[Federal Register Volume 65, Number 240 (Wednesday, December 13, 2000)]
[Notices]
[Pages 77913-77934]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: 00-31541]
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NUCLEAR REGULATORY COMMISSION
Biweekly Notice; Applications and Amendments to Facility
Operating Licenses Involving No Significant Hazards Considerations
I. Background
Pursuant to Public Law 97-415, the U.S. Nuclear Regulatory
Commission (the Commission or NRC staff) is publishing this regular
biweekly notice. Public Law 97-415 revised section 189 of the Atomic
Energy Act of 1954, as amended (the Act), to require the Commission to
publish notice of any amendments issued, or proposed to be issued,
under a new provision of section 189 of the Act. This provision grants
the Commission the authority to issue and make immediately effective
any amendment to an operating license upon a determination by the
Commission that such amendment involves no significant hazards
consideration, notwithstanding the pendency before the Commission of a
request for a hearing from any person.
This biweekly notice includes all notices of amendments issued, or
proposed to be issued from November 17, 2000, through December 1, 2000.
The last biweekly notice was published on November 29, 2000.
Notice of Consideration of Issuance of Amendments to Facility
Operating Licenses, Proposed No Significant Hazards Consideration
Determination, and Opportunity for a Hearing
The Commission has made a proposed determination that the following
amendment requests involve no significant hazards consideration. Under
the Commission's regulations in 10 CFR 50.92, this means that operation
of the facility in accordance with the proposed amendment would not (1)
involve a significant increase in the probability or consequences of an
accident previously evaluated; or (2) create the possibility of a new
or different kind of accident from any accident previously evaluated;
or (3) involve a significant reduction in a margin of safety. The basis
for this proposed determination for each amendment request is shown
below.
The Commission is seeking public comments on this proposed
determination. Any comments received within 30 days after the date of
publication of this notice will be considered in making any final
determination.
Normally, the Commission will not issue the amendment until the
expiration of the 30-day notice period. However, should circumstances
change during the notice period such that failure to act in a timely
way would result, for example, in derating or shutdown of the facility,
the Commission may issue the license amendment before the expiration of
the 30-day notice period, provided that its final determination is that
the amendment involves no significant hazards consideration. The final
determination will consider all public and State comments received
before action is taken. Should the Commission take this action, it will
publish in the Federal Register a notice of issuance and provide for
opportunity for a hearing after issuance. The Commission expects that
the need to take this action will occur very infrequently.
Written comments may be submitted by mail to the Chief, Rules and
Directives Branch, Division of Administrative 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's Public Document
Room, located at One White Flint North, 11555 Rockville Pike (first
floor), Rockville, Maryland 20852. The filing of requests for a hearing
and petitions for leave to intervene is discussed below.
By January 12, 2001, 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, located at One
White Flint North, 11555 Rockville Pike (first Floor), Rockville,
Maryland 20852. Publicly available records will be accessible
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
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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, located at One White Flint North, 11555 Rockville Pike
(first floor), Rockville, Maryland 20852, by the above date. A copy of
the petition should also be sent to the Office of the General Counsel,
U.S. Nuclear Regulatory Commission, Washington, DC 20555-0001, and to
the attorney for the licensee.
Nontimely filings of petitions for leave to intervene, amended
petitions, supplemental petitions and/or requests for a hearing will
not be entertained absent a determination by the Commission, the
presiding officer or the Atomic Safety and Licensing Board that the
petition and/or request should be granted based upon a balancing of
factors specified in 10 CFR 2.714(a)(1)(i)-(v) and 2.714(d).
For further details with respect to this action, see the
application for amendment which is available for public inspection at
the Commission's Public Document Room, located at One White Flint
North, 11555 Rockville Pike (first floor), Rockville, Maryland 20852.
Publicly available records will be accessible electronically from the
ADAMS Public Library component on the NRC Web site, http://www.nrc.gov
(the Electronic Reading Room).
Commonwealth Edison Company, Docket Nos. STN 50-454 and STN 50-455,
Byron Station, Unit Nos. 1 and 2, Ogle County, Illinois; Docket Nos.
STN 50-456 and STN 50-457, Braidwood Station, Unit Nos. 1 and 2, Will
County, Illinois
Date of amendment request: July 5, 2000.
Description of amendment request: The proposed amendment would
revise the maximum power level specified in each unit's license; revise
the value of rated thermal power of each unit from 3411 megawatts
thermal (MWt) to 3586.6 MWt and the reference source for conversion
factors in the calculation of Dose Equivalent Iodine 131 in the
technical specification (TS) definitions; add a Departure from Nucleate
Boiling Ratio (DNBR) limit specifically for a thimble cell; increase
the minimum limit for reactor coolant system (RCS) total flow; revise
the steam generator laser welded sleeve plugging limit; and reduce the
peak calculated containment internal pressure Pa for the
design basis loss-of-coolant accident (LOCA).
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. Does the proposed change involve a significant increase in
the probability or consequences of an accident previously evaluated?
A. Evaluation of the Probability of Previously Evaluated
Accidents.
Plant systems and components have been verified to be capable of
performing their intended design functions at uprated power
conditions. Where necessary, some components will be modified prior
to implementation of uprated power operations to accommodate the
revised operating conditions. The analysis has concluded that
operation at uprated power conditions will not adversely affect the
capability or reliability of plant equipment. Current TS
surveillance requirements ensure frequent and adequate monitoring of
system and component operability. All systems will continue to be
operated in accordance with current design requirements under
uprated conditions, therefore no new components or system
interactions have been identified that could lead to an increase in
the probability of any accident previously evaluated in the Updated
Final Safety Analysis Report (UFSAR). No changes were required to
the Reactor Trip or Engineered Safety Features (ESF) setpoints.
B. Evaluation of the Consequences of Previously Evaluated
Accidents.
The radiological consequences were reviewed for all design basis
accidents (DBAs) (i.e., both Loss of Coolant Accident (LOCA) and
non-LOCA accidents) previously analyzed in the UFSAR. The analysis
showed that the resultant radiological consequences for both LOCA
and non-LOCA accidents remain either unchanged or have not
significantly increased due to operation at uprated power
conditions. The radiological consequences of all DBAs continue to
meet established regulatory limits.
The proposed addition of Table E-7 of NRC Regulatory Guide
1.109, ``Calculation of Annual Doses to Man from Routine Releases of
Reactor Effluents for the Purpose of Evaluating Compliance with 10
CFR Part 50, Appendix I,'' Revision 1, 1977, or International
Commission on Radiological Protection (ICRP) 30, ``Limits for
Intakes of Radionuclides by Workers,'' Supplement to Part 1, page
192-212, Table titled, ``Committed Dose Equivalent in Target Organs
or Tissues per Intake of Unit Activity,'' for thyroid dose
conversion factors, will not significantly increase the consequences
of an accident previously evaluated. If Regulatory Guide 1.109, or
ICRP 30, Supplement to Part 1, are used to calculate maximum dose
equivalent iodine specific activity, the total RCS iodine activity
may increase, depending on the iodine nuclide mix, and this activity
is used to calculate the doses resulting from a Main Steam Line
Break (MSLB) or other analyzed accident. The calculated thyroid
doses resulting from an MSLB or other analyzed accident would not
increase as the corresponding dose conversion factors would be used
to calculate the offsite thyroid doses. For a given Dose Equivalent
I-131 concentration in the RCS, the offsite dose predicted using the
dose conversion factors in either Table E-7 of Regulatory Guide
1.109, or ICRP 30, Supplement to Part 1, is less than that predicted
by Table III of Atomic Energy Commission (AEC) Technical Information
Document TID-14844, ``Calculation of Distance Factors for Power and
Test Reactor Sites,'' which is currently referenced in the TS
definition of Dose Equivalent I-131.
ICRP-30 is the updated reference for thyroid dose conversion
factors used in the power uprate accident analysis radiological
evaluation. The current version of 10 CFR 20, ``Standards for
Protection Against Radiation,'' also utilizes ICRP-30 data.
Based on the analysis, it is concluded that the proposed TS
changes do not involve a significant increase in the probability or
consequences of an accident previously evaluated.
2. Does the proposed change create the possibility of a new or
different kind of accident from any accident previously evaluated?
The configuration, operation and accident response of the Byron
Station and the Braidwood Station systems, structures or components
are unchanged by operation at uprated power conditions or by the
associated proposed TS changes. Analyses of transient events have
confirmed that no transient event results in a new sequence of
events that could lead to a new accident scenario.
The effect of operation at uprated power conditions on plant
equipment has been
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evaluated. No new operating mode, safety-related equipment lineup,
accident scenario, or equipment failure mode was identified as a
result of operating at uprated conditions. In addition, operation at
uprated power conditions does not create any new failure modes that
could lead to a different kind of accident. Minor plant
modifications, to support implementation of uprated power
conditions, will be made as required to existing systems and
components. The basic design of all systems remains unchanged and no
new equipment or systems have been installed which could potentially
introduce new failure modes or accident sequences. No changes have
been made to any Reactor Trip or ESF actuation setpoints.
Based on this analysis, it is concluded that no new accident
scenarios, failure mechanisms or limiting single failures are
introduced as a result of the proposed changes. The proposed TS
changes do not have an adverse effect on any safety-related system.
Therefore, the proposed TS changes do not create the possibility of
a new or different kind of accident from any accident previously
evaluated.
3. Does the proposed change involve a significant reduction in a
margin of safety?
A comprehensive analysis was performed to support the power
uprate program at the Byron Station and the Braidwood Station. This
analysis identified and defined the major input parameters to the
NSSS [Nuclear Steam Supply System], reviewed NSSS design transients,
and reviewed the capabilities of the NSSS fluid systems, NSSS/BOP
[balance-of-plant] interfaces, NSSS control systems, and NSSS and
BOP components. All appropriate NSSS accident analysis was
reperformed to confirm acceptable results were maintained and that
the radiological consequences remained within regulatory limits. The
nuclear and thermal hydraulic performance of nuclear fuel was also
reviewed to confirm acceptable results. The analysis confirmed that
all NSSS and BOP systems and components are capable, some with minor
modifications, to safely support operations at uprated power
conditions.
To support the operation of Byron Station, Units 1 and 2, and
Braidwood Station, Units 1 and 2 at uprated power conditions,
nuclear fuel Departure from Nucleate Boiling Ratio (DNBR) reanalysis
was required to define new core limits, axial offset limits, and
Condition II, ``Faults of Moderate Frequency,'' acceptability. This
analysis included review of the following events: loss of RCS flow,
reactor coolant pump locked rotor, feedwater malfunction, dropped
control rod, steamline break, and control rod withdrawal from a
subcritical condition. DNB design criteria was met for all events.
NUREG-1431, ``Standard Technical Specifications, Westinghouse
Plants,'' Revision 1, dated April 1995, allows Dose Equivalent I-131
to be calculated using any one of three dose conversion factors;
Table III of TID-14844, 1962, Table E-7 of NRC Regulatory Guide
1.109, Revision 1, 1977, or ICRP 30, Supplement to Part 1. Using
thyroid dose conversion factors other than those given in TID-14844
results in lower doses and higher allowable activity but is
justified by the discussion given in the Federal Register (i.e.,
Federal Register (FR) page 23360 Vol. 56, May 21, 1991). This
discussion accompanied the final rulemaking on 10 CFR 20,
``Standards for Protection Against Radiation,'' by the NRC. In that
discussion, the NRC stated that it was incorporating modifications
to existing concepts and recommendations of the ICRP into NRC
regulations. Incorporation of the methodology of ICRP-30 into the 10
CFR 20 revision was specifically mentioned with the changes being
made resulting from changes and updates in the scientific techniques
and parameters used in calculating dose. This FR reference clearly
shows that the NRC was updating 10 CFR 20 to incorporate ICRP-30
recommendations and data. Regulatory Guide 1.109 thyroid dose
conversion factors are higher than the ICRP-30 thyroid dose
conversion factors for all five iodine isotopes of concern.
Therefore, using Regulatory Guide 1.109 thyroid dose conversion
factors to calculate Dose Equivalent I-131 is more conservative than
ICRP-30 and is therefore acceptable. For a given Dose Equivalent I-
131 concentration in the Reactor Coolant, the offsite dose predicted
using the dose conversion factors in either Table E-7 of Regulatory
Guide 1.109, Revision 1, NRC, 1977, or ICRP 30, Supplement to Part
1, is less than that predicted by Table III of TID-14844 which is
currently referenced in the TS definition of Dose Equivalent I-131.
ICRP-30 is the updated reference source used in the power uprate
accident analysis radiological evaluation. All regulatory acceptance
criteria continue to be met and adequate safety margin is
maintained.
Revising the minimum limit for RCS total flow from greater than
or equal to 371,400 gpm to greater than or equal to 380,900 gpm does
not represent a significant reduction in the margin of safety. The
reactor coolant pumps run at full flow and have a total flow
capacity greater than 380,900 gpm. The analysis has shown that DNBR
criteria has been met for all normal operational transients and loss
of flow accident scenarios.
The margin of safety of the reactor coolant pressure boundary is
maintained under uprated power conditions. The design pressure of
the reactor pressure vessel and reactor coolant system will not be
challenged as the pressure mitigating systems were confirmed to be
sufficiently sized to adequately control pressure under uprated
power conditions.
The proposed change revises the plugging limit for laser welded
sleeves from 40% to 38.7% of nominal wall thickness. The analysis
performed in support of the power uprate effort, indicated that it
is necessary to remove steam generator (SG) tubes with laser welded
sleeves from service upon discovering an imperfection depth of 38.7%
wall thickness to ensure the structural integrity of SG tubes which
have been sleeved thereby precluding the occurrence of an SG tube
rupture of sleeved tubes under all operating conditions. The
previous laser welded sleeve plugging limit was based on an analysis
that used lower tolerance limit material strength values. The new
analysis methodology, required for laser welded sleeves, uses
minimum strength properties from the American Society of Mechanical
Engineers Code. As determined by the new analysis, reducing the
plugging limit from 40% to 38.7% maintains a comparable margin of
safety to the previous analysis.
Reanalysis of containment structural integrity under Design
Basis Accident (DBA) conditions indicated that the safety margin
improved, even though the mass and energy release due to a LOCA
under uprated power conditions increases. Based on new and improved
analytical methodologies, Pa, the peak calculated
containment internal pressure for the design basis LOCA, is 42.8
psig as compared to the current value of 47.8 psig for Unit 1; and
is 38.4 psig as compared to the current value of 44.4 psig for Unit
2, for both Byron Station and Braidwood Station.
Radiological consequences of the following accidents were
reviewed: Main Steamline Break, Locked Reactor Coolant Pump (RCP)
Rotor, Locked RCP Rotor with Power-Operated Relief Valve Failure,
Rod Ejection, Small Line Break Outside Containment, Steam Generator
Tube Rupture, Large Break Loss of Coolant Accident, Small Break Loss
of Coolant Accident, Waste Gas Decay Tank Rupture, Liquid Waste Tank
Failure, and Fuel Handling Accident. The resultant radiological
consequences for each of these accidents did not show a significant
change due to uprated power conditions and 10 CFR 100 limits
continue to be met.
The analyses supporting the power uprate program have
demonstrated that all systems and components are capable of safely
operating at uprated power conditions. All design basis accident
acceptance criteria will continue to be met. Therefore, it is
concluded that the proposed TS changes do not involve a significant
reduction in the margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
requested amendments involve no significant hazards consideration.
Attorney for licensee: Ms. Pamela B. Stroebel, Senior Vice
President and General Counsel, Commonwealth Edison Company, P.O. Box
767, Chicago, Illinois 60690-0767.
NRC Section Chief: Anthony J. Mendiola.
Connecticut Yankee Atomic Power Company, Docket No. 50-213, Haddam Neck
Plant
Date of amendment request: July 7, 2000.
Description of amendment request: The proposed amendment would add
a new license condition which would approve the License Termination
Plan dated July 7, 2000, and allow the licensee to make changes to the
approved License Termination Plan without prior Nuclear Regulatory
Commission (NRC) approval if certain
[[Page 77916]]
criteria specified in the license condition are met.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), Connecticut Yankee
Atomic Power Company (CYAPCO) has provided its analysis of the issue of
no significant hazards consideration, which is presented below:
CYAPCO has reviewed the proposed change to the Operating License
in accordance with the requirements of 10 CFR 50.92, ``Issuance of
Amendment,'' and concluded that the change does not involve a
significant hazards consideration (SHC). The proposed change does
not involve an SHC because the change would not:
1. Involve a significant increase in the probability or
consequences of an accident previously evaluated.
Currently, the bounding airborne radioactivity event given in
the Haddam Neck Plant UFSAR [Updated Final Safety Analysis Report]
is the resin container accident. Whereas previously doses associated
with gaseous waste system accidents would have bounded those
associated with solid waste system failures, the small amount of
radioactivity contained within the gaseous radioactive waste system
with the plant in the permanently defueled condition results in this
system's failure no longer being bounding. The curie content of the
resin container was based on the actual radioactivity inventory
collected on the resin from the reactor coolant system
decontamination. This corresponded to approximately 90% of the NRC
Class C burial limits. Consistent with NUREG-0782 for a resin fire,
one percent of the activity of the container was assumed to be
released to the environment. The 1% bounds the potential airborne
release fraction from various resin incidents, such as an exothermic
reaction during dewatering, dropping of a high integrity container,
or a resin spill. Other airborne particulate radwaste or radioactive
material accidents considered in the UFSAR but bounded by the resin
container fire are as follows:
a fire in the radwaste storage facility,
a drop of a component (e.g., steam generator, reactor
vessel, or heat exchanger) being removed from the site,
a van of radioactive waste materials consumed by a fire
while stored in the yard area on-site,
a radiological HEPA [High-Efficiency Particulate Air]
filter rupture,
segmentation of components or structures during loss of
local engineering controls,
an oxyacetylene tank explosion, or
an explosion of liquid propane gas leaked from a front-
end loader.
The UFSAR also discusses a fuel handling accident in the fuel
building, involving the drop of a spent fuel assembly onto the fuel
racks. The postulated drop assumes the rupture of all fuel rods in
the associated assembly. The probability or consequences of this
accident would not be increased during any future fuel transfer
operations in the spent fuel pool related to decommissioning.
Transfer of the spent fuel to canisters for dry cask storage will
involve additional restrictions contained in the cask certificate of
compliance in order to maintain decommissioning activities within
the assumption and consequences of the fuel handling accident.
The requested license amendment is consistent with plant
activities described in the Post Shutdown Decommissioning Activities
Report (PSDAR) and the HNP [Haddam Neck Plant] Decommissioning
UFSAR. Accordingly, no systems, structures, or components that could
initiate the previously evaluated accidents or are required to
mitigate these accident are adversely affected by this proposed
change. Therefore, the proposed change does not involve an increase
in the probability or consequences of any previously evaluated
accident.
2. Create the possibility of a new or different kind of accident
from any accident previously evaluated.
Accident analyses related to decommissioning activities are
addressed in the UFSAR. The requested license amendment is
consistent with the plant activities described in the HNP
Decommissioning UFSAR and the PSDAR. Thus, the proposed change does
not affect plant systems, structures, or components in a way not
previously evaluated. No new failure mechanisms will be created by
this activity, and the proposed activity does not create the
possibility of a new or different kind of accident than those
previously evaluated.
3. Involve a significant reduction in a margin of safety.
The License Termination Plan (LTP) is a plan for demonstrating
compliance with the radiological criteria for license termination as
provided in 10 CFR 20.1402. The margin of safety defined in the
statements of consideration for the final rule on the Radiological
Criteria for License Termination is described as the margin between
the 100 mrem/yr public dose limit established in 10 CFR 20.1301 for
licensed operation and the 25 mrem/yr dose limit to the average
member of the critical group at a site considered acceptable for
unrestricted use (one of the criteria of 10 CFR 20.1402). This
margin of safety accounts for the potential effect of multiple
sources of radiation exposure to the critical group. Since the
License Termination Plan was designed to comply with the
radiological criteria for license termination for unrestricted use,
the LTP supports this margin of safety.
In addition, the LTP provides the methodologies and criteria
that will be used to perform remediation activities of residual
radioactivity to demonstrate compliance with the ALARA [as low as
reasonably achievable] criterion of 10 CFR 20.1402.
Additionally, the LTP was designed with recognition that (a) the
methods in MARSSIM (Multi-Agency Radiation Survey and Site
Investigation Manual) and (b) the building surface contamination
levels are not directly applicable to use with complex nonstructural
components. Therefore, the LTP states that nonstructural components
remaining in buildings (e.g., pumps, heat exchangers, etc.) will be
evaluated against the criteria of RG [Regulatory Guide] 1.86 to
determine if the components can be released for unrestricted use.
The LTP also states that materials, surveyed and evaluated as a part
of normal decommissioning activities and prior to implementation of
the final status survey, will be surveyed for release using current
site procedures to demonstrate compliance with the ``no detectable''
criteria. Such materials that do not pass these criteria will be
controlled as contaminated.
Also, as previously discussed, the bounding accident for
decommissioning is the resin container accident. Since the bounding
decommissioning accident results in more airborne radioactivity than
can be released from other decommissioning events, the margin of
safety associated with the consequences of decommissioning accidents
is not reduced by this activity.
Thus, 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. Robert K. Gad, III, Ropes & Gray, One
International Plaza, Boston, Massachusetts 02110-2624.
NRC Section Chief: Michael T. Masnik.
Energy Northwest, Docket No. 50-397, WNP-2, Benton County, Washington
Date of amendment request: November 2, 2000.
Description of amendment request: The proposed amendment would
revise the WNP-2 Technical Specifications (TS) to incorporate long-term
power stability solution requirements. The proposed changes reflect:
(1) The addition of a new TS Section 3.3.1.3, ``Oscillation Power Range
Monitoring (OPRM) Instrumentation,'' (2) a revision to TS Section
3.4.1, ``Recirculation Loops Operating,'' to remove monitoring
specifications that would no longer be necessary upon activation of the
automatic OPRM instrumentation, and (3) a revision to TS 5.6.5 to
include in the Core Operating Limits Report (COLR) the applicable
operating limits for the OPRMs, and also reference the topical report
which describes the analytical methods used to determine the setpoint
values for the OPRM.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
[[Page 77917]]
1. The proposed change does not involve a significant increase
in the probability or consequences of an accident previously
evaluated.
The proposed change specifies limiting conditions for operation,
required actions and surveillance requirements for the OPRM system
and allows operation in regions of the power-to-flow map currently
restricted by the requirements of interim corrective actions (ICAs)
and certain limiting conditions of operation of Technical
Specification 3.4.1. The restrictions of the ICAs and Technical
Specification 3.4.1 were imposed to ensure adequate capability to
detect and suppress conditions consistent with the onset of thermal-
hydraulic oscillations that may develop into a thermal-hydraulic
instability event. A thermal-hydraulic instability event has the
potential to challenge the minimum critical power ratio (MCPR)
safety limit. The OPRM system can automatically detect and suppress
conditions necessary for thermal-hydraulic instability. With the
installation of the OPRM system, the restrictions of the ICAs and
Technical Specification 3.4.1 are no longer required to prevent a
potential challenge to the MCPR safety limit during an anticipated
instability event.
The probability of a thermal-hydraulic event is dependent on
power-to-flow conditions such that only during operation inside
specific regions of the power-to-flow map, in combination with power
shape and inlet enthalpy conditions, can the occurrence of an
instability event be postulated to occur. Operation in these regions
may increase the probability that operation with conditions
necessary for a thermal-hydraulic instability can occur. When the
OPRM system is operable, conditions consistent with the imminent
onset of oscillations are automatically detected and the conditions
necessary for oscillations are suppressed, which decreases the
probability of an instability event. In the event the trip
capability of the OPRM is not maintained, the proposed change limits
the period of time before an alternate method to detect and suppress
thermal-hydraulic oscillations is required. The probability of a
thermal-hydraulic instability event may be increased during the
limited period of time that operation is allowed at conditions
otherwise requiring the trip capability of the OPRM to be
maintained. However, since the duration of this period of time is
limited, the increase in the probability of a thermal-hydraulic
instability event is not significant.
The proposed change requires the OPRM system to be operable and,
thereby, ensures mitigation of thermal-hydraulic instability events
with a potential to challenge the MCPR safety limit when initiated
from anticipated conditions, by detection of the onset of
oscillations and actuation of an RPS [reactor protection system]
trip signal. The OPRM also provides the capability of an RPS trip
being generated for thermal-hydraulic instability events initiated
from unanticipated, but postulated conditions. These mitigating
capabilities of the OPRM system will become available as a result of
the proposed change and have the potential to reduce the
consequences of anticipated and postulated thermal-hydraulic
instability events. The OPRM installation has been evaluated and
does not alter the function or capability of any other installed
equipment such as the average power range monitoring (APRM) system
or the RPS to mitigate the consequences of postulated events.
Therefore, operation of WNP-2 in accordance with the proposed
amendment will not involve a significant increase in the probability
or consequences of an accident previously evaluated.
2. The proposed change does not create the possibility of a new
or different kind of accident from any accident previously
evaluated.
The proposed change specifies limiting conditions for operation,
required actions and surveillance requirements of the OPRM system
and allows operation in regions of the power-to-flow map currently
restricted by the requirements of ICAs and Technical Specification
3.4.1. The OPRM system uses input signals shared with APRM and rod
block functions to monitor core conditions and generate an RPS trip
when required. Quality requirements for software design, testing,
implementation and module self-testing of the OPRM system provide
assurance that no new equipment malfunctions due to software errors
are created. The design of the OPRM system also ensures that neither
operation nor malfunction of the OPRM system will adversely impact
the operation of other systems and no accident or equipment
malfunction of these other systems could cause the OPRM system to
malfunction or cause a different kind of accident.
Operation in regions currently restricted by the requirements of
ICAs and Technical Specification 3.4.1 is within the nominal
operating domain and ranges of plant systems and components, and
within the range for which postulated equipment and accidents have
been previously evaluated.
Therefore, operation of WNP-2 in accordance with the proposed
amendment will not create the possibility of a new or different kind
of accident from any accident previously evaluated.
3. The proposed change does not involve a significant reduction
in a margin of safety.
The proposed change specifies limiting conditions for operation,
required actions and surveillance requirements of the OPRM system
and allows operation in regions of the power-to-flow map currently
restricted by the requirements of ICAs and Technical Specification
3.4.1.
The OPRM system monitors small groups of LPRM [local power range
monitor] signals for indication of local variations of core power
consistent with thermal-hydraulic oscillations and generates an RPS
trip when conditions consistent with the onset of oscillations are
detected. An unmitigated thermal-hydraulic instability event has the
potential to result in a challenge to the MCPR [minimum critical
power ratio] safety limit. The OPRM system provides the capability
to automatically detect and suppress conditions which might result
in a thermal-hydraulic instability event and, thereby, maintains the
margin of safety by providing automatic protection for the MCPR
safety limit while significantly reducing the burden on the control
room operators. In the event the trip capability of the OPRM is not
maintained, the proposed change limits the period of time before an
alternate method to detect and suppress thermal-hydraulic
oscillation is required. The alternate method to detect and suppress
oscillations would be comparable to current actions required by the
interim corrective actions and no significant reduction in the
margin of safety would result in the event that an unmitigated
instability event occurred.
Operation in regions currently restricted by the requirements of
ICAs and Technical Specification 3.4.1 is within the nominal
operating domain and ranges of plant systems and components, and
within the range assumed for initial conditions considered in the
analysis of anticipated operational occurrences and postulated
accidents.
Therefore, operation of WNP-2 in accordance with the proposed
amendment will not involve a significant reduction in the margin of
safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 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: Thomas C. Poindexter, Esq., Winston &
Strawn, 1400 L Street, NW., Washington, DC 20005-3502.
NRC Section Chief: Stephen Dembek.
Entergy Operations, Inc., Docket No. 50-313, Arkansas Nuclear One, Unit
No. 1, Pope County, Arkansas
Date of amendment request: August 29, 2000.
Description of amendment request: The proposed changes to the
Arkansas Nuclear One, Unit 1 (ANO-1) Technical Specifications (TS)
provide for the use of an Alternate Repair Criteria (ARC) for steam
generator tubes with indications of outer diameter intergranular attack
(ODIGA) within the upper tube sheet region of the once-through steam
generators (OTSGs). Amendment 202 to the ANO-1 TS dated October 4,
1999, allowed the ARC for ODIGA indications only during Operating Cycle
16 at ANO-1. The proposed change would allow continued operation beyond
Cycle 16 for ANO-1 with OTSG tubes that have ODIGA indications that are
located in a defined area of the upper tube sheet.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
[[Page 77918]]
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 criteria in 10 CFR 50.92(c). A discussion
of these criteria 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 purpose of the periodic surveillance performed on the OTSGs
in accordance with ANO-1 Technical Specification (TS) 4.18 is to
ensure that the structural integrity of this portion of the reactor
coolant system will be maintained. The TS plugging limit of 40% of
the nominal tube wall thickness requires tubes to be repaired or
removed from service because the tube may become unserviceable prior
to the next inspection. Unserviceable is defined in the TS as the
condition of a tube if it leaks or contains a defect large enough to
affect its structural integrity in the event of an operating basis
earthquake, a loss-of-coolant accident, or a steam line or feedwater
line break. The proposed TS change allows OTSG tubes with ODIGA
indications contained within a defined area of the UTS [upper tube
sheet] to remain in service with existing degradation exceeding the
existing 40% through-wall (TW) plugging limit.
Extensive testing and plant experience has illustrated that
ODIGA flaws confined to this area within the OTSG will not result in
tube burst and tube leakage is unlikely. Therefore, allowing ODIGA
flaws in this specific region to remain in service will not alter
the conditions assumed in the current ANO-1 accident analysis for
OTSG tube failures under postulated accident conditions. In
addition, the condition of the OTSG tubes in this region are
monitored during regular inspection intervals to assess for evidence
of growth. Any growth noted will be addressed through testing and
the operational assessment. Therefore, ANO-1 has determined that the
identification, testing, monitoring, assessment, and corrective
action programs provided in ANO [Arkansas Nuclear One] Engineering
Report No. 00-R-1005-01, sufficiently supports this change request.
Application of the ODIGA alternate repair criteria will allow
leaving tubes with ODIGA indications found in the defined area of
the UTS in service while ensuring safe operation by monitoring and
assessing the present and future conditions of the tubes. ANO-1 has
operated since 1984 with ODIGA affected tubes in service with no
appreciable effect on structural integrity or indications of tube
leakage from ODIGA sources within the UTS. Through the inspection,
testing, monitoring, and assessment program previously mentioned,
and the on-line leak detection capabilities available during plant
operation, continued safe operation of ANO-1 is reasonably assured.
Therefore, the application of the ODIGA alternate repair
criteria 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 implementation of the ODIGA alternate repair criteria will
not result in any failure mode not previously analyzed. The OTSGs
are passive components. The intent of the TS surveillance
requirements are being met by these proposed changes in that
adequate structural and leak integrity will be maintained.
Additionally, the proposed change does not introduce any new modes
of plant operation.
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 application of an alternate repair criteria for ODIGA
provides adequate assurance with margin that ANO-1 steam generator
tubes will retain their integrity under normal and accident
conditions. The structural requirements of ODIGA affected tubes have
been evaluated satisfactorily and meet or exceed regulatory
requirements. Leakage rates for these tubes within the defined
region of the upper tubesheet are essentially zero and are
reasonably assured to remain within the assumptions of the accident
analysis by proper application of the ODIGA alternate repair
criteria program. Assuming high differential pressures following an
ATWS [Anticipated Transient Without Scram] or MSLB [Main Steam Line
Break], if the ODIGA patches leak, the leakage would be less than
the normal makeup capacity of the reactor coolant system. Since no
appreciable impact is evidenced on the tubes structural integrity or
its resulting leak rate, the margin to safety remains effectively
unaltered.
Therefore, this 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: 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 No. 50-313, Arkansas Nuclear One, Unit
No. 1, Pope County, Arkansas
Date of amendment request: September 28, 2000.
Description of amendment request: The proposed changes to the
Arkansas Nuclear One, Unit 1 (ANO-1) technical specifications revise
the safety-related 4160 Volt (V) bus loss-of-voltage and 480 V bus
degraded voltage relay allowable values.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
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 two 4160 V vital bus loss-of-voltage protection relays that
are provided on each of the 4160 V safety buses act to mitigate the
consequences of an accident by detecting a loss of voltage,
isolating the safety buses, initiating load shedding schemes, and
starting the associated emergency diesel generator (EDG). The safety
function of the relays is unchanged by the proposed setpoint
revisions. The revised settings for the loss-of-voltage protection
relays will continue to provide the safety function with no
appreciable additional time delay. The proposed time delays are
within those assumed in the ANO-1 safety analyses. Additionally, the
lower voltage settings will aid in preventing unnecessary isolation
from the off-site power sources, which in turn will reduce the
probability of a loss of off-site power to the unit due to off-site
power system transients. Since the proposed change does not
adversely impact the mitigating function of the relays, the
consequences of an accident previously evaluated remains unchanged.
The two degraded voltage protection relays that are provided on
each of the 480 V safety buses act to mitigate the consequences of
an accident by detecting a sustained undervoltage condition,
isolating the safety buses from offsite power, and starting the
associated EDG. This safety function is unchanged by the proposed
setpoint revisions. The revised settings for the degraded voltage
protection relays will continue to provide the safety function of
protecting the associated Class IE equipment from the effects of a
low voltage condition. There is no proposed change to the existing
timer setting and the time delays remain within those assumed in the
ANO-1 safety analyses. Additionally, the revised allowable voltage
settings will not result in any unnecessary isolation from the off-
site power sources. Since the proposed change does not adversely
impact the mitigating function of the relays, the consequences of an
accident previously evaluated remains unchanged.
The ANO-1 technical specifications will continue to require the
4160 V bus loss-of-voltage functions and 480 V bus degraded voltage
functions to be surveillance tested at their present frequency
without changing the modes in which the surveillance is required or
the modes of applicability for these components. The technical
specifications will continue to require the same actions as
currently exist for the inoperability of one or more of the 4160 V
bus loss-of-voltage channels or the 480 V bus degraded voltage
channels.
[[Page 77919]]
Therefore, this change 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 introduces no new modes of plant operation
or new plant configuration that could lead to a new or different
kind of accident from any previously evaluated being introduced. The
4160 V vital bus loss-of-voltage protection relays are required to
operate following a complete loss of off-site power to initiate the
bus power source transfer to on-site power, i.e., the EDGs, to
prevent a loss of all AC power. Likewise, the 480 V bus degraded
voltage relays are required to operate upon detection of a sustained
undervoltage condition to protect the Class IE components from
damage from low voltage by initiating transfer of the 4160 V safety
bus power source to the EDG. These safety functions are unchanged by
the proposed setpoint revisions, and the proposed setpoints continue
to provide the required actions consistent with the ANO-1 safety
analysis.
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 two undervoltage relays located on each 4160 V safety bus
are provided to detect loss-of-voltage, isolate the safety buses,
initiate load shedding, and start the EDGs. The two undervoltage
relays located on each 480 V safety bus are provided to detect
sustained undervoltage, isolate the safety buses, and start the
EDGs. These safety functions are unchanged by the proposed setpoint
revisions. The proposed changes to the allowable values for both
loss-of-voltage and degraded voltage relays incorporate channel
uncertainties and calibration tolerances, while fully meeting their
required safety functions of loss-of-voltage and degraded voltage
protection without resulting in undesired tripping of the offsite
power source.
The lower loss-of-voltage values do not affect the margin of
safety since there is no appreciable time difference in reaching the
lower setpoints during a loss-of-voltage event. The maximum proposed
time delay allowable value with the minimum loss-of-voltage relay
allowable value is within that used in the ANO-1 safety analysis.
The revised allowable values for the loss-of-voltage relays will
continue to provide the safety function with no appreciable
additional time delay. Additionally, the lower voltage settings will
help to prevent unnecessary isolation from the off-site power
sources due to off-site perturbations in the electrical grid, and
thus contribute to increasing the margin of safety. Also, the
slightly higher range of allowable values for the degraded voltage
settings allows enhanced protection of the Class IE components, but
does not result in undesired tripping of the offsite power source
for the analyzed grid minimum normal condition. The degraded voltage
relays, therefore, also act to contribute to an increased margin of
safety.
Therefore, this change does not involve a significant reduction
in the margin of safety.
Therefore, based upon 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 Operations, Inc., Docket No. 50-313, Arkansas Nuclear One, Unit
No. 1, Pope County, Arkansas
Date of amendment request: September 28, 2000.
Description of amendment request: The proposed changes to Arkansas
Nuclear One, Unit 1 (ANO-1), Technical Specifications (TS) provide for
the implementation of a revised reroll repair process for ANO-1 Once-
Through Steam Generators (OTSG). The current TSs limit application of
the reroll repair process to repair tubes with defects in the upper
tubesheet area only, using a 1 inch roll length, and allow the reroll
repair process to be performed only once per steam generator tube. The
requested amendment would allow the reroll repair process to be used
multiple times for a single tube and would allow the repairs in both
the upper and lower tubesheets.
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 criteria in 10 CFR 50.92(c).
OTSG tubesheet areas where reroll installation is excluded are
specified in Appendix A of topical report BAW-2303P, Revision 4 [A
non-proprietary version of the report, BAW-2303NP, Revision 4,
``OTSG Repair Roll Qualification Report,'' was submitted on October
26, 2000.]. The following discussion applies to areas of the OTSG
tubesheets where installation of reroll repairs is permitted:
Criterion 1--Does Not Involve a Significant Increase in the
Probability or Consequences of an Accident Previously Evaluated.
Two types of repair rolls have been developed for installation
in the OTSGs, a single 1-inch roll expansion and an overlapping roll
consisting of two 1-inch roll expansions. The overlapping roll
provides a minimum of 1\5/8\ inch effective roll expansion. There is
an additional \1/4\-inch roll transition region on each end of the
roll expansion and a new leak-limiting pressure boundary is created
by the repair roll. Applicable OTSG transient conditions were
evaluated to develop a set of bounding test conditions for
application to both types of repair rolls. Testing included
examination of the effects of crevice deposits, cyclic loading, tube
yield strength, differential dilations, axial loads and internal
pressure.
Test results conclude that the single 1-inch minimum repair roll
is structurally adequate to prevent tube slip during all non-faulted
operating transients. A small amount of slippage is acceptable
provided the tube does not slip out of the tubesheet and tube bow
due to post-faulted transient heatup does not result in tube
failure. Exclusion areas are established in the tubesheets to
provide assurance that tube will not slip out of the tubesheet . The
1\5/8\ inch minimum overlapping roll is structurally acceptable
based on the bounding evaluation of the single 1-inch repair roll.
Bounding leak rates are applied based on tubesheet depth and
radial position. A post-slip leak rate is applied to any location
where there is potential for repair roll slip during a postulated
accident. The bounding leak rates are very conservative because the
leakage is based on test samples with a full circumferential sever
outboard of the repair roll. The majority of the degradation in the
tubesheets is comprised of short, axial cracks for which the leakage
would be much less under axial tensile loads than for the tested
severed tube. In addition, repair rolls will actually slip only if
the tube is severed outboard of the repair roll. Since the majority
of the degradation in the region of the roll joints has been
identified as small axial cracks, the probability of the repair roll
maintaining structural integrity is very high and the potential for
a joint to slip is very low. The leakage from each repair roll that
serves as a pressure boundary is added to the leakage from all other
sources and the total leakage must be within current accident
analysis limits.
The application of the reroll repair process as described in
topical report BAW-2303P, Revision 4 will not alter the conditions
assumed in the current ANO-1 accident analysis for OTSG tube
failures under postulated accident conditions. In addition, the
condition of the OTSG tubes in this region are monitored during
regular inspection intervals to assess for evidence of degradation.
Any degradation noted will be addressed in the operational
assessment and appropriate actions taken.
Therefore, this change does not involve a significant increase
in the probability or consequences of any accident previously
evaluated.
Criterion 2--Does Not Create the Possibility
[[Page 77920]]
of a New or Different Kind of Accident from any Previously
Evaluated.
The reroll process establishes a new pressure boundary for the
associated tube in the tubesheet region inboard of the flaw. The new
roll transition may eventually develop primary water stress
corrosion cracking (PWSCC) and require additional repair. Industry
experience with roll transition cracking has shown that PWSCC in
roll transitions are normally short axial cracks, with extremely low
leak rates. The standard MRPC eddy current inspection during the
refueling outages have proven to be successful in detecting these
defects.
In the unlikely event the rerolled tube failed and severed
completely at the heel transition of the reroll region, the tube
would retain engagement in the tubesheet bore, preventing any
interaction with neighboring tubes. In this case, leakage is
minimized and is well within the assumed leakage of the design basis
tube rupture accident. In addition, the possibility of rupturing
multiple steam generator tubes is unaffected. 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 repair roll is applicable to repairing axial, volumetric, or
circumferential indications. Testing was conservatively performed
with the assumption that the tube is severed at the heel transition
(360 degree and 100% through-wall circumferential defect). The joint
strength margin (actual load/limiting load) was calculated for each
tubesheet depth and radial position for the cooldown transient to
ensure margin against slip for non-faulted conditions. All locations
showed a joint strength margin less than 0.65 with an acceptable
margin less than 1.0.
A tube with degradation can be kept in service through the use
of the reroll process. The new roll expanded interface created with
the tubesheet satisfies all of the necessary structural and leakage
requirements. Since the joint is constrained within the tubesheet
bore, there is no additional risk associated with tube rupture.
Therefore, the analyzed accident scenarios remain bounding, and the
proposed modifications to the reroll process do not reduce the
margin of safety.
Therefore, based upon the reasoning presented above and the
previous discussion of the amendment request, Entergy Operations 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.
FirstEnergy Nuclear Operating Company, Docket No. 50-440, Perry Nuclear
Power Plant, Unit 1, Lake County, Ohio
Date of amendment request: November 9, 2000.
Description of amendment request: The proposed amendment requests
fourteen of the simpler, generic administrative/editorial/consistency
improvements agreed upon between the Nuclear Energy Institute Technical
Specification Task Force (TSTF) and the NRC, subsequent to the
conversion of the Perry Technical Specifications to the improved
Standard Technical Specifications. The proposed amendment requests
Perry-specific versions of TSTF 5, ``Delete Notification, Reporting,
and Restart Requirements if a Safety Limit is Violated;'' TSTF 32,
``Slow/Stuck Control Rod Separation Criteria;'' TSTF 38, ``Revise
Visual Surveillance of Batteries to Specify Inspection is for
Performance Degradation;'' TSTF 52, ``Implement 10 CFR Part 50,
Appendix J, Option B;'' TSTF 65, ``Use of Generic Titles for Utility
Positions;'' TSTF 104, ``Relocate to the Bases the Discussion of
Exceptions to Limiting Condition for Operation (LCO) 3.0.4;'' TSTF 106,
``Change to Diesel Fuel Oil Testing Program;'' TSTF 118,
``Administrative Controls Program Exceptions;'' TSTF 152, ``Revise
Reporting Requirements to be Consistent with 10 CFR Part 20;'' TSTF
153, ``Clarify Exception Notes to be Consistent with the Requirement
being Excepted;'' TSTF 166, ``Correct Inconsistency between LCO 3.0.6
and the Safety Functional Determination Program (SFDP) Regarding
Performance of an Evaluation;'' TSTF 258, ``Changes to Section 5.0,
Administrative Controls;'' TSTF 278, ``Battery Cell Parameters (LCO
3.8.6) includes more than Table 3.8.6-1 Limits;'' and TSTF 279,
``Remove the Words `Including Applicable Supports' from the Description
of the Inservice Testing Program.''
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration which is presented below:
1. This proposed amendment does not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
The proposed changes involve reformatting and rewording of the
existing Technical Specifications to be consistent with regulations
or other existing Technical Specifications, or the changes do not
involve a change in intent. The proposed changes also involve
Technical Specification requirements that are administrative rather
than technical in nature. As such, this change does not affect
initiators of previously evaluated events, or assumed mitigation of
accident or transient events. Therefore, this change does not
involve a significant increase in the probability or consequences of
an accident previously evaluated.
2. This proposed amendment 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 or different type of equipment will be installed) or
changes in methods governing normal plant operation. The proposed
changes will not impose new or eliminate old requirements on design
or operation of the plant. The administrative changes also do not
introduce new initiators of events. Thus, this change does not
create the possibility of a new or different kind of accident from
any accident previously evaluated.
3. This proposed amendment does not involve a significant
reduction in a margin of safety.
The proposed change has no impact on any safety analysis
assumptions or design basis margins. This change is administrative
in nature. The proposed changes will not impose new or eliminate old
requirements on design or operation of the plant. Therefore, the
change does not involve a significant reduction in a margin of
safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 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: Mary E. O'Reilly, Attorney, FirstEnergy
Corporation, 76 South Main Street, Akron, OH 44308.
NRC Section Chief: Anthony J. Mendiola.
GPU Nuclear, Inc., Three Mile Island Nuclear Station, Unit 2, Docket
No. 50-320, Dauphine County, Pennsylvania
Date of amendment request: July 25, 2000.
Description of amendment request: The proposed technical
specifications change request (TSCR) is to revise Three Mile Island
Nuclear Generating Station, Unit 2 (TMI-2), Technical Specification
(TS) Section 6.7.2 to eliminate a change associated with periodic
reviews of procedures. Currently, TS 6.7.2 states that required
procedures shall be reviewed periodically as required by American
National Standards Institute (ANSI) N18.7-1976 (a biennial review).
This TSCR proposes to revise the wording for TS 6.7.2 to be essentially
identical with the Three Mile Island, Unit 1 (TMI-1), TS requirements
for procedure reviews, which states that
[[Page 77921]]
required procedures shall be revised periodically, as set forth in
administrative procedures (currently a biennial review). This TSCR
would also be consistent with the TMI-2 Post-Defueling Monitored
Storage (PDMS) Quality Assuance (QA) Plan, which states that
``Procedural documentation shall be periodically reviewed for adequacy
as set forth in administrative procedures.''
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:
Applying the three standards set forth in 10 CFR 50.92, the
proposed changes to the Technical Specifications involve no
significant hazards consideration. The proposed changes would:
1a. Not involve a significant increase in the probability of an
accident previously evaluated because no accident initiators or
assumptions are affected. The proposed changes have no effect on any
plant systems. All Limited Conditions for PDMS and Safety Limits
specified in the Technical Specifications will remain unchanged.
1b. Not involve a significant increase in the consequences of an
accident previously evaluated because no accident conditions or
assumptions are affected. The proposed changes do not alter the
source term, containment isolation, or allowable radiological
consequences. The change in specified periodic procedure review
requirements will have no adverse effect on any plant system.
2. Not create the possibility of a new or different kind of
accident from any accident previously evaluated because no new
accident initiators or assumptions are introduced by the proposed
changes. The proposed changes have no direct effect on any plant
systems. The changes do not affect any system functional
requirements, plant maintenance, or operability requirements.
3. Not involve a significant reduction in the margin of safety
because the proposed changes do not involve significant changes to
the initial conditions contributing to accident severity or
consequences. The proposed changes have no direct effect on any
plant systems.
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: Ernest L. Blake, Jr., Esquire, Shaw,
Pittman, Potts & Trowbridge, 2300 N Street, NW., Washington, DC 0037.
NRC Section Chief: Michael T. Masnik.
Nuclear Management Company, LLC, Docket No. 50-331, Duane Arnold Energy
Center, Linn County, Iowa
Date of amendment request: October 16, 2000.
Description of amendment request: The proposed amendment would
incorporate new pressure and temperature (P/T) curves into the
Technical Specifications. The reactor pressure vessel P/T limit curves
would be updated for inservice leakage and hydrostatic testing, non-
nuclear heatup and cooldown, and criticality.
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 P/T [pressure and temperature] limits are
not derived from Design Basis Accident (DBA) analyses. They are
prescribed by the ASME [American Society of Mechanical Engineers]
Code and 10 CFR 50 Appendix G and H as restrictions on operation to
avoid encountering pressure, temperature, and temperature rate of
change conditions that might cause undetected flaws to propagate and
cause non-ductile failure of the reactor coolant pressure boundary.
The changes to the calculational methodology for the P/T limits
based upon Code Case N-640 continue to provide adequate margin in
the prevention of a non-ductile type fracture of the reactor
pressure vessel (RPV). The Code Case was developed based upon the
knowledge gained through years of industry experience. P/T curves
developed using the allowances of Code Case N-640 indeed yield more
operating margin. However, the experience gained in the areas of
fracture toughness of materials and pre-existing undetected defects
shows that some of the existing assumptions used for the calculation
of P/T limits are unnecessarily conservative and unrealistic.
Therefore, providing the allowances of the Code Case in developing
the P/T limit curves will continue to provide adequate protection
against non-ductile type fractures of the RPV.
The proposed change will not affect any other system or piece of
equipment designed for the prevention or mitigation of previously
analyzed events. The change does not adversely affect the integrity
of the reactor coolant system such that its function in control of
radiological consequences is affected.
(2) The proposed amendment will not create the possibility of a
new or different kind of accident from any accident previously
evaluated. The amendment will revise the P/T curves which are
established to the requirements of 10 CFR 50, Appendix G to assure
that non-ductile fracture of the reactor vessel is prevented.
The proposed change provides more operating margin in the P/T
limit curves for inservice leakage and hydrostatic pressure testing,
non-nuclear heatup and cooldown, and criticality, with benefits
being primarily realized during the pressure tests. The proposed
change does not result in any new or unanalyzed operation of any
system or piece of equipment important to safety, and as a result,
the possibility of a new type event is not created.
(3) The proposed amendment will not involve a significant
reduction in a margin of safety. 10 CFR 50, Appendix G specifies
fracture toughness requirements to provide adequate margins of
safety during operation over the service lifetime. The values of
adjusted reference temperature and upper shelf energy are expected
to remain within the limits of Regulatory Guide 1.99, Revision 2 and
Appendix G of 10 CFR 50 (less than 200 degrees F and greater than 50
ft-lbs respectively) for at least 32 effective full power years
(EFPY) of operation.
The proposed change reflects an update of P/T curves based on
the latest ASME guidance. The revised P/T curves provide more
operating margin and thus, more operational flexibility than the
current P/T curves. With the increased operational margin, a
reduction in the safety margin results with respect to the existing
curves. However, industry experience since the inception of the P/T
limits in 1974 confirms that some of the existing methodologies used
to develop P/T curves are unrealistic and unnecessarily
conservative. Accordingly, ASME Code Case N-640 takes into account
the acquired knowledge and establishes more realistic methodologies
for the development of P/T curves. Therefore, operational
flexibility is gained and an acceptable margin of safety to RPV non-
ductile type fracture is maintained.
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: Al Gutterman, Morgan, Lewis & Bockius, 1800
M Street, NW., Washington, DC 20036-5869.
NRC Section Chief: Claudia M. Craig.
Nuclear Management Company, LLC, Docket No. 50-305, Kewaunee Nuclear
Power Plant, Kewaunee County, Wisconsin
Date of amendment request: November 29, 1999, as supplemented on
November 10, 2000.
Description of amendment request: The licensee submitted a proposed
amendment to Kewaunee Nuclear Plant's Technical Specifications (TSs)
modifying the TSs to incorporate requested changes per Generic Letter
99-02, ``Laboratory Testing of Nuclear-Grade Activated Charcoal,''
dated June 3, 1999.
[[Page 77922]]
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 Shield Building Ventilation, the Auxiliary Building
Ventilation, the Spent Fuel Pool Sweep Systems and the Control Room
Post Accident Recirculation System are not accident initiators.
Therefore, the proposed change will not increase the probability of
an accident. The purpose of each of these systems is to mitigate the
consequences of an accident once it has occurred. Based upon a
comparison, the later version of ASTM D3803, ASTM D3803-89 was found
to test the efficiency of the charcoal material under more
conservative conditions. By testing the charcoal absorber material
under more conservative conditions, the charcoal will require
replenishment sooner. Therefore, the consequences will not be
increased.
The changes to the basis sections are to promote clarity and
uniformity. These statements were previously contained in the basis
section or clarify which revision of Regulatory Guide 1.52 that
should be used. This change provides acceptable guidelines for the
qualification of replacement charcoal absorbent. Therefore, these
changes will not 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.
This amendment request does not change any component at the
plant. It is changing the testing requirements for material already
installed. The material being tested has not changed. By testing the
charcoal material under this revised protocol the material will be
replaced with fresh charcoal sooner. This will ensure the equipment
performs as described in the USAR.
The changes to the basis sections are to promote clarity and
uniformity. These statements were previously contained in the basis
section or clarify which revision of Regulatory Guide 1.52 that
should be used. This change provides acceptable guidelines for the
qualification of replacement charcoal adsorbent. Therefore, these
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.
There is no reduction in the margin of safety. The efficiency of
the charcoal material assumed by the USAR will not change as a
result of this amendment and the functioning of the system will not
change. Therefore, the original margin of safety is maintained.
The changes to the basis sections are to promote clarity and
uniformity. These statements were previously contained in the basis
section or clarify which revision of Regulatory Guide 1.52 that
should be used. This change provides acceptable guidelines for the
qualification of replacement charcoal adsorbent. Therefore, these
changes will not involve a significant reduction in a margin of
safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: Bradley D. Jackson, Esq., Foley and Lardner,
P.O. Box 1497, Madison, WI 53701-1497.
NRC Section Chief: Claudia M. Craig.
Nuclear Management Company, LLC, Docket No. 50-305, Kewaunee Nuclear
Power Plant, Kewaunee County, Wisconsin
Date of amendment request: October 12, 2000.
Description of amendment request: In accordance with 10 CFR 50.59,
the topical report WPSRSEM-NP, ``Reload Safety Evaluation Methods for
Application to Kewaunee,'' Revision 3, is being submitted for the
staff's review and approval since the licensee determined the revision
of the report involved an unreviewed safety question. The topical
report is intended to be applicable to Kewaunee reload cycles after and
including Cycle 25, presently scheduled to commence in the fall of
2001. The topical report reflects:
Editorial changes, including corrections to the limiting
directions of core physics parameters and clarification of the
definition of core physics parameters.
Changes made to incorporate the CONTEMPT code for
containment analysis. CONTEMPT is currently described for this purpose
in the Kewaunee updated safety analysis report (USAR).
The adoption of the GOTHIC code for containment analysis.
Changes in Reload Safety Evaluation Methods due to Large
Break Loss-of-Coolant Accident Upper Plenum Injection Analysis.
The adoption of RETRAN-3D for use in the 2D mode for
system analysis.
The extension of the VIPRE-01 code to reflect changes in
fuel design.
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.
Analysis methods are not accident initiators, therefore, changes
in analysis methods will not increase significantly the probability
of occurrence or the consequences of an accident previously
evaluated.
The changed analysis methods are conservative and conform to
industry standards for analysis methods that are applied to design
basis safety analyses. Benchmark analyses have demonstrated good
agreement between the changed analysis methods and the current
analysis of record (AOR) methods. The safety analysis results using
the changed analysis methods are shown to satisfy all applicable
design and safety analysis acceptance criteria. The demonstrated
adherence to safety analysis acceptance criteria precludes new
challenges to components and systems that could adversely affect the
ability of existing components and systems to mitigate the
consequences of any accident or adversely affect the integrity of
any fission product barrier.
Analysis methods changes will not impact plant equipment
important to safety. Equipment important to safety will continue to
operate within its design capabilities. The analysis methods changes
also do not affect the plant configuration or the overall plant
performance capabilities.
Therefore, the changes will not increase probability or the
consequences of an accident previously evaluated.
(2) Create the possibility of a new or different kind of
accident from any previously evaluated.
The proposed change is a change to the analysis methods, which
are applied to Kewaunee. Analysis methods are not accident
initiators. The changed analysis methods are applied to the
accidents that are the established design basis accidents for
Kewaunee. Analysis methods changes will not impact plant equipment
important to safety. Equipment important to safety will continue to
operate within its design capabilities. The analysis methods changes
also do not affect the plant configuration or the overall plant
performance capabilities.
As demonstrated by the benchmark reports the methodologies
provide a more accurate but still conservative representation of
expected plant response following a design basis accident. Since the
new methodologies are conservative with respect to actual expected
plant response the changes will not create the possibility of an
accident of a different type than any previously evaluated.
(3) Involve a significant reduction in the margin of safety.
The proposed changes are changes to the analysis methods, which
are applied to Kewaunee design basis safety analyses. The revised
analysis methods have been verified through benchmark analyses
against the current Analysis of Record methods. The analysis methods
are conservative and appropriate for application to Kewaunee design
basis analyses. Safety analysis acceptance criteria are satisfied
when the changed analysis methods are applied to the Kewaunee design
basis safety analyses. Demonstrated adherence to safety analysis
acceptance criteria using the new analysis methods assures that
Technical Specification limits will be satisfied during operation
with the changed analysis methods.
[[Page 77923]]
Therefore, the margin of safety as defined in the basis of any
Technical Specification will not be reduced significantly because of
these changes.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: Bradley D. Jackson, Esq., Foley and Lardner,
P.O. Box 1497, Madison, WI 53701-1497.
NRC Section Chief: Claudia M. Craig.
Nuclear Management Company, LLC, Docket No. 50-305, Kewaunee Nuclear
Power Plant, Kewaunee County, Wisconsin
Date of amendment request: November 10, 2000.
Description of amendment request: The proposed amendment is to
revise several sections of the Kewaunee Nuclear Power Plant (KNPP)
Technical Specifications (TSs). These sections include administrative
changes, Table 4.1-1, and Sections 1.0, 6.4, and 6.10.
Administrative changes are submitted with this proposed amendment
to correct minor typographical errors in the Table of Contents and
among these changes are renumbering the index section pages and the
addition of previously omitted sections.
The proposed changes will modify TS Table 4.1-1, ``Minimum
Frequencies for Checks, Calibrations and Test of Instrument Channels.''
This proposed change will decrease the calibration frequency for
Turbine First Stage Pressure to support KNPP's 18-month operating
cycle, and modify the table to eliminate a note that could lead to non-
conservative calibration frequency.
The proposed TS Section 1.0, ``Definitions,'' will incorporate a
line item improvement to provide additional clarification on channel
calibration.
The proposed TS Section 6.4, ``Training,'' will remove the title of
director for the KNPP training program and relocate the title reference
to the Operational Quality Assurance Program Description (OQAPD).
The proposed TS Section 6.10, ``Record Retention,'' will revise the
off-site review committee title.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. Involve a significant increase in the probability or
consequences of an accident previously evaluated.
Table of Contents
The proposed changes are administrative in nature and,
therefore, have no impact on accident initiators or plant equipment,
and thus do not affect the probability or consequences of an
accident.
TS Section 1.0, ``Definitions''
A calibration will continue to ensure that a channel is within
specification. Furthermore, calibration methodology is not an
accident initiator. Therefore, the proposed change will not
significantly raise the probability or consequences of an accident
previously evaluated.
TS Table 4.1-1
The proposed change amends the calibration interval of the
turbine first stage pressure from 12 months to each refueling cycle
to coincide with KNPP's operating cycle. Calibration frequency would
not change the consequence of a failure of the first stage pressure
channel. Calibration frequency is not an accident initiator.
Therefore, the proposed changes will not significantly raise the
probability or consequences of an accident previously evaluated.
Additionally, this change is consistent with the turbine first stage
pressure calibration frequency stated in STS.
The proposed changes to the identified line items in Table 4.1-1
will require calibration of the instruments on a refueling cycle
interval without exception. These calibration frequencies are not
accident initiators and thus do not affect the probability of an
accident. These changes are more conservative than existing TS and,
therefore, will not increase the consequences of an accident.
TS Section 6.4, ``Training''
The proposed change will not change the intent of the TS.
Removing the title from the TS is administrative in nature and,
therefore, has no impact on accident initiators or plant equipment,
and thus does not affect the probability or consequences of an
accident.
TS Section 6.10, ``Record Retention''
The proposed change will not change the intent of the TS.
Changing the title of the off-site review committee is
administrative in nature and, therefore, has no impact on accident
initiators or plant equipment, and thus does not affect the
probability or consequences of an accident.
2. Create the possibility of a new or different kind of accident
from any accident previously evaluated.
Table of Contents
The proposed changes do not involve changes to the physical
plant or operations. Since these administrative changes do not
contribute to accident initiation, they do not produce a new
accident scenario or produce a new type of equipment malfunction.
Also, these changes do not alter any existing accident scenarios;
they do not affect equipment or its operation, and thus, do not
create the possibility of a new or different kind of accident.
TS Section 1.0, ``Definitions''
The proposed TS change to channel calibration will not introduce
any new equipment or result in existing equipment functioning
differently from that previously evaluated in the USAR or TS.
Calibration will continue to ensure that the channel is within
specification and capable of performing its design basis function.
No new accident is introduced and no safety-related equipment or
safety functions are altered. Therefore, the proposed change does
not affect any of the parameters or conditions that contribute to
initiation of any accident.
TS Table 4.1-1
The proposed TS change will not introduce any new equipment or
result in existing equipment functioning differently from that
previously evaluated in the USAR or TS. The proposed change amends
the calibration interval of the turbine first stage pressure from 12
months to each refueling cycle to coincide with KNPP's operating
cycle. Performing the surveillance during refueling will decrease
the likelihood for an induced transient. Expanding the calibration
frequency will not affect the performance of the first stage
pressure channel. A review of turbine first stage pressure
calibration results for the last three years concluded no adjustment
of the instrument was necessary due to little or no drift.
Furthermore, similar transmitters already calibrated on a refueling
basis have remained within acceptable limits. These results indicate
stable instrument performance to support extending calibration
frequency from 12 months to each refueling cycle.
The proposed changes will ensure that the affected channels are
calibrated on a refueling basis. These changes will not introduce
any new equipment or result in existing equipment functioning
differently from that previously evaluated in the USAR or TS. No new
accident is introduced and no safety-related equipment or safety
functions are altered. The proposed changes do not affect any of the
parameters or conditions that contribute to initiation of any
accident.
TS Section 6.4, ``Training''
The proposed change does not involve a change to the physical
plant or operations. Since an administrative change does not
contribute to accident initiation, it does not produce a new
accident scenario or produce a new type of equipment malfunction.
TS Section 6.10, ``Record Retention''
The proposed change does not involve a change to the physical
plant or operations. Since an administrative change does not
contribute to accident initiation, it does not produce a new
accident scenario or produce a new type of equipment malfunction.
3. Involve a significant reduction in the margin of safety.
Table of Contents
Administrative changes do not involve a significant reduction in
the margin of safety. The proposed changes do not affect plant
equipment or operation. Safety limits and limiting safety system
settings are not affected by these changes.
[[Page 77924]]
TS Section 1.0, ``Definitions''
Operation of the facility in accordance with this proposed TS
change would not involve a significant reduction in a margin of
safety. The specification will still ensure the operability of
channels requiring calibration.
TS Table 4.1-1
Operation of the facility in accordance with the proposed TS
changes would not involve a significant reduction in the margin of
safety. The calibration will continue to verify the operability of
the turbine first stage pressure channels. Therefore, the proposed
change does not involve a significant reduction in the margin of
safety.
Operation of the facility in accordance with the proposed TS
changes would not involve a significant reduction in a margin of
safety. The proposed changes will ensure the continued reliability
of the instruments. This change is more conservative than existing
TS and is consistent with STS.
TS Section 6.4, ``Training''
Administrative changes do not involve a significant reduction in
the margin of safety. The proposed change does not affect plant
equipment or operation. Safety limits and limiting safety system
settings are not affected by this change.
TS Section 6.10, ``Record Retention''
Administrative changes do not involve a significant reduction in
the margin of safety. The proposed change does not affect plant
equipment or operation. Safety limits and limiting safety system
settings are not affected by this change.
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: Bradley D. Jackson, Esq., Foley and Lardner,
P.O. Box 1497, Madison, WI 53701-1497.
NRC Section Chief: Claudia M. Craig.
PPL Susquehanna, LLC, Docket No. 50-388, Susquehanna Steam Electric
Station, Unit 2, Luzerne County, Pennsylvania
Date of amendment request: March 20, 2000.
Description of amendment request: The proposed amendment would
revise the Susquehanna Steam Electric Station (SSES), Unit 2, Technical
Specification 2.1.1.2, minimum critical power ratio (MCPR) safety
limits. These safety limits are being revised to reflect planned
changes to the core composition for the next operating cycle and to
support a separate license amendment proposing an increase in the SSES,
Unit 1 and 2, rated thermal power.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. The proposed change does not involve a significant increase
in the probability or consequences of an accident previously
evaluated.
The proposed changes in MCPR Safety Limits do not affect any
plant system or component (except the reactor core) and therefore
does not increase the probability of an accident previously
evaluated.
A Unit 2 Cycle 11 MCPR Safety Limit analysis was performed for
PPL by SPC [Siemens Power Corporation]. This analysis used NRC
approved methods as required by SSES Technical Specifications. For
Unit 2 Cycle 11 [U2C11], the critical power performance of the
ATRIUMTM-10 fuel was determined using the NRC approved
ANFB-10 correlation. Also, the analysis for U2C11 supports a Core
Thermal Power of 3493 MWt which is a 1.5% increase over U2C10 (3441
MWt). The Safety Limit MCPR calculations statistically combine
uncertainties on feedwater flow, feedwater temperature, core flow,
core pressure, core power distribution, and uncertainties in the
Critical Power Correlation. The SPC analysis used cycle specific
power distributions and calculated MCPR values such that at least
99.9% of the fuel rods are expected to avoid boiling transition
during normal operation or anticipated operational occurrences. The
resulting two-loop and single-loop MCPR Safety Limits are included
in the proposed Technical Specification change. Thus, the cladding
integrity and its ability to contain fission products are not
adversely affected. It is therefore concluded that the proposed
change does not increase the consequences of an accident previously
evaluated.
2. The proposed change does not create the possibility of a new
or different kind of accident from any accident previously
evaluated.
As discussed above, the proposed changes to the Unit 2 Technical
Specifications (MCPR Safety Limits) do not affect any plant system
or component and do not affect plant operation. The consequences of
transients and accidents will remain within the criteria approved by
the NRC. Therefore, the proposed change does not create the
possibility of a new or different kind of accident from any accident
previously evaluated.
3. The proposed change does not involve a significant reduction
in a margin of safety.
Since the proposed changes do not affect any plant system or
component, and do not have any impact on plant operation, the
proposed changes will not affect the function or operation of any
plant system or component. The consequences of transients and
accidents will remain within the criteria approved by the NRC. The
proposed MCPR Safety Limits do not involve a significant reduction
in the margin of safety as currently defined in the bases of the
applicable Technical Specification sections. 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: Bryan A. Snapp, Esquire, Assoc. General
Counsel, PPL Services Corporation, 2 North Ninth St., GENTW3,
Allentown, PA 18101-1179.
NRC Section Chief: Marsha Gamberoni.
Southern California Edison Company, et al., Docket No. 50-206, San
Onofre Nuclear Generating Station, Unit 1, San Diego County, California
Date of amendment requests: October 30, 2000-PCN 268.
Description of amendment requests: This amendment application
requests to delete license condition 2.C(3) related to fuel
transshipments between San Onofre Nuclear Generating Station, Unit 1
(SONGS 1), which is in the process of decommissioning, and SONGS Units
2 or 3 since such transshipments will no longer be made. In addition,
the amendment application requests revisions to the Unit 1 defueled
Technical Specifications to (1) remove the spent fuel pool (SFP)
temperature limits and related cooling system operability requirements,
(2) remove the SFP auxiliary feedwater storage tank makeup water
requirements and related surveillance requirements, (3) change the SFP
water level limit for conditions other than spent fuel movement, and
(4) change the operator staffing requirements for the decommissioning
control room. As a result of these proposed changes, the licensee also
proposes to delete the definitions of FUNCTIONAL and SPENT FUEL POOL
COOLING TRAIN and revise the table of contents and list of tables
according to the above 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?
No. The proposed change is a request to revise the San Onofre
Nuclear Generating Station, Unit 1 (SONGS 1) license and permanently
defueled technical specifications. The license condition for
transshipment is being deleted since there is no safety-related
equipment to protect and no plans for transshipment of SONGS 1 fuel
to SONGS 2 or 3. Since the purpose of removing this license
condition is that this activity will
[[Page 77925]]
no longer be performed, there is no impact on accident probability
or consequences. Deleting the technical specifications for spent
fuel pool temperature and makeup are based on the current benign
status of the spent fuel and spent fuel pool. The requirements and
surveillances provided by these technical specifications no longer
provide appropriate limits for the safe storage of the spent fuel.
The spent fuel temperature limit cannot be reached. Makeup water is
available from various sources onsite and offsite in a timely
manner. Deleting these technical specifications has no impact on the
probability or consequences of an accident. Modifying the spent fuel
pool water level requirements provides two levels for maintaining
water: One water level (elevation 28' [feet]) for just storage and a
higher water level (elevation 40' 3" [inches]) for fuel movement.
Lowering the water level for storage of spent fuel does not affect
the accident probability. The fuel handling accident will not occur
when the pool water level is at elevation 28 feet since spent fuel
will only be handled when the pool water level is at elevation 40'
feet 3". Removing the restrictions for having one individual of the
minimum shift crew located in the control room will not have any
impact on the fuel handling accident since a certified fuel handler
is still required to be present.
Therefore, this change does not involve an increase in the
probability or consequences of an accident previously evaluated.
2. Create the possibility of a new or different type of accident
from any accident previously evaluated?
No. This proposed change is a request to revise the SONGS 1
license and permanently defueled technical specifications. The
transshipment license condition is being deleted since there is no
safety-related equipment to protect and no plans for transshipment
of Unit 1 spent fuel to Units 2 or 3. The technical specifications
for spent fuel pool temperature and makeup are being deleted since
these requirements no longer provide limits appropriate for
maintaining the spent fuel pool. Removing these requirements does
not create the possibility for a new or different accident since the
associated limits are no longer attainable by the spent fuel pool.
The only potential accident remaining is the spent fuel handling
accident. Lowering the level of the spent fuel pool to elevation 28
feet has no impact on accident initiations since fuel handling will
not be allowed at this water level. Removing the restrictions in the
location of the minimum shift crew has no impact on accident
initiation, and the certified fuel handler will be present during
fuel handling operations.
Therefore, this change does not involve the possibility of a new
or different type of accident from any accident previously
evaluated.
3. Involve a significant reduction in a margin of safety?
No. This proposed change is a request to delete requirements
from the license and the technical specifications and modify the
spent fuel pool level requirements. Deleting the transshipment
license condition has no impact on margin since there no longer is
any safety-related equipment to protect and there are no plans for
transshipment of Unit 1 spent fuel to Units 2 or 3. Deleting the
spent fuel pool temperature and makeup requirements has no effect on
margin since the status of the spent fuel pool is such that the
margins associated with these requirements have increased and with
time will continue to increase. Modifying the level requirement to
allow the water level to be at elevation 28 feet for spent fuel
storage has no impact on margin since the spent fuel has cooled
significantly and fuel movement will not occur at this level. Since
the status of the spent fuel pool is such that the margins are
improving with time, removing the restrictions in the location of
the minimum shift crew has no effect on the margins.
Therefore, this change does not involve a reduction in a margin of
safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. The staff also reviewed the proposed administrative changes
to delete definitions and conform the table of contents and list of
tables to the proposed changes for no significant hazards
consideration. These administrative changes do not affect the design or
operation of the facility and satisfy the three standards of 10 CFR
50.92(c). Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: Douglas K. Porter, Esquire, Southern
California Edison Company, 2244 Walnut Grove Avenue, Rosemead,
California 91770.
NRC Section Chief: Michael Masnik (Unit 1).
Southern Nuclear Operating Company, Inc., et al., Docket Nos. 50-424
and 50-425, Vogtle Electric Generating Plant (VEGP), Units 1 and 2,
Burke County, Georgia
Date of application for amendments: October 5, 2000.
Brief description of amendments: The amendments revise the VEGP
Updated Final Safety Analysis Report (UFSAR) Chapters 11 and 15 to
incorporate changes due to an updated Dose Equivalent Iodine analysis.
The new analysis was performed in response to Westinghouse Nuclear
Safety Advisory Letter, ``NSAL-00-04: Nonconservatisms in Iodine
Spiking Calculations.''
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 changes do not significantly increase the
probability or consequences of an accident previously evaluated in
the [Updated Final Safety Analysis Report] UFSAR. The comprehensive
engineering review included evaluations or reanalysis of all
accident analyses. The letdown flow rate does not initiate any
accident; therefore, the probability of an accident has not been
increased. All dose consequences have been analyzed or evaluated
with respect to the proposed changes, and all acceptance criteria
continue to be met. Therefore, these changes do not involve a
significant increase in the probability or consequences of an
accident previously evaluated.
2. Create the possibility of a new or different kind of accident
from any accident previously evaluated?
The proposed changes do not create the possibility of a new or
different kind of accident than any accident already evaluated in
the UFSAR. No new accident scenarios, failure mechanisms or limiting
single failures are introduced as a result of the proposed changes.
The changes have no adverse effects on any safety-related system and
do not challenge the performance or integrity of any safety-related
system. Therefore, all accident analyses criteria continue to be
met, and these changes do not create the possibility of a new or
different kind of accident from any accident previously evaluated.
3. Involve a significant reduction in a margin of safety?
The proposed changes do not involve a significant reduction in a
margin of safety. All analyses and evaluations using these inputs
have been revised to reflect the proposed values. The evaluations
and analyses results demonstrate that applicable acceptance criteria
are met. Therefore, the proposed changes do not involve a
significant reduction in the margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 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. Arthur H. Domby, Troutman Sanders,
NationsBank Plaza, Suite 5200, 600 Peachtree Street, NE., Atlanta,
Georgia 30308-2216.
NRC Section Chief: Richard L. Emch, Jr.
Southern Nuclear Operating Company, Inc., Georgia Power Company,
Oglethorpe Power Corporation, Municipal Electric Authority of Georgia,
City of Dalton, Georgia, Docket Nos. 50-321 and 50-366, Edwin I. Hatch
Nuclear Plant, Units 1 and 2, Appling County, Georgia
Date of amendment request: November 3, 2000.
[[Page 77926]]
Description of amendment request: The proposed amendments would
revise Technical Specification 5.5.11, ``Technical Specification Bases
Control Program,'' to provide consistency with the changes to 10 CFR
50.59 which were published in the Federal Register (64 FR 53582) on
October 4, 1999.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. Does the change involve a significant increase in the
probability or consequences of an accident previously evaluated?
The proposed change deletes the reference to unreviewed safety
question as defined in 10 CFR 50.59. Deletion of the definition of
unreviewed safety question was approved by the NRC with the revision
of 10 CFR 50.59. Consequently, the probability of an accident
previously evaluated is not significantly increased. Changes to the
TS Bases are still evaluated in accordance with 10 CFR 50.59. As a
result, the consequences of any accident previously evaluated are
not significantly affected. Therefore, this change does not involve
a significant increase in the probability or consequences of an
accident previously evaluated.
2. Does the change create the possibility of a new or different
kind of accident from any accident previously analyzed?
The proposed change does not involve a physical alteration of
the plant (no new or different type of equipment will be installed)
or a change in the methods governing normal plant operation.
Therefore, this change does not create the possibility of a new or
different kind of accident from any accident previously evaluated.
3. Does the change involve a significant reduction in a margin
of safety?
The proposed change will not reduce a margin of safety because
it has no direct effect on any safety analyses assumptions. Changes
to the TS Bases that result in meeting the critieria in paragraph 10
CFR 50.59(c)(2) will still require NRC approval pursuant to 10 CFR
50.59. This change is administrative in nature based on the revision
to 10 CFR 50.59. Therefore, the proposed change does not involve a
significant reduction in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: Ernest L. Blake, Jr., Esquire, Shaw,
Pittman, Potts and Trowbridge, 2300 N Street, NW., Washington, DC
20037.
NRC Section Chief: Richard L. Emch, Jr.
Southern Nuclear Operating Company, Inc., et al., Docket Nos. 50-424
and 50-425, Vogtle Electric Generating Plant (VEGP), Units 1 and 2,
Burke County, Georgia
Date of application for amendments: November 6, 2000.
Brief description of amendments: The request revises the VEGP
Technical Specification (TS) Limiting Conditions for Operation 3.7.10,
3.7.11, and 3.7.13 to address degraded pressure boundaries. The changes
revise the TS to allow the pressure boundaries of ventilation systems
such as the Control Room Emergency Filtration System (CREFS) and the
Piping Penetration Area Filtration and Exhaust System (PPAFES) to be
opened intermittently under administrative controls. A new condition is
also added that allows 24 hours to restore inoperable CREFS and PPAFES
pressure boundaries before requiring the units to perform an orderly
shutdown.
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. Do the proposed changes involve a significant increase in the
probability or consequences of an accident previously evaluated?
No. The control room emergency filtration system (CREFS) and the
piping penetration area filtration and exhaust system (PPAFES) are
not assumed to be initiators of any analyzed accident. Therefore,
the proposed changes do not affect the probability of any accident
previously evaluated. The proposed changes for the CREFS and PPAFES
Technical Specifications (TS) would permit the subject pressure
boundaries to be opened intermittently under administrative control.
Based on the proposed compensatory measures in the form of a
dedicated individual who is in communication with the control room,
and his ability to rapidly restore the pressure boundary, the
capability to mitigate a design basis event will be maintained. In
addition, the proposed changes would add a new condition that would
permit a 24-hour period to take action to restore an inoperable
pressure boundary to operable status, modify existing conditions to
accommodate the new condition (so as to maintain the requirements of
the existing conditions), and correct a typographical error. With
respect to CREFS, the proposed changes do not involve a significant
increase in the consequences of an accident previously evaluated
based on the availability of a self-contained breathing apparatus to
minimize radiological dose due to iodine and the ability to operate
more than one train as the need arises to maintain positive pressure
or at least maintain an outflow of air from the control room
environment. With respect to the PPAFES, it has been demonstrated by
analysis that a breach of the pressure boundary will not result in
control room or offsite doses that exceed their respective limits.
The correction of the typographical error is an administrative
change that has no technical impact.
2. Do the proposed changes create the possibility of a new or
different kind of accident from any previously evaluated?
No. The proposed changes for the CREFS and PPAFES TS would
permit the subject pressure boundaries to be opened intermittently
under administrative control. In addition, the proposed changes
would add a new condition that would permit a 24-hour period to take
action to restore an inoperable pressure boundary to operable
status, modify existing conditions to accommodate the new condition
(so as to maintain the requirements of the existing conditions), and
correct a typographical error. The proposed changes do not alter the
operation of the plant or any of its equipment, introduce any new
equipment, or result in any new failure mechanisms or single
failures. Therefore, there is no potential for a new accident and no
changes to the way that an analyzed accident will progress. The
correction of the typographical error is an administrative change
that has no technical impact.
3. Do the proposed changes result in a significant reduction in
a margin of safety?
No. The proposed changes for the CREFS and PPAFES TS would
permit the subject pressure boundaries to be opened intermittently
under administrative control. In addition, the proposed changes
would add a new condition that would permit a 24-hour period to take
action to restore an inoperable pressure boundary to operable
status, modify existing conditions to accommodate the new condition
(so as to maintain the requirements of the existing conditions), and
correct a typographical error. The proposed changes do not adversely
affect the ability of the fission product barriers to perform their
functions. The only safety-related equipment affected by the
proposed changes is the CREFS and the PPAFES. It has been
demonstrated by analysis that a breach in the pressure boundary of
the PPAFES will not cause the control room or offsite doses to
exceed their respective limits. Adequate compensatory measures are
available to mitigate a breach in the CREFS pressure boundary. The
probabilities of design bases accidents that would place demands on
these systems during a period that the ventilation system pressure
boundaries would be allowed to be inoperable have been shown to be
negligible. In addition, the proposed changes avoid the potential of
placing one or both units in TS Limiting Condition for Operation
(LCO) 3.0.3 solely due to a breach of the ventilation system
pressure boundary. The correction of the typographical error is an
administrative change that has no technical impact.
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
[[Page 77927]]
involves no significant hazards consideration.
Attorney for licensee: Mr. Arthur H. Domby, Troutman Sanders,
NationsBank Plaza, Suite 5200, 600 Peachtree Street, NE., Atlanta,
Georgia 30308-2216.
NRC Section Chief: Richard L. Emch, Jr.
Southern Nuclear Operating Company, Inc., et al., Docket Nos. 50-424
and 50-425, Vogtle Electric Generating Plant (VEGP), Units 1 and 2,
Burke County, Georgia
Date of application for amendments: November 16, 2000.
Brief description of amendments: The request proposes to amend
Technical Specification 5.5.1, `` Technical Specification Bases Control
Program'' to provide consistency with the changes to 10 CFR 50.59 as
published in the Federal Register (64 FR 53582) dated October 4, 1999.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below.
1. Does the change involve a significant increase in the
probability or consequences of an accident previously evaluated?
The proposed change deletes the reference to unreviewed safety
question as defined in 10 CFR 50.59. Deletion of the definition of
unreviewed safety question was approved by the NRC with the revision of
10 CFR 50.59. Consequently, the probability of an accident previously
evaluated is not significantly increased. Changes to the TS Bases are
still evaluated in accordance with 10 CFR 50.59. As a result, the
consequences of any accident previously evaluated are not significantly
affected. Therefore, this change does not involve a significant
increase in the probability or consequences of an accident previously
evaluated.
2. Does the change create the possibility of a new or different
kind of accident from any accident previously analyzed?
The proposed change does not involve a physical alteration of the
plant (no new or different type of equipment will be installed) or a
change in the methods governing normal plant operation. Therefore, this
change does not create the possibility of a new or different kind of
accident from any accident previously evaluated.
3. Does the change involve a significant reduction in a margin of
safety?
The proposed change will not reduce a margin of safety because it
has no direct effect on any safety analyses assumptions. Changes to the
TS Bases that result in meeting the criteria in paragraph 10 CFR 50.59
(c)(2) will still require NRC approval pursuant to 10 CFR 50.59. This
change is administrative in nature based on the revision to 10 CFR
50.59. Therefore, the proposed change does not involve a significant
reduction in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 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. Arthur H. Domby, Troutman Sanders,
NationsBank Plaza, Suite 5200, 600 Peachtree Street, NE., Atlanta,
Georgia 30308-2216.
NRC Section Chief: Richard L. Emch, Jr.
Tennessee Valley Authority, Docket No. 50-260, Browns Ferry Nuclear
Plant, Unit 2, Limestone County, Alabama
Date of amendment request: November 21, 2000 (TSC-396).
Description of amendment request: The proposed amendment would
revise the reactor core Safety Limit Minimum Critical Power Ratio
(SLMCPR) specified in Technical Specification (TS) Section 2.1.1.2 from
1.10 to 1.07 for two reactor recirculation loop operation and from 1.12
to 1.10 for single loop operation. The change is based on use of newly
approved analytical methodology for the Cycle 12 reload analysis. This
methodology is described in Global Nuclear Fuels (GNF) licensing
document, ``General Electric Standard Application for Reactor Fuel,
GESTAR-II, Amendment 25,'' dated June 2000, which has been approved by
NRC.
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 proposed amendment establishes revised SLMCPR values for two
recirculation loop operation and for single recirculation loop
operation. The probability of an evaluated accident is derived from the
probabilities of the individual precursors to that accident. The
proposed SLMCPRs preserve the existing margin to transition boiling and
the probability of fuel damage is not increased. Since the change does
not require any physical plant modifications or physically affect any
plant components, no individual precursors of an accident are affected
and the probability of an evaluated accident is not increased by
revising the SLMCPR values.
The consequences of an evaluated accident are determined by the
operability of plant systems designed to mitigate those consequences.
The revised SLMCPRs have been performed using NRC-approved methods and
procedures. The basis of the MCPR [minimum critical power ratio] Safety
Limit is to ensure no mechanistic fuel damage is calculated to occur if
the limit is not violated. These calculations do not change the method
of operating the plant and have no effect on the consequences of an
evaluated accident. Therefore, the proposed TS change does not involve
an increase in the probability or consequences of an accident
previously evaluated.
B. The proposed amendment does not create the possibility of a new
or different kind of accident from any accident previously evaluated.
The proposed license amendment involves a revision of the SLMCPR
for two recirculation loop operation and for single loop operation
based on the results of an analysis of the Cycle 12 core. 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 TS change does not create
the possibility of a new or different kind of accident from any
accident previously evaluated.
C. The proposed amendment does not involve a significant reduction
in a margin of safety.
The margin of safety as defined in the TS bases will remain the
same. The new SLMCPRs are calculated using NRC-approved methods and
procedures which are in accordance with the current fuel design and
licensing criteria. The SLMCPRs remain high enough to ensure that
greater than 99.9% of all fuel rods in the core are
[[Page 77928]]
expected to avoid transition boiling if the limit is not violated,
thereby preserving the fuel cladding integrity. Therefore, the proposed
TS changes do not involve a reduction in the margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: General Counsel, Tennessee Valley Authority,
400 West Summit Hill Drive, ET 10H, Knoxville, Tennessee 37902.
NRC Section Chief: Richard P. Correia.
Vermont Yankee Nuclear Power Corporation, Docket No. 50-271, Vermont
Yankee Nuclear Power Station, Vernon, Vermont
Date of amendment request: October 25, 2000.
Description of amendment request: The proposed change would revise
the 125 volt DC (Vdc) station battery system Technical Specifications
(TSs) to reflect the availability of a second, fully qualified charger,
for each main station battery system. The licensee also proposed
corresponding changes to the listing of components in the TSs.
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 the 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.
There is no change in the method of operation of the 125 Vdc
main station battery systems by this change. The battery chargers
will function the same, except that an additional battery charger
will be available to each system. No change to accident assumptions
or precursors are involved with this change. Likewise, no change in
system operation or response to analyzed events is affected.
Therefore, the proposed change does not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
2. The operation of Vermont Yankee Nuclear Power Station in
accordance with the proposed amendment will not create the
possibility of a new or different kind of accident from any accident
previously evaluated.
The new chargers to be installed will provide additional
charging capability. No reduction in DC system equipment operation
or capability is involved. The methods by which the DC systems
perform their safety functions are unchanged and remain consistent
with current safety analysis assumptions. There is no change in
system or plant operation that involves failure modes other than
those 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. 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.
No adverse affect on equipment operation or capability will
result from this change. The installation of additional chargers in
fact enhances the reliability of the battery charging function. The
equipment fed by the DC systems involved in this change will
continue to provide adequate power to safety related loads in
accordance with analysis assumptions.
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: Mr. David R. Lewis, Shaw, Pittman, Potts and
Trowbridge, 2300 N Street, NW., Washington, DC 20037-1128.
NRC Section Chief: James W. Clifford.
Vermont Yankee Nuclear Power Corporation, Docket No. 50-271, Vermont
Yankee Nuclear Power Station, Vernon, Vermont
Date of amendment request: November 1, 2000.
Description of amendment request: The proposed change would revise
the operability requirement for high pressure coolant injection (HPCI)
and reactor core isolation cooling (RCIC) low steam line pressure
isolation instrumentation to coincide with system operability
requirements. The proposed change eliminates the need to open manual
containment isolation valves under administrative control during
reactor heatup, reduces the potential for operator error when closing
these valves (potential for leaving valve mispositioned) and clarifies
the steam line low pressure isolation function description. An
administrative change to correct the HPCI High Steam Line d/p
instrument component numbers was also proposed to ensure the accuracy
of isolation instrumentation information.
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 the 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 proposed change clarifies the equipment protection purpose
of the HPCI and RCIC low steam line isolation function. [The
proposed change would require] operability of the steam supply
pressure instrumentation [ ] whenever the systems are required to be
operable. This change does not significantly alter the function of
containment isolation actuation instruments nor does it
significantly alter containment integrity requirements. The proposed
change does not alter the basic operation of process variables,
systems, or components as described in the safety analysis. No new
equipment is being introduced.
The proposed change does not affect the ability of the primary
containment isolation system or high pressure core cooling systems
to perform their safety functions. The essential safety function of
providing primary containment integrity is maintained since
operability of the primary instrumentation associated with detection
of a HPCI or RCIC steam line break outside containment will continue
to be required when primary containment integrity is required. The
essential safety function of providing water to cool the core in the
event of a small break in the nuclear system is maintained. The
operational change being made would not alter the sequence of
events, plant response, or conclusions of existing safety analyses.
This proposed change results in no impact on analyzed accident event
precursors, initiators or effects.
Therefore, the proposed change does not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
2. The operation of Vermont Yankee Nuclear Power Station in
accordance with the proposed amendment will not create the
possibility of a new or different kind of accident from any accident
previously evaluated.
The proposed change does not involve any physical alteration of
plant equipment and does not change the method by which any safety-
related system performs its function. No new or different types of
equipment will be installed. Operation with the HPCI and RCIC steam
line isolation valves open between 212 deg.F and 150 psig does not
alter the input or result of existing accident analyses. The change
in plant operation does not involve failure modes other than those
previously evaluated. The methods governing plant operation and
testing remain consistent with current safety analysis assumptions.
Therefore, the proposed change 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.
[[Page 77929]]
The change involves operation with the HPCI and RCIC systems
with steam line isolation valves open between 212 deg.F and 150
psig. This change will not alter the basic operation of process
variables, systems, or components as described in the safety
analysis. No new equipment is introduced.
The proposed change maintains design margins of the primary
containment isolation system or high pressure core cooling systems
to perform their required safety functions. The essential safety
functions of providing primary containment integrity and providing
water to cool the core in the event of a small break in the nuclear
system are maintained. There is no physical or operational change
being made which would alter the sequence of events, plant response,
or margins in existing safety analyses. This proposed change results
in no impact on analyzed accident event precursors or effects.
This proposed change does not alter the physical design of the
plant. The change in method of operation results in no significant
impact on safety functions or assumed responses. The proposed change
does not alter the means by which primary containment isolation is
maintained and high pressure core cooling systems are operated.
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. David R. Lewis, Shaw, Pittman, Potts and
Trowbridge, 2300 N Street, NW., Washington, DC 20037-1128.
NRC Section Chief: James W. Clifford.
Virginia Electric and Power Company, Docket Nos. 50-338 and 50-339,
North Anna Power Station, Units No. 1 and No. 2, Louisa County,
Virginia
Date of amendment request: September 27, 2000, as supplemented
November 21, 2000.
Description of amendment request: The proposed changes will
increase the fuel enrichment limit from 4.3 weight percent to 4.6
weight percent Uranium\235\ (U\235\), establish Technical
Specifications to control the boron concentration in the spent fuel
pool (SFP) and impose restrictions on the storage locations for some
spent fuel assemblies, and change the method of criticality calculation
used to evaluate the effect of a fuel enrichment change on the SFP.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
[1.] Criterion 1. The proposed increase in maximum fuel
enrichment and the changes to the SFP design basis will not
significantly increase the probability of or consequences of an
accident previously evaluated in the North Anna Units 1 and 2 UFSAR
[Updated Final Safety Analysis Report].
The only accidents for which the probability of occurrence is
potentially affected by the fuel enrichment and SFP changes involve
criticality events during fuel handling and storage (e.g., fuel
mispositioning). The proposed Technical Specifications establish
additional restrictions on the placement of each fuel assembly in
the SFP to ensure subcriticality. However, criticality safety
analyses have been performed that demonstrate that the
Keff during the handling and storage of both new and
spent fuel remains low enough to ensure subcriticality during
postulated accident conditions. In addition, analyses of the
dilution of the spent fuel pool have been performed to ensure that
there is adequate time for a dilution event to be detected and
mitigated, such that the required subcritical margin is maintained
in the spent fuel pool. Therefore the probability of occurrence of
criticality during fuel handling or storage is not significantly
increased. In addition the consequences of the operating reactor
accident scenarios are also unchanged, because the source terms used
to determine the releases from fuel during accidents are a function
of burnup, rather than initial enrichment.
[2.] Criterion 2. The proposed increase in maximum fuel
enrichment or the change in the SFP design basis does not create a
new or different kind of accident from any already discussed in the
North Anna Units 1 and 2 UFSAR.
Although there are new restrictions on placement of fuel in the
SFP, the administrative controls on fuel movement to specified
locations in the pool are unchanged. The higher enrichment fuel and
the new Technical Specifications for the spent fuel pool do not
require any new or different plant equipment, and do not change the
manner in which currently installed equipment is operated. There are
no changes to normal core operation, and the units will meet all
applicable design criteria and will operate within existing
Technical Specifications limits. No new failure modes have been
created for any system, component, or piece of equipment, and no new
single failure mechanisms have been introduced. No new or different
plant equipment is introduced, and the operation of currently
equipment is not changed. The use of a higher maximum fuel
enrichment will not cause the design criteria for fuel operation or
storage to be exceeded. No new modes or limiting single failures are
created by the use of a higher fuel enrichment. Safety analyses for
the fuel storage area have demonstrated that subcriticality will be
maintained during fuel handling and storage, including fuel
mispositioning and pool dilution scenarios.
[3.] Criterion 3. The proposed increase in maximum fuel
enrichment and the changes to the SFP design basis will not
significantly reduce the margin of safety.
The use of higher enriched fuel and the changes to the SFP
design basis have the potential to affect only criticality events
during fuel handling and storage. Criticality analyses demonstrate
that the limits on Keff for the new and spent fuel
storage areas will be satisfied. Therefore, there is adequate margin
to ensure subcriticality during the storage and handling of fuel.
The requirements of 10 CFR 50 Appendix A General Design Criterion 62
are satisfied. Safety analyses demonstrated that Keff
will remain sufficiently low to ensure subcriticality, so no new
releases will result and there is no impact on radiological
consequences of accidents. The safety analyses of record will remain
applicable for the operation of fuel with a higher initial U\235\
enrichment and changes to the spent fuel pool. Therefore, the margin
of safety is not affected by the proposed increase in initial fuel
enrichment or changes to the spent fuel pool design basis.
Based on the evaluations and analyses results presented in the
foregoing safety significance evaluation, it has been demonstrated
that increasing the North Anna Units 1 and 2 maximum initial fuel
enrichment to 4.6 weight percent U\235\ and changing the design
basis of the spent fuel pool to eliminate any credit for Boraflex
but take credit for soluble boron in the pool will not result in 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 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. Donald P. Irwin, Esq., Hunton and
Williams, Riverfront Plaza East Tower, 951 E. Byrd Street, Richmond,
Virginia 23219.
NRC Section Chief: Richard L. Emch, Jr.
Notice of Issuance of Amendments to Facility Operating Licenses
During the period since publication of the last biweekly notice,
the Commission has issued the following amendments. The Commission has
determined for each of these amendments that the application complies
with the standards and requirements of the Atomic Energy Act of 1954,
as amended (the Act), and the Commission's rules and regulations. The
Commission has made appropriate findings as required by the Act and the
Commission's rules and regulations in 10 CFR Chapter I, which are set
forth in the license amendment.
Notice of Consideration of Issuance of Amendment to Facility
Operating License, Proposed No Significant Hazards Consideration
Determination,
[[Page 77930]]
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, located at One White Flint
North, 11555 Rockville Pike (first floor), Rockville, Maryland 20852.
Publicly available records will be accessible electronically from the
ADAMS Public Library component on the NRC Web site, http://www.nrc.gov
(the Electronic Reading Room).
Carolina Power & Light Company, et al., Docket Nos. 50-325 and 50-324,
Brunswick Steam Electric Plant, Units 1 and 2, Brunswick County, North
Carolina
Date of application for amendments: July 27, 2000.
Brief Description of amendments: The amendments change the
Technical Specifications to allow one of each unit's Direct Current
power subsystems to be inoperable when in Modes 4 and 5, and during
movement of irradiated fuel assemblies in the secondary containment.
Date of issuance: November 29, 2000.
Effective date: November 29, 2000.
Amendment Nos.: 211 and 238.
Facility Operating License Nos. DPR-71 and DPR-62: Amendments
change the Technical Specifications.
Date of initial notice in Federal Register: September 20, 2000 (65
FR 56948). The Commission's related evaluation of the amendments is
contained in a Safety Evaluation dated November 29, 2000.
No significant hazards consideration comments received: No.
Duke Energy Corporation, Docket Nos. 50-369 and 50-370, McGuire Nuclear
Station, Units 1 and 2, Mecklenburg County, North Carolina
Date of application for amendments: August 1, 2000.
Brief description of amendments: The amendments revise TS Section
3.7.15 and associated Bases, and Section 4.0 for the McGuire Nuclear
Stations, Units 1 and 2, to allow the use of credit for soluble boron
in spent fuel pool criticality analyses. The request is based on the
NRC-approved Westinghouse Owners Group Topical Report WCAP-14416-NP-A,
which provides generic methodology for crediting soluble boron.
Date of issuance: November 27, 2000.
Effective date: As of the date of issuance and shall be implemented
within 30 days from the date of issuance.
Amendment Nos.: 197 and 178.
Facility Operating License Nos. NPF-9 and NPF-17: Amendments
revised the Technical Specifications.
Date of initial notice in Federal Register: October 18, 2000 (65 FR
62385). The Commission's related evaluation of the amendments is
contained in a Safety Evaluation dated November 27, 2000.
No significant hazards consideration comments received: No.
Duke Energy Corporation, Docket Nos. 50-269, 50-270, and 50-287, Oconee
Nuclear Station, Units 1, 2, and 3, Oconee County, South Carolina
Date of application of amendments: October 18, 2000.
Brief description of amendments: The amendments revise the
implementation date of Amendment Nos. 312, 312, and 312 from November
30, 2000, so that implementation will be on or before implementation of
amendments resulting from the application that must be submitted by
April 5, 2001. This submittal will be based on an engineering study
that is being conducted to evaluate both the appropriate Keowee Hydro
Unit out-of-tolerance surveillance criteria and resolve overshoot
concerns.
Date of Issuance: November 27, 2000.
Effective date: As of the date of issuance.
Amendment Nos.: 317/317/317.
Facility Operating License Nos. DPR-38, DPR-47, and DPR-55:
Amendments revised the Implementation Date.
Date of initial notice in Federal Register: October 25, 2000 (65 FR
63896). The Commission's related evaluation of the amendments is
contained in a Safety Evaluation dated November 27, 2000.
No significant hazards consideration comments received: No.
Entergy Operations, Inc., Docket No. 50-368, Arkansas Nuclear One, Unit
No. 2, Pope County, Arkansas
Date of application for amendment: May 25, 2000.
Brief description of amendment: The amendment changed the action
statements for Technical Specification (TS) 3.8.2.2, A.C.
Distribution--Shutdown, and TS 3.8.2.4, DC Distribution--Shutdown, by
replacing the requirement to establish containment integrity within
eight hours with a requirement to immediately suspend core alterations,
the movement of irradiated fuel assemblies, and any operations
involving positive reactivity additions. Related changes to the
associated Bases were also made.
Date of issuance: November 28, 2000.
Effective date: As of the date of issuance to be implemented within
60 days from the date of issuance.
Amendment No.: 227.
Facility Operating License No. NPF-6: Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: July 12, 2000 (65 FR
43045). The Commission's related evaluation of the amendment is
contained in a Safety Evaluation dated November 28, 2000.
No significant hazards consideration comments received: No.
Florida Power Corporation, et al., Docket No. 50-302, Crystal River
Unit No. 3 Nuclear Generating Plant, Citrus County, Florida
Date of application for amendment: May 31, 2000.
Brief description of amendment: The Technical Specifications (TS)
were revised by adding an additional Condition to ITS 3.3.11, Emergency
Feedwater Initiation and Control System Instrumentation, regarding the
required action to be taken for one or more Emergency Feedwater
Initiation and Control System channels when up to two Reactor Coolant
Pump status signals are inoperable.
Date of issuance: November 21, 2000
Effective date: November 21, 2000
Amendment No.: 194.
Facility Operating License No. DPR-72: Amendment revised the TS.
Date of initial notice in Federal Register: July 12, 2000 (65 FR
43047). The Commission's related evaluation of the amendment is
contained in a Safety Evaluation dated November 21, 2000.
No significant hazards consideration comments received: No.
[[Page 77931]]
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: June 8, 2000.
Brief description of amendments: The amendments allow the use of
probabilistic risk assessment (PRA) techniques in evaluating the need
for tornado-generated missile barriers; this provides an alternative to
installing physical missile protection for those structures, systems,
and components that are not physically protected from tornado-generated
missiles.
Date of issuance: November 17, 2000.
Effective date: As of the date of issuance and shall be implemented
within 30 days.
Amendment Nos.: 247 and 228.
Facility Operating License Nos. DPR-58 and DPR-74: Amendments
approved revision of the UFSAR.
Date of initial notice in Federal Register: July 12, 2000 (65 FR
43049). The Commission's related evaluation of the amendments is
contained in a Safety Evaluation dated November 17, 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: September 1, 2000, as
supplemented October 27, 2000.
Brief description of amendments: The licensee proposed the
following three changes:
(1) A one-time change to Unit 1 Technical Specification (TS)
Surveillance Requirement (SR) 4.6.1.2 to add the following: ``A one-
time exception to the requirement to perform post-modification Type A
testing is allowed for the steam generators and associated piping, as
components of the containment barrier. For this case, American Society
of Mechnical Engineers (ASME) Section XI leak testing will be used to
verify leak tightness of the repaired or modified portions of the
containment barrier. Entry into MODES 3 and 4 following the extended
outage that commenced in 1997, may be made to perform this testing.''
(2) A change to Unit 1 and Unit 2 TS SR 4.6.1.2 to add the phrase
``except as modified by NRC-approved exemptions'' to the requirement to
perform testing in accordance with 10 CFR Part 50, Appendix J, Option
B, and the September 1995 version of Regulatory Guide 1.163.
(3) A change to the Unit 1 and Unit 2 Bases TS SR 4.6.1.2 to add
the phrase ``Regulatory Guide 1.163, dated September 1995, and Nuclear
Energy Institute (NEI) document NEI 94-01, except as modified'' after
the surveillance testing for measuring leakage rates are consistent
with the Appendix ``J'' of 10 CFR Part 50.
Date of issuance: November 17, 2000.
Effective date: As of the date of issuance and shall be implemented
within 45 days.
Amendment Nos.: 248 and 229.
Facility Operating License Nos. DPR-58 and DPR-74: Amendments
revised the Technical Specifications.
Date of initial notice in Federal Register: September 20, 2000 (65
FR 56953). The supplemental information contained clarifying
information and did not change the initial no significant hazards
consideration determination and did not expand the scope of the
original Federal Register notice.
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated November 17, 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: April 6, 2000, as supplemented
November 13, 2000.
Brief description of amendments: The amendments would approve
changes involving unreviewed safety questions to the Updated Final
Safety Analysis Report to incorporate new methodology to be used in the
analysis of high-energy line breaks at D. C. Cook.
Date of issuance: November 21, 2000.
Effective date: As of the date of issuance.
Amendment Nos.: 249 and 230.
Facility Operating License Nos. DPR-58 and DPR-74: Amendments
revised the Technical Specifications.
Date of initial notice in Federal Register: August 23, 2000 (65 FR
51355). The supplemental information contained clarifying information
and did not change the initial no significant hazards consideration
determination and did not expand the scope of the original Federal
Register notice. The Commission's related evaluation of the amendments
is contained in a Safety Evaluation dated November 21, 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: October 18, 2000, as
supplemented November 10, 2000.
Brief description of amendments: The amendments revise Technical
Specifications (TSs) 3/4.7.1.2, ``Auxiliary Feedwater [AFW] System,''
to change the description in the TSs surveillance requirement (SR)
4.7.1.2.d of the position for each automatic valve in the AFW system
from the ``fully open'' position to the ``correct'' position.
Date of issuance: November 30, 2000.
Effective date: As of the date of issuance and shall be implemented
within 30 days.
Amendment Nos.: 250 and 231.
Facility Operating License Nos. DPR-58 and DPR-74: Amendments
revised the Technical Specifications.
Date of initial notice in Federal Register: October 25, 2000 (65 FR
63899) The supplemental information contained clarifying information
and did not change the initial no significant hazards consideration
determination and did not expand the scope of the original Federal
Register notice.
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated November 30, 2000.
No significant hazards consideration comments received: No.
North Atlantic Energy Service Corporation, et al., Docket No. 50-443,
Seabrook Station, Unit No. 1, Rockingham County, New Hampshire
Date of amendment request: June 20, 2000, as supplemented on
September 25, 2000.
Description of amendment request: The amendment revises the
Technical Specifications (TS) by removing the prescriptive requirement
for determining the reactor coolant system flow rate by precision heat
balance in Surveillance Requirement 4.2.5.3. The amendment also revises
TS Table 2.2-1 to reflect the allowed calibration tolerance of the
protection racks and noting that the Trip Setpoint for Functional Unit
12, Reactor Coolant Flow-Low reactor trip is based on an indicated
value rather than a measured value.
Date of issuance: October 26, 2000.
Effective date: As of its date of issuance, and shall be
implemented at commencement of Cycle 8 operation (scheduled for
November 2000).
Amendment No.: 77.
[[Page 77932]]
Facility Operating License No. NPF-86: Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: August 9, 2000 (65 FR
48753) The supplemental letter provided clarifying information within
the scope of the original application 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 October 26, 2000.
No significant hazards consideration comments received: No.
Northeast Nuclear Energy Company, et al., Docket No. 50-245, Millstone
Nuclear Power Station, Unit No. 1, New London County, Connecticut
Date of application for amendments: June 6, 2000.
Brief description of amendment: The amendment deletes or modifies
license conditions and confirmatory orders to reflect the permanently
defueled condition of the unit.
Date of Issuance: November 15, 2000.
Effective date: As of the date of issuance and shall be implemented
within 90 days from the date of issuance.
Amendment No.: 108.
Facility Operating License No. DPR-21: The amendment revised the
Facility Operating License.
Date of initial notice in Federal Register: July 26, 2000 (65 FR
46010). The Commission's related evaluation of the amendment is
contained in a Safety Evaluation dated November 15, 2000.
No significant hazards consideration comments received: No.
Nuclear Management Company, Docket No. 50-263, Monticello Nuclear
Generating Plant, Wright County, Minnesota
Date of application for amendment: May 12, 2000.
Brief description of amendment: The amendment revises the Technical
Specification 4.6.E.1.d safety/relief valve bellows monitoring system
test frequency from quarterly to once per operating cycle.
Date of issuance: November 30, 2000.
Effective date: As of the date of issuance and shall be implemented
within 45 days.
Amendment No.: 114.
Facility Operating License No. DPR-22: Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: June 28, 2000 (65 FR
39959). The Commission's related evaluation of the amendment is
contained in a Safety Evaluation dated November 30, 2000.
No significant hazards consideration comments received: No.
Nuclear Management Company, LLC, Docket Nos. 50-266 and 50-301, Point
Beach Nuclear Plant, Units 1 and 2, Town of Two Creeks, Manitowoc
County, Wisconsin
Date of application for amendments: December 21, 1999, as
supplemented May 2, 2000
Brief description of amendments: These amendments incorporate
changes to the Technical Specifications (TSs) to more clearly define
the requirements for service water (SW) system operability in
accordance with the system configuration assumed in the SW system
analysis. The application dated December 21, 1999, as supplemented May
2, 2000, superceded an application dated July 30, 1998, in its
entirety. The December 21, 1999, application was submitted because the
licensee performed additional analyses of the SW system subsequent to
the submittal of the July 30, 1998, application, which necessitated
additional changes to the TSs.
Date of issuance: November 17, 2000.
Effective date: As of the date of issuance and shall be implemented
within 45 days.
Amendment Nos.: 199 and 204.
Facility Operating License Nos. DPR-24 and DPR-27: Amendments
revised the Technical Specifications.
Date of initial notice in Federal Register: February 23, 2000 (65
FR 9014). The May 2, 2000, supplemental letter provided additional
clarifying information that was within the scope of the original
application 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 November 17, 2000.
No significant hazards consideration comments received: No.
PSEG Nuclear LLC, Docket Nos. 50-272 and 50-311, Salem Nuclear
Generating Station, Unit Nos. 1 and 2, Salem County, New Jersey
Date of application for amendments: June 14, 2000, as supplemented
on October 12, 2000.
Brief description of amendments: The amendments modify the Salem
Unit Nos. 1 and 2 Technical Specifications (TS), and allow PSEG Nuclear
to use the Best Estimate Analyzer For Core Operations--Nuclear (BEACON)
system at Salem to fulfill certain TS surveillance requirements that
involve core power distribution measurements. BEACON is a core power
distribution monitoring and support system based on a three dimensional
nodal code. The system is used to provide data reduction for incore
neutron flux maps, core parameter analysis and follow, and core
prediction.
Date of issuance: November 6, 2000.
Effective date: As of the date of issuance, and shall be
implemented within 30 days of issuance.
Amendment Nos.: 237 and 218.
Facility Operating License Nos. DPR-70 and DPR-75: The amendments
revised the Technical Specifications.
Date of initial notice in Federal Register: July 26, 2000 (65 FR
46014). The Commission's related evaluation of the amendments is
contained in a Safety Evaluation dated November 6, 2000.
No significant hazards consideration comments received: No.
Public Service Electric & Gas Company, Docket No. 50-354, Hope Creek
Generating Station, Salem County, New Jersey
Date of application for amendment: November 24, 1999, as
supplemented September 14, 2000.
Brief description of amendment: The amendment revises the Technical
Specifications to implement Filtration, Recirculation, and Ventilation
System and Control Room Emergency Filtration System charcoal filter
testing requirements that are consistent with the U.S. Nuclear
Regulatory Commission guidance delineated in Generic Letter 99-02,
``Laboratory Testing of Nuclear-Grade Activated Charcoal.''
Date of issuance: November 17, 2000
Effective date: As of the date of issuance, and shall be
implemented within 60 days.
Amendment No.: 130.
Facility Operating License No. NPF-57: This amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: December 29, 1999 (64
FR 73096). The September 14, 2000, supplement provided clarifying
information that did not change the initial proposed no significant
hazards consideration determination or expand the scope of the original
application. The Commission's related evaluation of the amendment is
contained in a Safety Evaluation dated November 17, 2000.
No significant hazards consideration comments received: No.
[[Page 77933]]
Southern California Edison Company, et al., Docket Nos. 50-361 and 50-
362, San Onofre Nuclear Generating Station, Units 2 and 3, San Diego
County, California
Date of application for amendments: July 20, 2000 (PCN-488,
supplement 1; supersedes application dated August 11, 1999).
Brief description of amendments: The amendments revised Technical
Specifications surveillance requirements (SRs) related to the
acceptance criteria for TS 3.3.7, ``Diesel Generator (DG)--Undervoltage
Start,'' SR 3.3.7.3, which verifies operability of the loss of voltage
and degraded voltage actuation circuits. The amendments replaced the
analytical limits currently specified as acceptance criteria with
allowable values, and deleted SR 3.3.7.4 on the basis that it is
redundant with SR 3.3.7.3.
Date of issuance: November 29, 2000.
Effective date: November 29, 2000, to be implemented within 30 days
of issuance.
Amendment Nos.: Unit 2-174; Unit 3-165.
Facility Operating License Nos. NPF-10 and NPF-15: The amendments
revised the Technical Specifications.
Date of initial notice in Federal Register: August 23, 2000 (65 FR
51362). The Commission's related evaluation of the amendments is
contained in a Safety Evaluation dated November 29, 2000.
No significant hazards consideration comments received: No.
Tennessee Valley Authority, Docket Nos. 50-259, 50-260, and 50-296,
Browns Ferry Nuclear Plant, Units 1, 2, and 3, Limestone County,
Alabama
Date of application for amendments: August 28, 2000.
Description of amendment request: The amendments revise the Units
1, 2 and 3 Technical Specifications (TS) to incorporate TS Task Force
(TSTF) Items Nos. TSTF-71, TSTF-208, TSTF-222, TSTF-284, TSTF-258 and
TSTF-364. TSTFs are changes to the Improved Standard TS that were
initiated by the nuclear power industry and submitted to the NRC staff.
A description of each of the six TSTFs proposed for implementation at
Browns Ferry follows: (1) TSTF-71, Revision 2, adds an example of the
application of the Safety Function Determination Program to the Bases
for Limiting Conditions for Operation (LCO) 3.0.6. (2) TSTF-208,
Revision 0, extends the allowed time to reach MODE 2 in LCO 3.0.3 from
7 hours to 10 hours. The change is based on plant experience regarding
the time needed to perform a controlled shutdown in an orderly manner.
(3) TSTF-222, Revision 1, clarifies Improved Technical Specification
(ITS) Section 3.1.4, Control Rod Scram Times, Surveillance Requirements
(SRs) to better delineate the requirements for testing control rods
following refueling outages and for control rods requiring testing due
to work activities. (4) TSTF-258, Revision 4, revises TS Section 5.0,
Administrative Controls, to delete specific TS staffing requirement
provisions for Reactor Operators (ROs), eliminates TS details for
working hour limits, clarifies requirements for the Shift Technical
Advisor position, adds regulatory definitions for Senior ROs and ROs,
revises the Radioactive Effluent Controls Program to be consistent with
the intent of 10 CFR Part 20, deletes periodic reporting requirements
for mainsteam relief valve openings, and revises radiological area
control requirements for radiation areas to be consistant with those
specified in 10 CFR 20.1601(c). (5) TSTF-284, Revision 3, modifies
Improved TS Section 1.4, Frequency, to clarify the usage of the terms
``met'' and ``performed'' to facilitate the application of SR Notes.
Two new SR Examples, 1.4-5 and 1.4-6, are added to illustrate the
application of the terms. (6) TSTF-364, Revision 0, revises Section
5.5.10, TS Bases Control Program, to reference 10 CFR 50.59 rather than
``unreviewed safety question.'' Also, editorial change WOG-ED-24, which
substitutes ``require'' for ``involve'' in 5.5.10.b is made for
consistency in usage.
Date of issuance: November 21, 2000.
Effective date: November 21, 2000.
Amendment Nos.: 239, 266, and 226.
Facility Operating License Nos. DPR-33, DPR-52, and DPR-68:
Amendments revised the licenses.
Date of initial notice in Federal Register: October 4, 2000 (65 FR
59224)
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated November 21, 2000.
No significant hazards consideration comments received: No.
TXU Electric, Docket Nos. 50-445 and 50-446, Comanche Peak Steam
Electric Station, Unit Nos. 1 and 2, Somervell County, Texas
Date of amendment request: September 15, 2000.
Brief description of amendments: The proposed change replaces the
general references currently provided in Technical Specification 5.6.6
for determining the reactor coolant system pressure and temperature
limits with the requirement that the Pressure/Temperature Limits and
Low Temperature Overpressure Protection System Setpoints shall not be
revised without prior U.S. Nuclear Regulatory Commission approval.
Date of issuance: November 27, 2000.
Effective date: As of the date of issuance and shall be implemented
within 30 days from the date of issuance.
Amendment Nos.: 81 & 81.
Facility Operating License Nos. NPF-87 and NPF-89: The amendments
revised the Technical Specifications.
Date of initial notice in Federal Register: November 1, 2000 (65 FR
65351). The Commission's related evaluation of the amendments is
contained in a Safety Evaluation dated November 27, 2000.
No significant hazards consideration comments received: No.
Vermont Yankee Nuclear Power Corporation, Docket No. 50-271, Vermont
Yankee Nuclear Power Station, Vernon, Vermont
Date of application for amendment: September 19, 2000.
Brief description of amendment: The amendment revises the Technical
Specifications to establish operability requirements to ensure that
adequate reactor coolant inventory and sufficient heat removal
capability exist during cold shutdown and refueling conditions.
Date of Issuance: November 17, 2000.
Effective date: As of the date of issuance, and shall be
implemented within 60 days.
Amendment No.: 195.
Facility Operating License No. DPR-28: Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: October 18, 2000 (65 FR
62393). The Commission's related evaluation of this amendment is
contained in a Safety Evaluation dated November 17, 2000.
No significant hazards consideration comments received: No.
Virginia Electric and Power Company, et al., Docket Nos. 50-338 and 50-
339, North Anna Power Station, Units No. 1 and No. 2, Louisa County,
Virginia
Date of application for amendments: November 29, 1999, as
supplemented August 31, 2000.
Brief description of amendments: The amendments revise the testing
requirements in Technical Specification (TS) 4.7.7.1 and TS 4.7.8.1 to
incorporate the American Society for Testing and Materials D3803-1989
standard and the application of a safety factor of 2.0 for the charcoal
filter
[[Page 77934]]
efficiency assumed in Virginia Electric and Power Company's design-
basis dose analysis.
Date of issuance: November 20, 2000.
Effective date: As of the date of issuance and shall be implemented
within 30 days.
Amendment Nos.: 224 and 205.
Facility Operating License Nos. NPF-4 and NPF-7: Amendments revised
the Technical Specifications.
Date of initial notice in Federal Register: February 9, 2000 (65 FR
6413). The August 31, 2000, supplement provided clarifying information
only, and did not change the initial no significant hazards
consideration determination. The Commission's related evaluation of the
amendments is contained in a Safety Evaluation dated November 20, 2000.
No significant hazards consideration comments received: No.
Dated at Rockville, Maryland, this 6th day of December 2000.
For the Nuclear Regulatory Commission.
John A. Zwolinski,
Director, Division of Licensing Project Management, Office of Nuclear
Reactor Regulation.
[FR Doc. 00-31541 Filed 12-12-00; 8:45 am]
BILLING CODE 7590-01-P