[Federal Register Volume 65, Number 193 (Wednesday, October 4, 2000)]
[Notices]
[Pages 59218-59233]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: 00-25377]
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NUCLEAR REGULATORY COMMISSION
Applications and Amendments to Facility Operating Licenses
Involving No Significant Hazards Considerations; Biweekly Notice
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 September 11, 2000, through September 22,
2000. The last biweekly notice was published on September 20, 2000 (65
FR 56946, as corrected at 65 FR 57484 and 65 FR 58113).
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
[[Page 59219]]
take this action, it will publish in the Federal Register a notice of
issuance and provide for opportunity for a hearing after issuance. The
Commission expects that the need to take this action will occur very
infrequently.
Written comments may be submitted by mail to the Chief, Rules
Review and Directives Branch, Division of Freedom of Information and
Publications Services, Office of Administration, U.S. Nuclear
Regulatory Commission, Washington, DC 20555-0001, and should cite the
publication date and page number of this Federal Register notice.
Written comments may also be delivered to Room 6D22, Two White Flint
North, 11545 Rockville Pike, Rockville, Maryland from 7:30 a.m. to 4:15
p.m. Federal workdays. Copies of written comments received may be
examined at the NRC'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 November 3, 2000, the licensee may file a request for a hearing
with respect to issuance of the amendment to the subject facility
operating license and any person whose interest may be affected by this
proceeding and who wishes to participate as a party in the proceeding
must file a written request for a hearing and a petition for leave to
intervene. Requests for a hearing and a petition for leave to intervene
shall be filed in accordance with the Commission's ``Rules of Practice
for Domestic Licensing Proceedings'' in 10 CFR Part 2. Interested
persons should consult a current copy of 10 CFR 2.714 which is
available at the NRC'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 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 NRC's Public Document Room,
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 NRC'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).
Arizona Public Service Company, et al., Docket Nos. STN 50-528, STN 50-
529, and STN 50-530, Palo Verde Nuclear Generating Station, Units 1, 2,
and 3, Maricopa County, Arizona
Date of amendments request: June 8, 2000.
Description of amendments request: The licensee proposes to amend
Technical Specification (TS) 5.6.5, ``Core Operating Limits Report
(COLR),'' to add a methodology using the
[[Page 59220]]
CASMO-4 and SIMULATE-3 codes to the list of analytical methods used to
determine core operating limits contained in TS 5.6.5.b. The change
would allow the use of the CASMO-4 and SIMULATE-3 methodology to
perform nuclear design 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:
Standard 1--Does the proposed change involve a significant
increase in the probability or consequences of an accident
previously evaluated?
No. The proposed change does not involve a significant increase
in the probability or consequences of an accident previously
evaluated.
Arizona Public Service Company (APS) intends to replace the DIT/
ROCS/MC methodology with CASMO-4/SIMULATE-3 code package. The
proposed amendment would add methodology using CASMO-4 and SIMULATE-
3 codes to the list of analytical methods used to determine core
operating limits contained in Technical Specification 5.6.5.b. This
will allow the use of the CASMO-4 and SIMULATE-3 methodology to
perform all steady-state PWR [pressurized-water reactor] core
physics analyses.
The probability of occurrence of an accident previously
evaluated will not be increased by the proposed change in the
particular codes used for physics calculations for nuclear design
analysis. The results of nuclear design analyses are used as inputs
to the analysis of accidents that are evaluated in the Updated Final
Safety Analysis Report (UFSAR). These inputs do not alter the
physical characteristics or modes of operation of any system,
structure, or component involved in the initiation of an accident.
Thus, there is no significant increase in the probability of an
accident previously evaluated as a result of this change.
The consequences of an accident evaluated in the UFSAR are
affected by the value of inputs to the transient safety analysis. An
extensive benchmark of CASMO-4/SIMULATE-3 predictions with measured
data using a variety of fuel designs and operating conditions in
power reactors and critical experiments, was performed. The accuracy
of CASMO-4/SIMULATE-3 is similar to, and sometimes better than, the
accuracy of DIT/ROCS/MC. Furthermore, there is always the potential
for the value of the nuclear design parameters to change solely as a
result of the new reload fuel core loading pattern. Regardless of
the source of a change, an assessment is always made of changes to
the nuclear design parameters with respect to their effects on the
consequences of accidents previously evaluated in the UFSAR.
Refueling is an anticipated activity which is described in the
UFSAR. If increased consequences are anticipated, compensatory
actions are implemented to neutralize any expected increase in
consequences. These compensatory actions include, but are not
limited to, crediting any existing margins in the analysis or
redefining the operating envelope to avoid increased consequences.
Thus, the nuclear design parameters are intermediate results and by
themselves will not result in an increase in the consequence of an
accident evaluated in the UFSAR.
Therefore, the replacement of the DIT/ROCS/MC codes with the
CASMO-4/ SIMULATE-3 code package, which will perform the same
functions as the DIT/ROCS/MC codes with similar accuracy, does not
significantly increase the consequences of an accident previously
evaluated.
Standard 2--Does the proposed change create the possibility of a
new or different kind of accident from any accident previously
evaluated?
No. The proposed change does not create the possibility of a new
or different kind of accident from any accident previously
evaluated.
Arizona Public Service Company (APS) intends to replace the DIT/
ROCS/MC methodology with CASMO-4/SIMULATE-3 code package. The
proposed amendment would add methodology using CASMO-4 and SIMULATE-
3 codes to the list of analytical methods used to determine core
operating limits contained in Technical Specification 5.6.5.b.
The possibility for a new or different kind of accident
evaluated previously in the UFSAR will not be created by the
proposed change to the particular codes used for physics
calculations for nuclear design analyses. The change involves
replacing the NRC approved ABB Combustion Engineering Nuclear Power
(ABB/CE) DIT and ROCS/MC codes, with the Studsvik CASMO-4 and
SIMULATE-3 codes. The results of nuclear design analyses are used as
inputs to the analysis of accidents that are evaluated in the UFSAR.
These inputs do not alter the physical characteristics or modes of
operation of any system, structure or component involved in the
initiation of an accident.
Therefore, the replacement of the DIT/ROCS/MC codes with the
CASMO-4/SIMULATE-3 code package, which will perform the same
functions as the DIT/ROCS/MC codes with similar accuracy, does not
increase the possibility of a new or different kind of accident from
any accident previously evaluated.
Standard 3--Does the proposed change involve a significant
reduction in a margin of safety?
No. The proposed change does not involve a significant reduction
in a margin of safety.
The margin of safety as defined in the basis for any technical
specification will not be reduced nor increased by the proposed
change to the particular codes used for physics calculations for
nuclear design analyses. The change involves replacing the NRC
approved ABB/CE DIT and ROCS/MC codes, with the Studsvik CASMO-4 and
SIMULATE-3 codes. Extensive benchmarking of the CASMO-4/SIMULATE-3
computer codes has demonstrated that the values of those parameters
used in the safety analysis are not significantly changed relative
to the values obtained using the DIT/ROCS/MC computer codes. For any
changes in the calculated values that do occur, the application of
appropriate biases and uncertainties ensures that the current margin
of safety is maintained. Specifically, use of these code specific
biases and uncertainties in safety evaluations continues to provide
the same statistical assurance that the values of the nuclear
parameters used in the safety analysis are conservative with respect
to the actual values on at least a 95/95 probability/confidence
basis.
The NRC staff has reviewed the licensee's analysis and, based on
that review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
request for amendments involves no significant hazards consideration.
Attorney for licensee: Nancy C. Loftin, Esq., Corporate Secretary
and Counsel, Arizona Public Service Company, P.O. Box 53999, Mail
Station 9068, Phoenix, Arizona 85072-3999
NRC Section Chief: Stephen Dembek.
Arizona Public Service Company, et al., Docket Nos. STN 50-528, STN 50-
529, and STN 50-530, Palo Verde Nuclear Generating Station, Units 1, 2,
and 3, Maricopa County, Arizona
Date of amendments request: June 16, 2000.
Description of amendments request: The licensee proposes to amend
Technical Specification (TS) Table 3.3.10-1, ``Post Accident Monitoring
Instrumentation,'' to add the High Pressure Safety Injection cold leg
flow and hot leg flow instrumentation to this table. This change is
required because this instrumentation meets the criteria for a
Regulatory Guide 1.97, ``Instrumentation for Light-Water-Cooled Nuclear
Power Plants To Assess Plant and Environs Conditions During and
Following an Accident,'' Revision 2, Type A, Category 1 variable.
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:
Standard 1--Does the proposed change involve a significant
increase in the probability or consequences of an accident
previously evaluated?
No. The proposed change to TS Table 3.3.10-1, by adding the High
Pressure Safety Injection (HPSI) hot and cold leg flow
instrumentation, does not involve a significant increase in the
probability or consequences of an accident previously evaluated
because it does not represent a change to design configuration or
operation of the plant. The amendment does not affect
[[Page 59221]]
the operability or availability of the HPSI system or any other
safety related equipment. Additionally, there are no effects on the
failure modes associated with the probability of a failure of a
system important to safety.
Standard 2--Does the proposed change create the possibility of a
new or different kind of accident from any accident previously
evaluated?
No. The proposed amendment does not create the possibility of a
new or different kind of accident from any accident previously
evaluated because the change does not impact the response or
operation of the plant. The availability and operability of the
plant equipment is unchanged, as the design requirements have not
changed.
The proposed change revises only the Regulatory Guide (RG) 1.97,
Revision 2, classification of the HPSI hot leg and cold leg flow
indication loops. Regardless of RG classification the instruments
remain seismically, electrically, and otherwise qualified for the
application. Hence, the revised classification will not subject
these components to new modes of operation that could result in a
new failure mode, thus initiating an accident of a different type.
Standard 3--Does the proposed change involve a significant
reduction in a margin of safety?
No. This proposed amendment does not involve a significant
reduction in the margin of safety because neither of the following
PVNGS Technical Specification (TS) Bases (B 3.5.3 ECCS [emergency
core cooling system]--Operating, or 3.3.10 Post Accident Monitoring
(PAM) Instrumentation) is changed by the proposed amendment.
TS Bases B 3.5.3 ECCS--Operating--states that the function of
the ECCS is to provide core cooling and negative reactivity to
ensure that the reactor core is protected after any of the following
accidents:
a. Loss of Coolant Accident (LOCA);
b. Control Element Assembly (CEA) ejection accident;
c. Loss of secondary coolant accident, including uncontrolled
steam release or loss of feedwater; and
d. Steam Generator Tube Rupture (SGTR).
Changing the RG 1.97 Type and Category of these instruments does
not affect the ability of the ECCS to provide core cooling and
negative reactivity during these accidents.
TS Bases B 3.3.10--Post Accident Monitoring (PAM)
Instrumentation--states that the primary purpose of PAM
instrumentation is to display plant variables that provide
information required by the control room operators during accident
situations. This information provides the necessary support for the
operator to take the manual actions, for which no automatic control
is provided, that are required for safety systems to accomplish
their safety functions for Design Basis Events.
The OPERABILITY of PAM instrumentation ensures that there is
sufficient information available on selected plant parameters to
monitor and assess plant status and behavior following an accident.
These Type A variables are required to be included in this LCO
[Limiting Condition for Operation] because they provide the primary
information required to permit the control room operator to take
specific manually controlled actions, for which no automatic control
is provided, that are required for safety systems to accomplish
their safety functions for Design Basis Accidents (DBAs). The
addition of these instruments supports this TS Bases. 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
that review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
request for amendments involves no significant hazards consideration.
Attorney for licensee: Nancy C. Loftin, Esq., Corporate Secretary
and Counsel, Arizona Public Service Company, P.O. Box 53999, Mail
Station 9068, Phoenix, Arizona 85072-3999.
NRC Section Chief: Stephen Dembek.
Consolidated Edison Company of New York, Docket No. 50-247, Indian
Point Nuclear Generating Unit No. 2, Westchester County, New York
Date of amendment request: May 5, 1999, as supplemented on December
22, 1999, and September 18, 2000.
Description of amendment request: The proposed amendment would
revise Technical Specification (TS) 3.5, ``Instrumentation Systems,''
for the reactor protection system and engineered safety features
actuation system instrumentation. Specifically, the proposed amendment
would (1) change the allowed outage times for the instrumentation and
the analog channel test bypass time and (2) allow on-line testing and
maintenance of instrumentation. The proposed amendment also includes
several editorial changes to TS Tables 3.5-2 and 3.5-3. The proposed
amendment was originally noticed in the Federal Register on September
8, 1999 (64 FR 48861). It is now being noticed to correct errors made
in the original notice description of amendment request.
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 reactor protection and engineered safety features functions
are not initiators of any design basis accident or event and
therefore do not increase the probability of any accident previously
evaluated. The proposed changes to the AOTs [allowed outage times],
bypass times, and allowing on-line testing and maintenance have an
insignificant impact on plant safety based on the calculated CDF
[core damage frequency] increase being less than 1.0E-06. Therefore,
the proposed changes do not result in a significant increase in 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.
The proposed changes do not result in a change in the manner in
which the RPS [reactor protection system] and ESFAS [engineered
safety features actuation system] provide plant protection. No
change is being made which alters the functioning of the RPS and
ESFAS. Rather, the likelihood or probability of the RPS or ESF
functioning properly is affected as described above. Therefore, the
proposed changes do not create the possibility of a new or different
kind of accident nor involve a reduction in the margin of safety as
defined in the Safety Analysis Report.
3. The proposed change does not involve a significant reduction
in a margin of safety.
The proposed changes do not alter the manner in which safety
limits, limiting safety system setpoints or limiting conditions for
operations are determined. The impact of increased AOTs, testing
times, and allowing on-line testing and maintenance are expected to
result in an overall improvement in safety because:
The longer AOTs for the master relays, logic cabinets, and
analog channels will promote improved maintenance practices that
will provide improved component performance, improved availability
of the protection system, and a reduced number of spurious reactor
trips and spurious actuation of safety equipment.
The longer AOTs and bypass times for the analog channels will
provide additional time before being required to place the channel
in trip. With the channel in trip, the logic required to cause a
reactor trip or a safety system actuation is reduced to 1 of 2 (for
2 of 3 logic) and to 1 of 3 (for 2 of 4 logic). With the reduced
logic requirement, the potential for a spurious actuation is
increased. Leaving the channel in the bypass state for additional
time does reduce the availability of signals to initiate component
actuation for event mitigation when required, but as shown in this
analysis, the impact on plant safety is small due to the
availability of other signals or operator action to trip the reactor
or cause component actuation.
The longer allowed outage times will provide plant operators
additional flexibility in operating the plant. There will be
additional time available before an action needs to be taken to shut
down the plant or place a channel in the tripped state. This
additional flexibility will facilitate prioritizing component
repairs.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff
[[Page 59222]]
proposes to determine that the amendment request involves no
significant hazards consideration.
Attorney for licensee: Brent L. Brandenburg, Esquire, 4 Irving
Place, New York, New York 10003.
NRC Section Chief: Marsha Gamberoni.
Duke Energy Corporation, Docket Nos. 50-269, 50-270, and 50-287, Oconee
Nuclear Station, Units 1, 2, and 3, Oconee County, South Carolina
Date of amendment request: September 12, 2000.
Description of amendment request: The proposed amendments would
revise the Technical Specifications 5.5.10, Item e.6, Steam Generator
Tube Surveillance Program, by (1) removing the restriction on the lower
tube sheet area rolling, (2) removing the limitation of only one reroll
per steam generator tube, (3) eliminating the requirement that the
reroll be one inch in depth, and (4) changing the revision number
reference for Topical Report BAW-2303P, August 2000, ``OTSG Repair Roll
Qualification Report,'' from Revision 3 to Revision 4.
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:
Duke Energy Corporation (Duke) has made the determination that
this amendment request involves a No Significant Hazards
Consideration by applying the standards established by NRC
regulations in 10CFR50.92. This ensures operation of the facility in
accordance with the proposed amendment will not:
1. Involve a significant increase in the probability or
consequences of an accident previously evaluated?
No. The proposed change to the Technical Specifications
incorporates Revision 4 of Topical Report BAW-2303P, OTSG Repair
Roll Qualification Report. This document is also being submitted for
NRC review and approval. This revision addresses, and is consistent
with, the conclusions of all applicable Oconee licensing basis
analyses and ensures that previously evaluated accidents are
bounding. All the established acceptance criteria for the accidents
analyzed in the Oconee licensing basis continue to be met.
Therefore, no existing accident probabilities or consequences will
be impacted.
2. Create the possibility of a new or different kind of accident
from any accident previously evaluated?
No. Revision 4 of BAW-2303P addresses limiting events for steam
generator tube reroll repairs. These events include Main Steam Line
Break, the Small Break Loss of Coolant Accident, and other
transients on B&W Once-Through Steam Generators. For Oconee, the
Main Steam Line Break is the limiting event. This revised topical
report confirms the acceptability of the reroll repair techniques
previously used at Oconee. As a result, no new failure modes are
being created. BAW-2303P, as submitted for NRC review and approval,
does not create the possibility of a new or different kind of
accident.
3. Involve a significant reduction in a margin of safety?
No. Margin of safety is related to the confidence in the ability
of the fission product barriers to perform their design functions
during and following an accident situation. These barriers include
the fuel cladding, the reactor coolant system, and the containment
system. As part of the reactor coolant system pressure boundary, the
steam generator tubes are unique in that they are also relied upon
as a heat transfer surface between the primary and secondary
systems, such that residual heat can be removed from the primary
system. In addition, the steam generator tubes also isolate the
radioactive fission products in the primary coolant from the
secondary system. Finally, the steam generator tubes may be relied
upon to maintain their integrity under conditions resulting from
core damage severe accidents consistent with the containment
objectives of preventing uncontrolled fission product release. The
functions of the steam generator tubes will not be significantly
affected by the changes proposed in this license amendment request.
Implementation of BAW-2303P, Revision 4, as submitted for NRC review
and approval, at Oconee will result in assurance that parameters
affecting the integrity of the steam generator tubes continue to
meet applicable safety analyses and industry codes and standards.
Therefore, no safety margin will be significantly reduced.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: Anne W. Cottington, Winston and Strawn, 1200
17th Street, NW., Washington, DC 20005.
NRC Section Chief: Richard L. Emch, Jr.
Florida Power and Light Company, et al., Docket Nos. 50-335 and 50-389,
St. Lucie Plant, Unit Nos. 1 and 2, St. Lucie County, Florida
Date of amendment request: July 26, 2000.
Description of amendment request: The amendments would revise the
Technical Specifications (TS) Index to delete reference to the Bases
since, in accordance with 10 CFR 50.36(a), the Bases are not a part of
the TS. Future changes to the TS Bases will be evaluated per 10 CFR
50.59 and made under administrative control and reviews and in
accordance with the proposed TS Bases control program as described in
TS 5.5.14 of NUREG-1432, Revision 1, ``Standard Technical
Specifications Combustion Engineering Plants.''
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
(1) Operation of the facility in accordance with the proposed
amendments would not involve a significant increase in the
probability or consequences of an accident previously evaluated.
The proposed amendments are administrative in nature and do not
affect assumptions contained in plant safety analyses, the physical
design and operation of the plant, nor do they affect Technical
Specifications that preserve safety analysis assumptions. The
Technical Specification BASES, per 10 CFR 50.36(a), are not part of
the Technical Specifications. Changes to the TS BASES will be
controlled by a plant procedure under administrative controls and
reviews. Proposed changes to the TS BASES will be evaluated in
accordance with 10 CFR 50.59 and made under the programmatic
controls and requirements of the proposed Technical Specifications
(TS) Bases Control Program. Therefore, the proposed changes do not
increase the probability or consequences of accidents previously
analyzed.
(2) Operation of the facility in accordance with the proposed
amendments would not create the possibility of a new or different
kind of accident from any accident previously evaluated.
The proposed amendments are administrative in nature. The
proposed amendments will not create the possibility of a new or
different kind of accident from any accident previously evaluated
since the proposed amendments will not change the physical plant or
the modes of plant operation defined in the facility operating
license. No new failure mode is introduced due to the administrative
change, since the proposed change does not involve the addition or
modification of equipment nor does it alter the design or operation
of affected plant systems, structures, or components.
(3) Operation of the facility in accordance with the proposed
amendments would not involve a significant reduction in a margin of
safety.
The operating limits and functional capabilities of the affected
systems, structures, and components are unchanged by the proposed
amendments. The BASES information, per 10 CFR 50.36(a), is not a
part of the Technical Specifications. Changes to the TS BASES will
be controlled by a plant procedure under administrative controls and
reviews and made under the programmatic controls and requirements of
the proposed Technical Specifications (TS) Bases Control Program.
Proposed changes to the TS BASES will be evaluated in accordance
with 10 CFR 50.59 and the TS BASES will be maintained
[[Page 59223]]
in an FPL-controlled document. Therefore, the proposed changes do
not reduce any margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: M.S. Ross, Attorney, Florida Power & Light,
P.O. Box 14000, Juno Beach, Florida 33408-0420.
NRC Section Chief: Richard P. Correia.
Florida Power Corporation, et al., Docket No. 50-302, Crystal River
Nuclear Generating Plant, Unit No. 3, Citrus County, Florida
Date of amendment request: August 31, 2000.
Description of amendment request: This amendment would revise
Improved Technical Specification (ITS) Table 3.3.18-1, ``Remote
Shutdown System Instrumentation.'' The table would be updated to
reflect plant modifications and procedure changes regarding placing and
maintaining the plant in a safe shutdown condition if the control room
becomes inaccessible.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
The proposed changes do not involve a significant hazards
consideration. In support of this conclusion, the following analysis
is provided:
(1) Does not involve a significant increase in the probability
or consequences of an accident previously analyzed.
The instruments listed in Table 3.3.18-1 are used to provide
information on selected parameters to the operators that will allow
them to place and maintain the plant in a safe shutdown condition in
the event the control room becomes inaccessible. The proposed
license amendment revises Table 3.3.18-1, Remote Shutdown System
Instrumentation, to more accurately reflect the instruments that
would be used by the operators to perform abnormal operating
procedure AP-990, Shutdown from Outside the Control Room. The
proposed license amendment also revises ITS Bases Section B 3.3.18
to add a table that identifies, by equipment tag number, the
specific instruments used to satisfy the requirements of ITS 3.3.18
and ITS Table 3.3.18-1. The instruments identified in ITS Table
3.3.18-1 and ITS Bases Table B 3.3.18-1 are not initiators of any
design basis accidents. The design functions of the Remote Shutdown
System Instrumentation and the initial conditions for accidents that
require the Remote Shutdown System will not be effected by the
change. Therefore, the change will not increase the probability or
consequences of an accident previously evaluated.
(2) Does not create the possibility of a new or different kind
of accident from any accident previously analyzed.
The proposed amendment involves no changes to the CR-3 design or
to the functions or operation of the Remote Shutdown System. The
proposed amendment will ensure that sufficient and appropriate
instrumentation is available to allow the operators to place and
maintain the plant in a safe shutdown condition in the event the
control room becomes inaccessible. The proposed amendment will also
add information to Bases Section B 3.3.18 that will ensure timely
and accurate operability evaluations and entry into the appropriate
Conditions and Required Actions of ITS 3.3.18. The proposed
amendment will not create any new plant configurations different
from those already analyzed. Therefore, the proposed change will not
create the possibility of a new or different kind of accident from
any accident previously evaluated.
(3) Does not involve a significant reduction in the margin of
safety.
The proposed amendment revises Table 3.3.18-1, Remote Shutdown
System Instrumentation, to more accurately reflect the instruments
that would be used by the operators to perform a shutdown from
outside the control room. The proposed amendment will revise ITS
Bases Section B 3.3.18 to provide the operators with guidance that
will assist them in making timely and accurate operability
determinations and entries into the appropriate Conditions and
Required Actions for ITS 3.3.18. The proposed changes will not
reduce the ability of the Remote Shutdown System to monitor and
control reactivity, RCS [reactor coolant system] pressure, core heat
removal, or RCS inventory. Thus, the proposed amendment will not
result in 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: R. Alexander Glenn, General Counsel, Florida
Power Corporation, MAC--A5A, P.O. Box 14042, St. Petersburg, Florida
33733-4042.
NRC Section Chief: Richard P. Correia.
Pacific Gas and Electric Company, Docket No. 50-323, Diablo Canyon
Nuclear Power Plant, Unit No. 2, San Luis Obispo County, California
Date of amendment requests: June 19, 2000.
Description of amendment requests: The proposed license amendment
would revise Technical Specifications 5.5.9, ``Steam Generator (SG)
Tube Surveillance Program,'' and 5.6.10, ``SG Tube Inspection Report,''
to add new surveillance and reporting requirements associated with a SG
tube inspection and repair. The new requirements establish alternate
repair criteria for axial primary water stress corrosion cracking at
dented tube support plate intersections.
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.
Examination of crack morphology for primary water stress
corrosion cracking (PWSCC) at dented intersections has been found to
show one or two microcracks well aligned with only a few uncorroded
ligaments and little or no other inside diameter axial cracking at
the intersection. This relatively simple morphology is conducive to
obtaining good accuracy in nondestructive examination (NDE) sizing
of these indications. Accordingly, alternate repair criteria (ARC)
is established based on crack length and average and maximum depth
within the thickness of the tube support plate (TSP).
The application of the ARC requires a Monte Carlo condition
monitoring assessment to determine the as-found condition of the
tubing. The condition monitoring analysis described in WCAP-15128
Revision 3 is consistent with NRC Generic Letter 95-05 requirements.
The application of the ARC requires a Monte Carlo operational
assessment to determine the need for tube repair. The repair bases
are obtained by projecting the crack profile to the end of the next
operating cycle and determining the burst pressure and leakage for
the projected profile using Monte Carlo analysis techniques
described in WCAP-15128 Revision 3. The burst pressure and leakage
is compared to the requirements in WCAP-15128 Revision 3. Separate
analyses are required for the total crack length and the length
outside the TSP due to differences in requirements. If the projected
end of cycle (EOC) requirements are satisfied, the tube will be left
in service.
A steam generator (SG) tube rupture event is one of a number of
design basis accidents that are analyzed as part of a plant's
licensing basis. A single or multiple tube rupture event would not
be expected in a SG in which the ARC has been applied. The ARC
requires repair of any indication having a maximum crack depth
greater than or equal to 40 percent outside the TSP, thus limiting
the potential length of a deep crack outside the TSP at EOC
conditions and providing margin
[[Page 59224]]
against burst and leakage for free span indications.
For other design basis accidents such as a main steam line
break, main feed line break, control rod ejection, and locked
reactor coolant pump motor, the tubes are assumed to retain their
structural integrity.
Therefore, the proposed changes do not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
2. The proposed change does not create the possibility of a new
or different kind of accident from any accident previously
evaluated.
Implementation of the proposed SG tube ARC does not introduce
any significant changes to the plant design basis. A single or
multiple tube rupture event would not be expected in a SG in which
the ARC has been applied. Both condition monitoring and operational
assessments are completed as part of the implementation of ARC to
determine that structural and leakage margin exists prior to
returning SGs to service following inspections. If the condition
monitoring requirements are not satisfied for burst or leakage, the
causal factors for EOC indications exceeding the expected values
will be evaluated. The methodology and application of this ARC will
continue to ensure that tube integrity is maintained during all
plant conditions consistent with the requirements of Regulatory
Guide (RG) 1.121 and Revision 1 of RG 1.83. Therefore, a permanent
ARC is justified.
In the analysis of a SG tube rupture event, a bounding primary-
to-secondary leakage rate equal to the operational leakage limits in
the Technical Specifications (TS), plus the leak rate associated
with the double ended rupture of a single tube, is assumed. For
other design basis accidents, the tubes are assumed to retain their
structural integrity and exhibit primary-to-secondary leakage within
the limits assumed in the current licensing basis accident analyses.
Steam line break leakage rates from the proposed PWSCC ARC are
combined with leakage rates from other approved ARC (i.e., voltage-
based ARC and W* ARC). The combined leakage rates will not exceed
the limits assumed in the current licensing basis accident analyses.
The 40 percent maximum depth repair limit for free span
indications provides a very low likelihood of free span leakage
under design basis or severe accident conditions. Leakage from
indications inside the TSP is limited by the constraint of the TSP
even under severe accident conditions, and leakage behavior in a
severe accident would be similar to that found acceptable by the NRC
under approved ARC for axial outside diameter stress corrosion
cracking (ODSCC) at TSP intersections. Therefore, even under severe
accident conditions, it is concluded that application of the
proposed ARC for PWSCC at dented TSP locations results in a
negligible difference in risk of a tube rupture or large leakage
event, when compared to current 40 percent repair limits or
previously approved ARC.
DCPP continues to implement a maximum operating condition leak
rate limit of 150 gallons per day per SG to preclude the potential
for excessive leakage during all plant conditions.
The possibility of a new or different kind of accident from any
previously evaluated is not created because SG tube integrity is
maintained by inservice inspection, condition monitoring,
operational assessment, tube repair, and primary-to-secondary
leakage monitoring.
3. The proposed change does not involve a significant reduction
in a margin of safety.
Tube repair limits provide reasonable assurance that tubes
accepted for continued service without plugging or repair will
exhibit adequate tube structural and leakage integrity during
subsequent plant operation. The implementation of the proposed ARC
is demonstrated to maintain SG tube integrity consistent with the
criteria of draft NRC Regulatory Guide 1.121. The guidelines of RG
1.121 describe a method acceptable to the NRC staff for meeting
General Design Criteria (GDC) 2, 4, 14, 15, 31, and 32 by ensuring
the probability or the consequences of SG tube rupture remain within
acceptable limits. This is accomplished by determining the limiting
conditions of degradation of SG tubing, for which tubes with
unacceptable cracking should be removed from service.
Upon implementation of the proposed ARC, even under the worst-
case conditions, the occurrence of PWSCC at the tube support plate
elevations is not expected to lead to a SG tube rupture event during
normal or faulted plant conditions. The ARC involves a computational
assessment to be completed for each indication left in service
ensuring that performance criteria for tube integrity and leak
tightness are met until the next scheduled outage. Therefore, a
permanent ARC is justified.
As discussed below, certain tubes are excluded from application
of ARC. Existing tube integrity requirements apply to these tubes,
and the margin of safety is not reduced.
In addressing the combined loading effects of a loss-of-coolant
(LOCA) and safe shutdown earthquake (SSE) on the SGs (as required by
GDC 2), the potential exists for yielding of the TSP in the vicinity
of the wedge groups, accompanied by deformation of tubes and a
subsequent postulated in-leakage. Tube deformation could lead to
opening of pre-existing tight through wall cracks, resulting in
secondary to primary in-leakage following the event, which could
have an adverse affect on the Final Safety Analysis Report (FSAR)
results. Based on a DCPP analysis of LOCA and SSE, SG tubes located
in wedge region exclusion zones are susceptible to deformation, and
are excluded from application of ARC.
A DCPP tube stress analysis for feed line break (FLB)/steam line
break (SLB) plus SSE loading determined that high bending stresses
occur in certain SG tubes at the seventh TSP, because the stresses
exceed the maximum imposed bending stress for existing test data
(equal to approximately the lower tolerance limit yield stress).
These tubes are located in rows 11 to 15 and 36 to 46, and are
excluded from application of ARC.
Tube intersections that contain TSP ligament cracking are also
excluded from application of ARC.
Based on the above, it is concluded that the proposed license
amendment requires does not result in a significant reduction in
margin [of safety] with respect to the plant safety analyses as
defined in the FSAR or TS.
Therefore, the proposed changes do not involve a significant
reduction in the margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment requests involve no significant hazards consideration.
Attorney for licensee: Christopher J. Warner, Esq., Pacific Gas and
Electric Company, P.O. Box 7442, San Francisco, California 94120.
NRC Section Chief: Stephen Dembek.
Tennessee Valley Authority, Docket Nos. 50-259, 50-260 and 50-296,
Browns Ferry Nuclear Plant (BFN), Units 1, 2 and 3, Limestone County,
Alabama
Date of amendment request: August 28, 2000.
Description of amendment request: The proposed amendment would
revise the Units 1, 2 and 3 Technical Specifications (TS) to
incorporate Technical Specifications Task Force (TSTF) Nos.TSTF-71,
TSTF-208, TSTF-222, TSTF-284, TSTF-258 and TSTF-364. TSTFs are changes
that were submitted to the staff by the nuclear power industry TSTF
that have generic applicability. A description of each of the six TSTFs
follows: (1) TSTF-71, Revision 2, adds an example of the application of
the Safety Function Determination Program (SFDP) 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 (STA) 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
[[Page 59225]]
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 ITS 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.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analyses of the issue of no significant hazards
consideration, which are presented below:
(1) TSTF-71, Revision 2.
A. The proposed amendment does not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
The change adds an example of SFDP use to facilitate the
application of the TS, which serves to improve TS usefulness. The
proposed change is an administrative clarification of existing
requirements, and does not change TS requirements. Therefore, the
proposed amendment does not involve a significant increase in the
probability or consequences of an accident previously evaluated.
B. The proposed amendment does not create the possibility of a
new or different kind of accident from any accident previously
evaluated.
The proposed change does not involve a physical alteration of
the plant, add any new equipment, or require any existing equipment
to be operated in a manner different from the present design. The
proposed change will not impose any new or eliminate any existing
requirements. Therefore, the proposed amendment does not create the
possibility of a new or different kind of accident from any accident
previously evaluated.
C. The proposed amendment does not involve a significant
reduction in a margin of safety.
The proposed change will not reduce a margin of safety because
it has no effect on any safety analyses assumptions. This change is
administrative in nature. For these reasons, the proposed amendment
does not involve a significant reduction in the margin of safety.
(2) TSTF-208, Revision 0.
A. The proposed amendment does not involve a significant
increase in the probability or consequences of an accident
previously evaluated. The proposed change relaxes the Action time
for LCO 3.0.3. The subject Action time is not an initiating
condition for any accident previously evaluated and the accident
analyses do not assume that equipment is out of service (requiring
entry into LCO 3.0.3) prior to postulated events. Consequently, the
extended action time does not significantly increase the probability
of an accident previously evaluated. The consequences of an analyzed
accident during the extended action time are the same as the
consequences during the existing action time. As a result, the
consequences of an accident previously evaluated are not
significantly increased.
Therefore, the proposed amendment does not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
B. The proposed amendment does not create the possibility of a
new or different kind of accident from any accident previously
evaluated.
The proposed change does not involve a physical alteration of
the plant, add any new equipment, or require any existing equipment
to be operated in a manner different from the present design.
Therefore, the proposed amendment does not create the possibility of
a new or different kind of accident from any accident previously
evaluated.
C. The proposed amendment does not involve a significant
reduction in a margin of safety.
The proposed change will not reduce a margin of safety because
it has no effect on any safety analyses assumptions. The TS defines
specific time limits during which operation with degraded condition
is permitted. In this case, actual plant experience indicates that
the Action time in existing TS is too short to accomplish the
specified action to be in MODE 2 in an orderly manner. Extension of
the time would allow the reactor to be shutdown in a controlled
manner while minimizing risks associated with the initiation of
inadvertent transients. This maximizes reactor safety. For these
reasons, the proposed amendment does not involve a significant
reduction in the margin of safety.
(3) TSTF-222, Revision 1.
A. The proposed amendment does not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
The proposed change is an administrative clarification of
existing TS requirements which clarifies scram time testing
requirements for control rods. The rewording and reformatting
involves no technical changes to the existing TS. As such, there is
no effect on initiators of analyzed events or assumed mitigation of
accidents or transients. Therefore, the proposed amendment does not
involve a significant increase in the probability or consequences of
an accident previously evaluated.
B. The proposed amendment does not create the possibility of a
new or different kind of accident from any accident previously
evaluated.
The proposed change does not involve a physical alteration of
the plant, add any new equipment, or require any existing equipment
to be operated in a manner different from the present design.
Therefore, the proposed amendment does not create the possibility of
a new or different kind of accident from any accident previously
evaluated.
C. The proposed amendment does not involve a significant
reduction in a margin of safety.
The proposed change will not reduce a margin of safety because
it has no effect on any safety analyses assumptions. This change is
administrative in nature. For these reasons, the proposed amendment
does not involve a significant reduction in the margin of safety.
(4) TSTF-258, Revision 4.
A. The proposed amendment does not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
The proposed change is an administrative clarification of
existing TS requirements which clarifies and modifies administrative
controls in the areas of operator staffing requirements, working
hour limits, STA position, Radioactive Effluent Controls Program,
periodic reporting requirements for relief valve openings, and
radiological control requirements. These TS revisions do not affect
analysis inputs for analyzed accidents and transients. Therefore,
the proposed amendment does not involve a significant increase in
the probability or consequences of an accident previously evaluated.
B. The proposed amendment does not create the possibility of a
new or different kind of accident from any accident previously
evaluated.
The proposed change does not involve a physical alteration of
the plant, add any new equipment, or require any existing equipment
to be operated in a manner different from the present design.
Therefore, the proposed amendment does not create the possibility of
a new or different kind of accident from any accident previously
evaluated.
C. The proposed amendment does not involve a significant
reduction in a margin of safety.
The proposed changes are administrative type revisions and do
not reduce a margin of safety because they have no effect on any
safety analyses assumptions. For these reasons, the proposed
amendment does not involve a significant reduction in the margin of
safety.
(5) TSTF-284, Revision 3.
A. The proposed amendment does not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
The proposed change is an administrative clarification of
existing requirements. The change clarifies the TS terminology to
facilitate the use and application of Surveillance Requirement Notes
to improve TS use. Also, two additional examples of the application
of Surveillance Requirement Notes are incorporated. Therefore, the
proposed amendment does not involve a significant increase in the
probability or consequences of an accident previously evaluated.
B. The proposed amendment does not create the possibility of a
new or different kind of accident from any accident previously
evaluated.
[[Page 59226]]
The proposed change does not involve a physical alteration of
the plant, add any new equipment, or require any existing equipment
to be operated in a manner different from the present design. The
proposed change will not impose any new or eliminate any existing
requirements. Therefore, the proposed amendment does not create the
possibility of a new or different kind of accident from any accident
previously evaluated.
C. The proposed amendment does not involve a significant
reduction in a margin of safety.
The proposed change will not reduce a margin of safety because
it has no effect on any safety analyses assumptions. This change is
administrative in nature.
For these reasons, the proposed amendment does not involve a
significant reduction in the margin of safety.
(6) TSTF-364, Revision 0.
A. The proposed amendment does not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
The proposed change is an administrative modification of
existing TS requirements for the TS Bases change program to simply
reference changes pursuant to 10 CFR 50.59 rather than ``unreviewed
safety question''. This change is administrative and has no affect
on the current review and approval process for Final Safety Analyses
Report and Bases changes. As such, there is no effect on initiators
of analyzed events or assumed mitigation of accidents or transients.
Therefore, the proposed amendment does not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
B. The proposed amendment does not create the possibility of a
new or different kind of accident from any accident previously
evaluated. The proposed change does not involve a physical
alteration of the plant, add any new equipment, or require any
existing equipment to be operated in a manner different from the
present design. Therefore, the proposed amendment does not create
the possibility of a new or different kind of accident from any
accident previously evaluated.
C. The proposed amendment does not involve a significant
reduction in a margin of safety.
The proposed change modifies [sic] is an administrative
modification of existing TS requirements for the TS FSAR and Bases
change program to simply reference changes pursuant to 10 CFR 50.59
rather than ``unreviewed safety question.'' This change is
administrative and has no affect on the current review process for
FSAR and Bases changes, and will not reduce a margin of safety
because it has no effect on any safety analyses assumptions.
For these reasons, the proposed amendment does not involve a
significant reduction in the margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
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.
Tennessee Valley Authority, Docket Nos. 50-327 and 50-328, Sequoyah
Nuclear Plant, Units 1 and 2, Hamilton County, Tennessee
Date of application for amendments: August 31, 2000 (TS 00-05).
Brief description of amendments: The proposed amendment would
revise the Sequoyah Nuclear Plant (SQN) Technical Specifications (TSs).
The revision would relocate certain specifications related to
reactivity control that are not required to be contained in the TSs by
NRC regulations. These specifications include TSs 3.1.2.1 and 3.1.2.2
for boration flow paths, TSs 3.1.2.3 and 3.1.2.4 for boration charging
pumps, TSs 3.1.2.5 and 3.1.2.6 for borated water sources, TS 3.1.3.3
for position indication systems during shutdown, and TS 3.10.5 for
special test exceptions for the position indication system. These
specifications will be relocated in their entirety to the SQN Technical
Requirements Manual without changing the requirements currently
contained in the TSs.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), Tennessee Valley
Authority (TVA), 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 revision relocates the boration specifications and
one rod position indication system specification without a change to
the requirements and deletes a special test exception for rod
position indication that is no longer applicable to SQN. Relocation
to the TRM continues to provide an acceptable level of applicability
to plant operation and requires revisions to be processed in
accordance with the provisions in 10 CFR 50.59. Evaluations of
revisions in accordance with 10 CFR 50.59 will continue to ensure
that these specifications adequately control the functions for
boration and rod position indication systems to maintain safe
operation of the plant. The boration systems and the rod position
indication system is not postulated to be the initiator of a design
basis accident. Since there are no changes to these functions and
their operation will remain the same, the probability of an accident
is not increased by relocating these requirements to the TRM.
Additionally, the accident mitigation capability and offsite dose
consequences associated with accidents will not change because these
functions will not be altered by the proposed relocation. Therefore
the consequences of an accident are not increased by this relocation
to the TRM and the control of revisions to these specifications in
accordance with 10 CFR 50.59.
B. The proposed amendment does not create the possibility of a
new or different kind of accident from any accident previously
evaluated.
The proposed revision will not alter the functions for the
boration or the rod position indication systems such that accident
potential would be changed. The location of these specifications in
the TRM and the performance of revisions in accordance with 10 CFR
50.59 will continue to maintain acceptable operability requirements.
Therefore, the possibility of an accident of a new or different kind
is not created by the proposed relocation and deletion.
C. The proposed amendment does not involve a significant
reduction in a margin of safety.
The proposed specification relocation and special test deletion
will not affect plant setpoints or functions that maintain the
margin of safety. This is based on the relocation to the TRM
continuing to maintain the same level of operability requirements
and surveillance testing to adequately ensure functionality of the
boration and rod position indication systems. Control of TRM
requirements in accordance with 10 CFR 50.59 will ensure that
revisions to these functions will not inappropriately impact the
health and safety of the public without prior review and approval by
NRC. Therefore, the proposed relocation and deletion is acceptable
and will not reduce the margin of safety.
The NRC 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.
TXU Electric, Docket Nos. 50-445 and 50-446, Comanche Peak Steam
Electric Station, Units 1 and 2, Somervell County, Texas
Date of amendment request: May 2, 2000, as supplemented by letter
dated August 30, 2000.
Brief description of amendment: The proposed license amendment
would obtain approval from the Nuclear Regulatory Commission (NRC) of
changes to the Comanche Peak Steam Electric Station, Units 1 and 2
(CPSES)
[[Page 59227]]
Security Plans prior to their implementation. Prior approval is being
requested from the NRC, in accordance with the requirements of 10 CFR
50.54(p)(1), because some of the changes could have the potential of
reducing the effectiveness of the security plans.
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?
Response: No
The proposed changes involving Security activities, do not
reduce the ability for the Security organization to prevent
radiological sabotage and therefore does not increase the
probability or consequences of a radiological release previously
evaluated.
2. Do the proposed changes create the possibility of a new or
different kind of accident from any accident previously evaluated?
Response: No
The proposed changes involve functions of the Security
organization concerning intrusion detection, material search
requirements, alarm response and compensation, and vehicle control.
Analysis of the proposed changes has not indicated nor identified a
new or different kind of accident previously evaluated.
3. Do the proposed changes involve a significant reduction in a
margin of safety?
Response: No
Analysis of the proposed changes show that the proposed changes
affect only the functions of the Security organization and have no
impact upon nor cause a significant reduction in margin of safety
for plant operation. The failure points of key safety parameters are
not affected.
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: George L. Edgar, Esq., Morgan, Lewis and
Bockius, 1800 M Street, NW., Washington, DC 20036
NRC Section Chief: Robert A. Gramm
Wolf Creek Nuclear Operating Corporation, Docket No. 50-482, Wolf Creek
Generating Station, Coffey County, Kansas
Date of amendment request: September 15, 2000 (WO 00-0036).
Description of amendment request: The proposed amendment would
revise footnotes (b) and (c) to Table 1.1-1, ``Modes,'' of the Wolf
Creek Technical Specifications, and allow one reactor pressure vessel
(RPV) head closure bolt to not be fully tensioned in Mode 4 (hot
shutdown) and Mode 5 (cold shutdown). Each RPV head closure bolt is
composed of a stud, nut, and washer, and the bolts attach the vessel
head to the vessel body. The proposed revisions would allow the plant
(1) to be in Modes 4 and 5 with only 53 of 54 RPV head closure bolts
fully tensioned (i.e., to allow one head closure bolt to not be fully
tensioned), and (2) to operate with one RPV head closure bolt less than
fully tensioned. The proposed revision to footnote (b) requires the
proposed revision to the definition of refueling (i.e., footnote (c)
from the current ``one or more'' RPV head closure bolts detensioned to
the proposed ``two or more'' RPV head closure bolts detensioned). In
refueling, the RPV head closure bolts are detensioned and removed from
the vessel body, and the RPV head is removed from the vessel. The
licensee committed to the following program before operating with a not
fully tensioned RPV head closure bolt: (1) The circumstances for the
closure bolt not being fully tensioned will be reviewed to determined
that the analysis in the application is still applicable, (2) the RPV
will not be subject to hydrostatic test conditions before the closure
bolt is returned to service, and (3) the plant heatup rate will be held
to 50 deg.F per hour (half the normal rate) until the closure bolt is
returned to service.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. The proposed change does not involve a significant increase
in the probability or consequences of an accident previously
evaluated.
Overall protection system performance will remain within the
bounds of the accident analyses, since no hardware changes are
proposed. Since the stresses [in the RPV head and body] remain
within [the American Society of Mechanical Engineers Boiler and
Pressure Vessel (ASME)] Code allowables, the proposed change will
not affect the probability of any event initiators nor will the
proposed change affect the ability of any safety related equipment
to perform its intended function. There will be no degradation in
the performance of nor an increase in the number of challenges
imposed on safety related equipment assumed to function during an
accident situation.
Therefore, the proposed change does not involve a significant
increase in the probability or consequences of an accident
previously analyzed.
2. The proposed change does not create the possibility of a new
or different kind of accident from any accident previously
evaluated.
There are no hardware changes nor are there any changes in the
method by which any safety related plant system performs it[s]
safety function. The method of plant operation is unaffected.
Leakage would be precluded since, as noted in the evaluation [in the
application dated September 15, 2000,] adequate compression [of the
RPV head closure bolts] remains. However, if leakage were to result
from having less than the total number of closure studs fully
tensioned it would be detected by an increase in the temperature on
the leak-off line from the annular space between the inner and outer
vessel head o-rings. That temperature increase would be detected by
installed temperature indicators and alarmed in the control room.
Any leakage would be detected as an increase in RCS [reactor coolant
system] identified LEAKAGE. Since stresses remain within Code
allowables, no new accident scenarios, transient precursors, failure
mechanisms, or limiting single failures are introduced as a result
of this change.
Therefore, the proposed change will not create a new or
different kind of accident from any accident previously evaluated.
3. The proposed change does not involve a significant reduction
in a margin of safety.
As indicated in Section 4.0 above [of the application dated
September 15, 2000,] ASME Section III stress limits for effected
components are not exceeded. The evaluations indicate that the
reactor vessel will continue to meet ASME Code allowable stress
criteria with a single untensioned reactor vessel closure stud, or
with a single closure stud which fails in service. The proposed
change does not alter nor exceed the acceptance criteria for any
analyzed event. There will be no effect on the manner in which
safety limits or limiting safety system settings are determined nor
will there be any effect on those plant systems necessary to assure
the accomplishment of protection functions.
Therefore, the proposed change to the Technical Specifications
do not involve a significant reduction in the margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration. The
licensee's reference to ``closure studs'' in the above not significant
hazards consideration is a reference to the ``closure bolts'' in the
proposed amendment.
Attorney for licensee: Jay Silberg, Esq., Shaw, Pittman, Potts and
Trowbridge, 2300 N Street, NW., Washington, DC 20037.
NRC Section Chief: Stephen Dembek.
[[Page 59228]]
Previously Published Notices of Consideration of Issuance of
Amendments to Facility Operating Licenses, Proposed No Significant
Hazards Consideration Determination, and Opportunity for a Hearing
The following notices were previously published as separate
individual notices. The notice content was the same as above. They were
published as individual notices either because time did not allow the
Commission to wait for this biweekly notice or because the action
involved exigent circumstances. They are repeated here because the
biweekly notice lists all amendments issued or proposed to be issued
involving no significant hazards consideration.
For details, see the individual notice in the Federal Register on
the day and page cited. This notice does not extend the notice period
of the original notice.
Duke Energy Corporation, Docket Nos. 50-269, 50-270, and 50-287, Oconee
Nuclear Station, Units 1, 2, and 3, Oconee County, South Carolina
Date of amendment request: September 7, 2000.
Description of amendment request: The amendments revise
Surveillance Requirement 3.8.1.9.a by adding a note that states the
upper limits on frequency and voltage are not required to be met for
the annual test of the Keowee Hydro Units until the NRC issues an
amendment that removes the note in response to an amendment request to
be submitted no later than April 5, 2001.
Date of publication of individual notice in Federal Register:
September 19, 2000 (65 FR 56600).
Expiration date of individual notice: October 3, 2000, for
comments; October 19, 2000, for hearings.
Niagara Mohawk Power Corporation, Docket No. 50-410, Nine Mile Point
Nuclear Station Unit No. 2, Oswego County, New York
Date of application for amendment: July 14, 2000.
Brief description of amendment request: The proposed amendment
would revise License Condition 2.c.(10), ``Additional Condition 1,''
which was imposed by Amendment No. 91 dated February 15, 2000. License
Condition 2.c.(10) defines the meaning of implementation of Improved
Technical Specifications and specifies that implementation be completed
by August 31, 2000. The licensee has proposed to revise the
implementation date from August 31, 2000, to December 31, 2000.
Date of publication of individual notice in Federal Register: July
27, 2000 (65 FR 46183).
Expiration date of individual notice: August 28, 2000.
Tennessee Valley Authority, Docket No. 50-390 Watts Bar Nuclear Plant,
Unit 1, Rhea County, Tennessee
Date of application for amendments: June 7, as supplemented June
23, and August 24, 2000.
Brief description of amendment: Changes to Facility Operating
License and Technical Specifications to reflect an increase in
allowable thermal power from 3411 to 3459 megawatts.
Date of publication of individual notice in the Federal Register:
September 7, 2000 (65 FR 54322).
Expiration date of individual notice: October 10, 2000.
Notice of Issuance of Amendments to Facility Operating Licenses
During the period since publication of the last biweekly notice,
the Commission has issued the following amendments. The Commission has
determined for each of these amendments that the application complies
with the standards and requirements of the Atomic Energy Act of 1954,
as amended (the Act), and the Commission's rules and regulations. The
Commission has made appropriate findings as required by the Act and the
Commission's rules and regulations in 10 CFR Chapter I, which are set
forth in the license amendment.
Notice of Consideration of Issuance of Amendment to Facility
Operating License, Proposed No Significant Hazards Consideration
Determination, and Opportunity for A Hearing in connection with these
actions was published in the Federal Register as indicated.
Unless otherwise indicated, the Commission has determined that
these amendments satisfy the criteria for categorical exclusion in
accordance with 10 CFR 51.22. Therefore, pursuant to 10 CFR 51.22(b),
no environmental impact statement or environmental assessment need be
prepared for these amendments. If the Commission has prepared an
environmental assessment under the special circumstances provision in
10 CFR 51.12(b) and has made a determination based on that assessment,
it is so indicated.
For further details with respect to the action see (1) the
applications for amendment, (2) the amendment, and (3) the Commission's
related letter, Safety Evaluation and/or Environmental Assessment as
indicated. All of these items are available for public inspection at
the NRC'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, Docket No. 50-261, H. B. Robinson,
Steam Electric Plant, Unit No. 2, Darlington County, South Carolina
Date of application for amendment: November 30, 1999.
Brief description of amendment: This amendment revises the testing
requirements in Technical Specification 5.5.11, ``Ventilation Filter
Testing Program (VFTP),'' in response to Generic Letter 99-02,
``Laboratory Testing of Nuclear-Grade Activated Charcoal.''
Date of issuance: September 14, 2000.
Effective date: September 14, 2000.
Amendment No.: 189.
Facility Operating License No. DPR-23. Amendment revises the
Technical Specifications.
Date of initial notice in Federal Register: December 29, 1999 (64
FR 73087).
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated September 14, 2000.
No significant hazards consideration comments received: No.
Carolina Power & Light Company, et al., Docket No. 50-400, Shearon
Harris Nuclear Power Plant, Unit 1, Wake and Chatham Counties, North
Carolina.
Date of application for amendment: June 7, 2000, as supplemented on
August 23, 2000.
Brief description of amendment: This amendment revises the
surveillance test intervals and allowed outage times for Engineered
Safety Features Actuation System (ESFAS) instrumentation in Technical
Specification (TS) 3/4.3.2. It also revises the reactor trip system
instrumentation requirements in TS 3/4.3.1 associated with implementing
the ESFAS relaxations.
Date of issuance: September 13, 2000.
Effective date: September 13, 2000.
Amendment No.: 101.
Facility Operating License No. NPF-63. Amendment revises the
Technical Specifications.
Date of initial notice in Federal Register: July 12, 2000 (65 FR
43044).
The supplemental submittal dated August 23, 2000, provided
clarifying information only, and did not change the initial no
significant hazards consideration determination.
[[Page 59229]]
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated September 13, 2000.
No significant hazards consideration comments received: No.
Commonwealth Edison Company, Docket Nos. 50-237 and 50-249, Dresden
Nuclear Power Station, Units 2 and 3, Grundy County, Illinois
Date of application for amendments: February 23, 2000, as
supplemented by letters dated June 19 and July 17, 2000.
Brief description of amendments: The amendments changed Technical
Specification (TS) 3/4.6.K to revise the reactor pressure-temperature
(P-T) limits; changed TSs 1.0 and 3/4.12.C to delete a special test
exception that allowed the hydrostatic test to be performed above 212
degrees Fahrenheit while in Mode 4; and added a condition to the Unit 2
and 3 licenses to specify expiration dates for the P-T limits.
Date of issuance: September 19, 2000.
Effective date: Immediately, to be implemented within 30 days.
Amendment Nos.: 179 and 174.
Facility Operating License Nos. DPR-19 and DPR-25: The amendments
revised the Facility Operating Licenses and Technical Specifications.
Date of initial notice in Federal Register: April 5, 2000 (65 FR
17911).
The June 19 and July 17, 2000, letters are within the scope of the
original notice and did not change the original no significant hazards
consideration determination. The Commission's related evaluation of the
amendments is contained in a Safety Evaluation dated September 19,
2000.
No significant hazards consideration comments received: No.
Commonwealth Edison Company, Docket Nos. 50-237 and 50-249, Dresden
Nuclear Power Station, Units 2 and 3, Grundy County, Illinois
Date of application for amendments: August 3, 1999, as supplemented
by letter dated February 25, 2000.
Brief description of amendments: The amendments revised Technical
Specification (TS) Section 2.1.B to reflect a change to the Minimum
Critical Power Ratio for Unit 2; added an approved analytical method to
TS Section 6.9.A.6 for Units 2 and 3 for use in determining core
operating limits; and added conditions to the Unit 2 and 3 licenses to
limit the maximum rod average burnup for any rod to 60 GWD/MTU until
the staff has completed an environmental assessment supporting a
greater limit.
Date of issuance: September 21, 2000.
Effective date: Immediately, to be implemented within 30 days.
Amendment Nos.: 180 and 175.
Facility Operating License Nos. DPR-19 and DPR-25: The amendments
revised the Facility Operating Licenses and Technical Specifications.
Date of initial notice in Federal Register: September 8, 1999 (64
FR 48859). The February 25, 2000, submittal provided additional
clarifying information that did not change the initial proposed no
significant hazards consideration.
The Commission's related evaluation of the amendments is contained
in an Environmental Assessment dated September 19, 2000, and a Safety
Evaluation dated September 21, 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: February 29, 2000, as
supplemented by letter dated July 5, 2000.
Brief description of amendments: The amendments revised the
Technical Specifications Table 3.3.2-1, Engineered Safety Feature
Actuation System Instrumentation, Function 6.f, Auxiliary Feedwater
Pump Suction Pressure-Lo.
Date of issuance: September 13, 2000.
Effective date: As of the date of issuance and shall be implemented
within 30 days from the date of issuance.
Amendment Nos.: 193 and 174.
Facility Operating License Nos. NPF-9 and NPF-17: Amendments
revised the Technical Specifications.
Date of initial notice in Federal Register: August 23, 2000 (65 FR
51350).
The supplement dated July 5, 2000, provided clarifying information
that did not change the scope of the February 29, 2000, application and
the initial proposed no significant hazards consideration
determination.
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated September 13, 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: January 6, 2000, as
supplemented by letter dated July 20, 2000.
Brief description of amendments: The amendments revised the
Technical Specifications (TS) 3.3.1, ``Reactor Trip System
Instrumentation;'' TS 3.3.2, ``Engineered Safety Features Actuation
System Instrumentation;'' TS 3.3.5, ``Loss of Power Diesel Generator
Start Instrumentation;'' and TS 3.3.6, ``Containment Purge and Exhaust
Isolation Instrumentation.''
Date of issuance: September 18, 2000.
Effective date: As of the date of issuance and shall be implemented
within 30 days from the date of issuance.
Amendment Nos.: 194/175.
Facility Operating License Nos. NPF-9 and NPF-17: Amendments
revised the Technical Specifications.
Date of initial notice in Federal Register: March 22, 2000 (65 FR
15378).
The supplement dated July 20, 2000, provided clarifying information
that did not change the scope of the January 6, 2000, application and
the initial proposed no significant hazards consideration
determination.
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated September 18, 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: June 29, 2000, as supplemented
by letters dated July 27, and August 10, 2000. Other related
information was submitted by letters dated April 10, April 17, and June
19, 2000.
Brief description of amendments: The amendments revised the
Technical Specifications (TS) to reference the Westinghouse Best
Estimate Large Break Loss-of-Coolant Accident analysis methodology
described in WCAP-12945-P-A, March 1998. The changes also address
corresponding TS Bases changes.
Date of issuance: September 22, 2000.
Effective date: As of the date of issuance and shall be implemented
within 30 days from the date of issuance.
Amendment Nos.: 195 and 176.
Facility Operating License Nos. NPF-9 and NPF-17: Amendments
revised the Technical Specifications.
Date of initial notice in Federal Register: August 23, 2000 (65 FR
51349).
The supplements dated April 10, April 17, June 19, July 27, and
August 10, 2000, provided clarifying information that did not change
the scope of the June 29, 2000, application and the initial proposed no
significant hazards consideration determination.
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated September 22, 2000.
[[Page 59230]]
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: April 26, 1999; supplemented May
15, July 26, and August 23, 2000.
Brief description of amendments: The amendments revised various
provisions of the Technical Specifications and Final Safety Analysis
Report related to the steam generator tube loads following a main steam
line break and runout protection for the turbine-driven emergency
feedwater pump.
Date of Issuance: September 18, 2000.
Effective date: As of the date of issuance and shall be implemented
within 30 days from the date of issuance.
Amendment Nos.: 315, 315, & 315.
Facility Operating License Nos. DPR-38, DPR-47, and DPR-55:
Amendments revised the Technical Specifications and Final Safety
Analysis Report.
Date of initial notice in Federal Register: May 19, 1999 (64 FR
27320).
The supplements dated May 15, July 26, and August 23, 2000,
provided clarifying information that did not change the scope of the
April 26, 2000, application and the initial proposed no significant
hazards consideration determination.
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated September 18, 2000.
No significant hazards consideration comments received: No.
Energy Northwest, Docket No. 50-397, WNP-2, Benton County, Washington
Date of application for amendment: April 13, 2000, as supplemented
by letter dated May 15, 2000.
Brief description of amendment: The amendment revised Surveillance
Requirements 3.3.1.1.10 for Function 8 of Table 3.3.1.1-1 and
3.3.4.1.2.a. for reactor protection system and end of cycle
recirculation pump trip instrumentation of the WNP-2 technical
specifications. The amendment extends the frequency of these
surveillance requirements from 18 months to 24 months.
Date of issuance: September 15, 2000.
Effective date: September 15, 2000, to be implemented within 30
days from the date of issuance.
Amendment No.: 168.
Facility Operating License No. NPF-21: The amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: June 14, 2000 (65 FR
37423).
The May 15, 2000, supplemental letter provided clarifying
information, did not expand the scope of the application as originally
noticed and did not change the staff's original proposed no significant
hazards consideration determination.
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated September 15, 2000.
No significant hazards consideration comments received: No.
FirstEnergy Nuclear Operating Company, Docket No. 50-346, Davis-Besse
Nuclear Power Station, Unit 1, Ottawa County, Ohio
Date of application for amendment: July 28, 1999, as supplemented
on June 6, 2000.
Brief description of amendment: This amendment revised the ultimate
heat sink (UHS) average water temperature from 85 degrees Fahrenheit
( deg.F) to 90 deg.F and permits plant operations in
Operating Modes 1 through 4 with an average water temperature of
90 deg.F.
Date of issuance: September 12, 2000.
Effective date: Immediately, to be implemented within 90 days.
Amendment No.: 242.
Facility Operating License No. NPF-3: The amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: August 25, 1999 (64 FR
46438). The June 6, 2000, submittal provided additional clarifying
information that did not change the initial proposed no significant
hazards consideration determination. The Commission's related
evaluation of the amendment is contained in a Safety Evaluation dated
September 12, 2000.
No significant hazards consideration comments received: No.
FirstEnergy Nuclear Operating Company, Docket No. 50-346, Davis-Besse
Nuclear Power Station, Unit 1, Ottawa County, Ohio
Date of application for amendment: September 7, 1999, supplemented
July 14, 2000.
Brief description of amendment: This amendment revises Technical
Specification 3/4.3.2.1, Safety Features Actuation System
Instrumentation Trip Setpoints, to remove the ``Trip Setpoint'' values
for Instrument String Functional Unit ``b'', Containment Pressure-High,
and Functional Unit ``c'', Containment Pressure-High-High, and also
modifies the ``Allowable Values'' entry for these same Functional
Units, consistent with updated calculations using current setpoint
methodology. The changes also revise Limiting Conditions for Operation
(LCO) 3.3.2.1, and Bases 3/4.3.1 and 3/4.3.2 to reflect the removal of
the ``Trip Setpoint'' values for these Functional Units.
Date of issuance: September 14, 2000.
Effective date: As of the date of issuance and shall be implemented
within 90 days.
Amendment No.: 243.
Facility Operating License No. NPF-3: Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: December 15, 1999 (64
FR 70086).
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated September 14, 2000.
No significant hazards consideration comments received: No.
FirstEnergy Nuclear Operating Company, Docket No. 50-440, Perry Nuclear
Power Plant, Unit 1, Lake County, Ohio
Date of application for amendment: August 4, 1999, and as
supplemented by letter dated August 7, 2000.
Brief description of amendment: This amendment revised Technical
Specification 3.9.1, ``Refueling Equipment Interlocks,'' by introducing
an optional operator action when one or more required refueling
equipment interlocks are inoperable. The new operator action permits
continued in-vessel fuel movement under specific administrative
controls.
Date of issuance: September 12, 2000.
Effective date: As of the date of issuance and shall be implemented
within 90 days.
Amendment No.: 116.
Facility Operating License No. NPF-58: This amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: August 25, 1999 (64 FR
46439).
The August 7, 2000, supplement contained clarifying information
that was within the scope of the original application and Federal
Register notice and did not change the staff's initial proposed no
significant hazards consideration determination.
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated September 12, 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: September 16, 1999, as
supplemented
[[Page 59231]]
by letters dated May 3 and June 29, 2000.
Brief description of amendment: To allow an increase in the spent
fuel pool (SFP) storage capacity by replacing fuel racks in the ``B''
SFP with new high-density fuel racks.
Date of issuance: September 13, 2000.
Effective date: September 13, 2000.
Amendment No.: 193.
Facility Operating License No. DPR-31: Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: December 8, 1999 (64 FR
68702).
The May 3 and June 29, 2000 supplements provided clarifying
information and did not affect the initial no significant hazards
determination.
The Commission's related evaluation of the amendment is contained
in an Environmental Assessment dated September 5, 2000 and in a Safety
Evaluation dated September 13, 2000.
No significant hazards consideration comments received: No.
GPU Nuclear, Inc. et al., Docket No. 50-219, Oyster Creek Nuclear
Generating Station, Ocean County, New Jersey
Date of application for amendment: June 18, 1999, as supplemented
on June 22 and December 10, 1999, and February 10, and May 2, 2000.
Brief description of amendment: The amendment revised the Technical
Specifications to reflect the installation of additional spent fuel
pool storage racks. The additional new racks will provide 390
additional spent fuel assembly storage locations.
Date of issuance: September 15, 2000.
Effective date: As of the date of issuance to be implemented within
30 days.
Amendment No.: 215
Facility Operating License No. DPR-16: Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: August 17, 1999 (64 FR
44757).
The supplemental letters dated June 22 and December 10, 1999, and
February 10 and May 2, 2000, did not affect the proposed finding of no
significant hazards consideration, and was within the scope of the
amendment application as noticed.
The Commission's related evaluation of this amendment is contained
in a Safety Evaluation dated September 15, 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: February 18, 2000, as superseded by a
letter dated June 20, 2000.
Description of amendment request: The amendment changes the
Seabrook Station Technical Specifications to provide operational
flexibility during the shutdown modes of operation. These enhancements
include: (1) The ability to have a standby Safety Injection (SI) pump
available during Reactor Coolant System (RCS) reduced inventory
conditions with the RCS pressure boundary intact; (2) realigning a
footnote to clarify the allowance of an inoperable SI pump to be
energized for testing or filling accumulators; (3) allowance for an
additional charging pump to be made capable of injection during pump-
swap operations; (4) recognition that a substantial vent area exists
for cold overpressure protection when the reactor vessel head is on and
the studs are fully detensioned; (5) limit maneuvering the plant beyond
Hot Shutdown when one charging pump is operable; and (6) establishes a
new value for the open permissive interlock associated with the
Residual Heat Removal System suction isolation valves.
Date of issuance: September 11, 2000.
Effective date: As of its date of issuance, and shall be
implemented within 60 days.
Amendment No.: 74.
Facility Operating License No. NPF-86: Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: August 9, 2000 (65 FR
48752).
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated September 11, 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: December 3, 1999.
Description of amendment request: The amendment changes the
Technical Specifications by incorporating reference to the American
Society for Testing and Materials (ASTM) Standard D3803-1989,
``Standard Test Method for Nuclear-Grade Activated Charcoal,'' as the
test protocol for charcoal filter laboratory testing. In addition,
there is a change to Surveillance Requirement 4.7.6.1d.5) and
4.9.12d.4) specifying a minimum required heater output based on design
rated voltage.
Date of issuance: September 19, 2000.
Effective date: As of its date of issuance, and shall be
implemented within 60 days.
Amendment No.: 75
Facility Operating License No. NPF-86: Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: January 26, 2000 (65 FR
4282).
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated September 19, 2000.
No significant hazards consideration comments received: No.
Northeast Nuclear Energy Company, et al., Docket No. 50-336, Millstone
Nuclear Power Station, Unit No. 2, New London County, Connecticut
Date of application for amendment: February 1, 2000, as
supplemented on June 1 and July 13, 2000.
Brief description of amendment: The amendment modifies Technical
Specification (TS) 3.0.3 to state that this specification is not
applicable in MODES 5 or 6. The amendment also makes various changes to
TSs 3/4.1 ``Reactor Coolant System--Coolant Loops and Coolant
Recirculation'' and 3/4.9.8, ``Refueling Operations--Shutdown Cooling
and Coolant Circulation.'' In addition, various corrections and formats
are revised to achieve consistency of the structure and wording of the
TSs. The Bases for the affected TSs have also been revised accordingly.
Date of issuance: September 14, 2000.
Effective date: As of the date of issuance and shall be implemented
within 60 days from the date of issuance.
Amendment No.: 249.
Facility Operating License No. DPR-65: Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: July 31, 2000 (65 FR
46748).
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated September 14, 2000.
No significant hazards consideration comments received: No.
Northeast Nuclear Energy Company, et al., Docket No. 50-423, Millstone
Nuclear Power Station, Unit No. 3, New London County, Connecticut
Date of application for amendment: January 18, 1999, and
supplemented by letters dated April 5 and December 21, 1999; and May 2
and August 10, 2000.
[[Page 59232]]
Brief description of amendment: The amendment revises the Technical
Specifications (TSs) and the Final Safety Analysis Report for Millstone
3 to allow an entire reactor core to be offloaded to the spent fuel
pool (SFP) and an increase in the maximum design basis normal SFP water
temperature limit from 140 deg.F to 150 deg.F during planned
refueling outages. The increase in maximum design basis normal SFP
water temperature up to 150 deg.F affects certain Fuel Building area
TS temperature limits that require a revision to the TSs.
Date of issuance: September 12, 2000.
Effective date: As of the date of issuance and shall be implemented
within 60 days from the date of issuance.
Amendment No.: 182.
Facility Operating License No. NPF-49: Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: March 10, 1999 (64 FR
11962).
The letters dated April 5 and December 21, 1999, and May 2, and
August 10, 2000, provided clarifying information and did not change the
staff's initial proposed no significant hazards consideration
determination or expand the scope of the application as published in
the Federal Register.
The Commission's related evaluation of this amendment is contained
in a Safety Evaluation dated September 12, 2000.
No significant hazards consideration comments received: No.
Northeast Nuclear Energy Company, et al., Docket No. 50-423, Millstone
Nuclear Power Station, Unit No. 3, New London County, Connecticut
Date of application for amendment: February 3, 2000
Brief description of amendment: The amendment changes the Millstone
Unit No. 3 Final Safety Analysis Report to show that the configuration
of valves 3CHS*V61 and 3CHS*V62 takes exception to the American Society
of Mechanical Engineers Section III code requirements for class 2
components.
Date of issuance: September 15, 2000.
Effective date: As of the date of issuance and shall be implemented
within 60 days from the date of issuance.
Amendment No.: 183.
Facility Operating License No. NPF-49: Amendment revised the Final
Safety Analysis Report.
Date of initial notice in Federal Register: June 28, 2000 (65 FR
39958).
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated September 15, 2000.
No significant hazards consideration comments received: No.
PECO Energy Company, Docket Nos. 50-352 and 50-353, Limerick Generating
Station, (LGS) Units 1 and 2, Montgomery County, Pennsylvania
Date of amendment request: November 5, 1999, as supplemented July
17, 2000.
Description of amendment request: The amendments make changes to
Technical Specifications Sections 4.6.5.3.b.2 and 4.6.5.3.c, ``Standby
Gas Treatment System,'' 4.6.5.4.b.2 and 4.6.5.4.c, ``Reactor Enclosure
Recirculation System,'' and 4.7.2.c.2 and 4.7.2.d, ``Control Room
Emergency Air System.''
Date of Issuance: September 8, 2000.
Effective Date: September 8, 2000.
Amendment Nos.: 144 & 106.
Facility Operating License Nos. NPF-39 and NPF-85. The amendments
revised the Technical Specifications.
Date of Initial notice in Federal Register: January 26, 2000 (65 FR
4287)
The July 17, 2000, letter provided clarifying information that did
not change the initial no significant hazards consideration
determination or expand the scope of the original Federal Register
Notice.
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated September 8, 2000.
No significant hazards consideration comments received: No.
PECO Energy Company, Docket No. 50-352, Limerick Generating Station,
Unit 1, Montgomery County, Pennsylvania
Date of application for amendment: May 15, 2000, as supplemented
August 10, 2000.
Brief description of amendment: The amendment revises the pressure-
temperature limit curves.
Date of issuance: September 15, 2000.
Effective date: As of date of issuance, to be implemented within 30
days.
Amendment No.: 145.
Facility Operating License No. NPF-39. This amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: July 12, 2000 (65 FR
43051).
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated September 15, 2000. The August 10, 2000,
letter provided clarifying information that did not change the initial
no significant hazard consideration determination or expand the scope
of the original Federal Register notice.
No significant hazards consideration comments received: No.
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: April 20, 2000 (PCN-503), and
supplemented by letter dated June 6, 2000.
Brief description of amendments: The amendments revise TS 5.5.2.5,
``Reactor Coolant Pump Flywheel Inspection Program'' by changing the
volumetric examination frequency of the upper flywheel on each of the
primary reactor coolant pump motors from a 3-year to a 10-year cycle.
Date of issuance: September 8, 2000.
Effective date: September 8, 2000, to be implemented within 30 days
of issuance.
Amendment Nos.: Unit 2--170; Unit 3--161.
Facility Operating License Nos. NPF-10 and NPF-15: The amendments
revised the Technical Specifications.
Date of initial notice in Federal Register: May 17, 2000 (65 FR
31360). The supplemental letter dated June 6, 2000, provided clarifying
information that was within the scope of the April 20, 2000,
application and the Federal Register notice and did not change the
staff's initial proposed no significant hazards consideration
determination.
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated September 8, 2000.
No significant hazards consideration comments received: No.
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: September 12, 2000.
Brief description of amendments: The amendments revise TS 3.6.6.1,
``Containment Spray and Cooling Systems,'' to change the allowed outage
time (AOT) for a single inoperable train of the containment spray
system from 72 hours to 7 days. Also, the combined AOT that appears in
both Conditions A and C of TS 3.6.6.1 is revised from 10 days to 14
days.
Date of issuance: September 12, 2000.
Effective date: September 12, 2000, to be implemented within 30
days of issuance.
Amendment Nos.: Unit 2--171; Unit 3--162.
Facility Operating License Nos. NPF-10 and NPF-15: The amendments
revised the Technical Specifications.
[[Page 59233]]
Date of initial notice in Federal Register: May 3, 2000 (65 FR
25769).
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated September 12, 2000.
No significant hazards consideration comments received: No.
Southern Nuclear Operating Company, Inc., et al., Docket Nos. 50-424
and 50-425, Vogtle Electric Generating Plant, Units 1 and 2, Burke
County, Georgia
Date of application for amendments: March 6, 2000, as supplemented
by letter dated July 7, 2000.
Brief description of amendments: The amendments revised Technical
Specification (TS) 3.9.4, ``Containment Penetration,'' allowing the
equipment hatch to be open during core alteration and/or during
movement of irradiated fuel within the containment, provided the
capability for closure is maintained.
Date of issuance: September 11, 2000.
Effective date: As of the date of issuance and shall be implemented
within 30 days from the date of issuance.
Amendment Nos.: 115 and 93.
Facility Operating License Nos. NPF-68 and NPF-81: Amendments
revised the Technical Specifications.
Date of initial notice in Federal Register: June 28, 2000 (65 FR
39961), July 20, 2000 (65 FR 45115). The supplemental letter dated July
7, 2000, provided clarifying information that did not change the scope
of the March 6, 2000, application and the initial proposed no
significant hazards consideration determination.
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated September 11, 2000.
No significant hazards consideration comments received: No.
Tennessee Valley Authority, Docket No. 50-3a0, Watts Bar Nuclear Plant,
Unit 1, Rhea County, Tennessee
Date of application for amendment: July 10, 2000 (TS 00-08).
Brief description of amendment: Regarding the need to conduct
channel operational tests within 12 hours prior to physics tests and
the placing of a reactor trip instrumentation channel used in physics
tests in a bypassed condition instead of a tripped condition.
Date of issuance: September 13, 2000.
Effective date: September 13, 2000.
Amendment No.: 28.
Facility Operating License No. NPF-90: Amendment revises the
Technical Specifications.
Date of initial notice in Federal Register: August 9, 2000 (65 FR
48759).
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated September 13, 2000.
No significant hazards consideration comments received: No.
Wolf Creek Nuclear Operating Corporation, Docket No. 50-482, Wolf Creek
Generating Station, Coffey County, Kansas
Date of amendment request: June 23, 2000, and supplements dated
July 21 and 26, 2000.
Brief description of amendment: The amendment revises TS 3.9.4,
``Containment Penetrations,'' to allow containment penetrations (with
direct access to the outside atmosphere) to be unisolated under
administrative controls during refueling operations with core
alterations or irradiated fuel movement inside containment. The
amendment (1) revises the note in the Limiting Condition for Operation
3.4.9 for containment penetrations that may be unisolated under
administrative controls, deleting the reference to penetrations P-63
and P-98, and (2) deletes the exception for penetrations P-63 and P-98
in Surveillance Requirement 3.9.4.1. In addition, there are format and
editorial corrections to TS 3.8.3, ``Diesel Fuel Oil, Lube Oil, and
Start Air,'' and TS 5.2.2.b, ``Administrative Controls,'' to correct
errors issued in Amendment No. 123, issued March 31, 1999.
Date of issuance: September 12, 2000.
Effective date: September 12, 2000, to be implemented within 30
days of the date of issuance, including the completion of the
administrative procedures that ensure that open containment
penetrations, with direct access to the outside atmosphere during
refueling operations with core alterations and irradiated fuel movement
inside containment, will be promptly closed in the event of a fuel
handling accident inside containment.
Amendment No.: 135.
Facility Operating License No. NPF-42. The amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: July 12, 2000 (65 FR
43053). The July 21 and 26, 2000, supplements provided additional
clarifying information, did not expand the scope of the application as
originally noticed, and did not change the staff's original proposed no
significant hazards consideration determination.
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated September 12, 2000.
No significant hazards consideration comments received: No.
Dated at Rockville, Maryland, this 28th day of September 2000.
For the Nuclear Regulatory Commission.
Elinor G. Adensam,
Acting Director, Division of Licensing Project Management, Office of
Nuclear Reactor Regulation.
[FR Doc. 00-25377 Filed 10-3-00; 8:45 am]
BILLING CODE 7590-01-P