed change does not impact
any margin of safety.
[L-2--Less Restrictive]
1. Does the change involve a significant increase in the
probability or consequences of an accident previously evaluated?
This change does not result in any hardware changes. The ECCS
components covered by this TS are not assumed to be initiators of
any analyzed event. Therefore, this change does not involve a
significant increase in the probability of an accident previously
evaluated. The change would allow the required ECCS centrifugal
charging subsystem to be inoperable for up to 72 hours providing the
remaining operable ECCS components can provide the flow equivalent
to a single operable train which will ensure 100% of the flow
assumed in the safety analyses. Since the ability of the ECCS to
perform its safety function is not lost, this change does not
involve a significant increase in the consequences of an accident
previously evaluated.
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 necessitate a physical alteration
of the plant (no new or different type of equipment will be
installed) or changes in parameters governing normal plant
operation. The proposed change will only more accurately define the
minimum equipment required to be operable to perform the ECCS
function while in this Condition. Therefore, this change does not
create the possibility of a new or different kind of accident from
any previously evaluated.
3. Does this change involve a significant reduction in a margin
of safety?
The proposed change, which allows operation to continue for up
to 72 hours with components inoperable in the required ECCS
centrifugal charging subsystem, is acceptable based on the remaining
ECCS components providing 100% of the required ECCS flow, the small
probability of an event occurring in 72 hours that would require the
ECCS, and the reduced potential for a unit transient resulting from
the shutdown required by current TS for an inoperable required ECCS
centrifugal charging subsystem. The proposed allowed outage time of
72 hours for this condition is consistent with the time currently
allowed for one train of ECCS to be inoperable in Modes 1-3. The
exposure of the unit to the small probability of an event requiring
ECCS during this time is insignificant and offset by the benefit
gained through avoiding unnecessary plant transients. Therefore,
this change does not involve a significant reduction in margin of
safety.
[L-3--Less Restrictive]
1. Does the change involve a significant increase in the
probability or consequences of an accident previously evaluated?
The proposed changes extend the time allowed to adjust the Power
Range Neutron Flux-High trip setpoints for the case of two or more
inoperable MSSVs per SG and/or positive Moderator Temperature
Coefficient (MTC) and removes the requirement to adjust the Power
Range Neutron Flux-High trip setpoints only one MSSV is inoperable
and the MTC is zero or negative and do not result in any hardware or
operating procedure changes. The affected trip setpoints, the
requirement to reduce them or the time allowed to adjust them are
not assumed to be an initiator of any analyzed event. In addition,
the affected trip setpoints, the requirement to reduce them and the
time allowed to adjust them are not a precursor to any accident
analyses. Therefore, the proposed changes do not increase the
probability of an accident previously evaluated. The Power Range
Neutron Flux-High trip functions to mitigate the consequences of an
analyzed event by shutting down the reactor. The proposed changes
continue to provide assurance that the setpoints will be properly
adjusted to ensure the system functions as assumed in the applicable
safety analyses. Therefore, the consequences of an accident are not
significantly increased.
2. Does the change create the possibility of a new or different
kind of accident from any accident previously evaluated?
The proposed changes do not necessitate a physical alteration of
the plant (no new or different type of equipment will be installed)
or changes in parameters governing normal plant operation. The
proposed changes still ensure the operability of the trip function
at the correct setpoint and will facilitate the adjustment of the
setpoints such that the probability of error is minimized. Thus,
these changes do 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?
The time allowed to adjust the setpoints of the affected
instrumentation is not a specific assumption of any safety analysis.
For the case of a single inoperable MSSV with a zero or negative
MTC, a reactor power reduction alone is sufficient to limit primary
side heat generation such that overpressurization of the secondary
side is precluded for any RCS heatup event. Furthermore, for this
case there is sufficient total steam flow capacity provided by the
turbine and the remaining OPERABLE MSSVs to preclude
overpressurization in the event of an increased reactor power due to
reactivity insertion, such as in the event of an uncontrolled RCCA
bank withdrawal at power. The proposed changes still ensure the
setpoints are reduced consistent with the assumptions of the safety
analysis for the case of two or more inoperable MSSVs or a positive
MTC. The proposed changes also reduce the potential for an
inadvertent reactor trip that could result from adjusting the trip
setpoints too quickly. As such, any reduction in a margin of safety
will be insignificant and will likely be offset by the benefit
gained from the reduced potential for an inadvertent plant trip.
[L3a--Less Restrictive]
1. Does the change involve a significant increase in the
probability or consequences of an accident previously evaluated?
The proposed change clarifies the Action requirement to reduce
the power range neutron flux-high trip setpoint in Modes 2 and 3 and
does not result in any hardware or operating procedure changes. The
proposed change adds a note to the Action which specifies that the
Action is only required in Mode 1. In Modes 2 and 3, other reactor
trips (power range low and source range high) provide the required
protection consistent with the acceptance criteria of the safety
analysis. Therefore, the Action is not required in these Modes. The
affected trip
[[Page 46446]]
setpoints are not assumed to be an initiator of any analyzed event.
In addition, the affected trip setpoints are not a precursor to any
accident analyses. Therefore, the proposed change does not increase
the probability of an accident previously evaluated. The affected
reactor trip functions mitigate the consequences of an analyzed
event by shutting down the reactor. The proposed change continues to
provide assurance that the required reactor trip functions operate
as assumed in the applicable safety analyses. Therefore, the
consequences of an accident are not significantly increased.
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 necessitate a physical alteration
of the plant (no new or different type of equipment will be
installed) or changes in parameters governing normal plant
operation. The proposed change still ensures the operability of the
reactor trip function at the correct setpoint for the correct Mode
of operation. Thus, this change does 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?
The proposed change does not affect the ability of the MSSVs and
reactor trip system to mitigate the applicable transients consistent
with the assumptions of the safety analysis. The proposed change
continues to ensure the acceptance criteria of the applicable safety
analyses are met (primary and secondary system pressures are limited
to within the required values). As such, any reduction in a margin
of safety will be insignificant and will likely be offset by the
benefit gained from the reduced potential for an inadvertent plant
trip that could result from an error in adjusting the power range
neutron flux-high trip setpoint (unnecessary in Mode 2).
[L-4--Less Restrictive]
1. Does the change involve a significant increase in the
probability or consequences of an accident previously evaluated?
The proposed change revises the Actions of the MSIV LCO in order
to take credit for the redundant MSIV valves in each steam line.
This change does not result in any hardware or operating procedure
changes. The MSIVs are not assumed to be an initiator of any
analyzed event and function to isolate the steam lines to mitigate
analyzed events. As a result, the revision of this TS requirement
does not affect the probability of an accident previously evaluated.
The proposed change continues to provide adequate assurance that the
MSIVs are either capable of performing their intended safety
function or that the safety function has been performed (steam line
isolated) or that power is reduced. The proposed change continues to
limit plant operation when a single failure could prevent the
isolation function from being accomplished. Therefore, the proposed
change does not involve a significant increase in the consequences
of an accident previously evaluated.
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 necessitate a physical alteration
of the plant (no new or different type of equipment will be
installed) or changes in parameters governing normal plant
operation. The proposed change only affects the Actions of the MSIV
LCO. The proposed change continues to ensure the MSIVs are either
capable of isolating the steam lines or that the steam lines are
isolated or power reduced. Thus, this change does 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?
The proposed change continues to ensure the MSIVs are either
capable of isolating the steam lines or that the steam lines are
isolated or power reduced. The proposed change continues to limit
plant operation when a single failure could prevent the isolation
function from being accomplished. Therefore, the proposed change
also continues to preserve the assumptions of the applicable safety
analyses. As such, the proposed change does not impact the
assumptions of the applicable safety analyses. Therefore, the
proposed change does not involve a significant reduction in a margin
of safety.
[L-5--Less Restrictive]
1. Does the change involve a significant increase in the
probability or consequences of an accident previously evaluated?
The proposed change revises the Actions of the SWS LCO in order
to take credit for the redundant automatic turbine building
isolation valves in each train of SWS. This change does not result
in any hardware or operating procedure changes. The turbine building
isolation valves are not assumed to be an initiator of any analyzed
event and function to isolate the SWS flow to non-essential
components. As a result, the revision of this TS requirement does
not affect the probability of an accident previously evaluated. The
proposed change continues to provide adequate assurance that the
turbine building isolation valves are either capable of performing
their intended safety function and accommodate a single failure or
that the unit is placed in a condition where the function performed
by these valves is no longer required. The proposed change continues
to limit plant operation when a single failure could prevent the
isolation function of these valves from being accomplished.
Therefore, the proposed change does not involve a significant
increase in the consequences of an accident previously evaluated.
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 necessitate a physical alteration
of the plant (no new or different type of equipment will be
installed) or changes in parameters governing normal plant
operation. The proposed change only affects the Actions of the SWS
LCO. The proposed change continues to ensure the turbine building
isolation valves are either capable of isolating the SWS system and
accommodating a single failure or that the unit is placed in a
condition where this isolation function is no longer required. Thus,
this change does 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?
The proposed change continues to ensure the turbine building
isolation valves are either capable of isolating the non-essential
SWS loads and accommodating a single failure or that the unit is
placed in a condition where the isolation function is no longer
required. The proposed change continues to limit plant operation
when a single failure could prevent the isolation function from
being accomplished. Therefore, the proposed change also continues to
preserve the assumptions of the applicable safety analyses. As such,
the proposed change does not impact the assumptions of the
applicable safety analyses. Therefore, the proposed change does not
involve a significant reduction in a margin of safety.
[L-6--Less Restrictive]
1. Does the change involve a significant increase in the
probability or consequences of an accident previously evaluated?
The elimination of the requirement to meet surveillance tests
that verify functions which are not required in the Mode of
applicability of this TS will not increase the probability of any
accident previously evaluated. The proposed surveillance testing
continues to provide adequate assurance of the operability of the
required AC Source functions and therefore, does not involve an
increase in the consequences of any accident previously evaluated.
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 introduce a new mode of plant
operation and does not involve a physical modification to the plant.
Therefore, it does 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?
This change does not involve a significant reduction in a margin
of safety since the operability of the required AC Source functions,
continues to be determined in the same manner. Elimination of the
surveillance test requirements for AC Source functions not required
in these Modes does not impact the capability of the AC Sources to
perform their safety function in these Modes.
[L6a--Less Restrictive]
1. Does the change involve a significant increase in the
probability or consequences of an accident previously evaluated?
The inclusion of a note consistent with the STS to provide an
allowance not to perform certain surveillance tests on the AC Source
required operable by the TS will not increase the probability of any
accident previously evaluated. The required surveillance testing
must still be performed (but not on the AC
[[Page 46447]]
Source while it is required operable by the TS) and will continue to
provide adequate assurance of the operability of the required AC
Source functions. Therefore, this change does not involve an
increase in the consequences of any accident previously evaluated.
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 introduce a new mode of plant
operation and does not involve a physical modification to the plant.
Therefore, it does 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?
This change does not involve a significant reduction in a margin
of safety since the operability of the required AC Source functions,
continues to be determined in the same manner. The allowance not to
perform certain surveillance tests on the AC Source equipment when
that equipment serves to meet the TS minimum required power source
ensures a stable shutdown power supply to the unit and does not
impact the capability of the AC Sources to perform their safety
function in these Modes.
[L-7--Less Restrictive]
1. Does the change involve a significant increase in the
probability or consequences of an accident previously evaluated?
The proposed change effectively provides a longer allowed outage
time for the Service Water Intake Structure (SWIS) DC distribution
and battery systems. The proposed allowed outage time is consistent
with the time allowed for a Service Water train to be inoperable.
The DC power sources or their associated allowed outage times are
not assumed to be initiators of any analyzed event. As such, the
proposed change will not increase the probability of any accident
previously evaluated. The appropriate required actions consistent
with that for the equipment rendered inoperable must still be
performed. The proposed actions will continue to provide adequate
assurance of plant safety in the same manner as if the affected
equipment were inoperable for reasons other than power availability.
Therefore, this change does not involve an increase in the
consequences of any accident previously evaluated.
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 introduce a new mode of plant
operation and does not involve a physical modification to the plant.
Therefore, it does 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?
This change does not involve a significant reduction in a margin
of safety since the inoperability of the SWIS distribution and
battery systems affect only the Service Water system and the time
allowed for restoration of an inoperable Service Water train remains
unchanged. The allowance to declare the affected equipment
inoperable and take the associated equipment TS actions continues to
ensure plant safety by providing the same appropriate remedial
measures for the affected equipment as would be applicable if that
equipment were inoperable for reasons other than power availability.
Therefore, the proposed change does not significantly impact any
margin of safety.
[L-8--Less Restrictive]
1. Does the change involve a significant increase in the
probability or consequences of an accident previously evaluated?
The proposed change involves upgrading the ESFAS TS to more
closely agree with the FNP design and safety analysis and does not
result in any hardware changes. The proposed change revises the
applicability for the initiating functions of the main steam line
isolation function such that when a main steam line isolation valve
is closed and the isolation function is accomplished, the automatic
initiation of this function is no longer required operable. The
ESFAS is not assumed to be an initiator of any analyzed event. The
role of the ESFAS is in mitigating and thereby limiting the
consequences of accidents. The proposed change continues to
adequately ensure the operability of the ESFAS main steam line
isolation function when the lines are unisolated and thereby ensures
the protection provided by the function remains operable when
required. Therefore, the results of the analyses described in the
FSAR remain bounding. Additionally, the proposed changes do not
impose any new safety analyses limits or alter the plants ability to
detect and mitigate events. Therefore, this change does not involve
a significant increase in the probability or consequences of an
accident previously evaluated.
2. Does the change create the possibility of a new or different
kind of accident from any accident previously evaluated?
The proposed change involves upgrading the ESFAS TS to more
closely agree with the FNP design and safety analysis and does not
necessitate a physical alteration of the plant (no new or different
type of equipment will be installed) or changes in parameters
governing normal plant operation. Thus, this change does 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? The proposed change, which upgrades the ESFAS TS to be
more consistent with the FNP design and safety analysis does not
involve a significant reduction in a margin of safety. The proposed
change revises the Mode of applicability for the main steam line
isolation ESFAS function. The proposed change continues to
adequately ensure the operability of the isolation function when it
is required and thereby ensures the protection provided by the
function also remains available when required. As such, the results
of the analyses described in the FSAR remain bounding and this
change does not have a significant impact on any design basis safety
analysis.
[L-9--Less Restrictive]
1. Does the change involve a significant increase in the
probability or consequences of an accident previously evaluated?
The proposed change involves changing the CTS administrative
controls requirements regarding the Shift Supervisor (SS)
responsibility to more closely agree with the STS requirements and
does not result in any hardware changes. The requirement to issue
annual directives regarding the SS responsibilities is deleted. The
title Shift Supervisor is replaced with responsible SRO. In
addition, an allowance for an RO (in Modes 5 and 6) to temporarily
replace the SS is added. The proposed change also eliminates the
specific restriction against the STA temporarily replacing the SS.
The proposed changes do not impact the capability of the plant or
any equipment to provide the required safety function as described
in the FSAR. In addition, the results of the analyses described in
the FSAR remain bounding. Additionally, the proposed changes do not
impose any new safety analyses limits or alter the plants ability to
detect and mitigate events. Therefore, this change does not involve
a significant increase in the probability or consequences of an
accident previously evaluated.
2. Does the change create the possibility of a new or different
kind of accident from any accident previously evaluated?
The proposed change involves changing the TS administrative
controls regarding the responsibilities of the SS to more closely
agree with the STS requirements and eliminates the title Shift
Supervisor and does not necessitate a physical alteration of the
plant or changes in parameters governing normal plant operation.
Thus, this change does 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?
The proposed changes, which revise the TS administrative
controls requirements for SS responsibilities to be consistent with
the STS requirements and eliminate the title Shift Supervisor do not
involve a significant reduction in a margin of safety. The proposed
changes do not impact any safety analysis assumptions and do not
impose any new safety analyses limits or alter the plants ability to
detect and mitigate events. Therefore, the proposed changes do not
impact any margin of safety.
[L-10--Less Restrictive]
1. Does the change involve a significant increase in the
probability or consequences of an accident previously evaluated?
The proposed change affects only the interval allowed by the TS
surveillance to perform RCP flywheel inspections. The time allowed
between flywheel inspections is not specifically assumed to be a
precursor or initiator of any analyzed event. The studies performed
to justify the proposed time interval have shown it to be adequate
to detect any flaws or degradation in the RCP flywheel. As such, the
proposed change does not affect the probability of any initiating
events assumed in the accident analyses. The proposed change will
maintain an acceptable
[[Page 46448]]
level of safety by continuing to require RCP flywheel inspections at
an interval shown to be adequate. Consequently, the proposed change
will not have any affect on the consequences of any accident
previously evaluated.
2. Does the proposed change create the possibility of a new or
different kind of accident from any accident previously evaluated?
The proposed change does not involve a physical alteration of
the plant (no new or different types of equipment will be installed)
or changes in parameters governing normal plant operation. The
proposed change only affects the interval allowed by the TS to
inspect each RCP flywheel. The interval remains adequate to detect
any degradation. Therefore, the possibility of a new or different
kind of accident is not created by the proposed change.
3. Does the proposed change involve a significant reduction in a
margin of safety?
The proposed change affects the interval allowed by the TS to
inspect RCP flywheels. The proposed interval is based on the
findings of WCAP-14535A and the associated NRC SER. The WCAP
concludes that continued inspections of RCP flywheels are not
necessary and overall plant safety could be increased by eliminating
the inspections and reducing man rem dose as well as the potential
for flywheel damage during disassembly and reassembly for
inspection. The NRC SER requires the inspection of RCP flywheels be
retained but the interval increased to once every 10 years. As such,
the proposed change continues to conservatively assure the
operability of the RCP flywheel while reducing man rem exposure and
the potential for damage from disassembly and reassembly. Therefore,
the proposed change does not involve a significant reduction in a
margin of safety.
[L-11--Less Restrictive]
1. Does the change involve a significant increase in the
probability or consequences of an accident previously evaluated?
The proposed change involves the revision of the term health
physics supervisor to health physics supervision for the purpose of
specifying the frequency of radiation surveillances in RWPs. The
proposed change continues to provide adequate assurance that the
radiation surveillances are performed within acceptable frequencies.
The proposed change does not impact the capability of the plant or
any equipment to provide the required safety function as described
in the FSAR, or increase the potential radiation exposure of plant
personnel. In addition, the results of the analyses described in the
FSAR remain bounding. Additionally, the proposed change does not
impose any new safety analyses limits or alter the plants ability to
detect and mitigate events. Therefore, this change does not involve
a significant increase in the probability or consequences of an
accident previously evaluated.
2. Does the change create the possibility of a new or different
kind of accident from any accident previously evaluated?
The proposed change involves the supervisors who specify the
radiation surveillance frequencies in high radiation areas and does
not necessitate a physical alteration of the plant or changes in
parameters governing normal plant operation. Thus, this change does
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?
The proposed change, which revises the TS requirements for the
personnel who specify the frequencies of radiation surveillances in
high radiation areas. The proposed change allows additional
supervisory personnel to specify the required frequencies. The
proposed change does not impact any safety analysis assumptions and
does not impose any new safety analyses limits or alter the plants
ability to detect and mitigate events. In addition, the proposed
change continues to ensure adequate surveillances are performed in
high radiation areas. Therefore, the proposed change does not impact
any margin of safety.
[L-12--Less Restrictive]
1. Does the change involve a significant increase in the
probability or consequences of an accident previously evaluated?
The proposed change involves changing the CTS administrative
controls requirements regarding the Emergency Diesel Generator (EDG)
failure reporting requirement and does not result in any hardware
changes. The proposed change potentially reduces the number of
reports received by the NRC and revises the content to include valid
failures and demands. The proposed change continues to provide
adequate information to assess the EDG reliability at FNP. The
proposed change does not impact the capability of the plant or any
equipment to provide the required safety function as described in
the FSAR. In addition, the results of the analyses described in the
FSAR remain bounding. Additionally, the proposed change does not
impose any new safety analyses limits or alter the plants ability to
detect and mitigate events. Therefore, this change does not involve
a significant increase in the probability or consequences of an
accident previously evaluated.
2. Does the change create the possibility of a new or different
kind of accident from any accident previously evaluated?
The proposed change involves changing the TS administrative
controls regarding the required EDG report to more closely agree
with the STS requirements and does not necessitate a physical
alteration of the plant or changes in parameters governing normal
plant operation. Thus, this change does 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?
The proposed change, which revises the TS administrative
controls requirement for an annual EDG report to be consistent with
the STS requirement does not involve a significant reduction in a
margin of safety. The proposed change does not impact any safety
analysis assumptions and does not impose any new safety analyses
limits or alter the plants ability to detect and mitigate events. In
addition the proposed change continues to provide sufficient
information to assess the reliability of the EDG at FNP. Therefore,
the proposed change does not impact any margin of safety.
[L-13--Less Restrictive]
1. Does the proposed change involve a significant increase in
the probability or consequences of any accident previously
evaluated?
The proposed change extends the time allowed to restore an
inoperable RCS leakage detection instrument to operable status. The
CTS allow 7 days for restoration of the automatic RCS leak detection
instrument and the proposed change would allow 30 days for
restoration. However, adequate information continues to be furnished
to the plant staff to assure that RCS leakage does not go
undetected. In addition to the remaining operable automatic RCS leak
detection instrument, the TS required actions provide remedial
measures that ensure RCS leakage continues to be monitored by
diverse means. As such, potential RCS leakage will not go undetected
and operation with one required leak detection instrument inoperable
continues to be limited by the TS. Therefore, the proposed change
does not involve a significant increase in the probability or
consequences of any accident previously evaluated.
2. Does the proposed change create the possibility of a new or
different kind of accident from any accident previously evaluated?
The proposed change does not introduce any new equipment into
the plant or alter the manner in which existing equipment will be
operated. Therefore the proposed change will not create the
possibility of a new or different kind of accident from any accident
previously evaluated.
3. Does the proposed change involve a significant reduction in a
margin of safety?
The applicable required actions and remaining operable leakage
detection monitor provide adequate information to the plant staff to
ensure that RCS leakage does not go undetected. In addition,
operation with one required leak detection instrument inoperable
continues to be limited by the TS (30 days). As such, potential RCS
leakage will not go undetected and operation in the condition where
a single failure could cause a loss of automatic leakage detection
continues to be limited and therefore, this 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: Houston-Love Memorial Library,
212 W. Burdeshaw Street, Post Office Box 1369, Dothan, Alabama 36302.
Attorney for licensee: M. Stanford Blanton, Esq., Balch and
Bingham, Post
[[Page 46449]]
Office Box 306, 1710 Sixth Avenue North, Birmingham, Alabama.
NRC Section Chief: Richard L. Emch, Jr.
Southern Nuclear Operating Company, Inc., Georgia Power Company,
Oglethorpe Power Corporation, Municipal Electric Authority of Georgia,
City of Dalton, Georgia, Docket Nos. 50-321 and 50-366, Edwin I. Hatch
Nuclear Plant, Units 1 and 2, Appling County, Georgia
Date of amendment request: July 29, 1999.
Description of amendment request: The proposed amendments would
change the Limiting Condition for Operation 3.1.7, ``Standby Liquid
Control (SLC) System.'' The proposed amendments would change ``greater
than the Region B limits,'' which could be misleading, to ``within the
Region B limits.''
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. Does the change involve a significant increase in the
probability or consequences of an accident previously evaluated?
The proposed changes to the Unit 1 and Unit 2 Technical
Specifications do not increase the probability or consequences of
any previously evaluated accident or transient. These changes are
administrative in nature only and are intended to revise a
misleading statement in Condition A of Limiting Condition for
Operation (LCO) 3.1.7, ``Standby Liquid Control (SLC) System.'' The
change ensures the proper condition is entered when expected and the
sodium pentaborate solution temperature, concentration, and volume
limits are not exceeded without appropriate actions being taken. As
currently written, Condition A of LCO 3.1.7 could be entered
whenever the sodium pentaborate solution is not within Region A
limits, but is greater than Region B limits as depicted in Unit 1
and Unit 2 Technical Specifications Figures 3.1.7-1 and 3.1.7-2.
This is incorrect; Condition A should be entered whenever the
solution is not within Region A limits, but is within Region B
limits. If the solution is not within Region A limits and is greater
than Region B limits, both Standby Liquid Control subsystems are
inoperable and Condition C should be entered.
Technical Specifications Figure 3.1.7-1 displays the sodium
pentaborate solution volume versus concentration requirements;
Figure 3.1.7-2 displays the solution concentration versus
temperature requirements. Each figure contains three areas: Region
A, Region B, and the area not in either Region A or Region B. Region
A is the permissible region of continuous operation and is
represented by a four- or five-sided area. Region B is the original
licensing basis region and is represented by a four-sided area. If
the sodium pentaborate solution temperature, concentration, and
volume combinations are within Region A, the requirements of 10 CFR
50.62, ``Requirements for reduction of risk from anticipated
transients without scram (ATWS) events for light-water-cooled
nuclear power plants,'' are met, no condition applies, and no
actions need be taken. If solution temperature, concentration, and
volume combinations are not within Region A, but within Region B,
then the original licensing basis is met and operation within this
region is acceptable for up to 72 hours (Unit 1 FSAR, section 3.8.4,
Revision 6, page 3.8-6; Unit 2 FSAR, section 4.2.3.4.3, Revision 7,
page 4.2-98). If solution temperature, concentration, and volume
combinations are not within either region, then the ability of the
Standby Liquid Control system to shut down the reactor is not
assured and only eight hours is acceptable to restore the solution
to at least within Region B before the plant must be shut down.
Condition A contains misleading wording which could allow
operation outside both Region A and Region B for more than eight
hours. Specifically, it could be interpreted that Condition A allows
the sodium pentaborate solution temperature, concentration, and
volume to be greater than Region B limits for up to 72 hours.
Because Region B is demarcated by a four-side area, the terms
``within Region B'' and ``greater than Region B limits'' could be
interpreted to indicate different, and mutually exclusive, areas of
Figures 3.1.7-1 and 3.1.7-2. Indeed, ``greater than Region B
limits'' could be interpreted to refer to most or all of the area
neither in Region A nor Region B. For example, 20 weight percent
sodium pentaborate solution at 50 deg.F is a point on Figure 3.1.7-2
which is ``greater than the Region B limits,'' yet it is a point at
which the solution will precipitate in the storage tank rendering
the system incapable of injecting the proper amount of sodium
pentaborate into the reactor pressure vessel. Obviously, both
Standby Liquid Control subsystems would be inoperable if the
solution were at this point and Condition C should be entered to
limit severely the time the unit may continue to operate with the
solution in this state. However, the wording of Condition A could
cause an erroneous interpretation which would inappropriately extend
this time from eight to 72 hours.
The proposed changes correct the wording of Condition A to
ensure this condition is not entered inappropriately and to ensure
the proper condition is entered for those combinations of solution
temperature, concentration, and volume not within Region A or Region
B. These changes do not increase the probability of any previously
evaluated accident or transient because they are administrative in
nature and do not alter any plant operation or design features or
requirements which could result in systems or components performing
closer to their operational or design limits and thereby increasing
the possibility of a failure. These changes do not increase the
consequences of any previously evaluated accident or transient
because they ensure the sodium pentaborate solution limits are not
exceeded without appropriate actions being taken thereby ensuring
the Standby Liquid Control system is capable of mitigating the
consequences of an ATWS event.
2. Do the proposed changes create the possibility of a new or
different type of accident from any previously evaluated?
The proposed changes to the Unit 1 and Unit 2 Technical
Specifications do not create the possibility of a new or different
type of accident from any previously evaluated. The changes are
administrative in nature only and are intended to clarify Condition
A of LCO 3.1.7. They ensure the proper condition is entered when
expected and the sodium pentaborate solution temperature,
concentration, and volume limits are not exceeded without
appropriate actions being taken. Those limits, the conditions under
which the Standby Liquid Control system is required to be operable,
and the operation of the system remain unchanged and will continue
to be as described, assumed, and analyzed in the Unit 1 and Unit 2
Final Safety Analysis Reports, sections 3.8 and 4.2.3.4,
respectively. The only result of the proposed changes is to reduce
the time limit for continued unit operation with sodium pentaborate
solution temperature, concentration, or volume outside Region A and
Region B from 72 hours to eight hours. Consequently, the possibility
of a new or different type of accident can not be created by these
changes.
3. Do the proposed changes involve a significant reduction in
the margin of safety?
The proposed changes to the Unit 1 and Unit 2 Technical
Specifications do not involve a reduction in the margin of safety.
The changes are administrative in nature only and are intended to
clarify Condition A of LCO 3.1.7. They ensure the proper condition
is entered when expected and the sodium pentaborate solution
temperature, concentration, and volume limits are not exceeded
without appropriate actions being taken. Those limits, the
conditions under which the Standby Liquid Control system is required
to be operable, and the operation of the system remain unchanged by
the proposed changes and will continue to be as described, assumed,
and analyzed in the Unit 1 and Unit 2 Final Safety Analysis Reports.
Therefore, the margin of safety, that is, the ability to bring the
reactor to a subcritical condition under its most reactive
conditions with the Standby Liquid Control system, as embodied by
the sodium pentaborate solution temperature, concentration, and
volume limits and the system operability requirements will not be
reduced.
In conclusion, this proposed license amendment involves no
significant hazards consideration as determined by the standards set
forth by the NRC in 10 CFR 50.92(c). Specifically, it has been shown
in the preceding paragraphs that the proposed changes:
1. Do not involve a significant increase in the probability or
consequences of an accident previously evaluated,
2. Do not create the possibility of a new or different type of
accident from any previously evaluated, and
[[Page 46450]]
3. 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
amendment request involves no significant hazards consideration.
Local Public Document Room location: Appling County Public Library,
301 City Hall Drive, Baxley, Georgia
Attorney for licensee: Ernest L. Blake, Jr., Esquire, Shaw,
Pittman, Potts and Trowbridge, 2300 N Street, NW., Washington, DC.
NRC Section Chief: Richard L. Emch, Jr.
Tennessee Valley Authority, Docket No. 50-296, Browns Ferry Nuclear
Plant, Unit 3, Limestone County, Alabama
Date of amendment request: July 28, 1999.
Description of amendment request: The proposed amendment would add
to the Technical Specifications (TS), new limiting conditions for
operation and surveillance requirements for the Oscillation Power Range
Monitor (OPRM) instrumentation installed in response to Generic Letter
94-02.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
TVA has concluded that operation of BFN [Brown Ferry Nuclear
Plant] Unit 3 in accordance with the proposed change to the TS does
not involve a significant hazards consideration. TVA's conclusion is
based on its evaluation, in accordance with 10 CFR 50.91(a)(1), of
the three standards set forth in 10 CFR 50.92(c).
A. The proposed amendment does not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
The proposed amendment is to enable the OPRM Upscale trip
function which is contained in the previously installed PRNM [Power
Range Neutron Monitoring] equipment. Enabling the OPRM hardware
provides the long term stability solution required by Generic Letter
94-02. This hardware incorporates the Option III detect and suppress
solution reviewed and approved by the NRC in NEDO-31960, ``BWROG
Long Term Stability Solutions Licensing Methodology.'' The OPRM is
designed to meet all requirements of GDC 10 and 12 by automatically
detecting and suppressing design basis thermal-hydraulic power
oscillations prior to violating the fuel MCPR [minimum critical
power ratio] Safety Limit. The OPRM system provides this protection
in the region of the power-to-flow map where instabilities can
occur, including the region where ICAs [Interim Corrective Actions]
previously restricted operation because of stability concerns. Thus,
the ICA restrictions on plant operations are deleted from the TS,
including region avoidance and the requirement for the operator to
manually scram the reactor with no recirculation loops operating.
Operation at high core powers with low core flows may cause a
slight, but not significant, increase in the probability that an
instability can occur. This slight increase is acceptable because
subsequent to the automatic detection of a design basis instability,
the OPRM Upscale trip provides an automatic scram signal to the RPS
which is faster protection than the operator initiated manual scram
required by the current ICAs. Because of this rapid automatic
action, the consequences of an instability event are not increased
as a result of the installation of the OPRM system because it
eliminates operator actions.
Based on the above discussion, the proposed amendment does not
involve a significant increase in the probability or consequences of
an accident previously evaluated.
B. The proposed amendment does not create the possibility of a
new or different kind of accident from any accident previously
evaluated.
The proposed amendment permits BFN to enable the OPRM power
oscillation detect and suppress function provided in previously
installed PRNM hardware, and it simultaneously deletes certain
restrictions which preclude operation in regions of the power-to-
flow map where oscillations potentially may occur. Enabling the OPRM
Upscale trip function does not create any new system hardware
interfaces nor create any new system interactions. Potential
failures of the OPRM Upscale trip result either in failure to
perform a mitigation action or in spurious initiation of a reactor
scram. These failures would not create the possibility of a new or
different kind of accident. Based on the above discussion, the
proposed amendment does not create the possibility of a new or
different kind of accident from any accident previously evaluated.
C. The proposed amendment does not involve a significant
reduction in a margin of safety.
The OPRM Upscale trip function implements BWROG Stability Option
III, which was developed to meet the requirements of GDC 10 and GDC
12 by providing a hardware system that detects the presence of
thermal-hydraulic instabilities and automatically initiates the
necessary actions to suppress the oscillations prior to violating
the MCPR Safety Limit. The NRC has reviewed and accepted the Option
III methodology described in Licensing Topical Report NEDO-31960 and
concluded this solution will provide the intended protection.
Therefore, it is concluded that there will be no reduction in the
margin of safety as defined in TS as a result of enabling the OPRM
Upscale trip function and simultaneously removing the operating
restrictions previously imposed by the ICAs.
Based on the above discussion, the proposed amendment 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: Athens Public Library, 405 E.
South Street, Athens, Alabama 35611.
Attorney for licensee: General Counsel, Tennessee Valley Authority,
400 West Summit Hill Drive, ET 10H, Knoxville, Tennessee 37902.
NRC Section Chief: Sheri R. Peterson.
Vermont Yankee Nuclear Power Corporation, Docket No. 50-271, Vermont
Yankee Nuclear Power Station, Vernon, Vermont
Date of amendment request: July 20, 1999.
Description of amendment request: The licensee proposed the
following five changes: (1) Figure 2.1-1, average power range monitor
(APRM) Flow Reference Scram and APRM Rod Block Settings, the clarifying
statement ``Setpoints shall be [less than or equal to] values shown on
the graph'' is proposed to be added; (2) Bases Section 2.1.B, page 16,
and Bases Section 3.2 APRM rod block trip discussion, page 77, the
current Bases is proposed to be replaced with a more accurate
discussion of the function, as identified in the Vermont Yankee Nuclear
Power Station (VY) Final Safety Analysis Report (FSAR); (3) Table
3.1.1, Reactor Protection System (Scram) Instrument Requirements, APRM
Upscale (Flow Bias) function, it is proposed to add ``with a maximum of
120%'' to the APRM High Flux (Flow Bias) Trip Function equation; (4)
For Table 3.2.5, Control Rod-Block Instrumentation, Rod-Block Monitor
(RBM) Upscale (Flow Bias) function, the caveat ``with a maximum as
defined in the COLR'' [Core Operating Limits Report] is added to the
Trip Setting equation; (5) For Bases page 77, it is proposed to delete
the current paragraph describing the control rod-block systems and
replace it with the following: ``The trip logic for the nuclear
instrumentation control rod block logic is 1 out of n; i.e., any trip
on one of the six APRMs, six IRMs [intermediate range monitors] or four
SRMs [source range monitors] will result in a rod block. The minimum
instrument channel requirements for the IRM may be reduced by one for a
short period of time to allow for maintenance, testing, or calibration.
The RBM is
[[Page 46451]]
credited in the Continuous Rod Withdrawal During Power Range Operation
transient for preventing excessive control rod withdrawal before the
fuel cladding integrity safety limit [minimum critical power ratio]
(MCPR) or the fuel rod mechanical overpower limits are exceeded. The
RBM upper limit is clamped to provide protection at greater than 100%
rated core flow. The clamped value is cycle specific; therefore, it is
located in the Core Operating Limits Report.''
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration which is presented below:
1. The operation of Vermont Yankee Nuclear Power Station in
accordance with the proposed amendment will not involve a
significant increase in the probability or consequences of an
accident previously evaluated.
Changes 1 and 3 are administrative and have no impact on
technical content; therefore, they do not increase the probability
or consequences of an accident previously evaluated.
Changes 2 and 5 clarify ambiguities in the Bases. The wording is
descriptive only and does not change the meaning or intent of the
specification. Therefore, these changes do not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
Change 4 adds the Rod Block Monitor Upscale (Flow Bias) maximum
value limitation to the Technical Specifications. Limiting the
upscale trip setting at flows in excess of 100% of rated core flow
ensures the assumptions of the Continuous Rod Withdrawal During
Power Range Operation Transient are met. No other accident or
transient analyses are affected. Therefore, this change does not
involve a significant increase in the probability or consequences of
an accident previously evaluated.
2. The operation of Vermont Yankee Nuclear Power Station in
accordance with the proposed amendment will not create the
possibility of a new or different kind of accident from any accident
previously evaluated.
Change 4, limiting the maximum value for the Rod Block Monitor
Upscale (Flow Bias) function, is a change to plant design, in that
it clamps the upscale trip setting at flows in excess of 100% of
rated core flow at the 100% core flow value. This change ensures the
assumptions of the Continuous Rod Withdrawal During Power Range
Operation Transient are met and has no effect on any other accident
or transient analyses. Changes 1, 2, 3, and 5 do not involve a
change to the plant design.
None of the proposed changes affects any parameters or
conditions that could contribute to the initiation of any accident.
No new accident modes are created. No safety-related equipment or
safety functions, other than the Rod Block Monitor as discussed
above, are altered as a result of these changes.
Based on the above VY has concluded that the proposed change
will not create the possibility of a new or different kind of
accident from those previously evaluated.
3. The operation of Vermont Yankee Nuclear Power Station in
accordance with the proposed amendment will not involve a
significant reduction in a margin of safety.
Changes 1 and 3 are administrative and have no impact on
technical content. Therefore, they have no effect on margin of
safety.
Changes 2 and 5 clarify ambiguities in the Bases, using wording
taken directly from the FSAR. The wording is descriptive only and
does not change the meaning or intent of the specification.
Therefore, these changes do not involve a significant reduction in a
margin of safety.
Change 4 adds the Rod Block Monitor Upscale (Flow Bias) maximum
value limitation to the Technical Specifications. Limiting the
upscale trip setting at flows in excess of 100% of rated core flow
ensures the assumptions and, therefore the margin of safety, of the
Continuous Rod Withdrawal During Power Range Operation transient are
met. No other accident or transient analyses are affected.
Therefore, this 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: Brooks Memorial Library, 224
Main Street, Brattleboro, VT 05301.
Attorney for licensee: Mr. David R. Lewis, Shaw, Pittman, Potts and
Trowbridge, 2300 N Street, NW., Washington, DC 20037-1128.
NRC Section Chief: James W. Clifford.
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 notice was 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.
Carolina Power & Light Company, et al., Docket No. 50-261, H. B.
Robinson Steam Electric Plant, Unit 2, Darlington County, South
Carolina
Date of amendment request: July 30, 1999.
Brief Description of amendment: The proposed amendment would revise
Required Action A.1 of Technical Specification Limiting Condition for
Operation 3.7.8, ``Ultimate Heat Sink (UHS),'' to allow a Completion
Time of 72 hours to restore service water temperature to less than or
equal to 95 deg.F prior to entering the required actions for plant
shutdown. The amendment request was proposed as a temporary change to
be in effect until September 30, 1999.
Date of publication of individual notice in the Federal Register:
August 10, 1999 (64 FR 43406).
Expiration date of individual notice: August 24, 1999, for
comments; September 8, 1999, for hearings.
Local Public Document Room location: Hartsville Memorial Library,
147 West College Avenue, Hartsville, South Carolina 29550.
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 46452]]
provision in 10 CFR 51.12(b) and has made a determination based on that
assessment, it is so indicated.
For further details with respect to the action see (1) the
applications for amendment, (2) the amendment, and (3) the Commission's
related letter, Safety Evaluation and/or Environmental Assessment as
indicated. All of these items are available for public inspection at
the Commission's Public Document Room, the Gelman Building, 2120 L
Street, NW., Washington, DC, and at the local public document rooms for
the particular facilities involved.
Arizona Public Service Company, et al., Docket Nos. STN 50-528, STN 50-
529, and STN 50-530, Palo Verde Nuclear Generating Station, Units 1, 2,
and 3, Maricopa County, Arizona
Date of application for amendments: May 23, 1997, as supplemented
September 27, 1998, and May 26, 1999.
Brief description of amendments: The amendments revise the
Technical Specifications to allow the installation of ABB Combustion
Engineering leak tight sleeves in defective steam generator tubes as a
tube repair method.
Date of issuance: August 5, 1999.
Effective date: August 5, 1999, to be implemented within 45 days.
Amendment Nos.: Unit 1--120, Unit 2--120, Unit 3--120.
Facility Operating License Nos. NPF-41, NPF-51, and NPF-74: The
amendments revised the Technical Specifications.
Date of initial notice in Federal Register: June 16, 1999 (64 FR
32285).
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated August 5, 1999.
No significant hazards consideration comments received: No.
Local Public Document Room location: Phoenix Public Library, 1221
N. Central Avenue, Phoenix, Arizona 85004.
Commonwealth Edison Company, Docket Nos. STN 50-454 and STN 50-455,
Byron Station, Unit Nos. 1 and 2, Ogle County, Illinois
Date of application for amendments: March 14, 1997.
Brief description of amendments: The amendments deleted license
conditions which have been satisfied, revise others to delete parts
which are no longer applicable or to revise references, and make
editorial changes.
Date of issuance: August 10, 1999.
Effective date: Immediately, to be implemented within 30 days.
Amendment No.: 110.
Facility Operating License Nos. NPF-37 and NPF-66: The amendments
revised the Licenses.
Date of initial notice in Federal Register: April 22, 1998 (63 FR
19966).
The Commission's related evaluation of the amendments is contained
in an Environmental Assessment dated July 7, 1999 (64FR36722), and a
Safety Evaluation dated August 10, 1999.
No significant hazards consideration comments received: No.
Local Public Document Room location: Byron Public Library District,
109 N. Franklin, P.O. Box 434, Byron, Illinois 61010.
Commonwealth Edison Company, Docket Nos. 50-254 and 50-265, Quad Cities
Nuclear Power Station, Units 1 and 2, Rock Island County, Illinois
Date of application for amendments: March 30, 1999, as supplemented
June 30, 1999.
Brief description of amendments: The amendments revise the
Technical Specifications, Section 3/4.6.G, ``Leakage Detection
Systems,'' to allow an alternate methodology for quantifying Reactor
Coolant System (RCS) leakage when the normal RCS leakage detection
system is inoperable.
Date of issuance: August 4, 1999.
Effective date: Immediately, to be implemented within 60 days.
Amendment Nos.: 189 & 186.
Facility Operating License Nos. DPR-29 and DPR-30: The amendments
revised the Technical Specifications.
Date of initial notice in Federal Register: May 5, 1999 (64 FR
24194).
The June 30, 1999, submittal provided additional clarifying
information that did not change the original proposed no significant
hazards consideration determination.
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated August 4, 1999.
No significant hazards consideration comments received: No.
Local Public Document Room location: Dixon Public Library, 221
Hennepin Avenue, Dixon, Illinois 61021.
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: March 3, 1999, as supplemented
May 27, and June 22, 1999.
Brief description of amendments: The amendments change the required
qualifications for operations management specified in the technical
specifications (TSs) for the Beaver Valley Power Station, Units 1 and 2
(BVPS-1 and BVPS-2). The requirement that the operations manager hold a
Senior Reactor Operator (SRO) license at the time of appointment is
changed in the TSs to require that the assistant operations managers,
one for each unit, hold an SRO license on their assigned unit. The
revised TSs require the operations manager to hold, or have held, an
SRO license on a pressurized water reactor. Additionally, the Updated
Final Safety Analysis Report (UFSAR) for each unit is changed to
require the operations manager to ``hold, or have held,'' an SRO
license rather than ``hold'' a license. The revised UFSARs require the
same as the TSs; that the assistant operations managers hold an SRO
license on the unit to which they are assigned. Finally, the amendments
substitute generic personnel titles for plant-specific personnel titles
in the BVPS-1 and BVPS-2 TSs. The correlation between generic titles
and plant-specific titles is provided in the revised BVPS-2 UFSAR.
Date of issuance: August 10, 1999.
Effective date: Both units, as of date of issuance and shall be
implemented within 60 days.
Amendment Nos.: 224 and 100.
Facility Operating License Nos. DPR-66 and NPF-73: Amendments
revised the Technical Specifications.
Date of initial notice in Federal Register: April 21, 1999 (64 FR
19556).
The May 27, and June 22, 1999, letters provided additional
information but did not change the initial proposed no significant
hazards consideration determination or expand the amendment beyond the
scope of the initial notice.
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated August 10, 1999.
No significant hazards consideration comments received: No.
Local Public Document Room location: B. F. Jones Memorial Library,
663 Franklin Avenue, Aliquippa, PA 15001.
Florida Power Corporation, et al., Docket No. 50-302, Crystal River
Nuclear Generating Plant, Unit 3, Citrus County, Florida
Date of application for amendment: November 24, 1998, as
supplemented June 23, 1999.
Brief description of amendment: The amendment approves the addition
of a safety-related diesel-driven emergency feedwater pump (EFP-3) as a
functional replacement for the existing motor-driven pump, addition of
technical specifications and surveillances for this new pump, and
deletion of cycle specific interim technical specifications which would
not be required after the addition of the new pump.
[[Page 46453]]
Date of issuance: August 11, 1999.
Effective date: As of the date of issuance and shall be implemented
prior to commencing cycle 12 operation.
Amendment No.: 182.
Facility Operating License No. DPR-72: Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: January 13, 1999 (64 FR
2247).
The supplemental letter dated June 23, 1999, did not change the
original proposed no significant hazards consideration determination,
or expand the scope of the amendment request as originally noticed. The
Commission's related evaluation of the amendment is contained in a
Safety Evaluation dated August 11, 1999.
No significant hazards consideration comments received: No
Local Public Document Room location: Coastal Region Library, 8619
W. Crystal Street, Crystal River, Florida 34428.
No significant hazards consideration comments received: No
Local Public Document Room location: Coastal Region Library, 8619
W. Crystal Street, Crystal River, Florida 34428.
North Atlantic Energy Service Corporation, et al., Docket No. 50-443,
Seabrook Station, Unit No. 1, Rockingham County, New Hampshire
Date of amendment request: November 4, 1998.
Description of amendment request: To revise Technical
Specifications Surveillance Requirement 4.5.2b.1 to delete the
prescribed method of venting the Emergency Core Cooling System (ECCS)
which would allow an alternate method to verify that the ECCS piping is
full of water. In addition, the associated Bases are being revised to
reflect the intent of the surveillance requirement.
Date of issuance: August 12, 1999.
Effective date: As of its date of issuance, and shall be
implemented within 60 days.
Amendment No.: 61.
Facility Operating License No. NPF-86: Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: January 27, 1999 (64 FR
4157)
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated August 12, 1999.
No significant hazards consideration comments received: No
Local Public Document Room location: Exeter Public Library,
Founders Park, Exeter, NH 03833.
Power Authority of The State of New York, Docket No. 50-286, Indian
Point Nuclear Generating Unit No. 3, Westchester County, New York
Date of application for amendment: June 4, 1999.
Brief description of amendment: The amendment changes the Technical
Specifications by extending the allowed outage time for the 32
emergency diesel generator and its fuel oil storage tank.
Date of issuance: August 9, 1999.
Effective date: As of the date of issuance to be implemented within
30 days.
Amendment No.: 190.
Facility Operating License No. DPR-64: Amendment revises the
Technical Specifications.
Date of initial notice in Federal Register: July 6, 1999 (64 FR
36408).
No significant hazards consideration comments received: No.
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated August 9, 1999.
No significant hazards consideration comments received: No.
Local Public Document Room location: White Plains Public Library,
100 Martine Avenue, White Plains, New York 10610.
Power Authority of the State of New York, Docket No. 50-333, James A.
FitzPatrick Nuclear Power Plant, Oswego County, New York
Date of application for amendment: January 25, 1996, as
supplemented April 26, 1996, September 12, 1996, March 17, 1997,
September 9, 1997, December 30, 1998, and May 19, 1999.
Brief description of amendment: The amendment extends the allowed
outage time for an emergency diesel generator (EDG) system from 7 to 14
days, revises requirements for EDG testing at power, and revises
electrical power requirements for cold shutdown and refueling modes.
Date of issuance: July 30, 1999.
Effective date: As of the date of issuance to be implemented within
60 days.
Amendment No.: 253.
Facility Operating License No. DPR-59: Amendment revised the
Technical Specifications.
Date of initial notices in Federal Register: March 27, 1996 (61 FR
13532) and June 30, 1999 (64 FR 35208).
The licensee provided additional information on April 26, 1996,
September 12, 1996, March 17, 1997, September 9, 1997, and December 30,
1998, that provided clarifying information within the scope of the
initial Federal Register notice and did not change the staff's original
proposed no significant hazards consideration determination. The
changes proposed on May 19, 1999, were reflected in the staff's revised
proposed finding of no significant hazards consideration, and encompass
the additional information provided by the licensee.
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated July 30, 1999.
No significant hazards consideration comments received: No.
Local Public Document Room location: Reference and Documents
Department, Penfield Library, State University of New York, Oswego, New
York 13126.
PP&L, Inc., Docket Nos. 50-387 and 50-388, Susquehanna Steam Electric
Station, Units 1 and 2, Luzerne County, Pennsylvania
Date of application for amendments: June 19, 1998, (Unit 1) and
August 5, 1998, (Unit 2) as supplemented by letter dated November 23,
1998.
Brief description of amendments: The amendments to the Unit 1 and
Unit 2 Technical Specifications (TSs) involve the addition of a new
section entitled ``Oscillation Power Range Monitoring (OPRM)
Instrumentation'' and revisions to Section 3.4.1 ``Recirculation Loops
Operating'' to remove the specifications related to thermal power
stability which are no longer required after the installation of OPRM
instrumentation.
Date of issuance: July 30, 1999.
Effective date: Effective as of its date of issuance and is to be
implemented within 90 days following startup from the Unit 2 ninth
Refueling Inspection Outage, currently scheduled for April 16, 1999.
Amendment Nos.: 184 and 188.
Facility Operating License Nos. NPF-14 and NPF-22. The amendments
revised the Technical Specifications.
Date of initial notice in Federal Register: August 12, 1998 (63 FR
43210) and August 26, 1998 (63 FR 45528).
The November 23, 1998, 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 initial notice.
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated July 30, 1999.
No significant hazards consideration comments received: No.
Local Public Document Room location: Osterhout Free Library,
Reference Department, 71 South Franklin Street, Wilkes-Barre, PA 18701.
[[Page 46454]]
Sacramento Municipal Utility District, Docket No. 50-312, Rancho Seco
Nuclear Generating Station, Sacramento County, California
Date of application for amendments: April 23, 1999.
Brief description of amendments: The amendment changes Permanently
Defueled Technical Specification
D3/4.1, ``Spent Fuel Pool Level,'' to replace a specific reference to
spent fuel pool (SFP) level alarm switches with a generic reference to
SFP level instrumentation.
Date of issuance: August 13, 1999.
Effective date: August 13, 1999, to be implemented within 30 days.
Amendment No.: 126.
Facility Operating License Nos. NPF-37 and NPF-66: The amendment
revised the Technical Specifications.
Date of initial notice in Federal Register: June 30, 1999 (64 FR
35210).
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated August 13, 1999.
No significant hazards consideration comments received: No.
Local Public Document Room location: Central Library, Government
Documents, 828 I Street, Sacramento, California 95814.
STP Nuclear Operating Company, Docket Nos. 50-498 and 50-499, South
Texas Project, Units 1 and 2, Matagorda County, Texas
Date of amendment request: March 2, 1999, as supplemented by letter
dated July 13, 1999.
Brief description of amendments: The amendments allow the use of a
``check valve with flow through the valve secured'' as an additional
means to isolate an affected containment penetration (i.e., a
penetration with an inoperable penetration barrier) in Technical
Specification 3.6.3, Action b.
Date of issuance: August 3, 1999.
Effective date: As of the date of issuance to be implemented within
30 days from the date of issuance.
Amendment Nos.: Unit 1-113; Unit 2-101.
Facility Operating License Nos. NPF-76 and NPF-80: The amendments
revised the Technical Specifications.
Date of initial notice in Federal Register: April 7, 1999 (64 FR
17030).
The July 13, 1999, supplement provided additional clarifying
information within the scope of the original notice and did not change
the staff's initial proposed no significant hazards consideration
determination.
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated August 3, 1999.
No significant hazards consideration comments received: No.
Local Public Document Room location: Wharton County Junior College,
J. M. Hodges Learning Center, 911 Boling Highway, Wharton, Texas 77488.
Tennessee Valley Authority, Docket Nos. 50-260 and 50-296, Browns Ferry
Nuclear Plant, Units 2 and 3, Limestone County, Alabama
Date of application for amendments: March 12, 1997, as supplemented
by letters dated March 30, 1999, April 23, 1999, and June 18, 1999.
Brief description of amendments: The amendments revise the
Technical Specifications (TS) to extend, from 7 days to 14 days, the
Allowable Outage Time applicable to an inoperable emergency diesel
generator.
Date of issuance: August 2, 1999.
Effective date: August 2, 1999.
Amendment Nos.: 259 and 218.
Facility Operating License Nos. DPR-52 and DPR-68: Amendments
revised the TS.
Date of initial notice in Federal Register: June 30, 1999 (64 FR
35211)
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated August 2, 1999.
No significant hazards consideration comments received: None.
Local Public Document Room location: Athens Public Library, 405 E.
South Street, Athens, Alabama 35611.
TU Electric Company, Docket Nos. 50-445 and 50-446, Comanche Peak Steam
Electric Station, Unit Nos. 1 and 2, Somervell County, Texas
Date of amendment request: May 4, 1999, as supplemented by letter
dated June 4, 1999.
Brief description of amendments: The amendments correct a number of
editorial errors in the Technical Specifications that occurred with the
issuance of Amendment No. 64 to Facility Operating License Nos. NPF-87
and NPF-89, regarding the improved Technical Specifications conversion.
In addition, Surveillance Requirement (SR) 3.8.4.7 is revised to allow
the substitution of a modified performance discharge test, for a
service test, for the 125 VDC batteries and SRs 3.8.1.7, 3.8.1.12,
3.8.1.15, and 3.8.1.20 are revised to separate the voltage and
frequency acceptance criteria for the diesel generator start
surveillances into two sets of criteria; those criteria required to be
met within 10 seconds, and those criteria required to be met following
achievement of steady state conditions.
Date of issuance: August 3, 1999.
Effective date: As of the date of issuance and shall be implemented
within 30 days from the date of issuance.
Amendment Nos.: Unit 1-Amendment No. 66; Unit 2-Amendment No. 66.
Facility Operating License Nos. NPF-87 and NPF-89: The amendments
revised the Technical Specifications.
Date of initial notice in Federal Register: June 2, 1999 (64 FR
29715); and June 30, 1999 (64 FR 35212).
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated August 3, 1999.
No significant hazards consideration comments received: No.
Local Public Document Room location: University of Texas at
Arlington Library, Government Publications/Maps, 702 College, P.O. Box
19497, Arlington, Texas 76019.
Vermont Yankee Nuclear Power Corporation, Docket No. 50-271, Vermont
Yankee Nuclear Power Station, Vernon, Vermont
Date of application for amendment: February 1, 1999, as
supplemented on April 19 and April 23, 1999.
Brief description of amendment: The amendment totally replaces the
current Technical Specifications Section 6.0, ``Administrative
Controls.'' Administrative changes to certain other sections of the
Technical Specifications were made to conform to the changes resulting
from the re-write of Section 6.0.
The changes represent a comprehensive upgrade of Section 6.0 of the
Vermont Yankee Technical Specifications, incorporating improvements in
content and format based on industry standards. In accordance with
industry practice, some Technical Specifications requirements are being
relocated to the recently implemented Vermont Yankee Technical
Requirements Manual, Offsite Dose Calculation Manual, or Vermont Yankee
Operational Quality Assurance Manual and are being eliminated from the
Technical Specification.
Date of Issuance: July 19, 1999.
Effective date: As of the date of issuance, to be implemented
within 60 days.
Amendment No.: 171.
Facility Operating License No. DPR-28: Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: May 19, 1999 (64 FR
27326).
The Commission's related evaluation of this amendment is contained
in a Safety Evaluation dated July 19, 1999.
No significant hazards consideration comments received: No.
[[Page 46455]]
Local Public Document Room location: Brooks Memorial Library, 224
Main Street, Brattleboro, VT 05301.
Washington Public Power Supply System, Docket No. 50-397, Nuclear
Project No. 2, Benton County, Washington
Date of application for amendment: June 3, 1999, as supplemented by
letter dated July 22, 1999.
Brief description of amendment: The amendment updates the operating
license to reflect the name change of the licensee from ``Washington
Public Power Supply System'' to ``Energy Northwest'' and the name
change of the facility from ``WPPSS Nuclear Project No. 2'' to ``WNP-
2.''
Date of issuance: August 2, 1999.
Effective date: August 2, 1999.
Amendment No.: 157.
Facility Operating License No. NPF-21: The amendment revised the
operating license.
Date of initial notice in Federal Register: June 30, 1999. (64 FR
35214).
The July 22, 1999, supplemental letter provided additional
clarifying information, did not significantly expand the scope of the
application as originally noticed and did not change the staff's
original proposed no significant hazards consideration determination.
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated August 2, 1999.
No significant hazards consideration comments received: No.
Local Public Document Room location: Richland Public Library, 955
Northgate Street, Richland, Washington 99352.
Wisconsin Electric Power Company, Docket Nos. 50-266 and 50-301, Point
Beach Nuclear Plant, Units 1 and 2, Town of Two Creeks, Manitowoc
County, Wisconsin
Date of application for amendments: January 29, 1999. (TSCR 211),
as supplemented June 9 and July 15, 1999.
Brief description of amendments: These amendments reflect changes
to Sections 15.6 and 15.7 of the Point Beach Nuclear Plant, Units 1 and
2, Technical Specifications (TSs). The changes are considered
administrative in nature and reflect personnel title changes, an
increase in minimum operating crew shift staffing, relocation of the
Manager's Supervisory Staff composition and functional requirements to
owner-controlled documents, and revisions to the procedure review and
approval process.
Date of issuance: August 11, 1999.
Effective date: August 11, 1999. The TSs shall be implemented
within 90 days. Implementation also includes removal of selected
requirements from TS Section 15.6, Administrative Controls, and the
relocation of other requirements to licensee-controlled documents as
described in the licensee's application dated January 29, 1999, as
supplemented June 9 and July 15, 1999, and evaluated in the staff's
safety evaluation attached to the amendments. With respect to changes
to the final safety analysis report (FSAR), Wisconsin Electric Power
Company shall incorporate the revisions into the next FSAR update in
accordance with the schedule in 10 CFR 50.71(e).
Amendment Nos.: Unit 1-190; Unit 2-195.
Facility Operating License Nos. DPR-24 and DPR-27: Amendments
revised the Technical Specifications.
Date of initial notice in Federal Register: February 24, 1999 (64
FR 9202).
The June 9 and July 15, 1999, letters provided additional
clarifying information within the scope of the original Federal
Register notice and did not affect the staff's initial proposed no
significant hazards consideration determination.
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated August 11, 1999.
No significant hazards consideration comments received: No.
Local Public Document Room location: The Lester Public Library,
1001 Adams Street, Two Rivers, Wisconsin 54241.
Wolf Creek Nuclear Operating Corporation, Docket No. 50-482, Wolf Creek
Generating Station, Coffey County, Kansas
Date of amendment request: June 10, 1999.
Brief description of amendment: The amendment revised Technical
Specification Table 3.3-4, Functional Unit 7.b., Automatic Switchover
to Containment Sump (Refueling Water Storage Tank Level--Low-Low) to
reflect the results of calculations that were performed for the
associated instrumentation setpoints to consider the density variations
due to temperature and boric acid concentrations.
Date of issuance: August 9, 1999.
Effective date: August 9, 1999, and shall be implemented within 60
days from issuance of the amendment.
Amendment No.: 126.
Facility Operating License No. NPF-42. The amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: June 30, 1999 (64 FR
35215).
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated August 9, 1999.
No significant hazards consideration comments received: No.
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.
Wolf Creek Nuclear Operating Corporation, Docket No. 50-482, Wolf Creek
Generating Station, Coffey County, Kansas
Date of amendment request: June 11, 1999.
Brief description of amendment: The amendment revises TS 3.7.1.6,
Steam Generator Atmospheric Relief Valves, and associated Bases to (1)
require four atmospheric relief valves (ARVs) to be operable, (2)
eliminate the use of ``required'' in the action statements, (3) provide
action statements to address inoperability of two ARVs and three or
more ARVs due to causes other than excessive leakage, and (4) limit the
Limiting Condition for Operation 3.0.4 exception to when one ARV is
inoperable due to causes other than excessive seat leakage.
Date of issuance: August 12, 1999.
Effective date: August 12, 1999, and shall be implemented within 60
days from the date of issuance.
Amendment No.: 127.
Facility Operating License No. NPF-42. The amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: June 30, 1999 (64 FR
35215).
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated August 12, 1999.
No significant hazards consideration comments received: No.
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.
Dated at Rockville, Maryland, this 18th day of August 1999.
For the Nuclear Regulatory Commission.
John A. Zwolinski,
Director, Division of Licensing Project Management, Office of Nuclear
Reactor Regulation.
[FR Doc. 99-21914 Filed 8-24-99; 8:45 am]
BILLING CODE 7590-01-P