![[logo] US EPA](http://www.epa.gov/epafiles/images/logo_epaseal.gif){kind=logo}
Biweekly Notice; Applications and Amendments to Facility Operating Licenses Involving No Significant Hazards Considerations
Note: EPA no longer updates this information, but it may be useful as a reference or resource.
[Federal Register: January 26, 2000 (Volume 65, Number 17)]
[Notices]
[Page 4268-4295]
From the Federal Register Online via GPO Access [wais.access.gpo.gov]
[DOCID:fr26ja00-96]
[[Page 4268]]
-----------------------------------------------------------------------
NUCLEAR REGULATORY COMMISSION
Biweekly Notice; Applications and Amendments to Facility
Operating Licenses Involving No Significant Hazards Considerations
I. Background
Pursuant to Public Law 97-415, the U.S. Nuclear Regulatory
Commission (the Commission or NRC staff) is publishing this regular
biweekly notice. Public Law 97-415 revised section 189 of the Atomic
Energy Act of 1954, as amended (the Act), to require the Commission to
publish notice of any amendments issued, or proposed to be issued,
under a new provision of section 189 of the Act. This provision grants
the Commission the authority to issue and make immediately effective
any amendment to an operating license upon a determination by the
Commission that such amendment involves no significant hazards
consideration, notwithstanding the pendency before the Commission of a
request for a hearing from any person.
This biweekly notice includes all notices of amendments issued, or
proposed to be issued from December 31, 1999, through January 14, 2000.
The last biweekly notice was published on January 12, 2000.
Notice of Consideration of Issuance of Amendments to Facility
Operating Licenses, Proposed No Significant Hazards Consideration
of Determination, and Opportunity for a Hearing
The Commission has made a proposed determination that the following
amendment requests involve no significant hazards consideration. Under
the Commission's regulations in 10 CFR 50.92, this means that operation
of the facility in accordance with the proposed amendment would not (1)
involve a significant increase in the probability or consequences of an
accident previously evaluated; or (2) create the possibility of a new
or different kind of accident from any accident previously evaluated;
or (3) involve a significant reduction in a margin of safety. The basis
for this proposed determination for each amendment request is shown
below.
The Commission is seeking public comments on this proposed
determination. Any comments received within 30 days after the date of
publication of this notice will be considered in making any final
determination.
Normally, the Commission will not issue the amendment until the
expiration of the 30-day notice period. However, should circumstances
change during the notice period such that failure to act in a timely
way would result, for example, in derating or shutdown of the facility,
the Commission may issue the license amendment before the expiration of
the 30-day notice period, provided that its final determination is that
the amendment involves no significant hazards consideration. The final
determination will consider all public and State comments received
before action is taken. Should the Commission take this action, it will
publish in the Federal Register a notice of issuance and provide for
opportunity for a hearing after issuance. The Commission expects that
the need to take this action will occur very infrequently.
Written comments may be submitted by mail to the Chief, Rules
Review and Directives Branch, Division of Freedom of Information and
Publications Services, Office of Administration, U.S. Nuclear
Regulatory Commission, Washington, DC 20555-0001, and should cite the
publication date and page number of this Federal Register notice.
Written comments may also be delivered to Room 6D22, Two White Flint
North, 11545 Rockville Pike, Rockville, Maryland from 7:30 a.m. to 4:15
p.m. Federal workdays. Copies of written comments received may be
examined at the NRC Public Document Room, the Gelman Building, 2120 L
Street, NW., Washington, DC. The filing of requests for a hearing and
petitions for leave to intervene is discussed below.
By February 25, 2000, 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 electronically from
the ADAMS Public Library component on the NRC Web site, http:
//www.nrc.gov (the Electronic Reading Room). 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
[[Page 4269]]
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: Docketing and
Services Branch, or may be delivered to the Commission's Public
Document Room, the Gelman Building, 2120 L Street, NW., Washington DC,
by the above date. Where petitions are filed during the last 10 days of
the notice period, it is requested that the petitioner promptly so
inform the Commission by a toll-free telephone call to Western Union at
1-(800) 248-5100 (in Missouri 1-(800) 342-6700). The Western Union
operator should be given Datagram Identification Number N1023 and the
following message addressed to (Project Director): petitioner's name
and telephone number, date petition was mailed, plant name, and
publication date and page number of this Federal Register notice. A
copy of the petition should also be sent to the Office of the General
Counsel, U.S. Nuclear Regulatory Commission, Washington, DC 20555-0001,
and to the attorney for the licensee.
Nontimely filings of petitions for leave to intervene, amended
petitions, supplemental petitions and/or requests for a hearing will
not be entertained absent a determination by the Commission, the
presiding officer or the Atomic Safety and Licensing Board that the
petition and/or request should be granted based upon a balancing of
factors specified in 10 CFR 2.714(a)(1)(i)-(v) and 2.714(d).
For further details with respect to this action, see the
application for amendment which is available for public inspection at
the Commission's Public Document Room, the Gelman Building, 2120 L
Street, NW., Washington, DC, and electronically from the ADAMS Public
Library component on the NRC Web site, http:
//www.nrc.gov (the Electronic Reading Room).
Detroit Edison Company, Docket No. 50-341, Fermi 2, Monroe County,
Michigan
Date of amendment request: December 17, 1999.
Description of amendment request: The proposed amendment would
revise Technical Specification 2.1.1.2 to incorporate cycle-specific
safety limit minimum critical power ratios (SLMCPRs) for the core that
will be loaded during the upcoming refueling outage
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. The proposed change does not involve a significant increase
in the probability or consequences of an accident previously
evaluated.
The proposed license amendment establishes a revised SLMCPR
value of 1.07 for two recirculation loop operation and 1.09 for
single recirculation loop operation. The derivation of the cycle-
specific SLMCPRs was performed using NRC approved methods and
uncertainties described in Amendment Number 25 to NEDE-24011-P-A
(GESTAR II) and Licensing Topical Reports NEDC-32601P-A,
``Methodology and Uncertainties for Safety Limit MCPR Evaluations''
and NEDC-32694P-A, ``Power Distribution Uncertainties for Safety
Limit MCPR Evaluation.''
The probability of an evaluated accident is derived from the
probabilities of the individual precursors to that accident. The
consequences of an evaluated accident are determined by the
operability of plant systems designed to mitigate those
consequences. Limits have been established, consistent with NRC
approved methods, to ensure that fuel performance during normal,
transient, and accident conditions is acceptable.
The probability of an evaluated accident is not increased by
revising the SLMCPR values. The change does not require any physical
plant modifications or physically affect any plant components.
Therefore, no individual precursors of an accident are affected.
The proposed license amendment establishes a revised SLMCPR that
ensures that the fuel is protected during normal operation and
during any plant transients or anticipated operational occurrences.
Specifically, the reload analysis demonstrates that a SLMCPR value
of 1.07 (1.09 for single loop operation) ensures that less than 0.1
percent of the fuel rods will experience boiling transition during
any plant operation if the limit is not violated.
Based on (1) the determination of the new SLMCPR values using
NRC approved methods and uncertainties, and (2) the operability of
plant systems designed to mitigate the consequences of accidents not
having been changed; the consequences of an accident previously
evaluated have not been increased.
Therefore, the proposed Technical Specification change does not
involve an 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 license amendment involves a revision of the SLMCPR
from 1.11 to 1.07 for two recirculation loop operation and from 1.13
to 1.09 for single loop operation based on the results of analysis
of the Cycle 8 core which will once again be fully loaded with GE11
fuel. Creation of the possibility of a new or different kind of
accident would require the creation of one or more new precursors of
that accident. New accident precursors may be created by
modifications of the plant configuration, including changes in the
allowable methods of operating the facility. This proposed license
amendment does not involve any modifications of the plant
configuration or changes in the allowable methods of operation.
Therefore, the proposed Technical Specification change does not
create the possibility of a new or different kind of accident from
any accident previously evaluated.
3. The change does not involve a significant reduction in the
margin of safety.
The proposed license amendment establishes a revised SLMCPR
value of 1.07 for two recirculation loop operation and 1.09 for
single recirculation loop operation. The derivation of these revised
SLMCPRs was performed using NRC approved methods and uncertainties
described in Amendment Number 25 to NEDE-24011-P-A (GESTAR II) and
Licensing Topical Reports NEDC-32601P-A, ``Methodology and
Uncertainties for Safety Limit MCPR Evaluations'' and NEDC-32694P-A,
``Power Distribution Uncertainties for Safety Limit MCPR
Evaluation.'' Use of these methods ensures that the resulting SLMCPR
satisfies the fuel design safety criteria that less than 0.1 percent
of the fuel rods experience boiling transition if the safety limit
is not violated. Based on the assurance that the fuel design safety
criteria will be met, the proposed 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
[[Page 4270]]
proposes to determine that the amendment request involves no
significant hazards consideration.
Attorney for licensee: John Flynn, Esq., Detroit Edison Company,
2000 Second Avenue, Detroit, Michigan 48226.
NRC Section Chief: Claudia M. Craig.
Detroit Edison Company, Docket No. 50-341, Fermi 2, Monroe County,
Michigan
Date of amendment request: December 17, 1999.
Description of amendment request: The proposed amendment would
revise Technical Specification Surveillance Requirement (SR) 3.6.1.3.9
to relax the SR frequency by allowing a representative sample of excess
flow check valves (EFCVs) to be tested every 18 months, such that each
EFCV will be tested at least once every 10 years. Current SR 3.6.1.3.9
requires all EFCVs to be tested every 18 months.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. The change does not involve a significant increase in the
probability or consequences of an accident previously evaluated.
The current SR frequency requires each reactor instrumentation
line EFCV to be tested every 18 months. The EFCVs at Fermi 2 are
designed to close automatically in the event of a line break
downstream of the valve. Indicating lights on a control room panel
monitor EFCV positions. These valves may be reopened by actuation of
a solenoid valve, which is operated from a local control panel.
EFCVs at Fermi 2 are designed and installed following the guidance
of Regulatory Guide 1.11. This proposed change allows a reduced
number of EFCVs to be tested every 18 months. Industry operating
experience, documented in BWROG [Boiling Water Reactor Owners Group]
Report B21-00658-01, concludes that a change in surveillance test
frequency has a minimal impact on the reliability for these valves.
A failure of an EFCV to isolate cannot initiate previously evaluated
accidents; therefore, there can be no increase in the probability of
occurrence of an accident as a result of this proposed change.
Fermi 2 UFSAR [Updated Final Safety Analysis Report], Subsection
15.6.2 evaluates an instrument line pipe break within secondary
containment. The evaluation assumes that a small instrument line
instantaneously and circumferentially breaks at a location where it
may not be possible to isolate it and where immediate detection is
not automatic or apparent. The evaluation concluded that
pressurization of the secondary containment would not result from an
instrument line break and a failure of the associated EFCV to
isolate the ruptured line. The standby gas treatment system is not
impaired by this event, and the calculated offsite exposure is
substantially below the guidelines of 10 CFR 100. Additionally,
coolant lost from such a break is inconsequential when compared to
the makeup capabilities of the feedwater or RCIC [reactor core
isolation cooling] system. The BWROG report concludes that the risk
to the public with the extended testing interval is several orders
of magnitudes below the general public annual exposure limits in 10
CFR 20.105.
Although not expected to occur as a result of this change, the
postulated failure of an EFCV to isolate as a result of reduced
testing is bounded by the analysis in the UFSAR. Therefore, there is
no increase in the previously evaluated consequences of the rupture
of an instrument line and there is no potential increase in the
radiological consequences of an accident previously evaluated as a
result of this change.
2. The change does not create the possibility of a new or
different kind of accident from any accident previously evaluated.
This proposed change allows a reduced number of EFCVs to be
tested each operating cycle. No other changes in requirements are
being proposed. Industry operating experience as documented in the
BWROG report provides supporting evidence that the reduced testing
frequency will not affect the high reliability of these valves. The
potential failure of an EFCV to isolate as a result of the proposed
reduction in test frequency is bounded by the evaluation of an
instrument line pipe break described in Subsection 15.6.2 of the
UFSAR. This change is not a physical alteration of the plant and
will not alter the operation of the structures, systems and
components as described in the UFSAR. Therefore, a new or different
kind of accident will not be created.
3. The change does not involve a significant reduction in the
margin of safety.
The consequences of a postulated instrument line pipe break have
been evaluated in Subsection 15.6.2 of the UFSAR. The evaluation
assumed the line instantaneously and circumferentially breaks at a
location where it may not be possible to isolate it and that the
EFCV fails to isolate the break. Therefore, any potential failure of
an EFCV as a result of the reduced testing frequency is bounded by
this evaluation and 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.
Attorney for licensee: John Flynn, Esq., Detroit Edison Company,
2000 Second Avenue, Detroit, Michigan 48226.
NRC Section Chief: Claudia M. Craig
Entergy Operations, Inc., Docket No. 50-313, Arkansas Nuclear One,
Unit No. 1, Pope County, Arkansas
Date of amendment request: August 18, 1999.
Description of amendment request: The proposed change would
amend Technical Specification 3.5.3 and its associated Bases to
reflect a change in the reactor coolant system (RCS) low pressure
setpoint for Arkansas Nuclear One, Unit 1 (ANO-1). The RCS low
pressure setpoint has been adjusted in the conservative direction to
account for both the uncertainties associated with the actual value
and the current number of plugged steam generator tubes.
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:
An evaluation of the proposed change has been performed in
accordance with 10 CFR 50.91(a)(1) regarding no significant hazards
considerations using the standards in 10 CFR 50.92(c). A discussion of
these standards as they relate to this amendment request follows:
Criterion 1--Does Not Involve a Significant Increase in the Probability
or Consequences of an Accident Previously Evaluated
The proposed change to raise the current technical specification
(TS) ESAS [Engineered Safeguards Actuation Signal] setpoint for low RCS
pressure does not require new hardware or physical equipment
modifications to the plant design. By raising the setpoint, a more
prompt actuation of associated safeguards equipment will be achieved
for the accident scenarios previously analyzed in the ANO-1 Safety
Analyses Report (SAR). A more expeditious actuation will ensure a more
timely response to the accident and serve to potentially decrease the
consequences of an accident. The RCS Pressure LO LO [Low Low] alarm
setpoint has also been raised and applicable procedures revised to
provide the operator sufficient time to bypass the actuation during
controlled plant maneuvers.
Therefore, the raising of the low RCS pressure ESAS setpoint from
1526 psig [pounds per square inch, guage] to 1585 psig does not involve
a significant increase in the probability or consequences of any
accident previously evaluated.
Criterion 2--Does Not Create the Possibility of a New or Different Kind
of Accident from any Previously Evaluated
The proposed change is relevant to accident response and mitigation
and has no [a]ffect on accident initiation. An inadvertent actuation of
the HPI [high pressure injection] system could result in pressurizing
the RCS to the point where a pressurizer safety valve could open and
subsequently fail to close, resulting in a loss of coolant accident.
[[Page 4271]]
However, this event remains unaffected for normal power operations and
requires discussion of depressurization events only, such as a planned
cooldown, when an inadvertent actuation could occur earlier due to the
proposed higher setpoint. This concern is mitigated by the increase of
the RCS Pressure LO LO alarm setpoint from approximately 1550 psig to
1640 psig, thus providing the operator ample time to bypass the low RCS
pressure ESAS setpoint prior to inadvertent actuation. Therefore, no
new, previously unevaluated event has been introduced relating to the
inadvertent actuation of HPI components due to the proposed change.
Therefore, this change does not create the possibility of a new or
different kind of accident from any previously evaluated.
Criterion 3--Does Not Involve a Significant Reduction in the Margin of
Safety
The proposed change conservatively raises the existing low RCS
pressure ESAS setpoint to a new value using existing installed
equipment. The new value provides protection for the entire spectrum of
break sizes based on applicable evaluations and considers the effects
of projected steam generator tube plugging activities. The setpoint is
also sufficiently below normal operating pressure to aid in preventing
spurious initiation.
Therefore, this change does not involve a significant reduction in
the margin of safety.
Therefore, based on the reasoning presented above and the previous
discussion of the amendment request, Entergy Operations, Inc. has
determined that the requested change does not involve a significant
hazards consideration.
The Nuclear Regulatory Commission (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.
Attorney for licensee: Nicholas S. Reynolds, Esquire, Winston and
Strawn, 1400 L Street, NW., Washington, DC 20005-3502.
NRC Section Chief: Robert A. Gramm.
Entergy Operations, Inc., Docket Nos. 50-313 and 50-368, Arkansas
Nuclear One, Units 1 and 2 (ANO-1&2), Pope County, Arkansas
Date of amendment request: November 16, 1999.
Description of amendment request: The proposed changes to the
Arkansas Nuclear One, Units 1 and 2 (ANO-1 and ANO-2), Technical
Specifications (TSs) and associated Bases would provide a 30-day
allowable outage time (AOT) for Startup Transformer No. 2 (SU#2) which
is an offsite power source shared by both units. This 30-day AOT would
be used infrequently for the purpose of performing preventative
maintenance on the transformer to increase its reliability. The current
TS constraints would require both units to be in cold shutdown in order
to perform this maintenance. In addition, changes have been requested
to the requirements associated with demonstrating the operability of
the emergency diesel generators to increase the reliability of this
power supply.
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:
An evaluation of the proposed changes has been performed in
accordance with 10 CFR 50.91(a)(1) regarding no significant hazards
considerations using the standards in 10 CFR 50.92(c). A discussion
of these standards as they relate to this amendment request follows:
Criterion 1--Does Not Involve a Significant Increase in the Probability
or Consequences of an Accident Previously Evaluated
Based on existing methodologies, guidance, and procedures
utilized at ANO, including required assessments of risk associated
with any significant maintenance activity, the provision of a 30-day
AOT for preplanned preventative maintenance on SU#2 is acceptable.
The resulting increase in overall risk was considered to fall into
NRC Risk Region III (``Very Small Change''). Additionally, removal
of SU#2 from service in any plant mode of operation has been
previously evaluated and found acceptable given the existing
guidance and regulations associated with offsite power sources.
Five offsite power feeds are available to the ANO switchyard
with no more than two of the feeds in close proximity to one another
for a given length, except within the switchyard itself. Failure of
one feed, regardless of the cause, will result in no more than one
additional failure, leaving at least three offsite power sources yet
available, assuming the failure remains outside the ANO switchyard.
For events that pose a threat within the ANO switchyard, four
redundant Class 1E EDGs [emergency diesel generators] and one
Alternate AC [alternating current] diesel generator are capable of
supply power to the units. Upon loss of the remaining offsite power
transformer of a unit which may be off-line, offsite power may be
restored via backfeed operations from the Main Transformers to the
Unit Auxiliary transformer to supply in-house loads. This ensures
the availability of redundant power sources, including the
applicable contingencies established during safety-related equipment
maintenance performed at ANO, are sufficient in maintaining safe
unit operations during preplanned preventative maintenance on SU#2
transformer. Therefore, providing a 30-day AOT for preplanned
preventative maintenance on SU#2, not to be applied more than once
in any 10-year period, does not involve a significant increase in
the probability or consequences of any accident previously
evaluated.
The elimination of excessive EDG operability demonstrations
(cold starts) during periods when another required power source is
inoperable acts to enhance overall EDG reliability and is consistent
with guidance provided in NRC Generic Letter 84-15 ``Proposed Staff
Actions to Improve and Maintain Diesel Generator Reliability'' and
the Revised Standard Technical Specifications (NUREG-1430 and -
1432). Verification of the operability of the remaining EDG will be
performed within 24 hours should the failure mechanism that caused
the inoperability of the redundant EDG be concluded to be a common
cause type failure. The start test in the latter case acts to ensure
that an EDG source remains available when the cause of the failure
of the redundant EDG might impact the remaining EDG.
Therefore, eliminating excessive EDG cold starts does not
involve a significant increase in the probability or consequences of
any accident previously evaluated.
Criterion 2--Does Not Create the Possibility of a New or Different Kind
of Accident from any Previously Evaluated
The removal of SU#2 from service to support needed maintenance
activities has been previously evaluated for all modes of plant
operation. Extending the current AOT to 30 days on a limited basis
does not result in any new accident initiator. The EDGs are not
considered accident initiators, but are designed to support
mitigation of accident scenarios. The elimination of excessive EDG
cold starts acts to enhance overall EDG reliability and has no
effect on accident development.
Therefore, this change does not create the possibility of a new
or different kind of accident from any previously evaluated.
Criterion 3--Does Not Involve a Significant Reduction in the Margin of
Safety
The associated probabilistic risk assessments indicate that the
proposed 30-day AOT for SU#2 does not involve a significant increase
in overall site risk, nor reduce the margin to safety. Thorough
contingency action planning, which acts to maintain the operability
of other equipment important to safety during the SU#2 maintenance
window, additionally acts to ensure the margin to safety is
maintained. The EDGs are important to safety in that they are
designed to supply power to safety system components and equipment
during a loss of offsite power. The elimination of excessive cold
starts of the EDGs acts to enhance the overall reliability of the
EDGs and, therefore, proper mitigation of accident scenarios is
likewise enhanced.
[[Page 4272]]
Therefore, this change does not involve a significant reduction
in the margin of safety.
Therefore, based on the reasoning presented above and the
previous discussion of the amendment request, Entergy Operations,
Inc. has determined that the requested change does not involve a
significant hazards consideration.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c)
are satisfied. Therefore, the NRC staff proposes to determine that
the amendment request involves no significant hazards consideration.
Attorney for licensee: Nicholas S. Reynolds, Esquire, Winston
and Strawn, 1400 L Street, NW., Washington, DC 20005-3502.
NRC Section Chief: Robert A. Gramm.
Entergy Gulf States, Inc., and Entergy Operations, Inc., Docket No.
50-458, River Bend Station, Unit 1, West Feliciana Parish,
Louisiana
Date of amendment request: December 16, 1999.
Description of amendment request: The proposed license amendment
request would revise Fuel Handling Accident (FHA) dose calculations
for 3 scenarios documented in the River Bend Station, Unit 1 (RBS),
Updated Safety Analysis Report (USAR). The first is a FHA in the
fuel building, assumed to occur 24 hours post-shutdown. A second FHA
analysis was prepared to support Amendment 35 to RBS Technical
Specifications (TS) which assumed a FHA occurs in the primary
containment 80 hours post-shutdown during Local Leakage Rate Testing
(LLRT). A third analysis was prepared in support of Amendment 85 to
the RBS TS which assumed the containment is open at 11 days.
These analyses are being updated to account for several changes.
The primary reason for the revisions, as stated by the licensee, was
to update the analyses to reflect current RBS operating strategies
and make the analyses consistent with each other. Specifically,
Cases 1 and 2 of the three analyses assumed a Radial Peaking Factor
(RPF) of 1.5 consistent with Regulatory Guide (RG) 1.25. However,
current core design strategies could lead to an RPF as high as 1.65.
In addition, to account for the potential impact of extended burnup
fuel in future operating cycles, an increased iodine-131 gap
fraction of 0.12 was more conservatively assumed in lieu of the 0.10
recommended by RG 1.25. The revised analysis also includes a change
to the control room atmospheric dispersion factors (/Q)
for the Main Control Room (MCR) ventilation system. Credit is taken
for Standard Review Plan (SRP) Section 6.4 guidance for manual dual
control room air intakes in that the /Q's are divided by
4. The revised FHA analyses also credit this action at a 20 minute
delay to be consistent with the Loss of Coolant Accident (LOCA)
analysis.
Furthermore, an error was discovered in one of the FHA
calculations. The release rate assumed in the analysis did not
ensure that the RG 1.25 assumption of a 2-hour release was
preserved. The error is the result of an inherent bias in the
secondary mixing effects in the dose calculation. The results
continue to be bounded by the guidance contained in SRP 15.7.4 and
RG 1.25.
Reanalysis showed that the release rate error, compounded with
the other changes discussed above, resulted in calculated doses
greater than those currently found in the RBS USAR. In addition,
some of the doses were also greater than those presented in the
Amendment 85 submittal. However, the licensee has stated that the
results of the revised analyses remain ``well within'' 10 CFR 100,
the guidance contained in SRP 15.7.4, and RG 1.25. Since the
analyses results are above those reported in the RBS USAR, the
criterion of 10 CFR 50.59(a)(2)(i) is, therefore, satisfied.
Accordingly, the licensee has concluded that these changes involve
an unreviewed safety question.
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.
The analyses changes described by this proposed change to the
USAR are not initiators to events, and, therefore, do not involve
the probability of an accident. The changes to the FHA calculations
for radiological doses following a FHA reflect the current operating
strategies and make the analyses consistent. These changes included:
accounting for the impact of extended burnup fuel,
addressing a change to the control room atmospheric
dispersion factors assumed in the analysis, and
revising the Radial Peaking Factor (RPF) used in the
analysis. Current core design strategies could lead to a RPF higher
[than] that assumed in Regulatory Guide 1.25.
The TRANSACT code is used for offsite dose and control room dose
calculations. The TRANSACT code is derived from the TACT V code
documented in NUREG/CR-5109. RBS has benchmarked the TRANSACT code
as discussed in the request dated August 17, 1995, (RBG-41728) which
resulted in the NRC granting Amendment 85.
The revisions to the FHA are used to establish operational
conditions where specific activities represent situations where
significant radioactive releases can be postulated. These
operational conditions include:
initial fuel movement in the Fuel Building 24 hours
after shutdown,
fuel movement in Primary Containment after 80 hours
with leakrate testing being conducted, and
fuel movement in Primary Containment with the Primary
Containment open.
Because the analyses affected by the changes are not considered
an initiator to any previously analyzed accident, these changes
cannot increase the probability of any previously evaluated
accident. Therefore, this change does not increase the probability
of occurrence of an accident evaluated previously in the safety
analysis report (SAR).
This proposed change to the USAR does increase the consequences
of an accident, but the increase is within all regulatory limits and
guidance. While the calculated off-site and control room doses of a
FHA did increase, the dose consequences remain below the regulatory
limits of 10 CFR 100 and 10 CFR 50, Appendix A, GDC [General Design
Criterion]-19 as approved per NUREG-0989, and the guidance contained
in SRP 15.7.4 of less than 25% of the 10 CFR 100 limits. The cause
of these events remains the failure of the fuel assembly lifting
mechanism. These analyses demonstrated that for the worst case
bundle drop, the regulatory dose guidelines of SRP 15.7.4 continue
to be satisfied for the required decay periods.
This change accounts for the potential effects of current fuel
design and operating strategies including increased burnup of fuel,
increased iodine-131 fraction released, Main Control Room
ventilation system operation, and release rate timing assumptions.
Reanalysis of the off-site dose calculation demonstrates that the
revised doses are increased but remain less than the regulatory
limits of 10 CFR 100 and within the guidance of SRP 15.7.4.
Therefore, this change does not significantly increase the
consequences of an accident previously evaluated in the SAR.
The proposed changes, in conjunction with existing
administrative controls, bound the conditions of the current design
basis fuel handling accident analysis. The analysis also concludes
the limiting offsite radiological consequences are well within the
acceptance criteria of NUREG[-]0800, Section 15.7.4 and 10 CFR 50,
Appendix A, GDC[-]19. The analysis is also conducted in a
conservative manner containing margins in the calculation of
mechanical analysis, iodine inventory, and iodine decontamination
factor. Each of these conservatisms will further decrease the
consequences. Therefore, the proposed changes do not significantly
increase the probability or consequences of any previously evaluated
accident.
2. The proposed changes would not create the possibility of a
new or different kind of accident from any previous[ly] analyzed.
[[Page 4273]]
This change does not involve initiators to any events in the
SAR, nor does the activity create the possibility for any new
accidents. Rather, this change is a result of the evaluation of the
most limiting FHA, which can occur at River Bend.
The proposed changes to the dose analyses are consistent with
previous limits, only revising previous evaluations to account for
current operating strategies and assumptions. These changes
included:
accounting for the impact of extended burnup fuel,
addressing a change to the control room atmospheric
dispersion factors assumed in the analysis, and
revising the Radial Peaking Factor (RPF) used in the
analysis. Current core design strategies could lead to a RPF higher
[than] that assumed in Regulatory Guide 1.25.
The radiological consequences remain within accepted limits of
10 CFR 100 and guidance of the Standard Review Plan (NUREG-0800)
Section 15.7.4. Therefore, these changes 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[ly] analyzed.
3. The proposed changes do not involve a significant reduction
in a margin of safety.
The dose consequences are calculated in accordance with
regulatory guidance found in Regulatory Guide 1.25 and the SRP
[S]ection 15.7.4. The RBS analyses conservatively assumed that
failures are consistent with those in the standard General Electric
GESTAR II. These analyses result in a bounding number of fuel
failures. The RBS analyses are also consistent with those approved
by the NRC [Nuclear Regulatory Commission] in support of Technical
Specification Amendments 35 and 85 to the River Bend Station license
(NPF-47). The radiological dose consequences resulting from these
failures are therefore analyzed using accepted methods and criteria.
In addition, the analyses contain known conservatisms and margins to
ensure the results will remain bounding.
The revised limits are 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
regulatory limits and guidance. Safety margins and analytical
conservatisms have been evaluated and are well understood.
Conservative methods of analysis are maintained through the use of
accepted methodology and benchmarking the proposed methods to
previous analysis. Margins are retained to ensure that the analysis
adequately bounds all postulated event scenarios. The proposed
change only eliminates some excess conservatism from the analysis.
In addition, EOI [Entergy Operations, Inc.] has implemented
NUMARC [Nuclear Management and Resources Council (now NEI)] 91-06
guidelines for shutdown operations at RBS. Shutdown Operations
Protection Plan and Primary-Secondary Containment Integrity
procedures presently include guidance for closure of the containment
hatch and other significant openings in containment, in addition to
the requirements contained in the license and design basis. This
additional protection will enhance the ability to limit offsite
effects.
Acceptance limits for the fuel handling accident are provided in
10 CFR 100 with additional guidance provided in NUREG[-]0800,
Section 15.7.4. The proposed changes continue to ensure that the
whole-body and thyroid doses at the exclusion area and low
population zone boundaries, as well as control room doses, are below
the corresponding regulatory limits. These margins are unchanged,
therefore, the proposed changes do not involve a significant
reduction in a margin of safety.
The commission has provided guidance concerning the application
of the standards of 10 CFR 50.92 by providing certain examples (51
FR 7751, March 6, 1986) of amendments that are not considered likely
to involve a significant hazards consideration. This proposed
amendment is very similar to example (vi):
(vi) A change which either may result in some increase to the
probability or consequences of a previously-analyzed accident or may
reduce in some way a safety margin, but where the results of the
change are clearly within all acceptable criteria with respect to
the system or component specified in the Standard Review Plan: for
example, a change resulting from the application of a small
refinement of a previously used calculational model or design
method.
As we have shown in the preceding discussion, this refinement to
the FHA dose calculation results in a small increase to the
consequences of a previously analyzed accident, but the results of
the change remain clearly within the guidelines of 10 CFR 100,
Appendix A, GDC[-]19, and the guidance of SRP [S]ection 15.7.4,
without reducing 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.
Attorney for licensee: Mark Wetterhahn, Esq., Winston & Strawn,
1400 L Street, NW., Washington, DC 20005.
NRC Section Chief: Robert A. Gramm.
Entergy Gulf States, Inc., and Entergy Operations, Inc., Docket No. 50-
458, River Bend Station, Unit 1, West Feliciana Parish, Louisiana
Date of amendment request: December 20, 1999.
Description of amendment request: The proposed amendment would
change River Bend Station (RBS) Technical Specification (TS) 3.6.1.3,
``Primary Containment Isolation Valves (PCIVs),'' to allow the Inclined
Fuel Transfer System (IFTS) primary containment isolation blind flange
to be removed during MODE 1, 2, or 3. In its application, the RBS
licensee stated that, with the blind flange removed and certain
restrictions and administrative controls in place, the IFTS penetration
would not represent an uncontrolled breach of the containment boundary
and that the containment isolation function would continue to be
provided through implementation of these additional controls.
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.
The proposed change permits removal of the blind flange on the
Inclined Fuel Transfer System (IFTS) when primary containment
operability is required in MODE[S] 1, 2, and 3. This will permit
operation of the IFTS while the plant is operating. With respect to
the probability of an accident, this aspect of the containment
structure does not directly interface with the reactor coolant
pressure boundary. The removal of this blind flange does not involve
modifications to plant systems or design parameters that could
contribute to the initiation of any accidents previously evaluated.
Operation of IFTS is unrelated to the operation of the reactor, and
there is no aspect of IFTS operation that could lead to or
contribute to the probability of occurrence of an accident
previously evaluated. Removal of the blind flange and operation of
IFTS does not result in changes to procedures that could impact the
occurrence of an accident.
With respect to the issue of consequences of an accident, the
function of the containment is to mitigate the radiological
consequences of a loss of coolant accident (LOCA) or other
postulated events that could result in radiation being released from
the fuel inside containment. While the proposed change does not
change the plant design, it does permit alteration of the
containment boundary for the IFTS penetration. Altering the
containment boundary in this case (i.e., removing the blind flange)
results in some IFTS components possibly being subjected to
containment pressure in the event of a LOCA. However, the additional
post-accident peak pressure load to be imposed upon the components
in the IFTS if the blind flange is removed is a small fraction of
their design capability. Therefore, they are considered an
acceptable barrier to prevent uncontrolled release of post-accident
fission products for this proposed change.
The proposed change required examination of two potential
leakage pathways. The larger
[[Page 4274]]
is the IFTS transfer tube, itself. The other, much smaller one, is a
branch line used for draining the IFTS transfer tube during its
operation. It is clear that the gate valve at the bottom of the
transfer tube is always water sealed and maintained so by the
submergence of the water in the transfer tube and in the fuel
building spent fuel storage pool (the lower pool). The height of
this water seal is greater than that necessary to prevent leakage
from the bottom of the transfer tube during accidents that result in
the calculated peak post-DBA [design basis accident] LOCA pressure,
Pa . Furthermore, the hydraulically operated gate valve in
the lower end of the tube will remain closed, and has pressure
retaining capability greater than that of the containment structure
itself. The potential leakage pathway from the drain piping which
attaches to the transfer tube will be isolated if required, via
administrative controls on the drain piping isolation valve.
Additionally, the drain piping isolation valve will be added to the
Primary Containment Leakage Rate Testing Program (Technical
Specification 5.5.13) to ensure that leakage past this valve will be
maintained consistent with the leakage rate assumptions of the
accident analysis. Due to the test methodology, the portion of the
large transfer tube piping outboard of the blind flange (the portion
of the tube which becomes exposed to the containment atmosphere
during the draining portion of the IFTS operation) will also be part
of the leakage rate test boundary and will therefore also be tested.
Therefore, no unidentified leakage will exist from the piping and
components that are outboard of the blind flange, and the leakage
rate assumptions of the accident analysis will be maintained. Note
that the bottom gate valve in the IFTS transfer tube will remain
closed for this test evolution.
Therefore, the proposed change does not result in a significant
increase in the probability of the consequences of previously
evaluated accidents, provided the bottom gate valve remains closed
during MODE 1, 2, or 3 operation.
2. The proposed changes would not create the possibility of a
new or different kind of accident from any previous analyzed.
The proposed change consists of the removal of a passive
component which is not part of the primary reactor coolant pressure
boundary nor involved in the operation or shutdown of the reactor.
Being passive, its presence or absence does not affect any of the
parameters or conditions that could contribute to the initiation of
any incidents or accidents that are created from a loss of coolant
or an insertion of positive reactivity. Realigning the boundary of
the primary containment to include portions of the IFTS is also
passive in nature and therefore has no influence on, nor does it
contribute to the possibility of a new or different kind of
incident, accident or malfunction from those previously analyzed.
Furthermore, operation of the IFTS is unrelated to the operation of
the reactor and there is no mishap in the process that can lead to
or contribute to the possibility of losing any coolant from the
reactor or introducing the chance for an insertion of positive or
negative reactivity, or any other accidents different from and not
bounded by those previously evaluated.
Therefore, the proposed change does not result in creating the
possibility of a new or different kind of accident from any accident
previously evaluated, provided the bottom gate valve remains closed
during MODE 1, 2, or 3 operation.
3. The proposed changes do not involve a significant reduction
in a margin of safety.
The proposed change involves the realignment of the primary
containment boundary by removing the blind flange which is a passive
component. The margin of safety that has the potential of being
impacted by the proposed change involves the dose consequences of
postulated accidents which are directly related to potential leakage
through the primary containment boundary. The potential leakage
pathways due to the proposed change have been reviewed, and leakage
can only occur from the administratively controlled IFTS transfer
tube drain piping, and from the IFTS transfer tube itself. A
dedicated individual will be designated to provide timely isolation
of this drain piping during the duration of time when this proposed
change is in effect. The conservatively calculated dose which might
be received by the designated individual while isolating the drain
piping is calculated to be 3.8 rem TEDE [total effective dose
equivalent], which remains within the guidelines of General Design
Criterion (GDC) 19 (10 CFR 50, Appendix A, Criterion 19).
Furthermore, the drain piping isolation valve will be added to the
Primary Containment Leakage Rate Testing Program (Technical
Specification 5.5.13) to ensure that leakage from the piping and
components located outboard of the blind flange will be maintained
consistent with the leakage rate assumptions of the accident
analysis.
Studies of the capability of the IFTS system to withstand
containment pressurization under severe accident conditions have
been conducted. These studies conclude that IFTS, including the
transfer tube and its valves, has a capability to withstand beyond
design basis severe accident containment pressures which is greater
than that of the containment structure itself. The RBS Emergency
Operating Procedures (EOPs) are based on an ultimate containment
failure pressure capability of 53 psig [pounds per square inch--
gauge], which represents a margin of safety of 38 psi above the 15
psig containment design pressure. This margin of safety is not
impacted with the IFTS blind flange removed as long as the IFTS
bottom valve remains closed. This capability to withstand
containment pressurization under severe accident conditions envelops
other non-DBA LOCA scenarios, such as the small break LOCA. For the
large break LOCA, additional defense-in-depth is provided by
maintaining a water seal greater than Pa above the outlet
of the IFTS transfer tube in the lower pool.
Therefore, the proposed change does not involve a significant
reduction in a margin of safety.
The Nuclear Regulatory Commission (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.
Attorney for licensee: Mark Wetterhahn, Esq., Winston & Strawn,
1400 L Street, NW., Washington, DC 20005.
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: July 15, 1999 (NPF-38-216).
Description of amendment request: One proposed change adds a
Technical Specification (TS) Bases Control Program to the Waterford 3
TS Administrative Controls Section, modeled after the guidelines
contained in NUREG-1432. Additionally, the proposed change corrects an
editorial error identified in the TS following issuance of Amendment
146, dated October 19, 1998.
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: The proposed changes to the Waterford 3 [Waterford
Steam Electric Station, Unit 3] Technical Specifications add a TS
Bases Control Program and correctly reference the appropriate
document where administrative controls were relocated. The TS Bases
Control Program will provide administrative controls that ensure
changes to the TS Bases are appropriately reviewed and consistent
with the Updated Final Safety Analysis Report (UFSAR). The addition
of the proposed program does not affect any accident initiator or
mitigation of any events analyzed in Chapter 15 of the UFSAR. Also,
neither change has any affect on the operation of any structures,
systems, or components or the assumptions of any accident analyses.
The TS Bases Control Program will ensure that any change to the
Bases that involves an unreviewed safety question will receive prior
Nuclear Regulatory Commission approval. Changing the reference to
the Quality Assurance Program Manual (QAPM) for the item relocated
to the QAPM is purely administrative.
[[Page 4275]]
Therefore, the proposed changes 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: The proposed changes to the Waterford 3 TS add a TS
Bases Control Program and correctly reference the appropriate
document where administrative controls were relocated. The addition
of a TS Bases Control Program represents an administrative function
performed under existing regulatory controls consistent with 10 CFR
50.59. The proposed change to reference the appropriate document
where an administrative control was relocated is purely
administrative in nature. The change merely corrects the Technical
Specifications wording to reflect the actual location of the record
retention requirements for records of reviews performed on changes
to the Process Control Plan (PCP) and Offsite Dose Calculation
Manual (ODCM) in the QAPM.
These proposed changes do not involve a change in plant design
or affect the configuration or operation of any structure, system,
or component, nor does it involve any potential initiating events
that would create any new or different kind of accident. Therefore,
the proposed changes 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: The proposed changes to the Waterford 3 TS add a TS
Bases Control Program and correctly reference the appropriate
document where administrative controls were relocated. The addition
of a TS Bases Control Program is an administrative change and has no
[a]ffect on a margin of safety, as defined by Section 2 of the TS.
The only [a]ffect of the TS Bases Control Program is to establish
controls over how TS Bases changes are reviewed and implemented
consistent with 10 CFR 50.59.
The proposed change to a reference in the Administrative
Controls section merely corrects the TS wording to reflect the
actual location of the record retention requirements for records of
reviews performed on changes to the PCP and ODCM in the QAPM.
These proposed changes do not involve a change in plant design
or have any affect on the plant protective barriers. Therefore, the
proposed changes will not involve a significant reduction in a
margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c)
are satisfied. Therefore, the NRC staff proposes to determine that
the amendment request involves no significant hazards consideration.
Attorney for licensee: N. S. Reynolds, Esquire, Winston & Strawn
1400 L Street NW., 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: July 15, 1999 (NPF-38-217).
Description of amendment request: The proposed change creates a
new Technical Specification (TS) for the Main Feedwater Isolation
Valves Section modeled after the guidelines of TS 3.7.3 in NUREG-
1432. Additionally, the letter provides for Nuclear Regulatory
Commission (NRC) Staff review of an unreviewed safety question
regarding the crediting of the Reactor Trip Override feature and
Auxiliary Feedwater Pump high discharge pressure trip as assisting
the operation of the Main Feedwater Isolation Valves during their
required safety function, to close on a Main Steam Isolation Signal.
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: The proposed change to add the Main Feedwater
Isolation Valves (MFIVs) to the Technical Specifications (TS) and
provide an allowed outage time of 72 hours with appropriate required
ACTIONs does not affect the operation of any structures, systems, or
components or the assumptions of any accident analyses. The MFIVs
are primarily designed to mitigate the consequences of a Main Steam
Line Break (MSLB), and the Feedwater Line Break (FWLB). This TS
change ensures the 5 second closure time currently assumed in the
Waterford 3 [Waterford Steam Electric Station, Unit 3] analysis,
thus it preserves the current analysis. Hence, the consequences of
accidents previously evaluated do not change. Therefore, this change
does not involve an increase in the consequences of any accident
previously evaluated. Adding the MFIVs to the TS will not initiate
an accident. Providing a TS and allowed outage time makes no changes
to the plant and, thus, no increase in the probability of any
accident previously evaluated.
The accidents/events that may be affected by the proposed
resolution to credit the Reactor Trip Override (RTO) circuitry for
the Steam Generator [SG] Feed Pumps (SGFPs) during SGFP operation
and the crediting of the Auxiliary Feedwater (AFW) pump high
discharge pressure trip during AFW pump operation are the MSLB and
the FWLB.
The crediting of the RTO circuitry for the SGFPs and the
crediting of the AFW pump trip will not affect the probability of
occurrence of a MSLB or FWLB. Neither the SGFPs nor the AFW pump are
initiators of either line break.
The crediting of the RTO circuitry for the SGFPs and the
crediting of the AFW pump trip will not adversely affect the
consequences of a MSLB or FWLB. Ultimately, the RTO feature allows
more reliable MFIV closure by reducing the differential pressure
against which the MFIVs must close while not introducing a new
failure mechanism such as a Loss of Feedwater or water hammer event.
The RTO feature (which has always been a part of the Waterford 3
plant design) mitigates the consequences of the MSLB and MFLB by
reducing flow to the affected steam generator and containment.
The Loss of Feedwater Event can be initiated by the loss of a
SGFP. The currently analyzed Loss of Feedwater Event evaluates the
loss of both SGFPs, which bounds a potential loss of one SGFP.
Therefore, any modification that could increase the probability of a
pump trip could increase the probability of this event. Since the
proposed solution of crediting RTO features of the SGFPs and the
trip of the AFW pump for the MFIV margin issue uses existing
functions, no new features/trips will be added, and there is no
increase in the probability or consequences of a Loss of Feedwater
Event. The only plant modification being made is to enhance RTO such
that it will run the SGFPs back to a minimum speed on a reactor
trip, even when the FWCS [Feedwater Control System] is in manual.
Although this slows the pump down, feedwater and the SGFPs remain
available and the Loss of Feedwater Event probability is not
significantly increased. The modification to make RTO function when
the FWCS is in manual is not significant since the FWCS is in manual
such a short period of time during plant operation.
The AFW system is not credited in any accident analysis. The
Emergency Feedwater (EFW) system is relied upon in the safety
analyses to replenish SG inventory. Therefore, crediting the AFW
pump discharge pressure trip will not involve an increase in the
probability or consequences of any accident.
In conclusion, the proposed TS change and resolution to the MFIV
margin issue 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
[[Page 4276]]
accident from any accident previously evaluated?
Response: The proposed TS change in itself does not change the
design or configuration of the plant. No new or different equipment is
being installed by the TS. No new or different accidents result from
the addition of the MFIVs to the TS. Previously performed accident
analyses remain valid. The proposed allowed outage time and required
actions of the proposed TS do not change the procedural operation of
the plant, but specify the requirements for treatment of the MFIVs
under the plant TS. Therefore, no new or different type of accident
from any accident previously evaluated is created.
No new system interaction is created by crediting the existing RTO
and AFW pump trip. Failure to isolate feedwater would require two
failures, failure of the RTO or AFW circuitry, in addition to the
failure of the Main Feedwater Regulating Valves (MFRVs) and Startup
Feedwater Regulating Valves (SFRVs) to close, and is beyond single
failure criteria. If the RTO and AFW features were the single failure,
then closure of the regulating valves would be credited for MSIS [Main
Steam Isolation Signal] isolation since the regulating valves were
designed to close against SGFP shutoff head.
RTO and AFW pump trips would not be considered initiators of a MSLB
or FWLB, but could be considered initiators of a Loss of Feedwater
Event. However, this event is bounded by the analyzed Waterford 3 Loss
of Feedwater Events. No new event is created. The only hardware change
being made is the use of RTO for pump run back when the FWCS is in
manual. The existing signal will be used and routed through the same
methods as are currently installed, ensuring it will run the pump back
appropriately. Therefore, no new system interactions or events are
created.
The new method of potential failure that has not previously been
evaluated is in the fact that Waterford 3 would now be crediting a non-
safety related circuit for closure of the safety related MFIVs. Non-
safety features are not normally credited for the proper operation of a
safety related component. However, in this case, for the valve to close
in the 5 seconds assumed in safety analyses, the RTO and AFW pump trip
will be credited. Because this is new, different and not a previously
approved allowance, this resolution must be submitted for NRC Staff
approval. Entergy believes this resolution is acceptable based on the
high degree of reliability of these components.
The system design, as discussed above, does not increase the
potential for a Loss of Feedwater Event and current analyses bound all
potential accident scenarios. Therefore, the proposed TS change and
resolution to the MFIV margin issue 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: The MFIVs have no [a]ffect on a margin of safety as
defined by Section 2 of the TS. Their only [a]ffect is response to the
accidents described above, which will be enhanced by specifying an
allowed outage time, action requirements and surveillance requirements
in the TS. Therefore, no reduction in the margin of safety is involved
with the addition of these valves to the TS.
No new system interaction is created by the crediting of the RTO
feature or the AFW pump trip, or the addition of RTO operation in
manual.
The proposed resolution does affect a part of a protective
boundary, the MFIV, which serves to isolate the Main Feedwater system
from portions of the system inside containment. However, it does not
affect operation or function of the valve itself since no changes to
the valve are being made. The proposal allows increased margin for
valve closure; therefore, margins of safety are not affected. The valve
will close within the time limits required by safety analyses and
general design criteria.
Therefore, the proposed TS change and resolution to the MFIV margin
issue 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.
Attorney for licensee: N. S. Reynolds, Esquire, Winston & Strawn,
1400 L Street NW., 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: July 15, 1999 (NPF-38-218).
Description of amendment request: The proposed changes extend the
Reactor Coolant System Pressure Temperature Curves to 20 Effective Full
Power Years.
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: The proposed changes will not increase the probability
or consequences of any accident previously evaluated since the
proposed changes revise the pressure/temperature limits in
accordance with 10 CFR 50, Appendix G, utilizing the latest NRC
[Nuclear Regulatory Commission] guidelines in Regulatory Guide 1.99,
Revision 2, relative to estimating neutron irradiation damage to the
reactor vessel. The proposed changes also maintain the conservative
limits with respect to the low temperature overprotection (LTOP)
system and heatup and cooldown restrictions.
Therefore, the proposed change will not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
2. Will operation of the facility in accordance with this
proposed change create the possibility of a new or different kind of
accident from any accident previously evaluated?
Response: The proposed changes will not create the possibility
of a new or different kind of accident from any previously analyzed
since they do not introduce new systems, failure modes, or other
plant perturbations. The proposed changes revise the pressure/
temperature limits in accordance with 10 CFR 50, Appendix G,
utilizing the latest NRC guidelines in Regulatory Guide 1.99,
Revision 2, relative to estimating neutron irradiation damage to the
reactor vessel.
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: The proposed changes will not involve a significant
reduction in the margin of safety since equal or more stringent
pressure/temperature limitation requirements for reactor operation
will be applied. The proposed changes were derived in accordance
with approved NRC methodology which was developed to assure the
reactor coolant system pressure boundary is designed with sufficient
margin to withstand any condition during normal operation including
anticipated operational occurrences and system inservice leak and
hydrostatic tests.
These requirements were revised in accordance with 10 CFR 50,
Appendix G, utilizing the latest NRC guidance in Regulatory Guide
1.99, Revision 2, relative to estimating neutron irradiation damage
to the reactor vessel. The LTOP system limits were also reanalyzed
for the proposed changes.
[[Page 4277]]
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.
Attorney for licensee: N. S. Reynolds, Esquire, Winston & Strawn,
1400 L Street NW., 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: July 19, 1999. (NPF-38-219).
Description of amendment request: The proposed changes modify
Waterford Steam Electric Station, Unit 3 (Waterford 3) Technical
Specification (TS) 4.5.2.f.2 by increasing the performance requirement
for the low pressure safety injection (LPSI) pumps. The change revises
the LPSI pump Surveillance Requirements to measure pump developed head,
instead of pump discharge pressure. The associated changes to TS Bases
are included in the submittal.
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: Increasing the LPSI pump performance requirements will
not increase the probability or consequences of any accidents. There
are no physical changes to the pump. The only procedure changes
required are to Surveillance Procedure OP-903-030, ``Safety
Injection Pump Operability Evaluation.'' The changes do not impact
plant operating procedures. The LPSI system is primarily designed to
mitigate the consequences of a large break Loss of Coolant Accident
(LOCA). These proposed changes do not affect any of the assumptions
used in the deterministic LOCA analysis. Hence the consequences of
accidents previously evaluated do not change.
Therefore, the proposed change will not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
2. Will operation of the facility in accordance with this
proposed change create the possibility of a new or different kind of
accident from any accident previously evaluated?
Response: The proposed change does not alter plant operations,
nor does it alter the physical plant. The change only increases
existing equipment performance requirements. No different accidents
result from the increase in performance requirements. No change is
being made to the parameters within which the plant is operated. The
setpoints at which protective or mitigative actions are initiated
are unaffected by this change. No alteration in the procedures which
ensure the plant remains within analyzed limits is being proposed,
and no change is being made to the procedures relied upon to respond
to an off-normal event. As such, no new failure modes are being
introduced. The proposed change will only increase the performance
requirements of the LPSI pumps.
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: To the contrary, the change increases LPSI pump
performance requirements, increasing the margin between the TS
performance requirements and the analytical limit.
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.
Attorney for licensee: N. S. Reynolds, Esquire, Winston & Strawn,
1400 L Street NW., 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: August 4, 1999 (NPF-38-222).
Description of amendment request: The proposed change modifies
Technical Specifications (TS) 3.5.2 to extend the allowed outage time
(AOT) to seven days for one high pressure safety injection (HPSI) train
inoperable and TS 3.5.3 to change the end-state to HOT SHUTDOWN with at
least one OPERABLE shutdown cooling train in operation. Additionally,
an AOT of 72 hours in TS 3.5.2 is imposed for other conditions where
the equivalent of 100 percent emergency core cooling system (ECCS)
subsystem flow is available. If 100 percent ECCS flow is unavailable
due to two inoperable HPSI trains, an ACTION has been added to restore
at least one HPSI to OPERABLE status within one hour or place the plant
in HOT STANDBY in six hours and to exit the MODE of applicability in
the following six hours. In the event the equivalent of 100 percent
ECCS subsystem flow is not available due to other conditions, TS 3.0.3
is entered. The Limiting Condition for Operation terminology is being
changed for consistency with the ECCS requirements. Additionally, the
associated TS Bases are being changed.
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: The High Pressure Safety Injection System (HPSI) is
part of the Emergency Core Cooling System subsystem. Inoperable HPSI
components are not accident initiators in any accident previously
evaluated. Therefore, this change does not involve an increase in
the probability of any accident previously evaluated.
The HPSI system is primarily designed to mitigate the
consequences of a Loss of Coolant Accident (LOCA). These proposed
changes do not affect any of the assumptions used in the
deterministic LOCA analyses. Hence the consequences of accidents
previously evaluated do not change.
In order to fully evaluate the HPSI AOT extension, probabilistic
safety assessment (PSA) methods were utilized. The results of these
analyses show no significant increase in the core damage frequency.
These analyses are detailed in report CE NPSD-1041, ``Joint
Applications Report for High Pressure Safety Injection System
Technical Specification Modifications,'' March 1998.
The Configuration Risk Management Program is an Administrative
Program that assesses risk based on plant status. Adding the
requirement to implement this program for Technical Specification
3.5.2 does not affect the probability or the consequences of an
accident.
The proposed change allows a combination of equipment from
redundant trains to be inoperable provided that at least the
equivalent of a single ECCS subsystem remains operable. Analyzed
events are assumed to be initiated by the failure of plant
structures, systems or components. Allowing equipment from redundant
trains to constitute a single operable subsystem does not increase
the probability that a failure leading to an analyzed event will
occur. The ECCS components are passive until an actuation signal is
generated. This change does not increase the failure probability of
the ECCS components. This change reduces the plant's susceptibility
to common cause failures. As such, the probability of occurrence for
a previously analyzed accident are not significantly increased.
[[Page 4278]]
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: The proposed change does not change the design or
configuration of the plant. No new equipment is being introduced,
and installed equipment is not being operated in a new or different
manner. There is no change being made to the parameters within which
the plant is operated, and the setpoints at which protective or
mitigative actions are initiated are unaffected by this change. No
alteration in the procedures which ensure the plant remains within
analyzed limits is being proposed, and no change is being made to
the procedures relied upon to respond to an off-normal event. As
such, no new failure modes are being introduced. The proposed change
will only provide the plant some flexibility in maintaining the
minimum equipment required to be operable to perform the ECCS
function while in this condition. The change does not alter
assumptions made in the safety analysis and licensing basis.
Therefore, the change does not create the possibility of a new or
different kind of accident from any accident previously evaluated.
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: The proposed changes do not affect the limiting
conditions for operation or their bases used in the deterministic
analysis to establish the margin of safety. PSA evaluations were
used to evaluate these changes. These evaluations demonstrate that
the changes involve no significant increase in risk. These
evaluations are detailed in report CE NPSD-1041. The margin of
safety is established through equipment design, operating
parameters, and the setpoints at which automatic actions are
initiated. None of these are adversely impacted by the proposed
change. Sufficient equipment remains available to actuate upon
demand for the purpose of mitigating a transient event. The proposed
change, which allows operation to continue for up to 72 hours with
components inoperable in both ECCS subsystems, is acceptable based
on the remaining ECCS components providing 100% of the required ECCS
flow. The reduced potential for a self-induced plant transient
resulting from unit shutdown required for a second inoperable ECCS
train is minimized. Therefore, the change does not involve a
significant reduction in the margin of safety, and is offset by
minimizing the potential for a self induced plant transient.
Therefore, the proposed change will not involve a significant
reduction in a margin of safety.
The Nuclear Regulatory Commission (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.
Attorney for licensee: N. S. Reynolds, Esquire, Winston & Strawn,
1400 L Street NW., 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: August 4, 1999 (NPF-38-223).
Description of amendment request: The proposed change modifies
Technical Specification (TS) 3.5.2 to extend the allowed outage time
(AOT) to seven days for one low pressure safety injection (LPSI) train
inoperable. Additionally, an AOT of 72 hours is imposed for other
conditions where the equivalent of 100 percent emergency core cooling
system (ECCS) subsystem flow is available. If 100 percent ECCS flow is
unavailable due to two inoperable LPSI trains, an ACTION has been added
to restore at least one LPSI train to OPERABLE status within one hour
or place the plant in HOT STANDBY in six hours and to exit the MODE of
applicability in the following six hours. In the event the equivalent
of 100 percent ECCS subsystem flow is not available due to other
conditions, TS 3.0.3 is entered. The Limiting Condition for Operation
terminology is being changed for consistency with the ECCS
requirements. Additionally, the associated TS Bases are being changed.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. Will operation of the facility in accordance with this
proposed change involve a significant increase in the probability or
consequences of an accident previously evaluated?
Response: No. The Low Pressure Safety Injection System (LPSI) is
part of the Emergency Core Cooling System subsystem. Inoperable LPSI
components are not accident initiators in any accident previously
evaluated. Therefore, this change does not involve an increase in
the probability of an accident previously evaluated.
The LPSI system is primarily designed to mitigate the
consequences of a large Loss of Coolant Accident (LOCA). These
proposed changes do not affect any of the assumptions used in the
deterministic LOCA analysis. Hence, the consequences of accidents
previously evaluated do not change.
In order to fully evaluate the LPSI AOT extension, probabilistic
safety analysis (PSA) methods were utilized. The results of these
analyses show no significant increase in the core damage frequency.
As a result, there would be no significant increase in the
consequences of an accident previously evaluated. These analyses are
detailed in CE NPSD-995, Combustion Engineering Owners Group ``Joint
Applications Report for Low Pressure Safety Injection System AOT
Extension.''
The Configuration Risk Management Program is an Administrative
Program that assesses risk based on plant status. Adding the
requirement to implement this program for Technical Specification
3.5.2 does not affect the probability or the consequences of an
accident.
The proposed change allows a combination of equipment from
redundant trains to be inoperable provided that at least the
equivalent of single train of ECCS remains operable. Analyzed events
are assumed to be initiated by the failure of plant structures,
systems or components. Allowing equipment from redundant trains to
constitute a single operable train does not increase the probability
that a failure leading to an analyzed event will occur. The ECCS
components are passive until an actuation signal is generated. This
change does not increase the failure probability of the ECCS
components. This change reduces the plant's susceptibility to common
cause failures. As such, the probability of occurrence for a
previously analyzed accident are not significantly increased.
Therefore, the proposed change will not involve a significant
increase in the probability or consequences of any accident
previously evaluated.
2. Will operation of the facility in accordance with this
proposed change create the possibility of a new or different type of
accident from any accident previously evaluated?
Response: No. The proposed change does not change the design or
configuration of the plant. No new equipment is being introduced,
and installed equipment is not being operated in a new or different
manner. There is no change being made to the parameters within which
the plant is operated, and the setpoints at which protective or
mitigative actions are initiated are unaffected by this change. No
alteration in the procedures which ensure the plant remains within
analyzed limits is being proposed, and no change is being made to
the procedures relied upon to respond to an off-normal event. As
such, no new failure modes are being introduced. The proposed change
will only provide the plant some flexibility in maintaining the
minimum equipment required to be operable to perform the ECCS
function while in this Condition. The change does not alter
assumptions made in the safety analysis and licensing basis.
Therefore, the change does not create the possibility of a new or
different kind of accident from any accident previously evaluated.
Therefore, the proposed change will not create the possibility
of a new or different kind of accident from any accident previously
evaluated.
[[Page 4279]]
3. Will operation of the facility in accordance with this
proposed change involve a significant reduction in a margin of
safety?
Response: No. The proposed changes do not affect the limiting
conditions for operation or their bases used in the deterministic
analyses to establish the margin of safety. PSA evaluations were
used to evaluate these changes. These evaluations demonstrate that
the changes are either risk neutral or risk beneficial. These
evaluations are detailed in CE NPSD-995. The margin of safety is
established through equipment design, operating parameters, and the
setpoints at which automatic actions are initiated. None of these
are adversely impacted by the proposed change. Sufficient equipment
remains available to actuate upon demand for the purpose of
mitigating a transient event. The proposed change, which allows
operation to continue for up to 72 hours with components inoperable
in both ECCS trains, is acceptable based on the remaining ECCS
components providing 100% of the required ECCS flow. The reduced
potential for a self-induced plant transient resulting from unit
shutdown required for a second inoperable ECCS train is minimized.
Therefore, the change does not involve a significant reduction in
the margin of safety, and is offset by minimizing the potential for
a self induced plant transient.
Therefore, the proposed change will not involve a significant
reduction in a margin of safety.
The Nuclear Regulatory Commission (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.
Attorney for licensee: N. S. Reynolds, Esquire, Winston & Strawn
1400 L Street NW., Washington, DC 20005-3502.
NRC Section Chief: Robert A. Gramm.
Indiana Michigan Power Company, Docket Nos. 50-315 and 50-316, Donald
C. Cook Nuclear Plant, Units 1 and 2, Berrien County, Michigan
Date of amendment requests: December 3, 1998.
Description of amendment requests: The proposed amendments would
add a new Technical Specification (T/S) and associated Bases for the
distributed ignition system (DIS). The proposed change incorporates the
technical requirements of NUREG-1431, Revision 1, ``Standard Technical
Specifications, Westinghouse Plants.''
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. The proposed amendment will not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
The T/S being proposed for the DIS is consistent with its design
and operation as previously reviewed and approved, and therefore,
does not involve a significant increase in the probability or
consequences of an accident previously evaluated. The amendments
involve new requirements for the T/Ss and do not delete any existing
requirements.
2. The proposed amendment will not create the possibility of a
new or different kind of accident previously evaluated.
The T/S being proposed for the DIS is consistent with its design
and operation as previously reviewed and approved, and therefore,
does not create the possibility of a new or different kind of
accident from any accident previously evaluated.
3. The proposed amendment will not involve a significant
reduction in a margin of safety.
The T/S being proposed for the DIS is consistent with [the]
design and operation as previously reviewed and approved, and
therefore, does not involve a significant reduction in a margin of
safety. Compliance with the proposed T/S will provide additional
assurance of system availability to maintain a margin of safety for
containment integrity during degraded core events.
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 requests involve no significant hazards consideration.
Attorney for licensee: David W. Jenkins, Esq., 500 Circle Drive,
Buchanan, MI 49107.
NRC Section Chief: Claudia M. Craig.
Nebraska Public Power District, Docket No. 50-298, Cooper Nuclear
Station, Nemaha County, Nebraska
Date of amendment request: December 15, 1999.
Description of amendment request: This proposed technical
specification (TS) change will revise the average power range Monitors
(APRMs) neutron flux-high (flow biased) allowable value based on a
revised power to flow map. The revised power to flow map extends the
current plant operating domain to above the rated rod line, to within
an envelope referred to as the maximum extended load line limit (MELLL)
and adds the increased core flow (105%) region. The current power to
flow map is based on a region bounded by the extended load line limit
(ELLL) and evaluations prepared as part of the Core Operating Limits
Report (COLR).
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. The proposed amendment will not involve a significant
increase in the probability or consequences of an accident
previously evaluated. Attachment 3 [to the December 15,1999
application] (Reference 1) evaluates operation in the Maximum
Extended Load Line Limit (MELLL) and Increased Core Flow (ICF)
regions and the impact on equipment and safety system performance.
Impacts on containment, the reactor vessel, Recirculation System,
reactor vessel internals, limiting transients for the Cycle 20
reload (upcoming refuel outage), Loss of Coolant Accident (LOCA),
and Anticipated Transients Without SCRAM (ATWS) events were
evaluated. The conclusion is that for all events, accidents, and
equipment evaluated, operation and event response remain within
previously established design limits and acceptance criteria. No
changes in the initiators of accidents previously evaluated are
being made by this change. Because operation in the expanded regions
maintains adequate design margin and there are no changes in the
accident initiators, the proposed change does not involve a
significant increase in the probability of an accident previously
evaluated.
In support of operation in the MELLL region, the proposed change
modifies (increases) the Average Power Range Monitor (APRM) Neutron
Flux-High (Flow Biased) allowable value. Changes to the setpoint and
allowable value will be implemented in accordance with approved
setpoint methodology and plant procedures (References 7 and 8). As
noted in Technical Specifications (TS) Bases Section B.3.3.1.1.2.b:
``No specific safety analyses take credit for the APRM Neutron Flux-
High (Flow Biased) Function.'' The APRM allowable value credited in
accident analyses is based on the 120% fixed scram-allowable value
(TS Table 3.3.1.1-1, Function 2.c), which remains unchanged as a
result of this requested TS change. Though not credited in analyses,
the limiting flow biased value of 119% Reactor Thermal Power (RTP)
also remains unchanged. Evaluations presented in Attachment 3
demonstrate that operation in the MELLL envelope, with reliance on
the credited fixed scram allowable value (analytically assumed at
123% RTP to justify a 120% TS allowable value), results in event and
accident responses within design limits and established acceptance
criteria. Therefore, no significant increase in source term,
radiological consequences or other accident consequences occurs as a
result of the proposed change.
The proposed change has no affect on operation in the ICF
region. The allowable value, as part of the proposed change, will
reach its clamped upper limit value of 119% reactor thermal power.
Core flows at or above this level will result in the allowable value
reaching its current TS upper limit of 119%. As stated above, the
limiting value remains unchanged as part of this request.
The postulated failure mechanisms for the equipment are not
changed, nor are any
[[Page 4280]]
design limits exceeded. The proposed change will result in the need
to replace APRM equipment to allow operation in the extended power
to flow domain. These replacements will be evaluated per the
requirements of 10 CFR 50.59 as part of the Cooper Nuclear Station
(CNS) design change process to confirm no Unreviewed Safety Question
is created. Therefore, implementation of this proposed TS amendment
will not result in a significant increase in the probability or
consequences of an accident previously evaluated.
2. The proposed change will not create the possibility of a new
or different kind of accident than previously evaluated.
This proposed change does not modify the functional requirements
of the affected equipment, create any new system interfaces or
interactions, create any new process conditions that exceed design
limits, nor create any new system failure modes or sequences of
events that could lead to an accident.
The postulated failure mechanisms for the equipment are not
changed, nor are any design limits or acceptance criteria exceeded.
The proposed change will result in the need to replace APRM
equipment to allow operation in the extended power to flow domain.
These replacements will be evaluated per the requirements of 10 CFR
50.59 as part of the CNS design change process to confirm no
Unreviewed Safety Question is created. Therefore, the proposed
change does not create the possibility of a new or different kind of
accident from any accident previously evaluated.
3. The proposed change will not involve a significant reduction
in a margin of safety.
Change to the APRM Neutron Flux-High (Flow Biased) allowable
value is still limited by the 119% RTP value of TS. This value is
not credited in the safety analyses. In addition, the existing 120%
fixed scram allowable value (TS Table 3.3.1.1-1, Function 2.c) still
provides the same margin to the Analytical Limit of 123% RTP.
Analyses documented in Attachment 3 demonstrate that for operation
in the MELLL envelope or ICF region, adequate margin to design
limits is maintained and event acceptance criteria are met. Thus,
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.
Attorney for licensee: Mr. John R. McPhail, Nebraska Public Power
District, Post Office Box 499, Columbus, NE 68602-0499.
NRC Section Chief: Robert A. Gramm.
Nebraska Public Power District, Docket No. 50-298, Cooper Nuclear
Station, Nemaha County, Nebraska
Date of amendment request: December 22, 1999.
Description of amendment request: The proposed license amendment
requests Nuclear Regulatory Commission (NRC) review and approval of
revisions to the Cooper Nuclear Station (CNS) design basis accident
(DBA) radiological assessment calculational methodology used to
demonstrate compliance with the Exclusion Area Boundary and Low
Population Zone dose acceptance criteria specified in 10 CFR 100.11,
and the control room dose acceptance criteria discussed in General
Design Criteria (GDC) 19 of 10 CFR 50, Appendix A. The revisions entail
a complete rewrite of the radiological assessment calculational
methodology. The proposed changes do not revise the accident category,
general accident description, identification of accident cause,
frequency classification, starting conditions of the accident, accident
sequence of events, or system operation as described in the CNS Updated
Safety Analysis Report (USAR). The revised radiological assessment
calculational methodology does, however, involve changes to the
radiological consequence summary, fission product release from fuel
assumptions, fission product release to secondary containment
assumptions and conditions, fission product release to the environs
assumptions and initial conditions, and radiological effects summary
described in the CNS USAR. Additionally, the revised CNS DBA
radiological assessment calculational methodology incorporates the GDC
19 control room dose acceptance criteria determination as part of the
assessment. Previously the control room dose assessment was maintained
as separate design calculations and not included in the CNS USAR DBA
radiological assessment summaries.
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 not involve a significant increase in the probability or
consequences of an accident previously evaluated.
The proposed revisions to the Design Basis Accident (DBA)
radiological assessment calculational methodology do not affect the
accident initiators or precursors of accidents previously evaluated.
The proposed revisions to the methodology do not affect the existing
design, function or operation of systems, structures or components
in the facility. No new or different type of plant equipment is
installed by the revised radiological assessment calculational
methodology. Plant operating modes are not changed due to the
proposed revision to the DBA radiological assessment calculational
methodology. The proposed revisions are calculational in nature and
serve only to incorporate more recent site specific meteorological
data, reflect plant specific system operating parameters and design,
utilize more widely accepted accident assumptions for a facility of
Cooper Nuclear Station's vintage, incorporate the Technical
Information Document (TID-14844) source term to be consistent with
the accident assumptions used, update fuel parameter considerations
to include higher burnup fuel designs, and to utilize generic and
updated calculational and software methodologies to perform the
analysis. These revisions improve the consistency between the
accident dose calculation assumptions and improve the documentation
basis for each accident calculation. The revisions utilize
conservatively lower accident mitigation system filter efficiency
assumptions and incorporate plant specific accident mitigation
system operating parameter and design assumptions which result in a
calculated radiological consequence increase. Operation of accident
mitigation systems, structures and components is not altered by the
changes in accident mitigation assumptions. Due to the broad changes
in the calculational methodology and assumptions, and an increase in
the postulated accident source term, the calculated radiological
dose consequences of each design basis accident have changed and in
some cases increased. In each case, however, the calculated
radiological dose consequences satisfy the Exclusion Area Boundary
and Low Population Zone radiological dose acceptance criteria
specified in 10 CFR 100 and the control room dose acceptance
criteria discussed in General Design Criteria 19 (GDC 19) of 10 CFR
50, Appendix A. Therefore, the proposed revisions do not involve a
significant increase in the probability or consequences of an
accident previously evaluated.
2. Does not create the possibility for a new or different kind
of accident from any accident previously evaluated.
The proposed revisions to the DBA radiological assessment
calculational methodology do not change the existing design,
function or operation of systems, structures or components in the
facility. No new or different type of plant equipment is installed
by this change. There are no changes to existing design parameters
governing plant operation, plant operating modes, or changes in
system interfaces. No new types of accident initiators or precursors
are created by the proposed revision to the DBA radiological
assessment calculational methodology. The proposed revisions are
calculational in nature and serve only to incorporate more recent
site specific meteorological data, reflect plant specific system
operating parameters and design, utilize more widely accepted
accident assumptions for a facility of Cooper Nuclear Station's
vintage, incorporate the TID-14844 source term to be consistent with
the accident assumptions used, update fuel parameter considerations
to include higher burnup fuel designs, and to utilize generic and
updated calculational and software
[[Page 4281]]
methodologies to perform the analysis. These revisions improve the
consistency between the accident dose calculation assumptions and
improve the documentation basis for each accident calculation.
Therefore, the proposed change does not create the possibility of a
new or different kind of accident previously evaluated.
3. Does not create a significant reduction in the margin of
safety.
The proposed revisions to the DBA radiological assessment
calculational methodology do not involve a relaxation in the
criteria used to establish safety limits or a relaxation in the
limiting conditions for operation. The accident analysis sequence of
events remains unchanged. The proposed change will not result in any
challenges to plant equipment, fuel integrity, or the reactor
coolant system pressure boundary. The proposed revisions are
calculational in nature and serve only to incorporate more recent
site specific meteorological data, reflect plant specific system
operating parameters and design, utilize more widely accepted
accident assumptions for a facility of Cooper Nuclear Station's
vintage, incorporate the TID-14844 source term to be consistent with
the accident assumptions used, update fuel parameter considerations
to include higher burnup fuel designs, and to utilize generic and
updated calculational and software methodologies to perform the
analysis. These revisions improve the consistency between the
accident dose calculation assumptions and improve the documentation
basis for each accident calculation. 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.
Attorney for licensee: Mr. John R. McPhail, Nebraska Public Power
District, Post Office Box 499, Columbus, NE 68602-0499.
NRC Section Chief: Robert A. Gramm.
North Atlantic Energy Service Corporation, Docket No. 50-443, Seabrook
Station, Unit No. 1, Rockingham County, New Hampshire
Date of amendment request: November 19, 1999.
Description of amendment request: The licensee proposes to change
the Technical Specifications (TS) by relocating the specific
requirements of TS 6.4.3, ``Nuclear Safety Audit Review Committee
(NSARC),'' to the Quality Assurance Program located in the Updated
Final Safety Analysis 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 proposed change does not involve a significant increase
in the probability or consequences of an accident previously
evaluated.
The proposed change does not adversely affect accident
initiators or precursors nor alter the design assumptions,
conditions, configuration of the facility or the manner in which the
plant is operated. The proposed change does not alter or prevent the
ability of structures, systems, or components (SSCs) to perform
their intended function to mitigate the consequences of an
initiating event within the acceptance limits assumed in the Updated
Final Safety Analysis Report (UFSAR). The proposed change is
administrative in nature and does not decrease the effectiveness of
programmatic controls or the procedural details of assuring
operation of the facility in a safe manner.
The relocation of the Nuclear Safety Audit Review Committee
requirements from the Technical Specification to a new Appendix 17C
in UFSAR Chapter 17.2 does not alter the performance or frequency of
these activities. Future changes to the Quality Assurance Program
are subject to the 10 CFR 50.54(a) and 10 CFR 50.59 and change
processes.
The proposed change will not degrade the ability of systems,
structures and components important to safety to perform their
safety function. The proposed change will not change the response of
any system, structure or component important to safety as described
in the UFSAR. Since the plant response to an accident will not
change, there is no change in the potential for an increase in the
consequences of an accident previously analyzed. As such, the
proposed change does not involve a significant increase in the
probability or consequences of an accident previously evaluated.
2. The proposed change does not create the possibility of a new
or different kind of accident from any previously analyzed.
The proposed change does not alter the design assumptions,
conditions, configuration of the facility or the manner in which the
plant is operated. There are no changes to the source term,
containment isolation or radiological release assumptions used in
evaluating the radiological consequences in the Seabrook Station
UFSAR. Existing system and component redundancy is not being changed
by the proposed change. The proposed change has no adverse impact on
component or system interactions. The proposed change will not
adversely degrade the ability of systems, structures and components
important to safety to perform their safety function nor change the
response of any system, structure or component important to safety
as described in the UFSAR. The proposed change is administrative in
nature and does not change the level of programmatic controls and
procedural details of assuring operation of the facility in a safe
manner. The proposed changes involve the relocation of the
requirements of the Nuclear Safety Audit Review Committee from TS
6.4.3 to Updated Final Safety Analysis Report, Chapter 17.2,
``Quality Assurance Program'' in a new Appendix 17C. Future changes
to the Quality Assurance Program are subject to the 10 CFR 50.54(a)
and 10 CFR 50.59 and change processes.
Therefore, since there are no changes to the design assumptions,
conditions, configuration of the facility, or the manner in which
the plant is operated and surveilled, the proposed change does not
create the possibility of a new or different kind of accident from
any previously analyzed.
3. The proposed change does not involve a significant reduction
in a margin of safety.
The proposed changes involve the relocation of the requirements
of the Nuclear Safety Audit Review Committee from TS 6.4.3 to
Updated Final Safety Analysis Report, Chapter 17.2, ``Quality
Assurance Program'' in a new Appendix 17C. There is no adverse
impact on equipment design or operation and there are no changes
being made to the Technical Specification required safety limits or
safety system settings that would adversely affect plant safety. The
proposed change is administrative in nature and does not change the
level of programmatic controls and procedural details controls of
assuring operation of the facility in a safe manner.
Future changes to the Quality Assurance Program are subject to
the 10 CFR 50.54(a) and 10 CFR 50.59 change processes. Therefore,
relocation of the requirements contained in TS 6.4.3 to the Update
Final Safety Analysis Report 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.
Attorney for licensee: Lillian M. Cuoco, Esq., Senior Nuclear
Counsel, Northeast Utilities Service Company, P.O. Box 270, Hartford,
CT 06141-0270.
NRC Section Chief: James W. Clifford.
North Atlantic Energy Service Corporation, Docket No. 50-443, Seabrook
Station, Unit No. 1, Rockingham County, New Hampshire
Date of amendment request: November 29, 1999.
Description of amendment request: The licensee proposes to change
Technical Specification (TS) Surveillance Requirement (SR) 4.8.1.1.2f.,
to relocate sub requirement 4.8.1.1.2f.1 which requires inspection of
the emergency diesel generators (EDGs) on an 18-month cycle to be
subjected to an inspection in accordance with manufacturers
recommendations, to the Seabrook Station Technical Requirements Manual
(SSTRM).
Basis for proposed no significant hazards consideration
determination:
[[Page 4282]]
As required by 10 CFR 50.91(a), the licensee has provided its analysis
of the issue of no significant hazards consideration, which is
presented below:
1. The proposed change does not involve a significant increase
in the probability or consequences of an accident previously
evaluated.
The proposed change does not adversely affect accident
initiators or precursors nor alter the design assumptions,
conditions, and configuration of the facility or the manner in which
the plant is operated. The proposed change does not alter or prevent
the ability of structures, systems and components (SSCs) to perform
their intended function to mitigate the consequences of an
initiating event within the acceptance limits assumed in the Updated
Final Safety Analysis Report (UFSAR).
Performance of EDG inspection activities based on condition-
based maintenance rather than time-directed maintenance will neither
exacerbate nor significantly increase the probability or
consequences of an accident previously evaluated in the Seabrook
Station UFSAR. North Atlantic has extensive experience and expertise
in operating and maintaining the EDGs to determine the appropriate
maintenance activities for demonstrating operability of the EDGs.
North Atlantic will continue to use, in conjunction with
manufacturer recommendations, prudent engineering judgment when
conducting testing, preventive and corrective maintenance activities
on the EDGs. In addition, the other surveillance testing required by
SR 4.8.1.1.2f would continue to ensure that the EDGs are capable of
performing their safety function.
Throughout the first six fuel cycles, overall EDG condition has
steadily improved with the use of improved design, utilization of
better condition monitoring tools and procedures and the reduction
of intrusive preventative maintenance tasks made possible by the
improved on-line condition monitoring methods. These improvements
resolved problems that were recognized during the early years of EDG
operation.
North Atlantic has implemented the Maintenance Rule Program in
accordance with the provisions of 10 CFR 50.65, Regulatory Guide
(RG) 1.160, and NUMARC 93-01, ``Industry Guide for Monitoring the
Effectiveness of Maintenance at Nuclear Power Plants.''
North Atlantic's maintenance rule program establishes specific
performance criteria for SSCs. Reliability and unavailability
performance criteria have been assigned to risk significant and
standby safety-related non-risk significant SSCs. Other in-scope
SSCs have been assigned appropriate reliability and/or plant level
performance criteria. SSCs that are determined to not meet the
established performance criteria are designated as (a)(1) and are
subject to action plans, goal setting, and goal monitoring.
Performance of (a)(1) SSCs is compared to the established goals.
When it is determined that the performance goals have been achieved,
a SSC may be returned to the normal performance monitoring (a)(2)
status.
With regard to the EDGs, these components and the associated
support systems are risk significant and standby safety-related. The
experience to date, applying the Maintenance Rule Program to the
EDGs, has proven to be positive. Risk informed decision-making
concerning the benefits of maintenance and time out of service has
maintained reliable EDGs with unavailability consistent with the
assumptions in the Seabrook Station Probabilistic Risk Assessment
(PRA).
Furthermore, Operations Department personnel perform daily,
weekly, biweekly, monthly and quarterly walkdowns and inspections of
various items as well as the monthly surveillance run on each
diesel. These inspections, combined with system control panel
alarms, engine oil sampling and on-line monitoring of engine
vibration and running performance (cylinder firing, fuel delivery
and exhaust temperatures), enable expeditious response to a
developing degraded condition and provide a mechanism for failure
identification prior to performance of the refueling interval
surveillances.
Based on the reviews of the surveillance tests, inspections and
maintenance activities, it is concluded that there is no significant
impact on the reliability of the EDGs and, therefore, there is no
significant increase in the probability or consequences of any
previously analyzed accident.
2. Create the possibility of a new or different kind of accident
from any accident previously evaluated.
The proposed change does not alter the design assumptions,
conditions, and configuration of the facility or the manner in which
the plant is operated. There are no changes to the source term,
containment isolation or radiological release assumptions used in
evaluating the radiological consequences in the Seabrook Station
UFSAR. Existing system and component redundancy is not being changed
by the proposed change. The proposed change has no adverse affect on
component or system interactions. Therefore, since there are no
changes to the design assumptions, conditions, configuration of the
facility, or the manner in which the plant is operated, the proposed
change does not create the possibility of a new or different kind of
accident from any previously analyzed.
3. Involve a significant reduction in a margin of safety.
The proposed change does not adversely affect equipment design
or operation and there are no changes being made to the Technical
Specification required safety limits or safety system settings that
would adversely affect plant safety. The proposed change does not
adversely affect the EDG's ability to ensure that sufficient power
is available to supply the safety related equipment required for: 1)
the safe shutdown of the facility, and 2) the mitigation and control
of accident conditions within the facility.
Surveillance testing of the EDGs during normal plant operation
provides assurance that the proposed change will not adversely
affect the reliability of the EDGs. North Atlantic will continue to
use, in conjunction with manufacturer's recommendations, prudent
engineering judgment when conducting testing, preventive, and
corrective maintenance activities on the EDGs. In addition, the
other surveillance testing required by SR 4.8.1.1.2f would continue
to ensure that the EDGs are capable of performing their safety
function. Thus, it is concluded that the EDGs would continue to be
available upon demand to mitigate the consequences of an accident
and, therefore, there is no 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.
Attorney for licensee: Lillian M. Cuoco, Esq., Senior Nuclear
Counsel, Northeast Utilities.Service Company, P.O. Box 270, Hartford,
CT 06141-0270.
NRC Section Chief: James W. Clifford.
North Atlantic Energy Service Corporation, Docket No. 50-443, Seabrook
Station, Unit No. 1, Rockingham County, New Hampshire
Date of amendment request: December 3, 1999.
Description of amendment request: The licensee proposes to change
the technical specifications (TS) by incorporating reference to the
American Society for Testing and Materials (ASTM) Standard D3803-1989,
``Standard Test Method for Nuclear-Grade Activated Charcoal,'' as the
test protocol for charcoal filter laboratory testing. In addition,
there will be a change to Surveillance Requirements 4.7.6.1d.5) and
4.9.12d.4) specifying a minimum required heater output based on design
rated voltage.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. The proposed changes do not involve a significant increase in
the probability or consequences of an accident previously evaluated.
The proposed changes do not affect accident initiators or
precursors and do not alter the design assumptions, conditions or
configuration of the facility or the manner in which the plant is
operated and maintained. The proposed changes do not alter or
prevent the ability of structures, systems, or components (SSCs) to
perform their intended function to mitigate the consequences of an
initiating event within the acceptance limits assumed in the Updated
Final Safety Analysis Report (UFSAR).
The proposed changes modify the Technical Specifications to
reference
[[Page 4283]]
appropriate test parameters for performing laboratory testing of
nuclear-grade charcoal in ESF [engineered safety feature] filtration
systems in accordance with ASTM D3803-89. The testing methodology
associated with ASTM D3803-89 provides more stringent requirements
than what is currently employed. These more stringent requirements
will not result in operations that will increase the probability of
initiating an analyzed event and do not alter assumptions relative
to mitigation of an accident or transient event. The more
restrictive requirements continue to ensure process variables,
structures, systems, and components are maintained consistent with
the safety analyses and licensing basis.
The proposed change associated with verification of heater
capacity dissipation by specifying a minimum required output based
on design rated voltage does not affect continued operability of the
heater. Stipulating the design rated voltage ensures the heater(s)
remains capable of performing its safety function. Specifying an
upper kW range band is restrictive and has been determined to be
unnecessary. There is no safety concern with the heaters operating
at a higher kW output. Operating at a higher kW output improves
dehumidification. Should maximum operating bus voltage conditions be
experienced it does not pose a fire hazard or dry-out concern for
the charcoal filters.
There are no changes to previous accident analyses. The
radiological consequences associated with these analyses remain
unchanged. Therefore, the proposed changes do not involve a
significant increase in the probability or consequences of an
accident previously evaluated.
2. The proposed changes do not create the possibility of a new
or different kind of accident from any previously analyzed.
The proposed changes do not alter the design assumptions,
conditions or configuration of the facility or the manner in which
the plant is operated and maintained. The proposed changes have no
impact on component or system interactions.
The proposed changes modify the Technical Specifications to
reference appropriate test parameters for performing laboratory
testing of nuclear-grade charcoal in ESF filtration systems in
accordance with ASTM D3803-89. The changes do impose different, more
conservative testing requirements, on the ESF filtration systems
charcoal samples. However, there is no alteration in the methods
employed to obtain the charcoal sample and testing is performed
offsite.
The proposed change associated with verification of heater
capacity dissipation by specifying a minimum required output based
on design rated voltage does not affect continued operability of the
heater. The design function of the heater for humidity control
remains unchanged. Deletion of the upper kW range does not pose a
fire or dry-out concern for the charcoal filters.
These changes are consistent with the safety analyses and
licensing basis. The proposed changes do not introduce any new modes
of plant operation, or alter any operational setpoints.
Since the proposed changes do not involve the physical
alteration of SSCs (i.e., no new or different type of equipment to
be installed) or changes in the methods governing normal plant
operation, it is concluded that the proposed changes do not create
the possibility of a new or different kind of accident from any
previously analyzed.
3. The proposed changes do not involve a significant reduction
in a margin of safety.
There is no impact on equipment design or operation and there
are no changes being made to the Technical Specification required
safety limits or safety system settings that would adversely affect
plant safety. The proposed changes modify the Technical
Specifications to reference appropriate test parameters for
performing laboratory testing of nuclear-grade charcoal in ESF
filtration systems in accordance with ASTM D3803-89. The imposition
of the more conservative charcoal filter testing requirements
associated with ASTM D3803-89 has no significant impact on a margin
of safety. The conservative nature of ASTM D3803-89 is by
definition, providing additional restrictions to enhance plant
safety.
The proposed change associated with specifying a minimum
required heater output based on design rated voltage does not reduce
the ability of the heater to provide the minimum required kW output
for humidity control. Deletion of the upper kW range does not pose a
fire or dry-out concern for the charcoal filters.
The proposed changes maintain requirements within the safety
analysis and licensing basis. Therefore, the proposed changes do not
involve a significant reduction in any margin of safety.
The NRC staff has reviewed the licensee's analysis, and based on
this review, it appears that the three standards of 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: Lillian M. Cuoco, Esq., Senior Nuclear
Counsel, Northeast Utilities Service Company, P.O. Box 270, Hartford,
CT 06141-0270.
NRC Section Chief: James W. Clifford.
Northeast Nuclear Energy Company, et al., Docket No. 50-336, Millstone
Nuclear Power Station, Unit No. 2, New London County, Connecticut
Date of amendment request: September 7, 1999.
Description of amendment request: The proposed changes affect
Technical Specification 3/4.7.8, ``Plant Systems, Snubbers,'' by
removing the current special exception which precludes applying the
eighteen month functional testing surveillance to the Steam Generator
Hydraulic Snubbers.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. Involve a significant increase in the probability or
consequences of an accident previously evaluated.
The snubbers provide a restraint function to mitigate the
consequences of a Main Steam Line Break (MSLB) or to limit seismic
induced movements of the steam generators so as to protect the
attached Reactor Coolant System (RCS) piping and therefore prevent
the initiation of a Loss of Coolant Accident (LOCA).
While the proposed surveillance changes will extend the time
period required for 100% inspection of all steam generator snubbers
and also the actual service life of the snubber seals, the testing
of samples at reduced intervals will actually provide a more
reliable and timely indication of snubber functionality and provide
increased assurance that generic concerns associated with this
snubber set will be detected prior to any failure. The proposed
surveillance requirements are the same as currently used for the
balance of Millstone Unit No. 2 hydraulic snubbers. Given the
complete similarity of design and operation for these components,
the sampling approach is well suited for these snubbers. Given the
general acceptance of a 10% sampling approach in the general snubber
population, its use here for this homogenous set of components is
fully justified. In addition to the 10% sample that will be
functionally tested on an eighteen month interval, a concurrent 100%
visual inspection is conducted during each test period, providing
added assurance that no seal failures will go undetected for any
significant period. This visual inspection program is unchanged from
the existing surveillance program as currently documented in the
Millstone Unit No. 2 Technical Specification. The anticipated
reliability under the new surveillance frequency and testing methods
proposed for the steam generator snubbers will not affect the
probability of occurrence of a LOCA or a MSLB as the snubbers'
ability to perform their function will prevent over stressing of
either the Main Steam (MS) or RCS piping attached to the steam
generators. Furthermore, the anticipated reliability under the new
surveillance frequency and testing methods proposed for the steam
generator snubbers will ensure that the existing evaluated
consequences for these accidents will not be increased. Therefore,
these changes will not significantly increase the probability or
consequences of an accident previously evaluated.
The proposed change to Bases Section 3/4.7.8 will delete the
text associated with the current exception taken for steam generator
snubbers. This change will make the discussion in the Bases
consistent with the proposed Technical Specification changes.
Therefore, this change will not significantly increase the
probability or consequences of an accident previously evaluated.
The proposed changes do not alter how any structure, system, or
component functions. There will be no effect on equipment important
to safety. The proposed changes have no effect on any of the design
basis accidents previously evaluated. Therefore, this License
Amendment Request does not impact the probability of an
[[Page 4284]]
accident previously evaluated, nor does it 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 only accidents possible due to failure of the steam
generator snubbers to operate properly is increased stresses on both
the MS and RCS piping attached to the steam generator due to either
additional constraint in the case of premature lockup, or lack of
proper constraint in the case of failure to lock-up under dynamic
loading. Since the worst case scenario of such a failure would be
the initiation of a LOCA, which is currently evaluated in the SAR
[safety analysis report], there is no possibility of a new or
different kind of accident from any accident previously evaluated.
The proposed changes do not alter the plant configuration (no
new or different type of equipment will be installed) or require any
new or unusual operator actions. They do not alter the way any
structure, system, or component functions and do not alter the
manner in which the plant is operated. The proposed changes do not
introduce any new failure modes. Therefore, the proposed changes
will not create the possibility of a new or different kind of
accident from any accident previously evaluated.
3. Involve a significant reduction in a margin of safety.
The proposed changes will allow use of the preferred approach to
snubber surveillance which is in effect for the balance of Millstone
Unit No. 2 snubbers. The steam generator snubbers have been
previously exempt from the standard approach to snubber surveillance
due to the difficulty previously encountered in testing these large
and inaccessible components. Given the reliability of these snubbers
is not expected to change in that the same requirements as for all
other hydraulic snubbers will now consistently be met, there is no
significant reduction in a margin of safety. The proposed changes
will not alter any of the assumptions used in the accident analysis,
nor will they cause any safety system parameters to exceed their
acceptance limit. The proposed changes will not affect any
operability requirements for equipment important to plant safety.
Therefore, the proposed changes 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.
Attorney for licensee: Lillian M. Cuoco, Esq., Senior Nuclear
Counsel, Northeast Utilities Service Company, P.O. Box 270, Hartford,
Connecticut.
NRC Section Chief: James W. Clifford.
Northeast Nuclear Energy Company, et al., Docket No. 50-336, Millstone
Nuclear Power Station, Unit No. 2, New London County, Connecticut
Date of amendment request: November 23, 1999.
Description of amendment request: The proposed changes will update
the list of documents describing the analytical methods used to
determine the core operating limits, specified in Technical
Specification 6.9.1.8b.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. Involve a significant increase in the probability or
consequences of an accident previously evaluated.
The proposed change in document 1 of Technical Specification
6.9.1.8b is made to provide the most recent, Nuclear Regulatory
Commission (NRC) approved, methodology description and benchmarking
results of the reactor analysis system used in the core neutronics
analysis of cycle 14 and beyond. This change has no impact on plant
equipment operation. Since the change only affects the neutronics
analysis of the core, it cannot affect the likelihood or
consequences of accidents. Therefore, this change will not
significantly increase the probability or consequences of an
accident previously evaluated.
The proposed change in document 8 of Technical Specification
6.9.1.8b is made to include the most recent, NRC approved, Emergency
Core Cooling System (ECCS) model used in Large Break Loss of Coolant
Accident (LBLOCA) applications. This model contains resolution of
the deficiencies reported under 10 CFR 50.46(a) in a letter dated
May 20, 1999. The use of the revised methodology also constitutes an
improvement over the previous methodology. Therefore, this change
will not significantly increase the probability or consequences of
an accident previously evaluated.
The proposed changes in document 4 of Technical Specification
6.9.1.8b are administrative in nature. Therefore, these changes will
not significantly increase the probability or consequences of an
accident previously evaluated.
2. Create the possibility of a new or different kind of accident
from any accident previously evaluated.
The proposed change in document 1 of Technical Specification
6.9.1.8b is made to provide the most recent, NRC approved,
methodology description and benchmarking results of the reactor
analysis system used in the neutronics analysis of cycle 14 and
beyond. The proposed change in document 1 of Technical Specification
6.9.1.8b will not alter the plant configuration (no new or different
type of equipment will be installed) or require any new or unusual
operator actions. It does not alter the way any structure, system,
or component functions and does not alter the manner in which the
plant is operated.
The proposed change in the documents in number 8 of Technical
Specification 6.9.1.8b is made to include the most recent, NRC
approved, ECCS model used in LBLOCA applications. The proposed
change in document 8 of Technical Specification 6.9.1.8b will not
alter the plant configuration (no new or different type of equipment
will be installed) or require any new or unusual operator actions.
It does not alter the way any structure, system, or component
functions and does not alter the manner in which the plant is
operated.
The proposed changes in document 4 of Technical Specification
6.9.1.8b are administrative in nature. These changes do not alter
the way any structure, system, or component functions and do not
alter the manner in which the plant is operated.
These changes do not introduce any new failure modes. Therefore,
the proposed changes will not create the possibility of a new or
different kind of accident from any accident previously evaluated.
3. Involve a significant reduction in a margin of safety.
The proposed change in document 1 of Technical Specification
6.9.1.8b is made to provide the most recent, NRC approved,
methodology description and benchmarking results of the reactor
analysis system used in the neutronics analysis of cycle 14 and
beyond. It has no impact on plant equipment operation. The proposed
change in document 8 of Technical Specification 6.9.1.8b is made to
include the most recent, NRC approved, ECCS model used in LBLOCA
applications. This model contains resolution of the deficiencies
reported under 10 CFR 50.46(a) in a letter dated May 20, 1999. The
use of the revised methodology still provides a conservative
simulation of the LBLOCA and conservative core neutronics analysis.
The use of the revised methodology also constitutes an improvement
over the previous methodology. The new documents will clearly
identify the approved Siemens Topical Reports applicable to
Millstone Unit No. 2 and will ensure that methodology changes will
be identified and submitted to the NRC for approval, as required.
The proposed changes in document 4 of Technical Specification
6.9.1.8b are administrative in nature. Therefore, the proposed
changes 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.
Attorney for licensee: Lillian M. Cuoco, Esq., Senior Nuclear
Counsel, Northeast Utilities Service Company, P.O. Box 270, Hartford,
Connecticut.
NRC Section Chief: James W. Clifford.
[[Page 4285]]
Northeast Nuclear Energy Company, et al., Docket No. 50-336, Millstone
Nuclear Power Station, Unit No. 2, New London County, Connecticut
Date of amendment request: December 6, 1999.
Description of amendment request: The proposed changes will modify
the Technical Specification (TS) surveillance requirements associated
with ensuring a limited number of charging and high pressure safety
injection pumps are capable of injecting into the Reactor Coolant
System when the plant is shutdown. In addition, the TS Bases will be
modified to address these changes.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. Involve a significant increase in the probability or
consequences of an accident previously evaluated.
The proposed modifications to the surveillance requirements
(SRs) associated with Technical Specifications 3.1.2.3, 3.1.2.4, and
3.4.9.3 will remove information that specifies the methods to be
used to perform the associated SRs. These SRs verify the maximum
number of charging and high pressure safety injection (HPSI) pumps
capable of injecting into the RCS [Reactor Coolant System] when the
plant is shut down. This information will be transferred to the
associated Bases. Additional methods associated with the charging
pumps, which are technically equivalent to the current method, will
be included in the Bases change. This will not change the
requirement to verify that the associated pumps are not capable of
injecting into the RCS when the plant is shut down.
The proposed changes to the Technical Specifications and Bases
will have no adverse effect on plant operation, or the availability
or operation of any accident mitigation equipment. The plant
response to the design basis accidents will not change. In addition,
the proposed changes can not cause an accident. Therefore, there
will be no significant increase in the probability or consequences
of an accident previously evaluated.
2. Create the possibility of a new or different kind of accident
from any accident previously evaluated.
The proposed Technical Specification and Bases changes will not
alter the plant configuration (no new or different type of equipment
will be installed) or require any new or unusual operator actions.
They do not alter the way any structure, system, or component
functions and do not significantly alter the manner in which the
plant is operated. The proposed changes do not introduce any new
failure modes. Also, the response of the plant and the operators
following these accidents is unaffected by the changes. Therefore,
the proposed changes will not create the possibility of a new or
different kind of accident from any accident previously evaluated.
3. Involve a significant reduction in a margin of safety.
The proposed modifications to the surveillance requirements
associated with Technical Specifications 3.1.2.3, 3.1.2.4, and
3.4.9.3 will remove information that specifies the methods to be
used to perform the associated surveillance requirements. This will
not change the requirement to verify that the associated pumps are
not capable of injecting into the RCS when the plant is shut down.
The proposed changes to the Technical Specifications and Bases
will have no adverse effect on plant operation or equipment
important to safety. The plant response to the design basis
accidents will not change and the accident mitigation equipment will
continue to function as assumed in the design basis accident
analysis. Therefore, there will be no 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.
Attorney for licensee: Lillian M. Cuoco, Esq., Senior Nuclear
Counsel, Northeast Utilities Service Company, P.O. Box 270, Hartford,
Connecticut.
NRC Section Chief: James W. Clifford.
Northeast Nuclear Energy Company, et al., Docket No. 50-336, Millstone
Nuclear Power Station, Unit No. 2, New London County, Connecticut
Date of amendment request: December 7, 1999.
Description of amendment request: The proposed changes to the
Technical Specifications (TSs) are associated with the action
requirement to suspend positive reactivity additions. These changes
will remove the action requirement to suspend positive reactivity
additions from TS 3.4.2.1, ``Reactor Coolant System--Safety Valves,''
3.4.2.2, ``Reactor Coolant System--Safety Valves,'' and 3.7.6.1,
``Plant Systems--Control Room Emergency Ventilation System,'' and
provide guidance in the Bases for other TSs that require the suspension
of positive reactivity addition.
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.
Technical Specifications 3.4.2.1 and 3.4.2.2
The proposed changes to Technical Specifications 3.4.2.1 and
3.4.2.2, which address the pressurizer code safety valves in Modes 1
through 4, will combine these two specifications into one Technical
Specification, 3.4.2. The slight reduction in the Mode of
Applicability for the new Technical Specification, to be consistent
with the Mode of Applicability for Technical Specification 3.4.9.3,
which addresses the Low Temperature Overpressure Protection (LTOP)
System, is too small to result in a change in plant operations. The
LCO [limiting condition for operation] for the pressurizer code
safety valves in Mode 4 with all Reactor Coolant System (RCS) cold
leg temperatures > 275 deg.F will be expanded to require all
pressurizer code safety valves to be operable, instead of at least
one pressurizer code safety valve. This more restrictive change will
require additional accident mitigation equipment to be operable. The
proposed action requirements for plant operation in Modes 1, 2, and
3 have been expanded to require the plant to be in Mode 3 within 6
hours and in Mode 4 within the following 6 hours, instead of just
Mode 4 within 12 hours. In addition, the action requirements will be
modified to address 2 inoperable pressurizer code safety valves. An
entry into Technical Specification 3.0.3 will no longer be necessary
if both pressurizer code safety valves are inoperable. In addition,
the proposed action requirements are more restrictive than the
action requirements of Technical Specification 3.0.3. The proposed
action requirements for Mode 4 with all RCS cold leg temperatures >
275 deg.F are different. The new Mode 4 action requirements will
direct the plant to be cooled down to the applicability of Technical
Specification 3.4.9.3, which will require the LTOP System to be
placed in service to provide RCS overpressure protection. The
proposed action requirements will ensure that the plant is placed in
a condition where sufficient accident mitigation equipment will be
available.
The proposed Technical Specification, 3.4.2, will ensure the RCS
has adequate overpressure protection when operating above 275
deg.F. If the pressurizer code safety valves are not operable, the
proposed Technical Specification will require a plant shutdown that
will place the plant within the capability of the LTOP System to
provide RCS overpressure protection. The proposed changes will have
no adverse effect on plant operation, or the availability or
operation of any accident mitigation equipment. The plant response
to the design basis accidents will not change. In addition, the
proposed changes can not cause an accident. Therefore, there will be
no significant increase in the probability or consequences of an
accident previously evaluated.
Technical Specification 3.7.6.1
The proposed change to Technical Specification 3.7.6.1 will
remove the requirement to suspend positive reactivity additions if
both control room ventilation trains are inoperable in Modes 5 and
6. The Control Room Ventilation System is required
[[Page 4286]]
to be operable in Modes 5 and 6 to protect the control room
operators from an event that results in a rapid release of
radioactivity, such as a fuel handling accident. In Modes 5 and 6,
the positive reactivity addition methods of concern are boron
dilution, RCS cooldown (negative isothermal temperature
coefficient), and control rod withdrawal. Positive reactivity
additions associated with fuel handling are already addressed by the
additional action requirement in this specification to suspend core
alterations. Control rod withdrawal is prohibited by Technical
Specification 3.1.3.7, unless the RCS boron concentration is greater
than or equal to the refueling boron concentration of Technical
Specification 3.9.1. If the RCS is borated to the refueling
concentration, sufficient negative reactivity has been added to
compensate for the positive reactivity addition associated with
control rod withdrawal in Modes 5 and 6. Therefore, only boron
dilution and RCS temperature changes are of concern. However, both
of these methods will result in slow changes to core reactivity in
Modes 5 and 6, and since adequate shutdown margin (SDM) will have
been established prior to entering Mode 5 or 6 (Technical
Specifications 3.1.1.2 and 3.9.1), neither method will result in a
rapid release of radioactivity. Therefore, the requirement to
suspend positive reactivity additions is not necessary for the
protection of the control room operators.
The proposed change will have no adverse effect on plant
operation, or the availability or operation of any accident
mitigation equipment. The plant response to the design basis
accidents will not change. In addition, the proposed change can not
cause an accident. Therefore, there will be no significant increase
in the probability or consequences of an accident previously
evaluated.
2. Create the possibility of a new or different kind of accident
from any accident previously evaluated.
The proposed Technical Specification will not alter the plant
configuration (no new or different type of equipment will be
installed) or require any new or unusual operator actions. They do
not alter the way any structure, system, or component functions and
do not significantly alter the manner in which the plant is
operated. The proposed changes do not introduce any new failure
modes. Also, the response of the plant and the operators following
these accidents is unaffected by the changes. Therefore, the
proposed changes will not create the possibility of a new or
different kind of accident from any accident previously evaluated.
3. Involve a significant reduction in a margin of safety.
The proposed change to combine Technical Specifications 3.4.2.1
and 3.4.2.2 into a new Technical Specification, 3.4.2, will result
in a slight reduction in the Mode of Applicability for the new
Technical Specification, will require both pressurizer code safety
valves to be operable in Mode 4 with all RCS cold leg temperatures >
275 deg.F, will modify the action requirements in Modes 1, 2, and 3
to add a requirement to be in Mode 3 within 6 hours and to address
two inoperable pressurizer code safety valves, and will provide
different action requirements for Mode 4 with all RCS cold leg
temperatures > 275 deg.F. The reduction in Mode of Applicability is
too small to adversely impact plant operations. Requiring both
pressurizer code safety valves to be operable in Mode 4 with all RCS
cold leg temperatures > 275 deg.F will provide additional accident
mitigation equipment. The modified action requirement to be in Mode
3 within 6 hours will not change the requirement to be in Mode 4
within 12 hours. The action requirements added to address two
inoperable pressurizer code safety valves are more restrictive than
the action requirements of Technical Specification 3.0.3. The new
Mode 4 action requirements will direct the plant to be cooled down
to the applicability of Technical Specification 3.4.9.3, which will
require the LTOP System to be placed in service to provide RCS
overpressure protection. The proposed action requirements will
ensure that the plant is placed in a condition where sufficient
accident mitigation equipment will be available.
The proposed change to Technical Specification 3.7.6.1 will
remove the requirement to suspend positive reactivity additions if
both control room ventilation trains are inoperable in Modes 5 and
6. The Control Room Ventilation System is required to be operable in
Modes 5 and 6 to protect the control room operators from an event
that results in a rapid release of radioactivity, such as a fuel
handling accident. The proposed change will only impact slow methods
to change core reactivity, such as boron dilution and RCS
temperature changes. Therefore, the action requirement to suspend
positive reactivity additions is not necessary for the protection of
the control room operators.
The proposed changes will have no adverse effect on plant
operation or equipment important to safety. The plant response to
the design basis accidents will not change and the accident
mitigation equipment will continue to function as assumed in the
design basis accident analysis. Therefore, there will be no
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.
Attorney for licensee: Lillian M. Cuoco, Esq., Senior Nuclear
Counsel, Northeast Utilities Service Company, P.O. Box 270, Hartford,
Connecticut.
NRC Section Chief: James W. Clifford.
Northeast Nuclear Energy Company, et al., Docket Nos. 50-336 and 50-
423, Millstone Nuclear Power Station, Unit Nos. 2 and 3, New London
County, Connecticut
Date of amendment request: November 23, 1999.
Description of amendment request: The proposed change affects
Technical Specification 4.0.5, ``Limiting Conditions for Operation and
Surveillance Requirements'' by adding a biennial or 2-year surveillance
interval and incorporating a required frequency for performing
inservice testing activities of once per 731 days.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. Involve a significant increase in the probability or
consequences of an accident previously evaluated.
The proposed change amends Technical Specification Section
4.0.5.b by adding a biennial or 2 year surveillance to the existing
list. This surveillance interval is included as part of the current
Millstone Unit Nos. 2 and 3 Inservice Test (IST) surveillance
program. Inclusion of this surveillance interval in the facility
Technical Specifications clarifies the applicability of this
surveillance interval and affords operational flexibility in the
event a surveillance cannot be completed within the required
interval.
The proposed change will have no adverse effect on plant
operation, or the availability or operation of any accident
mitigation equipment. The plant response to the design basis
accidents will not change. In addition, the proposed change can not
cause an accident. Therefore, there will be no significant increase
in the probability or consequences of an accident previously
evaluated.
2. Create the possibility of a new or different kind of accident
from any accident previously evaluated.
The biennial surveillance relates to performing inservice
testing of plant components. The possibility of a new or different
kind of accident from any accident previously evaluated is not
created because the proposed Technical Specification change does not
introduce a new mode of plant operations and does not involve
physical modifications to the plant. Therefore, the proposed change
will not create the possibility of a new or different kind of
accident from any accident previously evaluated.
3. Involve a significant reduction in a margin of safety.
There is no impact on the margin of safety as defined in the
Technical Specifications. Performance of surveillance tests at
regular intervals provides assurance of reliability and availability
of accident mitigating equipment. The Technical Specifications
provide the required frequency for performing surveillance testing.
Adding a new surveillance frequency to the Technical Specifications
will provide consistent yet acceptable flexibility in scheduling
surveillance tests and provide additional assurance that testing
will be performed in a timely manner.
The proposed change will have no adverse effect on plant
operation or equipment
[[Page 4287]]
important to safety. The plant response to the design basis
accidents will not change and the accident mitigation equipment will
continue to function as assumed in the design basis accident
analysis. Therefore, there will be no 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.
Attorney for licensee: Lillian M. Cuoco, Esq., Senior Nuclear
Counsel, Northeast Utilities Service Company, P.O. Box 270, Hartford,
Connecticut.
NRC Section Chief: James W. Clifford.
Northeast Nuclear Energy Company, et al., Docket No. 50-423, Millstone
Nuclear Power Station, Unit No. 3, New London County, Connecticut
Date of amendment request: November 29, 1999.
Description of amendment request: The requested changes would
revise Technical Specification (TS) 3/4.6.6, ``Supplementary Leak
Collection and Release System,'' (SLCRS), TS 3/4.7.7, ``Control Room
Emergency Ventilation System,'' (CREVS), TS 3/4.7.9, ``Auxiliary
Building Filter System,'' (ABFS), and 3/4.9.12, ``Fuel Building Exhaust
System,'' (FBES), in response to Generic Letter (GL) 99-02,
``Laboratory Testing of Nuclear-Grade Activated Charcoal.'' The
requested changes require testing of nuclear-grade activated charcoal
to be conducted in accordance with American Society for Testing
Materials (ASTM) D3803-1989, ``Standard Test Method for Nuclear-Grade
Activated Carbon,'' as recommended by GL 99-02.
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:
NNECO [Northeast Nuclear Energy Company] has reviewed the
proposed revision in accordance with 10 CFR 50.92 and has concluded
that the revision does not involve any Significant Hazards
Consideration (SHC). The basis for this conclusion is that the three
criteria of 10 CFR 50.92(c) are not satisfied. The proposed TS
revision does not involve an SHC because the revision would not:
1. Involve a significant increase in the probability or
consequences of an accident previously evaluated.
The proposed change modifies the TS to reference ASTM D3803-
[19]89 for performing laboratory testing of nuclear-grade charcoal
in ESF [Engineered Safeguards Features] filtration systems. The
testing methodology associated with ASTM D3803-[19]89 provides more
stringent requirements than what is currently employed. These more
stringent requirements, along with a factor of safety of greater
than or equal to two in regards to the charcoal efficiency assumed
in the design bases dose analysis will not result in operations that
will increase the probability of initiating an analyzed event and do
not alter assumptions relative to mitigation of an accident or
transient event. The more restrictive requirements continue to
ensure process variables, structures, systems, and components are
maintained consistent with the safety analyses and licensing basis.
There are no related modifications to any systems. The proposed
change does not affect procedures governing plant operations.
Therefore there is no significant increase in the probability [or
consequences] of occurrence of a previously evaluated accident.
2. Create the possibility of a new or different kind of accident
from any accident previously evaluated.
The proposed change modifies the TS to reference ASTM D3803-
[19]89 for performing laboratory testing of nuclear-grade charcoal
in ESF filtration systems. The proposed change does not involve the
physical alteration of the plant (no new or different type of
equipment will be installed) or changes in the methods governing
normal plant operation. This change does impose different, more
conservative testing requirements on the ESF filtration system
charcoal samples. However there is no alteration in the methods
employed to obtain the charcoal sample and testing is performed
offsite. These changes are consistent with the safety analyses and
licensing basis. Furthermore, the proposed changes do not introduce
any new modes of plant operation, or alter any operational
setpoints. Thus the possibility of a new or different kind of
accident from any previously evaluated is not created.
3. Involve a significant reduction in the margin of safety.
The proposed change modifies the TS to reference ASTM D3803-
[19]89 for performing laboratory testing of nuclear-grade charcoal
in ESF filtration systems. The imposition of the more conservative
charcoal filter testing requirements associated with ASTM D3803-
[19]89 along with a factor of safety of greater than or equal to
two, in regards to the charcoal efficiency assumed in the design
bases dose analysis has no impact on, nor decreases the margin of
plant safety. The conservative nature of ASTM D3803-[19]89 is by
definition, providing additional restrictions to enhance plant
safety. This change maintains requirements within the safety
analysis and licensing basis. Therefore, there will be no
significant reduction in the margin of safety as defined in the
Bases for the TS affected by the proposed change.
As described above this TSCR [Technical Specification Change
Request] does not impact the probability of an accident previously
evaluated, does not involve a significant increase in the
consequences of an accident previously evaluated, does not create
the possibility of a new or different kind of accident from any
accident previously evaluated, and does not result in a significant
reduction in a margin of safety. Therefore, NNECO has concluded that
the proposed changes do not involve an SHC.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: Lillian M. Cuoco, Esq., Senior Nuclear
Counsel, Northeast Utilities Service Company, P.O. Box 270, Hartford,
Connecticut.
NRC Section Chief: James W. Clifford.
PECO Energy Company, Docket Nos. 50-352 and 50-353, Limerick Generating
Station, (LGS) Units 1 and 2, Montgomery County, Pennsylvania
Date of amendment request: November 5, 1999.
Description of amendment request: The proposed changes will revise
LGS Technical Specifications (TSs) to incorporate revised testing and
acceptance criteria for the performance of laboratory analysis of
safety-related nuclear-grade activated charcoal in response to Generic
Letter (GL) 99-02. ``Laboratory Testing of Nuclear-Grade Activated
Charcoal,'' dated June 3, 1999. In addition, minor editorial changes
are being proposed for wording consistency and to correct a
typographical error.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. The proposed TS changes do not involve a significant increase
in the probability or consequences of an accident previously
evaluated.
Changing the methodology for the performance of the laboratory
testing of nuclear-grade activated charcoal samples from Reg.
[Regulatory] Guide 1.52 to ASTM D3803-1989 in accordance with
Generic Letter 99-02, and establishing a new methyl iodide
penetration acceptance criteria does not involve any physical
changes or modifications to the function or operation of any safety-
related structure, system, or component. The new testing methodology
will enable a more accurate, conservative and reliable determination
of the charcoal decontamination efficiencies associated with the
SGTS [Standby Gas Treatment System], RERS [Reactor Enclosure
Recirculation System], and CREFAS [Control Room Emergency Fresh Air
System] which will better assure that the assumed charcoal
efficiencies credited in the licensed accident
[[Page 4288]]
analysis are adequately maintained. Implementing this change will
only involve revisions to existing procedures.
The SGTS, RERS, and CREFAS are standby systems that are designed
to mitigate the consequences of the analyzed accidents. No analyzed
accident initiating events are impacted, no new accident initiators
or new failure modes are created and the credited charcoal
efficiency for each system in the licensed accident analyses is not
changing as a result of the proposed changes. The ability of the
SGTS, RERS, and CREFAS to perform all of their safety-related
mitigation functions as designed will not be affected by the
proposed changes. Furthermore, the change in the testing methodology
and acceptance criteria will not result in increasing the dose rates
currently calculated in the existing accident analyses.
In addition, the proposed minor editorial changes are
administrative in nature and do not impact the operation, physical
configuration, or function of plant equipment or systems. The
proposed editorial changes do not impact the initiators or
assumptions of analyzed events, nor do they impact mitigation of
accidents or transient events.
Therefore, the proposed changes do not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
2. The proposed TS changes do not create the possibility of a
new or different kind of accident from any accident previously
evaluated.
Changing the methodology for the performance of the laboratory
testing of nuclear-grade activated charcoal in accordance with
Generic Letter 99-02, and establishing new methyl iodide penetration
acceptance criteria is not an accident initiator, does not create
any new failure modes, nor does it result in the occurrence of an
accident. This change does not result in any physical plant
modification and does not affect the safety-related function,
assigned charcoal efficiency assumed in the accident analyses, or
operation of the SGTS, RERS, and CREFAS. This change will only
involve revisions to existing procedures.
In addition, the proposed minor editorial changes are
administrative in nature and do not alter plant configuration,
require that new equipment be installed, alter assumptions made
about accidents previously evaluated, or impact the operation or
function of plant equipment.
Therefore, the proposed changes do not create the possibility of
a new or different kind of accident from any accident previously
evaluated.
3. The proposed TS changes do not involve a significant
reduction in a margin of safety.
The safety-related air cleaning units used in ESF [Engineered
Safety Feature] ventilation systems reduce the potential onsite and
offsite consequences of a radiological accident by adsorbing
radioiodine. Changing the methodology for the performance of the
laboratory testing of nuclear-grade activated charcoal samples from
Reg. Guide 1.52 to ASTM D3803-1989 in accordance with Generic Letter
99-02, and the establishment of new methyl iodide penetration
acceptance criteria does not increase the dose rates above what is
currently calculated in the accident analyses.
In addition, the proposed minor editorial changes are
administrative in nature and do not involve any physical changes to
plant structures, systems or components (SCCs), or the manner in
which SSCs are operated, maintained, modified, tested, or inspected.
The proposed editorial changes do not involve a change to any safety
limits, limiting safety system settings, limiting conditions of
operation, or design parameters for any SSC. The proposed editorial
changes do not impact any safety analysis assumptions and do not
involve a change in initial conditions, system response times, or
other parameters affecting any accident analysis.
Therefore, 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 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: J. W. Durham, Sr., Esquire, Sr. V.P. and
General Counsel, PECO Energy Company, 2301 Market Street, Philadelphia,
PA 19101.
NRC Section Chief: James W. Clifford.
PECO Energy Company, Public Service Electric and Gas Company, Delmarva
Power and Light Company, and Atlantic City Electric Company, Dockets
Nos. 50-277 and 50-278, Peach Bottom Atomic Power Station, Units Nos. 2
and 3, York County, Pennsylvania
Date of application for amendments: November 17, 1999.
Description of amendment request: The proposed changes will revise
the Peach Bottom Units 2 and 3 Technical Specifications (TSs) Section
5.5.7.c., Ventilation Filter Testing Program (VFTP), in accordance with
Generic Letter (GL) 99-02, ``Laboratory Testing of Nuclear-Grade
Activated Charcoal.'' This TS change will (1) specify that the
laboratory testing for methyl iodide penetration be performed
referencing ASTM D3803-1989 at a temperature of 30 deg.C (86 deg.F),
and (2) revise the acceptance criteria for methyl iodide penetration.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. The proposed changes do not involve a significant increase in
the probability or consequences of an accident previously evaluated.
Changing the methodology for the performance of the laboratory
testing of nuclear grade activated charcoal samples from RG
[Regulatory Guide] 1.52 to ASTM [American Society for Testing and
Materials] D3803-1989 and the establishment of new methyl iodide
penetration acceptance criteria and test temperature in accordance
with Generic Letter 99-02, do not involve any changes or
modifications to the function or operation of any safety related
structure, system, or component. The new testing methodology enables
a more accurate and conservative charcoal decontamination efficiency
to be determined which better assures that the assumed charcoal
efficiency credited in the licensed accident analysis is being
adequately maintained. Implementing this change only involves
revisions to existing procedures.
The SGTS [Standby Gas Treatment System] and MCREVS [Main Control
Room Emergency Ventilation System] are standby systems that are
designed to mitigate the consequences of the analyzed accidents. No
analyzed accident initiating events are impacted, no new accident
initiators or new failure modes are created and the credited
charcoal efficiency for each system in the licensed accident
analyses is not changing. The change in laboratory testing
methodology does not degrade the ability of these systems to perform
all of their safety related mitigation functions as designed.
Therefore, the proposed changes described above do not involve a
significant increase in the probability or consequences of an
accident previously evaluated.
2. The proposed changes do not create the possibility of a new
or different kind of accident from any accident previously
evaluated.
Changing the methodology for the performance of the laboratory
testing of nuclear grade activated charcoal in accordance with
Generic Letter 99-02 and establishing new methyl iodide penetration
acceptance criteria is not an accident initiator, does not create
any new failure modes, nor does it result in the occurrence of an
accident. This change does not result in any physical plant
modification and does not affect the safety related function,
charcoal efficiency, or operation of the SGTS or MCREVS. This change
only involves revisions to existing procedures to comply with NRC
guidance from GL 99-02.
Therefore, the possibility of a new or different kind of
accident than previously evaluated is not created.
3. The proposed changes do not involve a significant reduction
in a margin of safety.
The safety related air cleaning units used in ESF [Engineered
Safety Feature] ventilation systems reduce the potential onsite and
offsite consequences of a radiological accident by absorbing
radioiodine. Changing the methodology for the performance of the
laboratory testing of nuclear-grade activated charcoal samples from
RG 1.52 to ASTM D3803-1989 in accordance with Generic Letter 99-02,
and the establishment of new methyl iodide penetration acceptance
criteria does not increase the dose rates above what is currently
calculated in the accident analyses.
Therefore, the above change does not involve a significant
reduction in a margin of safety.
[[Page 4289]]
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.
Attorney for Licensee: J. W. Durham, Sr., Esquire, Sr. V.P. and
General Counsel, PECO Energy Company, 2301 Market Street, Philadelphia,
PA 19101.
NRC Section Chief: James W. Clifford.
Portland General Electric Company, et al., Docket No. 50-344, Trojan
Nuclear Plant, Columbia County, Oregon
Date of amendment request: November 16, 1999.
Description of amendment request: The proposed amendment would
revise the Trojan Nuclear Plant (TNP) Permanently Defueled Technical
Specifications by removing Figure 4.1-1, ``Site and Exclusion Area
Boundaries,'' from Section 4.0, ``Design Features,'' and incorporate
the applicable portion of this figure in the Trojan Nuclear Plant
Defueled Safety Analysis Report. Other associated administrative
changes resulting from the deletion of Figure 4.1-1, as well as an
editorial change to the table of contents, are also proposed.
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 requested license amendment consists of changes that are
administrative and/or editorial in nature, in that the physical and
operational characteristics of the TNP site are unchanged. As such,
the requested amendment does not in any way affect systems,
structures, or components that could initiate or be required to
mitigate the consequences of an accident previously evaluated.
Therefore, the proposed changes do not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
2. The proposed change does not create the possibility of a new
or different kind of accident from any accident previously
evaluated.
The requested license amendment consists of changes that are
administrative and/or editorial in nature, in that the physical and
operational characteristics of the TNP site are unchanged. As such,
the requested amendment does not affect systems, structures, or
components in any way not previously evaluated, and no new or
different failure modes will be created. Therefore, 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 does not involve a significant reduction
in a margin of safety.
The requested license amendment consists of changes that are
administrative and/or editorial in nature, in that the physical and
operational characteristics of the TNP site are unchanged.
Therefore, 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 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: Douglas R. Nichols, Esq., Portland General
Electric Company, 121 S.W. Salmon Street, Portland, Oregon 97204.
NRC Section Chief: Michael T. Masnik.
Public Service Electric & Gas Company, Docket No. 50-354, Hope Creek
Generating Station, Salem County, New Jersey
Date of amendment request: December 27, 1999.
Description of amendment request: The proposed amendment would
revise Technical Specifications (TS) 4.6.2.2.b, ``Suppression Pool
Spray,'' and 4.6.2.3.b, ``Suppression Pool Cooling,'' to modify the
acceptance criteria associated with flow rate testing of the Residual
Heat Removal (RHR) system pumps.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. The proposed changes do not involve a significant increase in
the probability or consequences of an accident previously evaluated.
The proposed TS change does not involve any physical changes to
plant structures, systems or components (SSC). The RHR system will
continue to function as designed. The RHR system is designed to
mitigate the consequences of an accident, and therefore, cannot
contribute to the initiation of any accident. The proposed TS
surveillance requirement changes implement testing methods that more
appropriately control and reflect RHR operation and establish
acceptance criteria, which ensure that Hope Creek's licensing and
design basis assumptions are met. In addition, this proposed TS
change will not increase the probability of occurrence of a
malfunction of any plant equipment important to safety, since the
manner in which the RHR system is operated is not affected by these
proposed changes. The proposed surveillance requirement acceptance
criteria ensure that the RHR safety functions will be accomplished.
Therefore, the proposed TS changes would not result in the increase
of the consequences of an accident previously evaluated, nor do they
involve an increase in the probability 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 TS changes do not involve any physical changes to
the design of any plant SSC. The design and operation of the RHR
system is not changed from that currently described in Hope Creek's
licensing basis. The RHR system will continue to function as
designed to mitigate the consequences of an accident. Implementing
the proposed changes does not result in plant operation in a
configuration that would create a different type of malfunction to
the RHR system than any previously evaluated. In addition, the
proposed TS changes do not alter the conclusions described in Hope
Creek's licensing basis regarding the safety related functions of
this system.
Therefore, the proposed TS change does not create the
possibility of a new or different kind of accident from any
previously evaluated.
3. The proposed change does not involve a significant reduction
in a margin of safety.
The proposed changes contained in this submittal would implement
testing methods that adequately demonstrate RHR pump capability and
establish acceptance criteria consistent with Hope Creek's licensing
basis. The ability of RHR to perform its safety functions is not
adversely affected by these proposed changes. Therefore, the
proposed TS 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.
Attorney for licensee: Jeffrie J. Keenan, Esquire, Nuclear Business
Unit-N21, P.O. Box 236, Hancocks Bridge, NJ 08038.
NRC Section Chief: James W. Clifford.
Public Service Electric & Gas Company, Docket Nos. 50-272 and 50-311,
Salem Nuclear Generating Station, Unit Nos. 1 and 2, Salem County, New
Jersey
Date of amendment request: December 29, 1999.
Description of amendment request: The proposed amendments would
revise the Salem Nuclear Generating Station Technical Specification
requirements for instrumentation in the reactor trip system by adding
tolerances to certain setpoint values.
[[Page 4290]]
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 not involve a significant increase in the probability or
consequences of an accident previously evaluated.
The accidents of concern affected by the over-temperature or
over-power delta temperature [trip signal] which have been evaluated
are unaffected by the proposed editorial changes thus the changes do
not significantly increase the probability or consequences of an
accident previously evaluated.
2. Does not create the possibility of a new or different kind of
accident from any accident previously analyzed.
The changes proposed are editorial in nature and do not alter
physical configuration, replace or modify existing equipment, affect
operating practices or create any new or different accident
precursors which could impact on the accident analysis. Thus there
is no possibility of a new or different kind of accident as a result
of the proposed changes.
3. Does not involve a significant reduction in a margin of
safety.
No margin of safety will be reduced by the proposed changes. The
proposed changes do not adversely affect the ability of the trip
systems to operate when called upon. Rather, these changes should
result in clarity regarding the proper calibration of the trip
instrumentation and therefore the margin of safety is preserved for
those events in which there is a dependence upon an over-temperature
or over-power delta temperature trip signal.
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.
Attorney for licensee: Jeffrie J. Keenan, Esquire, Nuclear Business
Unit--N21, P.O. Box 236, Hancocks Bridge, NJ 08038.
NRC Section Chief: James W. Clifford.
Rochester Gas and Electric Corporation, Docket No. 50-244, R. E. Ginna
Nuclear Power Plant, Wayne County, New York
Date of amendment request: November 30, 1999.
Description of amendment request: The proposed amendment would
allow the Rochester Gas and Electric Corporation to revise Sections
5.5.10 (a.3), (c.5), and (d.3) of the Ginna Station Improved Technical
Specifications (ITS) to provide a reference to American Society for
Testing and Materials (ASTM) Standard Procedure D3803-1989 as the
procedure for performing laboratory testing of charcoal adsorbers that
are installed in the Ginna Control Room Emergency Air Treatment System
(CREATS), Containment Post-Accident Sampling System (CPASS), and Spent
Fuel Pool Charcoal Absorber System (SFPCAS). These charcoal adsorbers
for the CREATS and CPASS are installed for the purpose of reducing the
levels of radioactive iodide species released to the containment and
control room during a postulated design basis, while the charcoal
adsorbers in the SFPCAS are installed for reducing the levels of
radioactive iodide species released to the auxiliary building during a
postulated fuel handling accident. The changes to ITS Sections (a.3),
(c.5), and (d.3) will also provide a specific test temperature and
humidity level for performing the testing of the charcoal adsorbers,
and to increase the allowable penetration of methyl iodide to these
systems from 10% to 14.5%. The requests for the changes are consistent
with the staff's position stated in NRC Generic Letter 99-02,
``Laboratory Testing of Nuclear-Grade Activated Charcoal,'' dated June
3, 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. With respect to the more restrictive proposals
associated with providing a reference to ASTM D3803-1989, ``Standard
Test Method for Nuclear-Grade Activated Carbon,'' and providing a
specific test temperature and relative humidity for testing the
charcoal adsorbers, the proposed changes do not involve a significant
hazards consideration as discussed below:
(1) Operation of Ginna Station in accordance with the proposed
changes does not involve a significant increase in the probability
or consequences of an accident previously evaluated. The changes add
a reference to the latest approved test protocol and provide for
specific test conditions. This does not increase the probability of
an accident previously evaluated since the tests are of themselves
not an accident initiator. The proposed changes are in accordance
with NUREG-1431 guidance and provide a higher assurance of the
ability of the charcoal adsorbers to perform as assumed in the
accident analysis. Therefore, the probability or consequences of an
accident previously evaluated is not significantly increased.
(2) Operation of Ginna Station in accordance with the proposed
changes does not create the possibility of a new or different kind
of accident from any accident previously evaluated. The proposed
changes add specific details of charcoal adsorber testing and do not
of themselves involve a physical alteration of the plant (ie. no new
or different type of equipment will be added to perform the required
testing) or changes in the methods governing normal plant operation.
The changes only involve implementing currently approved test
methodology. Therefore, the possibility for a new or different kind
of accident from any accident previously evaluated is not created.
(3) Operation of Ginna Station in accordance with the proposed
changes does not involve a significant reduction in a margin of
safety. The proposed changes only add conservatism in the test
requirements for the charcoal adsorbers credited in the accident
analysis. ASTM D3803-1989 is considered to be the most accurate and
most realistic protocol for testing charcoal in ventilation systems
because it offers the greatest assurance of accurately and
consistently determining the capability of the charcoal. Therefore,
this change does not involve a significant reduction in a margin of
safety.
With respect to the less restrictive proposal to increase the
allowable test limit for methyl iodide penetration of charcoal
adsorbers, the changes do not involve a significant hazards
consideration as discussed below:
(4) Operation of Ginna Station in accordance with the proposed
changes does not involve a significant increase in the probability
or consequences of an accident previously evaluated. The changes
revise the acceptance criteria for the allowed penetration of methyl
iodide during the testing of charcoal adsorbers in the plant
ventilation systems. This does not increase the probability of an
accident previously evaluated since the tests are of themselves not
an accident initiator. Because ASTM D3803-1989 is a more accurate
and demanding test than older tests this new protocol will allow the
use a safety factor of 2 for determining the acceptance criteria for
charcoal filter efficiency. The new acceptance criteria continue to
ensure that the efficiency assumed in the accident analysis is still
valid. Therefore, the probability or consequences of an accident
previously evaluated is not significantly increased.
(2) Operation of Ginna Station in accordance with the proposed
changes does not create the possibility of a new or different kind
of accident from any accident previously evaluated. The proposed
changes of revising charcoal adsorber testing acceptance criteria do
not of themselves involve a physical alteration of the plant (ie. no
new or different type of equipment will be added to perform the
required testing) or changes in the methods governing normal plant
operation. Therefore, the possibility for a new or different kind of
accident from any accident previously evaluated is not created.
(3) Operation of Ginna Station in accordance with the proposed
changes does not involve a significant reduction in a margin of
safety. The proposed changes only revise the test acceptance
criteria of charcoal adsorbers as the result of implementing testing
in accordance with ASTM D3803-1989. ASTM D3803-1989 is considered to
be the most accurate and most realistic protocol for testing
charcoal in ventilation systems because it offers the greatest
assurance of
[[Page 4291]]
accurately and consistently determining the capability of the
charcoal. Therefore, this change does not involve a significant
reduction in a margin of safety.
Based upon the preceding information, the Rochester Gas and
Electric Corporation determined that the proposed changes do not
involve a significant increase in the probability or consequences of an
accident previously evaluated, create the possibility of a new or
different kind of accident from any accident previously evaluated, or
involve a significant reduction in a margin of safety.
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.
Attorney for licensee: Nicholas S. Reynolds, Winston & Strawn, 1400
L Street, NW., Washington, DC 20005.
NRC Section Chief: Marsha Gamberoni, Acting.
Tennessee Valley Authority, Docket No. 50-390 Watts Bar Nuclear Plant,
Unit 1, Rhea County, Tennessee
Date of amendment request: September 30, 1999 (TS 98-005).
Description of amendment request: The proposed amendment would
revise the Watts Bar Nuclear Plant Unit 1 Technical Specifications (TS)
analytical methods for core operating limits to implement an analysis
supporting a more negative moderator temperature coefficient (MTC) for
the end of cycle condition. This alternate methodology is based on a
Westinghouse Electric Company analysis documented in reports WCAP-
15088-P, Revision 1 (proprietary), ``Safety Evaluation Supporting a
More Negative EOL Moderator Temperature Coefficient Technical
Specification for the Watts Bar Nuclear plant,'' and WCAP-15099-P,
Revision 1 (non-proprietary, same title).
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
A. The proposed amendment does not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
The more negative EOL [end-of-life] MTC does not increase the
probability of an accident previously evaluated in the FSAR [Final
Safety Analysis Report]. No new performance requirements are being
imposed on any system or component such that any design criteria
will be exceeded. The conservative MDC [moderator density
coefficient] assumption in the current analyses of record has been
confirmed to remain bounding for the more negative proposed TS
values. Therefore, no change in the modeling of the accident
analysis conditions or response is necessary in order to implement
this change. The consequences of an accident previously evaluated in
the FSAR are not increased due to the more negative EOL MTC. The
dose predictions presented in the FSAR remain valid such that no
more severe consequences will result.
B. The proposed amendment does not create the possibility of a
new or different kind of accident from any accident previously
evaluated.
The more negative EOL MTC does not create the possibility of an
accident which is different than any already evaluated in the FSAR.
No new failure modes have been defined for any system or component
nor has any new limiting single failure been identified.
Conservative assumptions for MDC have already been modeled in the
FSAR analyses and it has been determined that the more negative MTC
values to be implemented in the TS will continue to be bounded by
these assumptions.
C. The proposed amendment does not involve a significant
reduction in a margin of safety.
The evaluation of the more negative EOL MTC has taken into
account the applicable technical specifications and has bounded the
conditions under which the specifications permit operation. The
applicable technical specification is Section 5.9.5.b which lists
methods approved by the NRC for use in determining the core
operating limits. The values of the LCO [limiting condition for
operation] and SRs [surveillance requirements] are located in the
COLR [core operating limits report]. The analyses which support
these technical specifications have been evaluated. The results as
presented in the FSAR remain bounding for the more negative EOL MTC.
Therefore, the margin of safety, as defined in the bases to these
technical specifications, is not reduced.
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.
Attorney for licensee: General Counsel, Tennessee Valley Authority,
400 West Summit Hill Drive, ET 10H, Knoxville, Tennessee 37902.
NRC Section Chief: Richard Correia.
Vermont Yankee Nuclear Power Corporation, Docket No. 50-271, Vermont
Yankee Nuclear Power Station, Vernon, Vermont
Date of amendment request: December 21, 1999.
Description of amendment request: This proposed change revises the
control rod block requirements consistent with the BWR/4 Standard
Technical Specifications. Some functions are proposed to be relocated
to the Technical Requirements Manual, the requirements for the retained
functions are clarified, and two functions are added to the Technical
Specifications.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration which is presented below:
1. The 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.
The relocated functions are not assumed as initial conditions
for, nor are they credited in the mitigation of, any design basis
accident or transient previously evaluated. Since reactor operation
with these revised and relocated Specifications is fundamentally
unchanged, no design or analytical acceptance criteria will be
exceeded. As such, this change does not impact initiators of
analyzed events nor assumed mitigation of design basis accident or
transient events.
More stringent and purely administrative changes do not affect
the initiation of any event, nor do they negatively impact the
mitigation of any event. Therefore, the proposed changes do 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.
None of the proposed changes affects any parameters or
conditions that could contribute to the initiation of an accident.
No new accident modes are created since the manner in which the
plant is operated is unchanged. No safety-related equipment or
safety functions are altered as a result of these changes.
Therefore, the proposed changes will not create the possibility of a
new or different kind of accident from any accident 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.
There is no impact on equipment design or operation, and there
are no changes being made to safety limits or safety system settings
that would adversely affect plant safety as a result of the proposed
changes. Since the changes have no effect on any safety analysis
assumption or initial condition, the margins of safety in the safety
analyses are maintained. In addition, neither administrative changes
with no technical impact, nor the imposition of more stringent
requirements have a negative impact on a margin of safety.
Therefore, the proposed
[[Page 4292]]
changes do not involve a significant reduction in a margin of
safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
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.
Wolf Creek Nuclear Operating Corporation, Docket No. 50-482, Wolf Creek
Generating Station, Coffey County, Kansas.
Date of amendment request: December 15, 1999 (ET 99-0050).
Description of amendment request: The proposed amendment would
modify the Improved Technical Specifications (ITSs) that were issued in
Amendment No. 123 on March 31, 1999, and implemented on December 18,
1999. The proposed changes would expand the region of acceptable seal
injection flow in Figure 3.5.5-1 of ITS 3.5.5 and provide the following
10 editorial changes: (1) delete the redundant ``%'' sign in the
allowable value for function 4 in Table 3.3.1-1 on reactor trip system
instrumentation, (2) delete the extra spacing in the description of
function 20 in Table 3.3.1-1, (3) insert periods at the end of the text
for Conditions M and N in the actions for limiting condition for
operation (LCO) 3.3.2 on engineered safety features actuation system
instrumentation (ESFASI), (4) spell ``requirements'' correctly in
function 5.c of Table 3.3.2-1 for ESFASI, (5) delete the unneeded ``SR
3.3.2.6'' from the surveillance requirements column for Function 7.a in
Table 3.3.2-1, (6) align the wording ``Coincident with Safety
Injection'' with the title of Function 7.b in Table 3.3.2-1, (7) align
the data in the 4 columns of Table 3.3.7-1, CREVS [control room
emergency ventilation system] Actuation Instrumentation, for Function 3
with the first line of the title of the function, (8) align the
specified completion time in Condition B of the actions for LCO 3.7.1
for main steam safety valves with text for the Required Action B.2, (9)
add the acronym ``EES'' to Emergency Exhaust System in the table of
contents and use the acronym in the upper right-hand-corner of the 4
ITS pages for LCO 3.7.13 on the emergency exhaust system, and (10)
uncapitalize the word ``Associated'' in Condition B of the actions for
LCO 3.8.4 on DC sources--operating because it should not be
capitalized. The licensee would also add text to the Bases to the
applicable safety analyses for the seal injection flow of LCO 3.5.5.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. The proposed change does not involve a significant increase
in the probability or consequences of an accident previously
evaluated.
The restriction on RCP [reactor coolant pump] seal injection
flow limits the amount of ECCS [emergency core cooling system] flow
that would be diverted from the injection path following an
accident. This limit is based on safety analysis assumptions that
are required because RCP seal injection flow is not isolated during
SI [safety injection]. The intent of the LCO 3.5.5 limit on seal
injection flow is to make sure that flow through the RCP seal water
injection line is low enough to ensure that sufficient centrifugal
charging pump injection flow is directed to the RCS [reactor coolant
system] via the injection points. The expansion of the Acceptable
Range for the flow limits does not impact the assumed ECCS flow that
would be available for injection into the RCS following an accident.
There are no hardware changes nor are there any changes in the
method by which any safety related plant system performs its safety
function. Since the change continues to ensure 100 percent of the
assumed charging flow is available, the proposed change does not
involve a significant increase in the probability or consequences of
an accident previously evaluated.
The proposed editorial changes involve corrections to the
improved Technical Specifications that are associated with the
original conversion application and supplements or the certified
copy of the improved Technical Specifications. As such, these
changes are considered as administrative changes and do not modify,
add, delete, or relocate any technical requirements of the Technical
Specifications.
Therefore, the proposed changes do not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
2. The proposed change does not create the possibility of a new
or different kind of accident from any accident previously
evaluated.
The proposed changes do not involve a physical alteration of the
plant (no new of different type of equipment will be installed) or
changes in methods governing normal plant operation. The proposed
changes will not impose any new or eliminate any old requirements.
The expansion of the Acceptable Range for the [seal injection] flow
limits does not impact the assumed ECCS flow that would be available
for injection into the RCS following an accident.
The proposed editorial changes involve corrections to the
improved Technical Specifications that are associated with the
original conversion application and supplements or the certified
copy of the improved Technical Specifications. As such, these
changes are considered as administrative changes and do not modify,
add, delete, or relocate any technical requirements of the Technical
Specifications.
Thus, the changes do 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 change [for seal injection flow] does not affect
the acceptance criteria for any analyzed event. There will be no
effect on the manner in which safety limits or limiting safety
system settings are determined nor will there be any effect on those
plant systems necessary to assure the accomplishment of protection
functions. The expansion of the Acceptable Range for the flow limits
does not impact the assumed ECCS flow that would be available for
injection into the RCS following an accident.
The proposed editorial changes involve corrections to the
improved Technical Specifications that are associated with the
original conversion application and supplements or the certified
copy of the improved Technical Specifications. As such, these
changes are considered as administrative changes and do not modify,
add, delete, or relocate any technical requirements of the Technical
Specifications.
Therefore, the 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
amendment request involves no significant hazards consideration.
Attorney for licensee: Jay Silberg, Esq., Shaw, Pittman, Potts and
Trowbridge, 2300 N Street, N.W., Washington, D.C. 20037.
NRC Section Chief: Stephen Dembek.
Previously Published Notices of Consideration of Issuance of
Amendments to Facility Operating Licenses, Proposed No Significant
Hazards Consideration Determination, and Opportunity for a Hearing
The following notices were previously published as separate
individual notices. The notice content was the same as above. They were
published as individual notices either because time did not allow the
Commission to wait for this biweekly notice or because the action
involved exigent circumstances. They are repeated here because the
biweekly notice lists all amendments issued or proposed to be issued
involving no significant hazards consideration.
For details, see the individual notice in the Federal Register on
the day and
[[Page 4293]]
page cited. This notice does not extend the notice period of the
original notice.
Indiana Michigan Power Company, Docket, Nos. 50-315 and 50-316, Donald
C. Cook Nuclear Plant, Units 1 and 2, Berrien County, Michigan
Date of application for amendments: December 22, 1999.
Brief description of amendments: The amendments would delete the
Donald C. Cook (D.C. Cook), Unit 1 and 2, Technical Specification (TS)
5.4.2, ``Reactor Coolant System Volume,'' because the information
regarding the reactor coolant system (RCS) is not required by TS
Section 5.0, ``Design Features,'' for compliance with 10 CFR
50.36(c)(4). Changes to the RCS volume information are included in the
D.C. Cook Updated Final Safety Analyses Report, and are controlled in
accordance with 10 CFR 50.59.
Date of publication of individual notice in Federal Register:
January 13, 1999 (65 FR 2199).
Expiration date of individual notice: February 14, 2000.
Power Authority of the State of New York, Docket No. 50-333, James A.
FitzPatrick Nuclear Power Plant, Oswego County, New York
Date of amendment request: March 31, 1999, as supplemented by
letters dated May 20, June 1, July 14, and October 14, 1999.
Description of amendment request: The amendment converts the
current Technical Specifications (TSs) for the James A. FitzPatrick
Nuclear Power Plant, to a set of improved TSs based upon NUREG-1433,
``Standard Technical Specifications for General Electric Plants BWR/4''
Revision 1 dated April 1995.
Date of publication of individual notice in Federal Register:
November 8, 1999 (64 FR 60854).
Expiration date of individual notice: December 8, 1999.
Notice of Issuance of Amendments to Facility Operating Licenses
During the period since publication of the last biweekly notice,
the Commission has issued the following amendments. The Commission has
determined for each of these amendments that the application complies
with the standards and requirements of the Atomic Energy Act of 1954,
as amended (the Act), and the Commission's rules and regulations. The
Commission has made appropriate findings as required by the Act and the
Commission's rules and regulations in 10 CFR Chapter I, which are set
forth in the license amendment.
Notice of Consideration of Issuance of Amendment to Facility
Operating License, Proposed No Significant Hazards Consideration
Determination, and Opportunity for A Hearing in connection with these
actions was published in the Federal Register as indicated.
Unless otherwise indicated, the Commission has determined that
these amendments satisfy the criteria for categorical exclusion in
accordance with 10 CFR 51.22. Therefore, pursuant to 10 CFR 51.22(b),
no environmental impact statement or environmental assessment need be
prepared for these amendments. If the Commission has prepared an
environmental assessment under the special circumstances provision in
10 CFR 51.12(b) and has made a determination based on that assessment,
it is so indicated.
For further details with respect to the action see (1) the
applications for amendment, (2) the amendment, and (3) the Commission's
related letter, Safety Evaluation and/or Environmental Assessment as
indicated. All of these items are available for public inspection at
the Commission's Public Document Room, the Gelman Building, 2120 L
Street, NW., Washington, DC, and electronically from the ADAMS Public
Library component on the NRC Web site, http://www.nrc.gov (the
Electronic Reading Room).
AmerGen Energy Co., LLC, Docket No. 50-289, Three Mile Island Nuclear
Station, Unit 1, Dauphin County, Pennsylvania
Date of application for amendment: June 11, 1999.
Brief description of amendment: The amendment made various title
changes to the plant organization.
Date of issuance: January 7, 2000.
Effective date: As of the date of issuance and shall be implemented
within 30 days.
Amendment No.: 219.
Facility Operating License No. DPR-50: Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: July 14, 1999 (64 FR
38027).
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated January 7, 2000.
No significant hazards consideration comments received: No.
Commonwealth Edison Company, Docket Nos. 50-295 and 50-304, Zion
Nuclear Power Station, Units 1 and 2, Lake County, Illinois
Date of application for amendments: October 2, 1998, as
supplemented by letters dated April 13, 1999, and September 15, 1999.
Information in Commonwealth Edison correspondence dated July 8, 1999,
and August 30, 1999, was also considered during the review of the
amendments.
Brief description of amendments: The amendments replace the custom
operational technical specifications with a set of permanently defueled
technical specifications that reflect the permanently shutdown and
defueled status of the Zion Nuclear Power Station, Units 1 and 2. The
amendments also delete certain license conditions from the operating
licenses that are no longer applicable to the facility in its
permanently shutdown and defueled condition. Information supplied in
Commonwealth Edison letters dated July 8, 1999, August 30, 1999, and
September 15, 1999, provided clarifying information and did not expand
the scope of the original Federal Register notice dated June 2, 1999,
and did not change the staff's proposed no significant hazards finding.
Date of issuance: December 30, 1999.
Effective date: December 30, 1999.
Amendment Nos.: Unit 1--180; Unit 2--167.
Facility Operating License Nos. DPR-39 and DPR-48: The amendments
revised the Technical Specifications and the operating licenses.
Date of initial notice in Federal Register: June 2, 1999 (64 FR
29709).
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated December 30, 1999.
No significant hazards consideration comments received: No.
Detroit Edison Company, Docket No. 50-341, Fermi 2, Monroe County,
Michigan
Date of application for amendment: September 10, 1999 (NRC-99-
0072), as supplemented November 19, 1999 (NRC-99-0107).
Brief description of amendment: The amendment revises the Technical
Specification surveillance requirements for the Division I 130/260-volt
dc battery to accommodate the design of the replacement battery.
Date of issuance: January 12, 2000.
Effective date: As of the date of issuance and shall be implemented
prior to the startup from the seventh refueling outage.
Amendment No.: 136.
Facility Operating License No. NPF-43: Amendment revises the
Technical Specifications.
Date of initial notice in Federal Register: November 3, 1999 (64 FR
59800). The November 19, 1999, letter provided clarifying information
that was within
[[Page 4294]]
the scope of the original Federal Register notice and did not change
the staff's initial proposed no significant hazards consideration
determination.
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated January 12, 2000.
No significant hazards consideration comments received: No.
Duke Energy Corporation, et al., Docket Nos. 50-413 and 50-414, Catawba
Nuclear Station, Units 1 and 2, York County, South Carolina
Date of application for amendments: September 16, 1999,
supplemented November 3, 1999.
Brief description of amendments: The amendments revise Section
3.8.4, ``DC Sources--Operating'' of the Technical Specifications.
Specifically, the amendments modify Surveillance Requirements (SRs)
3.8.4.8 and 3.8.4.9 and the associated Bases SR 3.8.4.8 and 3.8.4.9 to
allow testing of the direct current (dc) channel batteries with the
units on line. The change to SR 3.8.4.8 would also prohibit the diesel
generator batteries from being service tested while the units are on
line.
Date of issuance: January 7, 2000.
Effective date: As of the date of issuance and shall be implemented
within 45 days from the date of issuance.
Amendment Nos.: Unit 1-183; Unit 2-175.
Facility Operating License Nos. NPF-35 and NPF-52: Amendments
revised the Technical Specifications.
Date of initial notice in Federal Register: October 20, 1999 (64 FR
56529). The Commission's related evaluation of the amendments is
contained in a Safety Evaluation dated January 7, 2000.
No significant hazards consideration comments received: No.
First Energy Nuclear Operating Company, Docket No. 50-440, Perry
Nuclear Power Plant, Unit 1, Lake County, Ohio.
Date of application for amendment: September 9, 1999.
Brief description of amendment: This amendment revised the Perry
Nuclear Power Plant Environmental Protection Plan by eliminating the
requirement to sample Lake Erie sediment in the Perry and Eastlake
Plant area for Corbicula, since Corbicula and zebra mussels have
already been identified, and control and treatment plans have been
implemented which are effective for both species.
Date of issuance: January 5, 2000.
Effective date: As of the date of issuance and shall be implemented
within 30 days.
Amendment No.: 110.
Facility Operating License No. NPF-58: This amendment revised the
Environmental Protection Plan.
Date of initial notice in Federal Register: November 3, 1999 (64 FR
59802). The Commission's related evaluation of the amendment is
contained in a Safety Evaluation dated January 5, 2000.
No significant hazards consideration comments received: No.
Indiana Michigan Power Company, Docket Nos. 50-315 and 50-316, Donald
C. Cook Nuclear Plant, Units 1 and 2, Berrien County, Michigan
Date of application for amendments: November 3, 1999.
Brief description of amendments: The amendments allow use of fuel
rods with ZIRLO cladding, specify an alternate methodology to determine
the integral fuel burnable absorber (IFBA) requirements for
Westinghouse fuel assemblies stored in the new fuel storage racks, and
delete the designation of the fuel assembly types allowed in the spent
fuel storage racks and the new fuel storage racks.
Date of issuance: January 6, 2000.
Effective date: As of the date of issuance and shall be implemented
within 45 days.
Amendment Nos.: 239 and 220.
Facility Operating License Nos. DPR-58 and DPR-74: Amendments
revised the Technical Specifications.
Date of initial notice in Federal Register: December 1, 1999 (64 FR
67335).
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated January 6, 2000.
No significant hazards consideration comments received: No.
Pacific Gas and Electric Company, Docket Nos. 50-275 and 50-323, Diablo
Canyon Nuclear Power Plant, Unit Nos. 1 and 2, San Luis Obispo County,
California
Date of application for amendments: March 18, 1998, as supplemented
by letters dated March 25, September 29, and November 3, 1999.
Brief description of amendments: The amendments change the way
passive failures in the auxiliary saltwater (ASW) and component cooling
water (CCW) systems are mitigated during the long-term recovery period
following a loss-of-coolant accident (LOCA). Specifically, plant
procedures will no longer require ASW and CCW system train separation
after the transfer to hot leg recirculation following a LOCA.
Date of issuance: January 13, 2000.
Effective date: January 13, 2000, and shall be implemented in the
next periodic update to the FSAR Update in accordance with 10 CFR
50.71(e).
Amendment Nos.: Unit 1--138, Unit 2--138.
Facility Operating License Nos. DPR-80 and DPR-82: The amendments
revised the Final Safety Analysis Report Update.
Date of initial notice in Federal Register: October 7, 1998 (63 FR
53953).
The supplemental letters dated March 25, September 29, and November
3, 1999, provided additional clarifying information, did not expand the
scope of the application as originally noticed, and did not change the
staff's initial no significant hazards consideration determination.
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated January 13, 2000.
No significant hazards consideration comments received: No.
PP&L, Inc., Docket No. 50-387, Susquehanna Steam Electric Station, Unit
1, Luzerne County, Pennsylvania
Date of application for amendment: March 12, 1999, as supplemented
by letter dated November 1, 1999.
Brief description of amendment: This amendment revised the Minimum
Critical Power Ratio safety limits in TS Section 2.1.1.2 and modified
the references in TS Section 5.6.5 of a critical power correlation
applicable to Siemens Power Corporation Atrium-10 fuel.
Date of issuance: December 30, 1999.
Effective date: As of date of issuance and shall be implemented
upon startup from the Unit 1 eleventh refueling and inspection outage
currently scheduled for spring 2000.
Amendment No.: 186.
Facility Operating License No. NPF-14: This amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: April 7, 1999 (64 FR
17029).
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated December 30, 1999.
No significant hazards consideration comments received: No.
Southern Nuclear Operating Company, Inc., Docket Nos. 50-348 and 50-
364, Joseph M. Farley Nuclear Plant, Units 1 and 2, Houston County,
Alabama
Date of amendments request: December 1, 1998, as supplemented by
your letters of April 21, July 19, October 18, and November 11, 1999.
Brief Description of amendments: The proposed amendments would
revise the Technical Specifications to reflect replacing the current
Model 51 steam generators with Westinghouse Model
[[Page 4295]]
54F steam generators. The replacement program includes re-analyzing and
evaluating loss-of-coolant-accident (LOCA) and non-LOCA mass and energy
releases, containment and sub-compartment pressure and temperature
responses, dose analyses, and the effects on nuclear steam supply and
balance of plant systems.
Date of issuance: December 29, 1999.
Effective date: As of the date of issuance and shall be implemented
prior to Unit 1 entering Mode 5 for Cycle 17 (Spring 2000) and prior to
Unit 2 entering Mode 5 for Cycle 15 (Spring 2001).
Amendment Nos.: Unit 1-147; Unit-238.
Facility Operating License Nos. NPF-2 and NPF-8: Amendments revise
the Improved Technical Specifications.
Date of initial notice in Federal Register: October 20, 1999 (64 FR
56533). The supplemental letters dated October 18, and November 11,
1999, provided clarifying information that did not change the initial
proposed no significant hazards consideration determinations.
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated December 29, 1999.
No significant hazards consideration comments received: No.
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: July 22, 1998, as supplemented by
letters dated June 16, October 21 and 27, November 17, and December 9,
1999.
Brief description of amendments: The amendments revised the
Technical Specifications to reflect the steam generator water level
low-low trip setpoint differences between the existing Model E and the
replacement Model Delta-94 steam generators for the reactor trip system
and the engineered safety features actuation system instrumentation.
Date of issuance: December 29, 1999.
Effective date: December 29, 1999, to be implemented following
replacement of Unit 1 Model E steam generators with Model Delta-94
steam generators and prior to entry into Operational Mode 3.
Amendment Nos.: Unit 1--120; Unit 2--108.
Facility Operating License Nos. NPF-76 and NPF-80: The amendments
revised the Technical Specifications.
Date of initial notice in Federal Register: September 9, 1998 (63
FR 48268).
The June 16, October 21 and 27, November 17, and December 9, 1999,
supplements provided additional clarifying information that was within
the scope of the original application and Federal Register 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 December 29, 1999.
No significant hazards consideration comments received: No.
Dated at Rockville, Maryland, this 19th day of January 2000.
For the Nuclear Regulatory Commission.
John A. Zwolinski,
Director, Division of Licensing Project Management, Office of Nuclear
Reactor Regulation.
[FR Doc. 00-1732 Filed 1-25-00; 8:45 am]
BILLING CODE 7590-01-P