Federal Register, Volume 66 Issue 85 (Wednesday, May 2, 2001)

[Federal Register Volume 66, Number 85 (Wednesday, May 2, 2001)]
[Notices]
[Pages 22021-22040]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: 01-10822]

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NUCLEAR REGULATORY COMMISSION

Biweekly Notice; Applications and Amendments to Facility
Operating Licenses Involving No Significant Hazards Considerations

I. Background

Pursuant to Public Law 97-415, the U.S. Nuclear Regulatory
Commission (the Commission or NRC staff) is publishing this regular
biweekly notice. Public Law 97-415 revised section 189 of the Atomic
Energy Act of 1954, as amended (the Act), to require the Commission to
publish notice of any amendments issued, or proposed to be issued,
under a new provision of section 189 of the Act. This provision grants
the Commission the authority to issue and make immediately effective
any amendment to an operating license upon a determination by the
Commission that such amendment involves no significant hazards
consideration, notwithstanding the pendency before the Commission of a
request for a hearing from any person.
This biweekly notice includes all notices of amendments issued, or
proposed to be issued from April 9, 2001, through April 20, 2001. The
last biweekly notice was published on April 18, 2001 (66 FR 19998).

Notice of Consideration of Issuance of Amendments to Facility
Operating Licenses, Proposed No Significant Hazards Consideration
Determination, and Opportunity for a Hearing

The Commission has made a proposed determination that the following
amendment requests involve no significant hazards consideration. Under
the Commission's regulations in 10 CFR 50.92, this means that operation
of the facility in accordance with the proposed amendment would not (1)
involve a significant increase in the probability or consequences of an
accident previously evaluated; or (2) create the possibility of a new
or different kind of accident from any accident previously evaluated;
or (3) involve a significant reduction in a margin of safety. The basis
for this proposed determination for each amendment request is shown
below.
The Commission is seeking public comments on this proposed
determination. Any comments received within 30 days after the date of
publication of this notice will be considered in making any final
determination.
Normally, the Commission will not issue the amendment until the
expiration of the 30-day notice period. However, should circumstances
change during the notice period such that failure to act in a timely
way would result, for example, in derating or shutdown of the facility,
the Commission may issue the license amendment before the expiration of
the 30-day notice period, provided that its final determination is that
the amendment involves no significant hazards consideration. The final
determination will consider all public and State comments received
before action is taken. Should the Commission take this action, it will
publish in the Federal Register a notice of issuance and provide for
opportunity for a hearing after issuance. The Commission

[[Page 22022]]

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 Administrative
Services, Office of Administration, U.S. Nuclear Regulatory Commission,
Washington, DC 20555-0001, and should cite the publication date and
page number of this Federal Register notice. Written comments may also
be delivered to Room 6D22, Two White Flint North, 11545 Rockville Pike,
Rockville, Maryland from 7:30 a.m. to 4:15 p.m. Federal workdays.
Copies of written comments received may be examined at the NRC 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 June 1, 2001, the licensee may file a request for a hearing with
respect to issuance of the amendment to the subject facility operating
license and any person whose interest may be affected by this
proceeding and who wishes to participate as a party in the proceeding
must file a written request for a hearing and a petition for leave to
intervene. Requests for a hearing and a petition for leave to intervene
shall be filed in accordance with the Commission's ``Rules of Practice
for Domestic Licensing Proceedings'' in 10 CFR Part 2. Interested
persons should consult a current copy of 10 CFR 2.714 which is
available at the Commission's Public Document Room, located at One
White Flint North, 11555 Rockville Pike (first floor), Rockville,
Maryland 20852. Publicly available records will be accessible and
electronically from the ADAMS Public Library component on the NRC Web
site, http://www.nrc.gov (the Electronic Reading Room). If a request
for a hearing or petition for leave to intervene is filed by the above
date, the Commission or an Atomic Safety and Licensing Board,
designated by the Commission or by the Chairman of the Atomic Safety
and Licensing Board Panel, will rule on the request and/or petition;
and the Secretary or the designated Atomic Safety and Licensing Board
will issue a notice of a hearing or an appropriate order.
As required by 10 CFR 2.714, a petition for leave to intervene
shall set forth with particularity the interest of the petitioner in
the proceeding, and how that interest may be affected by the results of
the proceeding. The petition should specifically explain the reasons
why intervention should be permitted with particular reference to the
following factors: (1) The nature of the petitioner's right under the
Act to be made a party to the proceeding; (2) the nature and extent of
the petitioner's property, financial, or other interest in the
proceeding; and (3) the possible effect of any order which may be
entered in the proceeding on the petitioner's interest. The petition
should also identify the specific aspect(s) of the subject matter of
the proceeding as to which petitioner wishes to intervene. Any person
who has filed a petition for leave to intervene or who has been
admitted as a party may amend the petition without requesting leave of
the Board up to 15 days prior to the first prehearing conference
scheduled in the proceeding, but such an amended petition must satisfy
the specificity requirements described above.
Not later than 15 days prior to the first prehearing conference
scheduled in the proceeding, a petitioner shall file a supplement to
the petition to intervene which must include a list of the contentions
which are sought to be litigated in the matter. Each contention must
consist of a specific statement of the issue of law or fact to be
raised or controverted. In addition, the petitioner shall provide a
brief explanation of the bases of the contention and a concise
statement of the alleged facts or expert opinion which support the
contention and on which the petitioner intends to rely in proving the
contention at the hearing. The petitioner must also provide references
to those specific sources and documents of which the petitioner is
aware and on which the petitioner intends to rely to establish those
facts or expert opinion. Petitioner must provide sufficient information
to show that a genuine dispute exists with the applicant on a material
issue of law or fact. Contentions shall be limited to matters within
the scope of the amendment under consideration. The contention must be
one which, if proven, would entitle the petitioner to relief. A
petitioner who fails to file such 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: Rulemaking and
Adjudications Branch, or may be delivered to the Commission's Public
Document Room, located at One White Flint North, 11555 Rockville Pike
(first floor), Rockville, Maryland 20852, by the above date. A copy of
the petition should also be sent to the Office of the General Counsel,
U.S. Nuclear Regulatory Commission, Washington, DC 20555-0001, and to
the attorney for the licensee.
Nontimely filings of petitions for leave to intervene, amended
petitions, supplemental petitions and/or requests for a hearing will
not be entertained absent a determination by the Commission, the
presiding officer or the Atomic Safety and Licensing Board that the
petition and/or request should be granted based upon a balancing of
factors specified in 10 CFR 2.714(a)(1)(i)-(v) and 2.714(d).
For further details with respect to this action, see the
application for amendment which is available for public inspection at
the Commission's Public Document Room, located at One White Flint
North, 11555 Rockville Pike (first floor), Rockville, Maryland 20852.
Publicly available records will be accessible and 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: April 1, 2001 (102-04552).
Description of amendments request: The amendments would revise the
requirements on the following programs in the administrative controls
section of the technical specifications (TSs): (1) Section 5.5.13,
``Diesel Fuel Oil Testing Program,'' (2) Section 5.5.14, ``TS Bases
Control Program,'' (3) Section 5.5.15, ``Safety Functions Determination

[[Page 22023]]

Program (SFDP),'' and (4) Section 5.6.5, ``Core Operating Limits Report
(COLR).'' The proposed changes clarify the program requirements in
Section 5.5.13 without changing testing methods or limits, revise the
program in Section 5.5.14 based on changes to 10 CFR 50.59 in the
regulations, clarify the program requirements in Section 5.5.15
including changing the program name to the plant-specific name for the
program, and add the CENTS code to the list of analytical methods used,
including the use of CENTS for control element assembly ejection
analyses, to determine core operating limits and revise the list of
referenced topical reports in the COLR in Section 5.6.5.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:

1. The proposed change does not involve a significant increase
in the probability or consequences of an accident previously
evaluated.
Technical Specification (TS) 5.5.13, Diesel Generator Fuel Oil
Program. TS 5.5.13.a.3 currently states, ``Water and sediment are
within the limits of ASTM D1796,'' for the acceptability of new
diesel fuel oil. This is an incorrect reference for the limits of
water and sediment of new fuel oil. The water and sediment limits
for new fuel oil are contained within the Technical Specification
Bases. ASTM D1796 contains testing methods used for analysis of new
fuel oil for water and sediment. This proposed amendment changes the
wording of TS 5.5.13.a.3 to state, ``Water and sediment within
limits when tested in accordance with ASTM D1796.'' This proposed
change is an administrative change and will have no affect on plant
design, operation, or maintenance. Additionally, this proposed
change does not result in any hardware changes or affect plant
operating practices. The water and sediment testing methods and
limits are not affected by this change. Thus, this proposed change
does not involve a significant increase in the probability or
consequences of an accident previously evaluated.
TS 5.5.14, TS Bases Control Program, requires a program for
processing changes to the Bases of the TS [...]
In the initial sentence to TS 5.5.14.b, the word ``involve''
will be replaced with ``require.'' Additionally, the second
allowance for changing TS Bases as described in TS 5.5.14.b will be
revised to state, ``A change to the updated FSAR or Bases that
requires NRC approval pursuant to 10 CFR 50.59.'' This change is
based on the changes to 10 CFR 50.59 published in the Federal
Register (Volume 64, Number 191) dated October 4, 1999. This change
is consistent with NRC approved Technical Specifications Task Force
(TSTF) traveler number 364-revision 0.
This change will also numerically format the two options listed
in TS 5.5.14.b. This is consistent with other listings contained in
Section 5.0 of the TS.
This proposed change deletes the reference to ``unreviewed
safety question'' as previously used in 10 CFR 50.59[, before the
rule change published October 4, 1999, in the Federal Register.]
Deletion of this definition was approved by the NRC with the
revision to 10 CFR 50.59.
[These] proposed change[s to TS 5.5.14 are] administrative
change[s] and will have no affect on plant design, operation, or
maintenance. Additionally, [these] change[s do] not result in any
hardware changes or affect plant operating practices. Therefore,
[the] proposed change[s to TS 5.5.14 do] not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
TS 5.5.15, Safety Functions Determination Program (SFDP).
Clarification is being added to TS 5.5.15. The second paragraph of
TS 5.5.15 will be changed to read: ``A loss of safety function
exists when, assuming no concurrent single failure, no concurrent
loss of offsite power, or no concurrent loss of onsite diesel
generator(s), a safety function assumed in the accident analysis
cannot be performed. For the purpose of this program, a loss of
safety function may exist when a support system is inoperable, and *
* *''
An additional paragraph will be added to the end of TS 5.5.15
stating, ``When a loss of safety function is caused by the
inoperability of a single Technical Specification support system,
the appropriate Conditions and Required Actions to enter are those
of the support system.''
Additionally, clarification will be added to limiting conditions
for operation (LCO) 3.0.6 Bases of the ``appropriate LCO for loss of
safety function.'' The Bases will also clarify the requirement for
the SFDP that consideration does not have to be made for a loss of
power in determining loss of function. This change is consistent
with NRC approved TSTF traveler number 273-revision 2, as amended by
editorial change WOG-ED-23.
In addition, an editorial change to remove the ``s'' from the
word ``Functions'' in the title for TS 5.5.15 will occur. The change
reflects the plant specific name for this program.
[These] proposed change[s to TS Section 5.5.15 are]
administrative change[s] and will have no affect on plant design,
operation, or maintenance. The change[s] clarif[y] the requirements
for determining loss of safety function and the correct LCO to enter
for loss of safety function. The proposed change[s do] not result in
any hardware changes or affect plant operating practices. The
program will still determine when a safety function has been lost
and will direct the appropriate action. Therefore, [the] proposed
change[s to TS 5.5.15 do] not involve a significant increase in the
probability or consequences of an accident previously evaluated.
TS 5.6.5, Core Operating Limits Report (COLR) is being revised
to add the option to use the CENTS computer code in licensing
analysis by adding CENTS to the list of approved core operating
limit analytical methods contained in TS 5.6.5.b. The CENTS computer
code has been generally approved for the calculation of transient
behavior in Pressurized Water reactors (PWRs) designed by Combustion
Engineering (CE). PVNGS intends to qualify CENTS for use in future
Palo Verde licensing analyses by following the guidelines prescribed
in Generic Letter (GL) 83-11, Supplement 1.
CENTS is a best-estimate code designed to provide realistic
simulation of Nuclear Steam Supply System (NSSS) behavior during
normal and transient conditions. The CENTS Safety Evaluation (SE)
documents the generic NRC approval of the CENTS code for use in the
licensing analyses for PWRs designed by CE. The CENTS SE is
described in letter, ``Acceptance for Referencing of Licensing
Topical Report CE-NPD 282-P, ``Technical Manual for the CENTS Code''
dated March 17, 1994, from USNRC to S. A. Toelle, ABB Combustion
Engineering.
The proposed change does not immediately alter any methodology
used in [an] reload analysis. It only provides the option to replace
the CESEC transient simulation code with an alternate NRC approved
code. Providing the option to substitute the NRC approved CESEC code
with another NRC approved code (CENTS) will not alter the physical
characteristics of any component involved in the initiation or
mitigation of an accident. The actual implementation of the CENTS
code will be performed by following the guidance provided in Generic
Letter (GL) 83-11, Supplement 1. This proposed change does not
result in any hardware changes or affect plant operating practices.
Thus, this proposed change does not involve a significant increase
in the probability or consequences of an accident previously
evaluated.
TS 5.6.5, core operating limits report (COLR) which identifies
the methodology report(s) by number, title, date, and NRC staff
approval document, will be revised to allow the reports to be
identified by number and title only. A note will be added to TS
5.6.5.b to specify that a complete citation be included in the COLR
for each report, including the report number, title, revision, date,
and any supplements.
This change has previously been reviewed and accepted by the NRC
in letter, ``Acceptance for Siemens References to Approved Topical
Reports in Technical Specifications'' from S.A. Richards, NRC to
J.F. Mallay, Siemens Power Corporation dated December 15, 1999. This
change is also consistent with NRC accepted TSTF 363-revision 0.
Additionally, TS 5.6.5.b.6 and 5.6.5.b.7 both list the same
topical report (Calculative Methods for the CE Small Break LOCA
Evaluation Model, CENPD-137). TS 5.6.5.b.7 is the supplement to the
topical report listed in [TS] 5.6.5.b.6. TS 5.6.5.b.7 will be
deleted and the ``Calculative Methods for the CE Small Break LOCA
Evaluation Model, CENPD-137'' topical report (along with its
supplement) will be listed in full text within the COLR.
[The] proposed change[s related to the listing of topical
reports in TS 5.6.5.b are] administrative change[s] and will have no

[[Page 22024]]

affect on plant design, operation, or maintenance. Thus, [these]
proposed change[s 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.
TS 5.5.13, Diesel Generator Fuel Oil Program. The proposed
change is an administrative change. This change would have no affect
on the physical plant. Consequently, plant configuration and the
operational characteristics remain unchanged and the proposed
changes do not create the possibility of a new or different kind of
accident from any accident previously evaluated.
TS 5.5.14, TS Bases Control Program. The proposed changes
associated with TS 5.5.14.b do not involve any physical changes.
These changes allow PVNGS to be in compliance with NRC approved
changes to 10 CFR 50.59. This change is an administrative change.
Plant configuration and the operational characteristics remain
unchanged and thus, the proposed changes do not create the
possibility of a new or different kind of accident from any accident
previously evaluated.
TS 5.5.15, SFDP. The proposed change to TS 5.5.15 does not
involve any physical changes to the plant[s]. This change is an
administrative change. The loss of function of the specific
component is addressed in its specific TS LCO and plant
configuration will be governed by the required actions of those
LCOs. Since this proposed change is a clarification that does not
degrade the availability or capability of safety related equipment,
this change does not create the possibility of a new or different
kind of accident from any accident previously evaluated.
TS 5.6.5, COLR is being revised to add the option to use the
CENTS computer code in licensing analysis by adding CENTS to the
list of approved core operating limit analytical methods contained
in TS 5.6.5.b. The proposed change will not affect reload analysis
other than providing an option to replace the CESEC transient
simulation code with an equivalent code. Providing this option in
and of itself will not alter the physical characteristics of any
component in the plant. Since providing the option to use the CENTS
code will not alter the physical characteristics of any component in
the plant, this proposed change does not create the possibility of a
new or different kind of accident from any accident previously
evaluated.
TS 5.6.5, Core Operating Limits Report (COLR) which identifies
the methodology report(s) by number, title, date, and NRC staff
approval document, will be revised to allow the reports to be
identified by number and title only. This is an administrative
change. This change has no affect on the physical plant. Plant
configuration and the operational characteristics remain unchanged
and thus, this change does not create the possibility of a new or
different kind of accident from any accident previously evaluated.
3. The proposed change does not involve a significant reduction
in a margin of safety.
TS 5.5.13, Diesel Generator Fuel Oil Program. The proposed
change to TS 5.5.13.a.3 is an administrative change. This change
would have no affect on the physical plant and has no effect on any
safety analyses assumptions. Therefore, this proposed change does
not involve a significant reduction in a margin of safety.
TS 5.5.14, TS Bases Control Program. The proposed changes
associated with TS 5.5.14.b will not reduce a margin of safety
because it has no direct effect on any safety analyses assumptions.
Changes to the TS Bases that result in meeting the criteria in
paragraph (c)(2) of 10 CFR 50.59 will still require NRC approval
pursuant to 10 CFR 50.59. This change is administrative in nature
and is based on NRC reviewed and approved changes to 10 CFR 50.59.
Therefore, this proposed change does not involve a significant
reduction in a margin of safety.
TS 5.5.15, SFDP. The proposed change to TS 5.5.15 are
clarifications only. No changes are made in the LCO, the time
required for the TS required actions to be completed, or the out of
service time for the components involved. The NRC has approved the
proposed administrative changes (TSTF 273-revision 2, as amended by
editorial change WOG-ED-23). Safety-related equipment controlled by
the TS will still perform as credited in the safety analysis.
Therefore, this proposed change does not involve a significant
reduction in a margin of safety.
TS 5.6.5, COLR is being revised to add the option to use the
CENTS computer code in licensing analysis by adding CENTS to the
list of approved core operating limit analytical methods. The
proposed change will allow running existing analyses with a
different method that has been reviewed and approved by NRC. The
actual implementation of the CENTS code will be performed by
following the guidance provided in Generic Letter (GL) 83-11,
Supplement 1. Thus, this proposed change does not involve a
significant reduction in a margin of safety.
TS 5.6.5, Core Operating Limits Report (COLR) which identifies
the methodology report(s) by number, title, date, and NRC staff
approval document, will be revised to allow the reports to be
identified by number and title only. This is an administrative
change. This change has no affect on the physical plant. Plant
configuration and the operational characteristics remain unchanged.
Therefore, change does not involve a significant reduction in a
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.

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: April 4, 2001 (102-04554).
Description of amendments request: The amendments would revise
Specification 3.3.12, ``Boron Dilution Alarm System (BDAS),'' and
Specification 3.9.2, ``Refueling Operations--Nuclear Instrumentation''
of the technical specification (TSs). Specification 3.9.2 applies to
the required operability of startup range monitors (SRMs). The
applicability modes for limiting condition for operation (LCO) 3.3.12
would be extended to Mode 6, refueling. A note to ``Enter applicable
Conditions and Required Actions of LCO 3.3.12, ``Boron Dilution Alarm
System (BDAS),'' for BDAS made inoperable by SRMs'' would be added to
the Actions for LCO 3.9.2 and the Required Action B.2 on performing
surveillance requirement (SR) 3.9.1.1, and associated completion time,
for LCO 3.9.2 would be deleted.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:

1. Does the proposed change involve a significant increase in
the probability or consequences of an accident previously evaluated?
No. The proposed changes to Technical Specifications 3.3.12 and
3.9.2 do not significantly increase the probability or consequences
of an accident previously evaluated in the Updated Final Safety
Analysis Report (UFSAR) [for Palo Verde Nuclear Generating Station].
The proposed amendment[s] would add MODE 6 Applicability to TS
3.3.12 for the BDAS. In addition, the proposed amendment[s] would
add a note to the Actions of TS 3.9.2 which directs the operator to
enter the applicable Conditions and Required Actions of TS 3.3.12 in
the event that the BDAS is made inoperable by inoperable startup
range monitors (SRMs). Finally, the proposed amendment[s] would
delete the TS 3.9.2 Required Action B.2.
The boron dilution alarm system (BDAS) and chemical monitoring
of the reactor coolant system (RCS) boron concentration are
established in the MODE 6 inadvertent deboration analysis in UFSAR
Section 15.4.6 to alert the operator of a boron dilution event at
least 30 minutes prior to a loss of subcriticality. The BDAS and RCS
boron monitoring are not accident initiators. The proposed changes
will ensure that the assumptions of UFSAR Section 15.4.6, for
mitigating an inadvertent deboration event, are met. In addition,
the proposed changes do

[[Page 22025]]

not alter the design or configuration of the plant but establish
requirements for operating the plant as analyzed and designed. The
amendment[s do] not physically affect the operability or
availability of the boron dilution alarm system (BDAS), but ensures
it is available as required or that sufficient actions are taken if
it becomes inoperable. Furthermore, the inadvertent deboration event
analysis does not involve dose consequences since the acceptance
criteria is to provide operator notification at least 30 minutes
prior to the loss of subcriticality such that the operator may
terminate the event before subcriticality is achieved [and
exceeded,] and the RCS and fuel clad boundaries are challenged.
Therefore, the proposed amendment[s] to TS 3.3.12 and TS 3.9.2 [do]
not involve a significant increase in the probability or
consequences of an accident previously evaluated.
2. Does the proposed change create the possibility of a new or
different kind of accident from any accident previously evaluated?
No. The proposed amendment[s] to Technical Specifications 3.3.12
and 3.9.2 [do] not create the possibility of an accident of a new or
different kind from any accident previously evaluated. The proposed
amendment[s] would add MODE 6 Applicability to TS 3.3.12 for the
BDAS. In addition, the proposed amendment[s] would add a note to the
Actions of TS 3.9.2 which directs the operator to enter the
applicable Conditions and Required Actions of TS 3.3.12 in the event
that the BDAS is made inoperable by inoperable startup range
monitors (SRMs). Finally, the proposed amendment[s] would delete the
TS 3.9.2 Required Action B.2. The proposed changes do not alter the
design or configuration of the plant but establish requirements for
operating the plant as analyzed and designed.
In MODE 6, the BDAS and the startup range monitors (SRM) are the
primary means to monitor reactivity changes during core alterations
and to alert the operator of a boron dilution event in time to
prevent a loss of subcriticality. Chemical sampling to monitor RCS
boron concentration is used when the BDAS is unavailable. Accidents
involving reactivity anomalies are evaluated in UFSAR Section 15.4,
Reactivity and Power Distribution Anomalies. Inadvertent deboration
is described in UFSAR Section 15.4.6 as requiring the BDAS or
chemical monitoring of the RCS boron concentration to alert the
operator at least 30 minutes prior to the loss of subcriticality in
MODE 6. The proposed changes to TS 3.3.12 and 3.9.2 will require the
BDAS to be OPERABLE in MODE 6 or perform RCS boron concentration
monitoring if the BDAS is inoperable.
The BDAS and RCS boron concentration monitoring are means to
detect a boron dilution event. The proposed changes ensure this
detection occurs as required. The proposed amendment[s do] not
physically affect the response or operation of the plant. Therefore,
the proposed changes do not create the possibility of a new or
different kind of accident from any accident previously evaluated.
3. Does the proposed change involve a significant reduction in a
margin of safety?
No. The proposed changes to Technical Specifications 3.3.12 and
3.9.2 do not involve a significant reduction in a margin of safety.
The proposed amendment[s] would add MODE 6 Applicability to TS
3.3.12 for the BDAS. In addition, the proposed amendment[s] would
add a note to the Actions of TS 3.9.2 which directs the operator to
enter the applicable Conditions and Required Actions of TS 3.3.12 in
the event that the BDAS is made inoperable by inoperable startup
range monitors (SRMs). Finally, the proposed amendment[s] would
delete the TS 3.9.2 Required Action B.2. These changes ensure the
adequate detection of a boron dilution event.
The current Technical Specifications 3.3.12 satisfies the
inadvertent deboration safety analysis requirements to have the BDAS
OPERABLE in MODES 3, 4,and 5. In accordance with UFSAR Section
15.4.6, Inadvertent Deboration, the same requirements and actions
apply for MODE 6. Therefore, it is proposed that MODE 6
Applicability for the BDAS be added to TS 3.3.12. In addition, the
Action section of TS 3.9.2 would be modified with a note to ensure
the safety analysis assumptions are satisfied in MODE 6, since the
SRM must be OPERABLE for the corresponding BDAS channel to be
OPERABLE. Technical Specification Bases 3.3.12 and UFSAR Section
15.4.6 indicate that the BDAS is necessary to alert the operator of
an inadvertent deboration event at least 15 minutes before the
reactor loses subcriticality in MODES 3, 4, and 5. UFSAR Section
15.4.6 also indicates that 30 minutes is required in MODE 6. These
criteria are in agreement with the guidance of NUREG 0800, [NRC's]
Standard Review Plan. Therefore, the margin of safety being
considered for [these] proposed amendment[s] is the 30 minutes
before the loss of subcriticality that the operator must be notified
[...] in the event of a boron dilution event. The proposed changes
to TS 3.3.12 and TS 3.9.2 will require the BDAS to be OPERABLE in
MODE 6 and, if the BDAS is inoperable, will require that the RSC
boron concentration be monitored at pre-analyzed frequencies via
chemical sampling in order to satisfy the 30 minute acceptance
criteria. Finally, the proposed change[s] also serve to clarify that
an inoperable SRM will cause the corresponding BDAS channel to be
inoperable, thus requiring action in accordance with TS 3.3.12, in
addition to TS 3.9.2. [The proposed changes add a requirement to TS
3.3.12 and account for the BDAS being inoperable because of
inoperable SRMS] Therefore, the proposed change[s do] not involve a
significant reduction in a margin of safety.

Consumers Energy Company, Docket No. 50-155, Big Rock Point Plant,
Charlevoix, County, Michigan

Date of amendment request: October 26, 2000, as supplemented by
letters dated February 9, February 28, March 14, March 15, and March
23, 2001.
Description of amendment request: The proposed amendment reflects
the replacement of the original 75-ton reactor building gantry crane by
an upgraded single-failure proof 125-ton crane designed to meet Crane
Manufacturers Association of America (CMAA) Specification 70 and
American Society of Mechanical Engineers (ASME) B30.2. The proposed
amendment to the Technical Specifications (TSs) would revise (1)
Definition 1.8 on fuel handling, (2) the applicability of TS 3/4.2.1 on
fuel handling support system requirements, and (3) Section 3.2.2.d of
the limiting conditions for operation for TS 3/4.2.2 on fuel handling
general requirements, and would delete TS 3/4.3.1 on control of heavy
loads. The licensee also submitted revisions to the bases for TSs 3/
4.2.2 and 3/4.3.1. The crane has a Design Rated Load (DRL) of 125 tons;
however, it has been analyzed to safely retain a load of 105 tons under
the site-specific earthquake and the Maximum Critical Load (MCL) for
the crane is 105 tons.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee provided
its analysis in its letters dated October 26, 2000, and March 14, 2001,
which address the issue of no significant hazards consideration, and is
presented below:

The proposed [amendment] does not:
(1) Involve a significant increase in the probability or
consequences of an accident previously evaluated.
A significant increase in the probability of an accident is not
created because:
The replacement crane will not be utilized for a
greater number of fuel handling evolutions than was the case for the
existing 75-ton crane. The existing crane was utilized for each
transfer of fuel assemblies between the reactor and the Spent Fuel
Pool; in the case of full-core offloads, which was the normal
practice during refueling outages at Big Rock Point [Plant], the
existing crane would make 84 transfers of irradiated fuel from the
reactor to the Spent Fuel Pool, and a nominal 62 transfers of
irradiated fuel from the Spent Fuel Pool back to the reactor. The
replacement crane will handle fuel only after it has been placed
into the W100 Transfer Cask. It is anticipated that the W100 Fuel
Transfer Cask will be handled 14 times while it contains fuel (one
movement from the

[[Page 22026]]

Spent Fuel Pool to a staging area in Room 444, and one movement from
Room 444 to a W150 Storage Cask), during loading of seven W150
Storage Casks. Additional moves of the W100 Transfer Cask when it is
loaded with fuel would be required only if an off-normal condition
required a loaded cask to be returned to the Spent Fuel Pool.
The replacement crane has been analyzed to safely
handle the 105-ton W100 Fuel Transfer Cask under seismic conditions
that include the Big Rock Point [Plant] site-specific safe shutdown
earthquake of 0.104g. The UFHSR [Updated Final Hazards Summary
Report] is being revised to limit the weight of loads being moved
over the Spent Fuel Pool to 105 tons.
The existing crane has been used to lift the properly
rigged 24-ton fuel transfer cask over fuel; the probability of
dropping the 24-ton fuel transfer cask was minimized by the proper
rigging that consisted of attaching a safety catch device to the
transfer cask. In the case of the replacement crane, loads will be
prevented from dropping by the design of the single-failure proof
Ederer X-SAM hoist, which prevents loads from dropping more than 18
inches in the event of any single failure. Administrative controls
will be instituted on the use of the replacement crane to require
lifts of any heavy loads over fuel or over structures, the failure
of which would jeopardize safe storage of fuel, to be done at a
height of greater than 18 inches. Administrative controls will be
instituted to prohibit use of the replacement crane for movement of
any cask over fuel; these controls will be specified in the Big Rock
Point [Plant] UFHSR. Administrative controls that apply to our [the
licensee's] current 75-ton crane will be maintained, and
strengthened, as appropriate, to provide greater assurance that
heavy loads transported over fuel will be safely transported.
Strengthened administrative controls include limiting the number of
crane operators to approximately 12 individuals, and requiring that
they receive Operator Engineer training in the use of the upgraded
crane.
The existing crane met single-failure proof criteria
only when it was used to handle the properly-rigged 24-ton fuel
transfer cask; the 105-ton single-failure proof crane will be
single-failure proof for all lifts of loads which are 105 tons or
less.
The replacement of the existing crane with a 105-ton
single-failure proof crane is being performed as a safety-related
modification, and 10 CFR [Part] 50 Appendix B [Quality Assurance]
criteria are being applied to all critical elements of design,
purchasing, installation and testing. Therefore, the replacement
crane and trolley can be expected to perform in accordance with
their design specifications. As a result, the probability of a
trolley failure on the replacement crane is considered to be no
greater than the probability of a failure of the safety catch device
which was employed with the existing crane when it was used to
handle the 24-ton fuel transfer cask.
A significant increase in the consequences of an accident is not
created because:
This change affects fuel handling, and fuel handling
accidents have already been analyzed and bound all other categories
of accidents at the Big Rock Point Plant. Analysis indicates that
the dose from the bounding fuel accident (a 24-ton fuel transfer
cask drop), assuming a free release path without isolation of
ventilation from containment, falls below the Protective Action
Guidelines (PAGs) of Environmental Protection Agency-400 (EPA-400)
68 days following plant shutdown (the reactor was shutdown [as] of
8/29/1997). The analysis assumed a total of 500 damaged assemblies
in the Spent Fuel Pool, with 84 of them being freshly discharged
from the reactor. The Spent Fuel Pool contains 441 fuel assemblies.
With more than three years of radiological and heat decay since the
plant was shutdown, the potential source terms for gaseous and
volatile radionuclides associated with the remaining design basis
accidents has continued to decrease; therefore, the doses at the
site boundary associated with a postulated accident involving any
number of the available fuel assemblies have also decreased. The
design of the Ederer X-SAM trolley and hoist is such that upon a
single failure of the trolley that would allow the suspended load to
free-fall, the load could fall for a maximum of 18 inches before the
drum brake mechanism would engage to stop the downward travel. An
18-inch drop of the 105-ton dry fuel storage system fuel transfer
cask has been analyzed and has been determined not to result in
failure of the floors of the Spent Fuel Pool, Room 444 or the
laydown area at the 599-foot 5-inch elevation of containment. These
are the only floors in containment over which the cask will [be]
moved with the 105-ton single-failure proof crane at a height of
less than 18 inches. The 105-ton W100 Transfer Cask is the largest
load that will be handled over the Spent Fuel Pool, when fuel is
being stored in the Pool. For other floors/structures, (i.e., the
Reactor Deck at elevation 632' 6" [632 feet 6 inches]),
administrative controls will be imposed to require the 105-ton cask
to be suspended at least 18 inches above the floor/structure.
[The proposed amendment reflects the replacement of the original
75-ton non single-failure proof crane, with a single-failure proof
crane. The replacement crane addresses malfunctions (e.g., dropping
loads under single-failure conditions) that were possible with the
original crane.]
Based on this discussion, it is concluded that this proposed
change to the Defueled Technical Specifications does not involve a
significant increase in the probability or consequences of an
accident previously evaluated.
(2) Create the possibility of a new or different kind of
accident from any accident previously evaluated.
The proposed change is requested to reflect the removal of the
original 75-ton reactor building non single-failure proof semi-
gantry crane and its replacement with a single-failure proof 105-ton
crane, which will be designed to meet the applicable criteria and
guidelines of NUREG-0554 and NUREG-0612. The change results from
installation of a crane that replaces another crane. The general
functions performed by the replacement crane (cask handling and
movement of heavy loads) do not differ from those performed by the
original crane. Therefore, new or different accidents will not be
created by elimination of restrictions associated with the original
75-ton crane, since the design of the replacement crane addresses
malfunctions (for example, dropping loads under single-failure
conditions) that were possible with the original crane.
Based on this discussion, it is concluded that this proposed
change to the Defueled Technical Specifications does not create the
possibility of a new or different kind of accident from any accident
previously evaluated.
(3) Involve a significant reduction in the margin of safety.
To prevent failure of the Spent Fuel Pool structure when
handling loads over the Pool with the existing 75-ton crane, loads
were limited to 24 tons, and cask handling evolutions were limited
to the southwest corner of the Spent Fuel Pool. These measures
ensured that the Spent Fuel pool would not fail as a result of a
load being dropped into it. The replacement crane has been designed
such that a load will not drop more than 18 inches if a single
failure should occur in its trolley. A drop of the 105-ton W100 Fuel
Transfer Cask from a height of 18 inches to the Spent Fuel Pool
floor has been determined not to result in failure of the Spent Fuel
Pool; loads handled by the replacement crane will be restricted to
105 tons to ensure that the structural integrity of the Spent Fuel
Pool will not be compromised by a postulated drop of the 105-ton
W100 Fuel Transfer Cask.
The existing crane is designed to handle loads up to 75 tons
Because the existing crane was not designed as a single-failure
proof crane, restrictions were placed on load paths, load weights,
and the configuration of the 24-ton fuel transfer cask (the cask was
required to have a safety catch device attached between the cask and
the crane structure to prevent dropping the transfer cask in the
event of a trolley failure) to ensure that a margin of safety
existed with respect to dropping heavy loads on spent fuel and to
prevent a dropped load from causing structural failure of the Spent
Fuel Pool. The replacement crane is designed to withstand the Big
Rock Point [Plant] site-specific safe shutdown earthquake of 0.104g
while safely retaining a load equal to 105 tons. Therefore, handling
the 105-ton W100 transfer cask with this crane provides equivalent
margins with respect to crane failure as the current restriction
that limits loads being handled over the Spent Fuel Pool to 24 tons.
The UFHSR will restrict handling of loads over the Spent Fuel Pool
to 105 tons whenever fuel is stored in the Pool. The trolley and
hoist for the replacement crane are designed to be single-failure
proof, and provide a margin of safety for dropping a suspended load
equivalent to the safety catch employed with 24-ton transfer cask
safety catch.
[The proposed amendment reflects the replacement of the original
75-ton non single-failure proof crane, with a single-failure proof
125-ton crane having an MCL of 105 tons.. The replacement crane is
designed to meet the applicable criteria of NUREG-0554 and

[[Page 22027]]

NUREG-0612, CMAA Specification 70, and ASME B30.2.]
Based on this discussion, it is concluded that this proposed
change to the Defueled Technical Specifications does not involve a
significant decrease in the margin of safety.

The NRC staff has reviewed the licensee's analysis in both letters
of October 26, 2000, and March 14, 2001, 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: David A. Mikelonis, Esquire, Consumers
Energy Company, 212 West Michigan Avenue, Jackson, Michigan 49201.
NRC Section Chief: Robert A. Gramm.

Consumers Energy Company, Docket No. 50-255, Palisades Plant, Van Buren
County, Michigan

Date of amendment request: March 5, 2001, as revised March 30,
2001.
Description of amendment request: The proposed amendment would
revise Technical Specification (TS) 5.5.12, ``Technical Specifications
(TS) Bases Control Program,'' to be consistent with the changes to 10
CFR 50.59 published in the Federal Register on October 4, 1999 (64 FR
53582), as reflected in the Nuclear Energy Institute's Technical
Specification Task Force (TSTF) Standard TS Change Traveler, TSTF-364,
``Revision to TS Bases Control Program to Incorporate Changes to 10 CFR
50.59.'' Specifically, Palisades TS 5.5.12b currently states, in part,
that licensees may make changes to Bases without prior NRC approval
provided the changes do not ``involve * * * [a] change to the updated
FSAR [Final Safety Analysis Report] or Bases that involves an
unreviewed safety question as defined in 10 CFR 50.59.'' The proposed
amendment would change this quoted portion of TS 5.5.12b to state
``require * * * [a] change to the updated FSAR or Bases that requires
NRC approval pursuant to 10 CFR 50.59.''
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 following evaluation supports the finding that operation of
the facility in accordance with the proposed changes would not:
a. Involve a significant increase in the probability or
consequences of an accident previously evaluated.
The proposed change deletes the reference to unreviewed safety
question as defined in 10 CFR 50.59. Deletion of the definition of
unreviewed safety question was approved by the NRC with the revision
of 10 CFR 50.59. Consequently, the probability of an accident
previously evaluated is not significantly increased. Changes to the
TS Bases are still evaluated in accordance with 10 CFR 50.59. As a
result, the consequences of any accident previously evaluated are
not significantly affected. Therefore, this change does not involve
a significant increase in the probability or consequences of an
accident previously evaluated.
b. Create the possibility of a new or different kind of accident
from any previously evaluated.
The proposed change does not involve a physical alteration of
the plant (no new or different type of equipment will be installed)
or a change in the methods governing normal plant operation.
Therefore, this change does not create the possibility of a new or
different kind of accident from any accident previously evaluated.
c. Involve a significant reduction in the margin of safety.
The proposed change will not reduce a margin of safety because
it has no direct effect on any safety analyses assumptions. Changes
to the TS Bases that result in meeting the criteria in paragraph 10
CFR 50.59(c)(2) will still require NRC approval pursuant to 10 CFR
50.59. This change is administrative in nature based on the revision
to 10 CFR 50.59. Therefore, the proposed change does not involve a
significant reduction in a margin of safety.

The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 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: Arunas T. Udrys, Esquire, Consumers Energy
Company, 212 West Michigan Avenue, Jackson, Michigan 49201.
NRC Section Chief: Claudia M. Craig.

Consumers Energy Company, Docket No. 50-255, Palisades Plant, Van Buren
County, Michigan

Date of amendment request: April 2, 2001.
Description of amendment request: The proposed amendment would
revise the Technical Specifications (TS) by removing all requirements
for, and references to, the ``Assembly Radial Peaking Factor,''
(FR^A). Consequently, in TS Section 1.0, the
definition of Assembly Radial Peaking Factor would be deleted and the
definition of the Total Radial Peaking Factor
(FR^T) would be corrected to read:
``FR^T shall be the maximum ratio of the
individual fuel pin power to the core average pin power integrated over
the total core height, including tilt.'' In Limiting Condition for
Operation (LCO) 3.2.2, the title would be changed to ``TOTAL RADIAL
PEAKING FACTOR (FR^T);'' the wording would state
``FR^T shall be within the limits specified in the
[Core Operating Limits Report] COLR;'' Condition A would state
``FR^T not within limits specified in the COLR;''
Required Action A.1 would state ``Restore FR^T to
within limits;'' and Surveillance Requirement (SR) 3.2.2.1 would state
``Verify FR^T is within limits specified in the
COLR.'' In LCO 3.2.3, Required Action A.1 would state: Verify
FR^T is within the limits of LCO 3.2.2, ``Total
Radial Peaking Factor (FR^T)'.'' Associated
changes would be made to the TS Bases and table of contents.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:

A. Does the change involve a significant increase in the
probability or consequences of an accident previously evaluated?
There are no changes in plant systems, plant control operating
procedures or instrument alarm or trip settings associated with this
[TS Change Request] TSCR. Because neither physical equipment, nor
operating methods for that equipment change, the probability of
accident initiation would not change. Therefore, the proposed
technical specification change would not involve a significant
increase in the probability of an accident previously evaluated.
The assembly radial peaking (FR^A) has been
used in the past safety analyses and radiological consequence
analyses. These analyses utilized the assumption that
FR^A would remain within the Technical
Specifications limit during plant operations. These analyses verify,
for Anticipated Operational Occurrences (AOOs) and Postulated
Accidents (PAs), that:
(1) The Departure from Nucleate Boiling Ratio (DNBR) remains
above the appropriate Technical Specifications Safety Limit, and
(2) The calculated offsite doses and control room dose for the
affected events remained within the guidelines of 10 CFR 100,
Section 11, ``Determination of exclusion area, low population zone
and population center distance,'' and 10 CFR 50, Appendix A, General
Design Criteria (GDC) 19, ``Control room.''
Improved DNB correlations and better spacer grid design have
allowed the safety analysis calculations to be performed using only
the total radial peaking factor (FR^T) limit
(which remains unchanged), without exceeding the specified Safety
Limits. The radiological consequence events that previously used the
FR^A limit have been re-analyzed using the
slightly higher FR^T limit to determine the
source strength. The revised calculated offsite dose and control
room dose for the affected events remained within the guidelines of
10 CFR 100 and GDC 19.
Because the results of the transient analyses, which were
performed without FR^A

[[Page 22028]]

assumptions, continue to meet the Safety Limits, and because the
dose consequences of all analyzed events, which were also performed
without FR^A assumptions, continue to be within
the guidelines of 10 CFR 100 and GDC 19, the proposed technical
specification change would not involve a significant increase in the
consequences of an accident previously evaluated.
Therefore, operation of the plant in accordance with the
proposed Technical Specifications would not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
B. Does the change create the possibility of a new or different
kind of accident from any accident previously evaluated?
Operation of the plant in accordance with the proposed Technical
Specifications would not add any new equipment, settings, or alter
any plant operating practices. The only change is the deletion of
all Technical Specifications references to the Assembly Radial
Peaking Factor, FR^A, (a peaking factor no
longer used in core design or safety analyses). Since there will be
no change in operating plant equipment, settings, or normal
operating practices, operation in accordance with the proposed
Technical Specifications would not create the possibility of a new
or different kind of accident from any accident previously
evaluated.
C. Does this change involve a significant reduction in a margin
of safety?
The disposition of the [Standard Review Plan] SRP Chapter 15
events, the setpoint verification, the [fuel centerline melt] FCM
and the [minimum departure from nucleate boiling ratio] MDNBR
analyses documented in Siemens report EMF-2259 Revision 1,
``Palisades Cycle 15 Safety Analysis Report'' dated August 1999
considered the impact of several changes in fuel design and plant
operations for Cycle 15. A detailed and simplified XCOBRA-IIIC model
that incorporated limiting radial and axial power distributions, as
well as the removal of the FR^A peaking limit,
were developed for Cycle 15. This model was applied to all DNB event
analyses for Cycle 15 and the MDNBR values for limiting AOOs and PAs
were evaluated with the [High Thermal Performance] HTP DNB
correlation. The limiting MDNBR is calculated for SRP event 15.3.3
Reactor Coolant Pump Rotor Seizure and the limiting FCM is
calculated for SRP event 15.4.3 Single Rod Withdrawal. The
calculated results for the limiting events meet the Safety Limits
specified in TS LCO 2.1.
The SRP events were dispositioned in accordance with Siemens
approved methodologies listed in Palisades TS Section 5.6.5,
Amendment 189. The completed safety analysis supports Palisades
plant operation at 2530 Mwt.
The results of the transient analyses, which were performed
without FR^A assumptions, continue to meet the
Safety Limits, and the dose consequence of all analyzed events,
which were also performed without FR^A
assumptions, continue to be within the guidelines of 10 CFR 100 and
GDC 19 * * * [Therefore] operation of the Facility in accordance
with the proposed technical specification change would not involve a
significant reduction in the margin of safety.
Therefore, operation of the plant in accordance with the
proposed Technical Specifications would not involve a significant
reduction in a margin of safety.

Duke Energy Corporation, et al., Docket Nos. 50-413 and 50-414, Catawba
Nuclear Station, Units 1 and 2, York County, South Carolina

Date of amendment request: March 1, 2001.
Description of amendment request: The amendments would allow
implementation of 10 CFR Part 50, Appendix J, Option B, which governs
performance-based containment leakage testing requirements for Types B
and C testing. Catawba has previously implemented 10 CFR Part 50,
Appendix J, Option B requirements for Type A testing. In addition to
the changes associated with the adoption of 10 CFR Part 50, Appendix J,
Option B, the licensee is also proposing the following two changes: (1)
Technical Specification (TS) 3.6.3 will be modified to delete the
requirement for conducting soap bubble tests of welded penetrations
during Type A tests which are not individually Type B or Type C
testable, and (2) the Bases for TS 3.6.2 will be modified to clarify
that for the purpose of certain TS 3.6.2 Required Actions, the air lock
door bulkhead is considered to be part of the door.
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 following discussion is a summary of the evaluation of the
changes contained in this proposed amendment against the 10 CFR
50.92(c) requirements to demonstrate that all three standards are
satisfied. A no significant hazards consideration is indicated if
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.

First Standard

The proposed amendment will not involve a significant increase
in the probability or consequences of an accident previously
evaluated. Implementation of these changes will provide continued
assurance that specified parameters associated with containment
integrity will remain within acceptance limits as delineated in 10
CFR [Part] 50, Appendix J, Option B. The changes are consistent with
current safety analyses. Although some of the proposed changes
represent minor relaxation to existing TS requirements, they are
consistent with the requirements specified by Option B of 10 CFR
[Part] 50, Appendix J. The systems affecting containment integrity
related to this proposed amendment request are not assumed in any
safety analyses to initiate any accident sequence. Therefore, the
probability of any accident previously evaluated is not increased by
this proposed amendment. The proposed changes maintain an equivalent
level of reliability and availability for all affected systems. In
addition, maintaining leakage within analyzed limits assumed in
accident analyses does not adversely affect either onsite or offsite
dose consequences. Therefore, the proposed amendment does not
increase the consequences of any accident previously evaluated.

Second Standard

The proposed amendment will not create the possibility of a new
or different kind of accident from any accident previously
evaluated. No changes are being proposed which will introduce any
physical changes to the existing plant design. The proposed changes
are consistent with the current safety analyses. Some of the changes
may involve revision in the testing of components; however, these
are in accordance with the Catawba current safety analyses and
provide for appropriate testing or surveillance that is consistent
with 10 CFR [Part] 50, Appendix J, Option B. The proposed changes
will not introduce new failure mechanisms beyond those already
considered in the current safety analyses. No new modes of operation
are introduced by the proposed changes. The proposed changes
maintain, at minimum, the present level of operability of any system
that affects containment integrity.

Third Standard

The proposed amendment will not involve a significant reduction
in a margin of safety. The provisions specified in Option B of 10
CFR [Part] 50, Appendix J allow changes to Type B and Type C test
intervals based upon the performance of past leak rate tests. 10 CFR
[Part] 50, Appendix J, Option B allows longer intervals between
leakage tests based on performance trends, but does not relax the
leakage acceptance criteria. Changing test intervals from those
currently provided in the TS to those provided in 10 CFR [Part] 50,
Appendix J, Option B does not increase any risks above and beyond
those that the NRC has deemed acceptable for the performance

[[Page 22029]]

based option. In addition, there are risk reduction benefits
associated with reduction in component cycling, stress, and wear
associated with increased test intervals. The proposed changes
provide continued assurance of leakage integrity of containment
without adversely affecting the public health and safety and will
not significantly reduce existing safety margins. Similar proposed
changes have been previously reviewed and approved by the NRC, and
they are applicable to Catawba.
Based upon the preceding discussion, Duke Energy has concluded
that the proposed amendment does not involve a significant hazards
consideration.

The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendments request involves no significant hazards consideration.
Attorney for licensee: Ms. Lisa F. Vaughn, Legal Department
(PB05E), Duke Energy Corporation, 422 South Church Street, Charlotte,
North Carolina 28201-1006.
NRC Section Chief: Richard L. Emch, Jr..

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: March 29, 2001.
Description of amendment request: The proposed amendments would
revise the Keowee Hydro Unit (KHU) Technical Specifications
Surveillance Requirements (SRs) to address concerns related to voltage
and frequency overshoot during surveillance testing. This would be
accomplished by removing the note that had been implemented by
Amendment Nos. 316, 316, and 316 (October 4, 2000) for Oconee Nuclear
Station, Units 1, 2, and 3, respectively, to temporarily waive the
upper limits specified in Surveillance Requirement 3.8.1.9, thereby
reinstalling the original SR. In addition, as a result of an upgrade of
the KHU governors, the proposed amendments would reduce the time delay
specified in Technical Specification 3.8.1 and SR 3.8.1.17 from 12
seconds 1 second to 5 seconds 1 second. In
addition, related Bases changes have been proposed.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:

1. Involve a significant increase in the probability or
consequences of an accident previously evaluated.
No. The License Amendment Request (LAR) removes a Note to
Surveillance Requirement (SR) 3.8.1.9 that temporarily waived the
surveillance requirements associated with the upper limits for
Keowee Hydro Unit (KHU) voltage and frequency. The waiver of these
requirements allowed Duke to avoid an unplanned forced shutdown of
all three Oconee units, and the potential safety consequences and
operational risks associated with that action.
This LAR also changes the arming time delay associated with the
out-of-tolerance logic that had been approved for installation in
Amendment Nos. 312, 312, and 312. This change lowers the allowed
time delay, thereby resulting in the activation of the out-of-
tolerance logic more quickly after KHU startup.
Since this LAR assures that each KHU reaches its required
operating band within the required time, and that if maloperation of
a unit occurs, the KHU will be taken off line, the probability or
consequences of an accident previously evaluated is not
significantly increased.
2. Create the possibility of a new or different kind of accident
from any accident previously evaluated.
No. The LAR involves removing a Note that temporarily waived SR
3.8.1.9.a associated with the KHUs. This LAR also changes the time
delay associated with the activation of out-of-tolerance logic that
had been approved for installation in Amendment Nos. 312, 312, and
312. This change lowers the allowed time delay, thereby resulting in
the activation of the out-of-tolerance logic more quickly after KHU
startup.
Since this LAR restores Technical Specification SR 3.8.1.9 to
the condition prior to Amendment Nos. 316, 316, and 316 and provides
a shortened arming delay for the out-of-tolerance logic that was
approved in Amendment Nos. 312, 312, and 312, no new failure
mechanism or accident sequence is introduced. Therefore, the
possibility of a new or different kind of accident from any kind of
accident previously evaluated is not created.
3. Involve a significant reduction in a margin of safety.
No. The LAR involves removing a Note that allowed temporary
waiver of the requirements to meet SR 3.8.1.9.a and shortens the
arming time delay associated with the activation of out-of-tolerance
logic that had been approved for installation in Amendment Nos. 312,
312, and 312.
This LAR, therefore, improves the margin of safety by assuring
that SR 3.8.1.9.a can be implemented. The change to a shorter arming
time delay for the out-of-tolerance circuit activation also improves
the margin of safety by limiting the time that a KHU would be
carrying safety loads in an out-of-tolerance condition.
Therefore, this request does not involve a significant reduction
in a margin of safety.

The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: Anne W. Cottington, Winston and Strawn, 1200
17th Street, NW., Washington, DC 20005.
NRC Section Chief: Richard L. Emch, Jr.

Entergy Operations Inc., Docket No. 50-382, Waterford Steam Electric
Station, Unit 3, St. Charles Parish, Louisiana

Date of amendment request: April 13, 2001.
Description of amendment request: The proposed change relaxes the
allowable cooldown rate in the Reactor Coolant System (RCS) Technical
Specifications (TS) 3.4.8.1, ``Pressure / Temperature Limits.''
Specifically, the change eliminates the limitation of a 10 deg.F per
hour cooldown rate when the RCS temperature is below 135 deg.F. The
proposed limitations permit a 100 deg.F per hour cooldown rate to
continue down to an RCS temperature of 110 deg.F, at which point the
rate is reduced to 30 deg.F per hour.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:

1. Will the operation of the facility in accordance with this
proposed change involve a significant increase in the probability or
consequences of an accident previously evaluated?
Response:
Limitations have been imposed on cooldown of the Reactor Coolant
System (RCS) to assure compliance with the minimum temperature
requirements of 10 CFR [Part] 50, Appendix G. The proposed changes
revise the allowable cooldown limits in a way such that operation
remains consistent with the design assumptions and satisfies the
stress limits for cyclic operation. By ensuring operation remains
within the bounds of the existing design basis and assumptions, the
probability of a brittle fracture of the reactor vessel has not been
increased.
Therefore, the proposed change will not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
2. Will the operation of the facility in accordance with this
proposed change create the possibility of a new or different kind of
accident from any accident previously evaluated?
Response:
The proposed changes will not create the possibility of a new or
different kind of accident from any previously analyzed since they
do not introduce new systems, failure

[[Page 22030]]

modes, or other plant perturbations. The proposed changes revise the
cooldown limitations based on the fact the conservatively estimated
peak pressure that can occur when the RCS cold leg temperature is
below 200 deg.F is less than the proposed pressure limit. The
limits assure that operation remains consistent with the design
assumptions and satisfies the stress limits for cyclic operation.
Therefore, the proposed change will not create the possibility
of a new or different kind of accident from any accident previously
analyzed.
3. Will the operation of the facility in accordance with this
proposed change involve a significant reduction in a margin of
safety?
Response:
The margin of safety provided by Technical Specification 3.4.8.1
is based on assuring that the maximum cooldown rates are consistent
with the design assumptions and satisfy the stress limits for cyclic
operation. The proposed changes will not involve a significant
reduction in the margin of safety since equivalent pressure and
temperature limit requirements for reactor operation will be
applied. The proposed changes were derived in accordance with
approved NRC methodology which was developed to assure the reactor
coolant system pressure boundary is designed with sufficient margin
to withstand any condition during normal operation including
anticipated operational occurrences and system in-service leak and
hydrostatic tests.
These requirements were revised in accordance with 10 CFR [Part]
50, Appendix G utilizing the latest NRC guidance in Regulatory Guide
1.99, Revision 2 relative to estimating neutron irradiation damage
to the reactor vessel. In addition, the 16 EFPY [effective full
power year] basis for these pressure/temperature limits has been
found to include sufficient margin to account for the limits of
uncertainty described in Draft Regulatory Guide DG-1053.
Therefore, the proposed change will not involve in a significant
reduction in a margin of safety.

The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: N.S. Reynolds, Esquire, Winston & Strawn,
1400 L Street NW., Washington, DC 20005-3502.
NRC Section Chief: Robert A. Gramm.

Exelon Generation Company, LLC, Docket Nos. STN 50-456 and STN 50-457,
Braidwood Station, Unit Nos. 1 and 2, Will County, Illinois

Date of amendment request: February 9, 2001.
Description of amendment request: The proposed amendment (one-time
change) revises the Steam Generator (SG) inspection frequency
requirements in TS 5.5.9.d.2, ``Steam Generator (SG) Tube Surveillance
Program, Inspection Frequencies,'' for the Braidwood Station, Unit 1,
Fall 2001 refueling outage, to allow a 40 month inspection interval
after one SG inspection, rather than after two consecutive inspections
resulting in C-1 classification. This one-time change is proposed to
eliminate unnecessary SG inspections during the upcoming Unit 1, Fall
2001 refueling outage, thus, resulting in significant dose, schedule,
and cost savings.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:

1. Does the proposed change involve a significant increase in
the probability or consequences of an accident previously evaluated?
The proposed one-time change revises the Steam Generator (SG)
inspection interval requirements in Technical Specifications (TS)
5.5.9.d.2, ``Steam Generator (SG) Tube Surveillance Program,
Inspection Frequencies,'' for the Braidwood Station, Unit 1, Fall
2001 refueling outage, to allow a 40 month inspection frequency
after one inspection, rather than after two consecutive inspections
results that are within the C-1 category. C-1 category is defined as
``5% of the total tubes inspected are degraded tubes and none of the
inspected tubes are defective.''
The proposed one-time extension of the Unit 1, SG tube inservice
inspection interval does not involve changing any structure, system,
or component, or affect reactor operations. It is not an initiator
of an accident and does not change any existing safety analysis
previously analyzed in the Byron/Braidwood Stations' Updated Final
Safety Analysis Report (UFSAR). As such, the proposed changes do not
involve a significant increase in the probability of an accident
previously evaluated.
Since the proposed change does not alter the plant design, there
is no direct increase in SG leakage. Industry experience indicates
that the probability of increased SG tube degradation would not go
undetected. Additionally, steps described below will further
minimize the risk associated with this extension. For example, the
scope of inspections performed during the last Braidwood Station,
Unit 1, refueling outage (i.e., the first refueling outage following
SG replacement) exceeded the TS requirements for the first two
refueling outages after SG replacement. That is, more tubes were
inspected than were required by TS. Currently, Braidwood Station,
Unit 1, does not have an active SG damage mechanism, and will meet
the current industry examination guidelines without performing SG
inspections during the next refueling outage. Additionally, as part
of our SG Tube Surveillance Program, both a Condition Monitoring
Assessment and an Operational Assessment are performed after each
inspection and compared to the Nuclear Energy Institute (NEI) 97-06,
``Steam Generator Program Guidelines,'' performance criteria. The
results of the Condition Monitoring Assessment demonstrated that all
performance criteria were met during the Braidwood Station, Unit 1,
Spring 2000 refueling outage, and the results of the Operational
Assessment show that all performance criteria will be met over the
proposed operating period. Considering these actions, along with the
improved SG design and reliability of Babcock and Wilcox
International (BWI) replacement SGs, extending the SG tube
inspection frequency does not involve a significant increase in the
consequences of an accident previously evaluated.
Therefore, the proposed change does not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
2. Does the proposed change create the possibility of a new or
different kind of accident from any accident previously evaluated?
The proposed change revises the SG inspection frequency
requirements in TS 5.5.9.d.2, ``Steam Generator (SG) Tube
Surveillance Program, Inspection Frequencies,'' for the Braidwood
Station, Unit 1, Fall 2001 refueling outage, to allow a 40 month
inspection interval after one inspection, rather than after two
consecutive inspections with inspection results within the C-1
category.
The proposed change will not alter any plant design basis or
postulated accident resulting from potential SG tube degradation.
The scope of inspections performed during the last Braidwood
Station, Unit 1, refueling outage (i.e., the first refueling outage
following SG replacement) significantly exceeded the TS requirements
for the scope of the first two refueling outages after SG
replacement.
Primary to secondary leakage that may be experienced during all
plant conditions is expected to remain within current accident
analysis assumptions. The proposed change does not affect the design
of the SGs, the method of SG operation, or reactor coolant chemistry
controls. No new equipment is being introduced, and installed
equipment is not being operated in a new or different manner. The
proposed change involves a one-time extension to the SG tube
inservice inspection frequency, and therefore will not give rise to
new failure modes. In addition, the proposed change does not impact
any other plant system or components.
Therefore, the proposed change does not create the possibility
of a new or different kind of accident from any accident previously
evaluated.
3. Does the proposed change involve a significant reduction in a
margin of safety?
The SG tubes are an integral part of the Reactor Coolant System
(RCS) pressure boundary that are relied upon to maintain the RCS
pressure and inventory. The SG tubes isolate the radioactive fission
products in the reactor coolant from the secondary system.

[[Page 22031]]

The safety function of the SGs is maintained by ensuring the
integrity of the SG tubes. In addition, the SG tubes comprise the
heat transfer surface between the primary and secondary systems such
that residual heat can be removed from the primary system.
SG tube integrity is a function of the design, environment, and
current physical condition. Extending the SG tube inservice
inspection frequency by one operating cycle will not alter the
function or design of the SGs. SG inspections conducted during the
first refueling outage following SG replacement demonstrated that
the SGs do not have an active damage mechanism, and the scope of
those inspections significantly exceeded those required by the TS.
These inspection results were comparable to similar inspection
results for the same model of replacement SGs installed at other
plants, and subsequent inspections at those plants yielded results
that support this extension request. The improved design of the
replacement SGs also provides reasonable assurance that significant
tube degradation is not likely to occur over the proposed operating
period.
Therefore, the proposed changes do not involve a significant
reduction in a margin of safety.

The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
requested amendments involve no significant hazards consideration.
Attorney for licensee: Mr. Robert Helfrich, Senior Counsel,
Nuclear, Mid-West Regional Operating Group, Exelon Generation Company,
LLC, 1400 Opus Place, Suite 900, Downers Grove, Illinois 60515.
NRC Section Chief: Anthony J. Mendiola.

Nuclear Management Company, LLC, Docket No. 50-305, Kewaunee Nuclear
Power Plant, Kewaunee County, Wisconsin

Date of amendment request: April 6, 2001.
Description of amendment request: The proposed amendment would
change the Kewaunee Nuclear Power Plant Technical Specification Section
6.2, ``Organization,'' and Section 6.13, ``High Radiation Area'' to
reflect the title change from Shift Supervisor to Shift Manager.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:

(1) Involve a significant increase in the probability or
consequences of an accident previously evaluated.
The proposed change will not alter the intent of the TS.
Changing the title from Shift Supervisor to Shift Manager is
administrative in nature. It has no impact on accident initiators or
plant equipment, and thus, does not affect the probability or
consequences of an accident.
(2) Create the possibility of a new or different kind of
accident from any accident previously evaluated.
The proposed change does not involve a change to the physical
plant or operations. Since this is an administrative change it does
not contribute to accident initiation. Therefore, it does not
produce a new accident scenario or produce a new type of equipment
malfunction.
(3) Involve a significant reduction in the margin of safety.
Since this is an administrative change, it does not involve a
significant reduction in the margin of safety. The proposed change
does not affect plant equipment or operation. Safety limits and
limiting safety system settings are not affected by this change.

The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: Bradley D. Jackson, Esq., Foley and Lardner,
P.O. Box 1497, Madison, WI 53701-1497.
NRC Section Chief: Claudia M. Craig.

Omaha Public Power District, Docket No. 50-285, Fort Calhoun Station,
Unit No. 1, Washington County, Nebraska

Date of amendment request: February 7, 2001.
Description of amendment request: The proposed amendment would
revise the technical specifications to replace the accident source term
used in all design basis site boundary and control room dose analysis
with the alternate source term. Additionally, the proposed amendment
would implement regulatory guidance provided in Regulatory Guide (RG)
1.183, ``Alternative Radiological Source Terms for Evaluating Design
Basis Accidents at Nuclear Power Reactors,'' regarding the licensing
basis source term for design basis events.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:

1. The proposed change does not involve a significant increase
in the probability or consequences of an accident previously
evaluated.
The proposed changes to FCS [Fort Calhoun Station] TS [Technical
Specifications] modify requirements to: place the control room
ventilation system in operation and in filtered air mode during
refueling operations in the containment or spent fuel pool, place a
spent fuel pool area radiation monitor in operation during refueling
operations at the spent fuel pool, delete a specification that
requires a ventilation isolation actuation signal (VIAS) and two
radiation monitors to be operable, increase the volume of trisodium
phosphate (TSP) in the reactor containment building, include both
internal and external leakage for the residual heat removal (RHR)
system leakage test, perform an internal leakage test on the RHR
system, and credit the alternative source term (AST) for the design
basis site boundary and control room dose analyses. These TS changes
do not impact operation of other equipment or systems important to
safety. The proposed TS changes reflect the parameters used in the
radiological consequences calculations described in Attachment E [to
the licensee's February 7, 2001, letter].
The current TS 3.16 limits RHR system leakage to 1243 cc/hour
from external sources and does not provide a limit for leakage from
internal sources due to valve seat back leakage to the safety
injection refueling water tank (SIRWT) or require an internal
leakage test to be performed. The re-analysis for LOCA [loss-of-
coolant accident] assumed a total leakage from all RHR sources of
3800 cc/hour. The internal leakage would leak back into the water
remaining in the SIRWT. While it appears the allowable leakage is
being increased, the limit is more inclusive, and therefore, more
conservative than the current leakage limit. The internal leakage
test performed on the RHR system will measure and quantify the back
leakage into the SIRWT.
The proposed changes to TSs 2.3 and 3.6 are necessary to ensure
the post-LOCA pH of the recirculation water is equal to or greater
than 7.0. Radiation levels in containment following a LOCA may cause
the generation of hydrochloric and nitric acids from radiolysis of
cable insulation and sump water. TSP will neutralize these acids.
The radiolysis analysis performed demonstrates that the sump pH will
be greater than or equal to 7.0 post design basis accident (DBA),
which meets the intent of RG 1.183 regarding iodine revolatization.
Therefore, there is no increase in the probability or consequences
of an accident previously evaluated due to radiolysis concerns.
The proposed change to TS 2.8.2(4) requires the control room
ventilation system to be in operation and in the Filtered Air mode.
This is a conservative action to reduce control room operator
exposure. This action is credited in the fuel handling accident
analysis. 10 CFR 50.36 requires, in part, that if an operating
restriction is an initial condition of a DBA, then a Limiting
Condition for Operation (LCO) should be established. Therefore, this
action, which will reduce operator exposure, will not involve a
significant increase in the probability or consequences of an
accident previously evaluated.
The proposed change to TS 2.8.3(5) will delete the requirement
for the ventilation

[[Page 22032]]

isolation actuation signal (VIAS) to be operable with two radiation
monitors operable, and require the control room ventilation system
to be in operation and in the Filtered Air mode and a spent fuel
pool area radiation monitor to be in operation during refueling
operations in the spent fuel pool. The current basis for TS 2.8.3(5)
is to ensure the control room ventilation system is operated in
Filtered Air mode upon receipt of a VIAS. The proposed change will
require the control room ventilation system placed in the Filtered
Air mode during refueling operations, thereby eliminating the need
for the VIAS to be operable. Therefore, this change will not involve
a significant increase in the probability or consequences of an
accident previously evaluated.
The changes proposed do not affect the precursors for accidents
or transients analyzed in Chapter 14 of the FCS USAR. Therefore,
there is no increase in the probability of accidents previously
evaluated. The probability remains the same since the accident
analyses performed and discussed in the basis for the TS changes,
involve no change to a system, component or structure that affects
initiating events for any USAR Chapter 14 accident evaluated. A re-
analysis of USAR Chapter 14 events was conducted with respect to
radiological consequences. This re-analysis was performed in
accordance with current accepted methodology, and consequences were
expressed in terms of TEDE [total effective dose equivalent] dose.
The current methodology is no longer exactly comparable to the
previous methods used for dose consequences. The previous dose
calculations analyzed the dose consequences to thyroid and whole
body as a result of postulated DBA events. The previous dose
calculations were shown to be well below the regulatory limits of 10
CFR 100.11 (25 percent) with respect to thyroid and whole body dose.
The current accepted NRC methodology, as described in 10 CFR 50.67,
specifies new dose acceptance criteria in terms of TEDE dose. The
revised analyses for all evaluated DBA events meet the applicable
TEDE dose acceptance criteria (specified also in RG 1.183) for
alternative source term implementation. The most current analyses do
not credit several engineered safeguards features (ESF) filtration
systems as the previous analyses did, and hence, are more
conservative in that aspect. If a comparison is performed between
the previous calculations (thyroid and whole body dose) and revised
analyses TEDE results (per method shown in footnote 7 of RG 1.183),
a slight increase in dose consequences is exhibited but is not
significant, and the TEDE results are below regulatory acceptance
criteria.
The changes proposed do not increase the probability of an
accident previously evaluated. Because of the new regulatory
requirements related to AST implementation, the dose consequences,
if compared to previous ones, are only slightly increased (using
guidance in footnote 7 of RG 1.183). However, the dose consequences
of the revised analyses are below the AST regulatory acceptance
criteria.
2. The proposed change does not create the possibility of a new
or different kind of accident from any accident previously
evaluated.
The implementation of the proposed changes does not create the
possibility of an accident of a different type than was previously
evaluated in the USAR. The proposed changes to FCS TS modify
requirements to: place the control room ventilation system in
operation and in filtered air mode during refueling operations in
the containment or spent fuel pool, place a spent fuel pool area
radiation monitor in operation during refueling operations at the
spent fuel pool, delete a specification that requires a ventilation
isolation actuation signal (VIAS) and two radiation monitors to be
operable, increase the volume of trisodium phosphate (TSP) in the
reactor containment building, include both internal and external
leakage for the residual heat removal (RHR) system leakage test,
perform an internal leakage test on the RHR system, and credit the
alternative source term (AST) for the design basis site boundary and
control room dose analyses[.]
The changes proposed do not change how DBA events were
postulated nor do the changes themselves initiate a new kind of
accident with a unique set of conditions. The changes proposed were
based on a complete re-analysis of offsite and control room operator
doses, where the system requirements being revised were not credited
in the calculations. The revised analyses are consistent with the
regulatory guidance established in RG 1.183. The revised analyses
utilize the most current understanding of source term timing and
chemical forms as a more appropriate mitigation technique. Not
crediting filtration systems and only crediting natural forces is
conservative from the aspect of dose consequences. Through this re-
analysis, no new accident initiator or failure mode was identified.
3. The proposed change does not involve a significant reduction
in a margin of safety.
The implementation of the proposed changes does not reduce the
margin of safety. The radiological analyses results, with the
proposed changes, remain within the regulatory acceptance criteria
(10 CFR 50 Appendix A, 10 CFR 50.67) utilizing the TEDE dose
acceptance criteria directed in RG 1.183. These criteria have been
developed for application to analyses performed with alternative
source terms. These acceptance criteria have been developed for the
purpose of use in design basis accident analyses such that meeting
these limits demonstrates adequate protection of public health and
safety. An acceptable margin of safety is inherent in these
licensing limits. Therefore, there is no significant reduction in
the margin of safety as a result of the proposed changes.

The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: James R. Curtiss, Esq., Winston & Strawn,
1400 L Street, NW., Washington, DC 20005-3502.
NRC Section Chief: Stephen Dembek.

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 amendment requests: April 6, 2001.
Description of amendment requests: The amendment application
proposes to revise the Facility Operating License No. NPF-10, and
Facility Operating License No. NPF-15 for San Onofre Nuclear Generating
Station, Units 2 and 3, respectively. The licensee proposed to add
annotations to technical specification Surveillance Requirements
3.8.1.2, 3.8.1.3, 3.8.1.9, 3.8.1.10 and 3.8.1.19 that provide guidance
to ensure a diesel generator sub-component, an automatic voltage
regulator (AVR), is operable and regularly tested. The proposed
annotations clarify that only one AVR is required for the associated
diesel generator to be operable and only one AVR can be in service at
any one time.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:

1. Involve a significant increase in the probability or
consequences of an accident previously evaluated?
Response: No.
The diesel generators provide emergency power to accident
mitigation equipment in the event of a loss of offsite power. They
cannot cause an accident. The San Onofre Nuclear Generating Station
(SONGS) emergency diesel generators (EDG) each have two automatic
voltage regulators (AVRs) that are 100% redundant to each other.
Maintaining both AVRs for each diesel in a high state of readiness,
while minimizing unnecessary testing on the diesels, optimizes the
overall availability of the diesel generator systems to perform
their function if required.
This change allows testing the two AVRs for each diesel on a
staggered monthly basis. In addition, it clarifies that each AVR
only needs to be subjected to a dynamically challenging test once
every 24 months provided that its dynamic performance is measured
and determined to be acceptable. These testing requirements
demonstrate a high level of assurance that each AVR will be capable
of performing its design function while minimizing unnecessary wear
on the diesels. The reliability of the diesel generators to provide
emergency power will not be degraded as a result of this change.
Therefore, this change does not involve a significant increase in
the probability or consequences of any accident previously
evaluated.

[[Page 22033]]

2. Create the possibility of a new or different kind of accident
from any accident previously evaluated?
Response: No.
The AVRs are a subcomponent of the EDGs. This change to the
surveillance test frequency does not physically change the use,
function, or design of the EDG or its subcomponent, the AVR.
This change ensures both 100% capacity AVRs are adequately
tested to ensure operability without increasing the number of test
starts of the EDGs.
Therefore, this change does not create the possibility of a new
or different kind of accident from any previously evaluated.
3. Involve a significant reduction in a margin of safety?
Response: No.
This change allows testing the two AVRs for each diesel on a
staggered monthly basis. In addition, it clarifies that each AVR
only needs to be subjected to a dynamically challenging test once
every 24 months provided that its dynamic performance is measured
and determined to be acceptable. These testing requirements
demonstrate a high level of assurance that each AVR will be capable
of performing its design function while minimizing unnecessary wear
on the diesels. This proposed change does not involve an alteration
of the SONGS 2 and 3 design. The reliability of the diesel
generators to provide emergency power will not be degraded as a
result of this change.
Therefore, this change does not involve a significant reduction
in a margin of safety.

The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment requests involve no significant hazards consideration.
Attorney for licensee: Douglas K. Porter, Esquire, Southern
California Edison Company, 2244 Walnut Grove Avenue, Rosemead,
California 91770.
NRC Section Chief: Stephen Dembek.

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 amendment request: January 11, 2001.
Description of amendment request: The proposed amendments would
revise Technical Specifications 5.5.17, ``Containment Leakage Rate
Testing Program,'' to add an exception to Regulatory Guide (RG) 1.163,
``Performance-Based Containment Leak-Testing Program.'' Specifically,
the licensee proposes to use America Society of Mechanical Engineering,
Subsections IWL and IWE to meet the intent of RG 1.163. The proposed
change will affect the frequency of containment concrete visual
examinations and allow the examinations to be preformed during power
operation instead of exclusively during refueling outages.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:

1. Does the proposed change involve a significant increase in
the probability or consequences of an accident previously evaluated?
No. The proposed change affects the frequency of visual
examinations that will be performed for the concrete surfaces of the
Vogtle Electric Generating Plant (VEGP) Unit 1 and Unit 2
containments for the purpose of the Containment Leakage Rate Testing
Program. In addition, the proposed change allows those examinations
to be performed during power operation as opposed to during a
refueling outage. The frequency of visual examinations of the
concrete surfaces of the containments and the mode of operation
during which those examinations are performed has no relationship to
or adverse impact on the probability of any of the initiating events
assumed for the accident analyses. Therefore, the proposed change
does not involve a significant increase in the probability of any
accident previously evaluated. The proposed change would allow
visual examinations that are performed pursuant to NRC-approved
[American Society of Mechanical Engineering] ASME Section XI Code
requirements (except where relief has been granted by the NRC) to
meet the intent of visual examinations required by Regulatory Guide
1.163, without requiring additional visual examinations pursuant to
the Regulatory Guide. The intent of early detection of deterioration
will continue to be met by the more rigorous requirements of the
Code-required visual examinations. Therefore, the safety function of
the VEGP containments as a fission product barrier will be
maintained, and there will not be a significant increase in the
consequences of an accident previously evaluated.
2. Does the proposed change create the possibility of a new or
different kind of accident from any previously evaluated?
No. The proposed change affects the frequency of visual
examinations that will be performed for the concrete surfaces of the
Vogtle Electric Generating Plant (VEGP) Unit 1 and Unit 2
containments for the purpose of the Containment Leakage Rate Testing
Program. In addition, the proposed change allows those examinations
to be performed during power operation as opposed to during a
refueling outage. The proposed change does not adversely affect or
otherwise alter plant operation. No new equipment is introduced, and
no new limiting single failures are created. Therefore, the proposed
change will not create the possibility of a new or different kind of
accident from any previously evaluated.
3. Does the proposed change involve a significant reduction in a
margin of safety?
No. The proposed change affects the frequency of visual
examinations that will be performed for the concrete surfaces of the
Vogtle Electric Generating Plant (VEGP) Unit 1 and Unit 2
containments for the purpose of the Containment Leakage Rate Testing
Program. In addition, the proposed change allows those examinations
to be performed during power operation as opposed to during a
refueling outage. The proposed change would allow visual
examinations that are performed pursuant to NRC-approved ASME
Section XI Code requirements (except where relief has been granted
by the NRC) to meet the intent of visual examinations required by
Regulatory Guide 1.163, without requiring additional visual
examinations pursuant to the Regulatory Guide. The intent of early
detection of deterioration will continue to be met by the more
rigorous requirements of the Code-required visual examinations.
Therefore, the safety function of the VEGP containments as a fission
product barrier will be maintained, and there will not be a
significant reduction in a margin of safety.

The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: Mr. Arthur H. Domby, Troutman Sanders,
NationsBank Plaza, Suite 5200, 600 Peachtree Street, NE., Atlanta,
Georgia 30308-2216.
NRC Section Chief: Richard L. Emch, Jr.

Tennessee Valley Authority, Docket Nos. 50-327 and 50-328, Sequoyah
Nuclear Plant, Units 1 and 2, Hamilton County, Tennessee

Date of application for amendments: April 12, 2001 (TS 00-02).
Brief description of amendments: The proposed amendment would
change the Sequoyah Nuclear Plant Technical Specification (TS)
surveillance requirements for the ice condenser. The request would
change the method and frequency for determining boron concentration and
pH of the ice and proposes an additional test requirement for ice that
is to be added to the ice condenser.
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 only analyzed accidents of possible consideration in regards
to changes potentially affecting the ice condenser are a

[[Page 22034]]

loss of coolant accident (LOCA) and a main steam line break (MSLB)
inside containment. However, the ice condenser is not postulated as
being the initiator of any LOCA or MSLB. This is because it is
designed to remain functional following a design basis earthquake,
and the ice condenser does not interconnect or interact with any
systems that interconnect or interact with the reactor coolant or
main steam systems. Since the proposed changes to the TS and TS
bases are solely to revise and provide clarification of the ice
sampling and chemical analysis requirements, and are not the result
of or require any physical change to the ice condenser, there can be
no change in the probability of an accident previously evaluated in
the Safety Analysis Report.
In order for the consequences of any previously evaluated event
to be changed, there would have to be a change in the ice
condenser's physical operation during a LOCA or MSLB, or in the
chemical composition of the stored ice. The proposed changes do not
alter either from existing requirements, except to add an upper
limit on boron concentration, which is the bounding value for the
hot leg switchover timing calculation. Though the frequency of the
existing surveillance requirement (SR) for sampling the stored ice
is changed from once every 18 months to once every 54 months, the
sampling requirements are strengthened overall with: (1) the
requirement to obtain one randomly selected sample from each ice
condenser bay (24 total samples) rather than 9 ``representative''
samples, and (2) the addition of a new SR to verify each addition of
ice meets the existing requirements for boron concentration and pH
value. The only other change is to clarify that each sample of
stored ice is individually analyzed for boron concentration and pH,
but that the acceptance criteria for each parameter is based on the
average values obtained for the 24 samples. This is consistent with
the bases for the boron concentration of the ice, which is to ensure
the accident analysis assumptions for containment sump pH and boron
concentration are not altered following complete melting of the ice
condenser. Historically, chemical analysis of the stored ice has had
a very limited number of instances where an individual sample did
not meet the boron or pH requirements, with all subsequent
evaluations (follow-up sampling) showing the ice condenser as a
whole was well within these requirements. Requiring chemical
analysis of each sample is provided to preclude the practice of
melting all samples together before performing the analysis, and to
ensure the licensee is alerted to any localized anomalies for
investigation and resolution without the burden of entering a 24-
hour action, provided the averaged results are acceptable. Thus,
based on the above, the proposed changes do 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.
Because the TS and TS bases changes do not involve any physical
changes to the ice condenser, any physical or chemical changes to
the ice contained therein, or make any changes in the operational or
maintenance aspects of the ice condenser as required by the TS,
there can be no new accidents created from those already identified
and evaluated.
C. The proposed amendment does not involve a significant
reduction in a margin of safety.
The ice condenser TSs ensure that during a LOCA or MSLB the ice
condenser will initially pass sufficient air and steam mass to
preclude over pressurizing lower containment, that it will absorb
sufficient heat energy initially and over a prescribed time period
to assist in precluding containment vessel failure, and that it will
not alter the bulk containment sump pH and boron concentration
assumed in the accident analysis. Since the proposed changes do not
physically alter the ice condenser, but rather only serve to
strengthen and clarify ice sampling and analysis requirements, the
only area of potential concern is the effect these changes could
have on bulk containment sump pH and boron concentration following
ice melt. However, this is not affected because there is no change
in the existing requirements for pH and boron concentration, except
to add an upper limit on boron concentration. This upper limit is
the bounding value for the hot leg switchover timing calculation.
Averaging the pH and boron values obtained from analysis of the
individual samples taken is not a new practice, just one that was
not consistently used by all ice condenser plants. Using the
averaged values provides an equivalent bulk value for the ice
condenser, which is consistent with the accident analysis for the
bulk pH and boron concentration of the containment sump following
ice melt. Changing the performance frequency for sampling the stored
ice does not reduce any margin of safety because: (1) the newly
proposed surveillance (SR 4.6.5.1.f) ensures ice additions meet the
existing boron concentration and pH requirements, (2) there are no
normal operating mechanisms, including sublimation, that reduce the
ice condenser bulk pH and boron concentration, and (3) the number of
required samples has been increased from 9 to 24 (1 randomly
selected ice basket per bay), which is approximately the same number
of samples that would have been taken in the same time period under
the existing requirements. Thus, it can be concluded that the
proposed TS and TS bases changes do not involve a significant
reduction in 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: April 3, 2001.
Brief description of amendments: The proposed change would revise
Technical Specification (TS) 3.3.6, ``Containment Ventilation Isolation
Instrumentation'' to modify the Note for Required Action B.1 such that
it applies only to ``Required Action and associated Completion Time of
Condition A not met.'' The proposed change is the result of the
discovery of an error which occurred when the TS was converted to the
``improved TS format'' with issuance of License Amendment Nos. 64 and
64, for Comanche Peak Steam Electric Station, Units 1 and 2 on February
26, 1999.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:

1. Do the proposed changes involve a significant increase in the
probability or consequences of an accident previously evaluated?
Response: No.
The proposed change removes an allowance to open containment
pressure relief valves under administrative controls when one train
of Automatic Actuation Logic and Actuation Relays is inoperable. The
proposed change corrects a non-conservative technical specification
and thus makes the technical specifications consistent with the
previously evaluated accident analyses. Therefore, the proposed
change does not involve a significant increase in the probability or
consequences of an accident 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 change makes the technical specifications
consistent with the previously evaluated accident analyses.
Therefore, the proposed change does not create the possibility of a
new or different kind of accident from any previously evaluated.
3. Do the proposed changes involve a significant reduction in a
margin of safety?
Response: No.
The proposed change makes the technical specifications
consistent with the previously evaluated accident analyses.
Therefore the proposed change does not involve a reduction in a
margin of safety.

The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff

[[Page 22035]]

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: March 22, 2001 (ET 01-0012).
Description of amendment request: The amendment would (1) decrease
the allowable values for Function 8, pressurizer pressure-low,
pressurizer pressure-high, in Table 3.3.1-1, ``Reactor Trip System
Instrumentation,'' and (2) decrease the allowable value for pressurizer
pressure-low for safety injection in Table 3.3.2-1, ``Engineered Safety
Feature Actuation System Instrumentation.'' The changes are needed
because the licensee will be replacing the existing Tobar pressurizer
pressure transmitters with Rosemount transmitters in the next refueling
outage.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:

1. The proposed change does not involve a significant increase
in the probability or consequences of an accident previously
evaluated.
The existing safety related pressurizer pressure transmitters
are being replaced with ones of similar characteristics and
functions, and without changing the design or functional basis of
the system, structure, or components associated with the pressure
transmitters.
The protection system performance will remain within the bounds
of the previously performed accident analysis. The Reactor Trip
System (RTS) and Engineered Safety Feature Actuation System (ESFAS)
instrumentation will continue to function in a manner consistent
with the plant design basis. The replacement of the pressurizer
pressure transmitters and proposed changes to the affected Allowable
Values will not affect any of the analysis assumptions for any of
the accidents previously evaluated, since the changes are consistent
with the setpoint methodology and ensure adequate margin to the
Safety Analysis Limit. The proposed changes will not affect any
event initiators nor will the proposed changes 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. There will be no change to
normal plant operating parameters or accident mitigation
capabilities.
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.
A review of the failure modes and effects in Updated Safety
Analysis Report Section 7.7.2 found that failure of the replacement
pressure transmitters will be the same as for the existing pressure
transmitters. As such, the effects of such failures on [the safety]
functions of the other equipment are concluded to be similar to
those previously evaluated.
There are no changes in the method by which any safety related
plant system performs its safety function. The normal manner of
plant operation remains unchanged. The increase [or decrease] in the
pressurizer pressure functions Allowable Values still provides
acceptable margin between the nominal Trip Setpoint and Allowable
Value. The changes in Allowable Value does not impact the systems
capability to provide both control and protection functions. No new
accident scenarios, transient precursors, failure mechanisms, or
limiting single failures are introduced as a result of the proposed
changes.
Therefore, the proposed changes do 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.
The proposed changes do not affect the acceptance criteria for
any analyzed event nor is there a change in any Safety Analysis
Limit. There will be no effect on the manner in which safety limits
or RTS and ESFAS settings are determined nor will there be any
affect on those plant systems necessary to assure the accomplishment
of protection functions. [The proposed changes to the pressurizer
pressure Allowable Values will maintain the accident analyses in the
Updated Safety Analysis Report.]
Therefore, the proposed changes do not involve a significant
reduction in a margin of safety.

The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: Jay Silberg, Esq., Shaw, Pittman, Potts and
Trowbridge, 2300 N Street, NW., Washington, DC 20037.
NRC Section Chief: Stephen Dembek.

Wolf Creek Nuclear Operating Corporation, Docket No. 50-482, Wolf Creek
Generating Station, Coffey County, Kansas

Date of amendment request: March 23, 2001 (CO 01-0013).
Description of amendment request: The amendment proposes to (1)
delete certain license conditions from the operating license, and (2)
revise Table 5.5.9-2, ``Steam Generator Tube Inspection,'' in Section
5.5.9, ``Steam Generator (SG) Tube Surveillance Program,'' of the
technical specifications (TSs). License Conditions 2.C.(4) and 2.C.(6)
through 2.C.(14) of the facility operating license are considered to
have been completed and obsolete, or to duplicate other license
requirements, and are proposed to be deleted. Attachments 2 and 3 to
the facility operating license are also proposed to be deleted. Section
2.F of the facility operating license is considered to duplicate the
reporting requirements in 10 CFR 50.72 and 50.73 and is proposed to be
deleted. The reporting requirements in two ``Action Required'' columns
of TS Table 5.5.9-2 are also considered to duplicate the reporting
requirements in 10 CFR 50.72 and 50.73 and are proposed to be deleted.
The list of the attachments and appendices to the facility operating
license would also be revised to reflect the proposed deletion of
Attachments 2 and 3.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:

1. The proposed change does not involve a significant increase
in the probability or consequences of an accident previously
evaluated.
This request involves administrative changes only. No actual
plant equipment or accident analyses will be affected by the
proposed changes. 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.
This request involves administrative changes only. No actual
plant equipment or accident analyses will be affected by the
proposed change and no failure modes not bounded by previously
evaluated accidents will be created. Therefore, the proposed changes
do 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.
Margin of safety is associated with confidence in the ability of
the fission product barriers (i.e., fuel and fuel cladding, Reactor
Coolant System pressure boundary, and containment structure) to
limit the level of radiation dose to the public. This request
involves administrative changes only [and does not change these
barriers].
No actual plant equipment or accident analyses will be affected
by the proposed change. Additionally, the proposed changes

[[Page 22036]]

will not relax any criteria used to establish safety limits, will
not relax any safety system settings, or will not relax the bases
for any limiting conditions of operation [in the TSs]. Therefore,
the proposed changes do not involve a significant reduction in a
margin of safety.

The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: Jay Silberg, Esq., Shaw, Pittman, Potts and
Trowbridge, 2300 N Street, NW., Washington, DC 20037.
NRC Section Chief: Stephen Dembek.

Wolf Creek Nuclear Operating Corporation, Docket No. 50-482, Wolf Creek
Generating Station, Coffey County, Kansas

Date of amendment request: April 3, 2001 (ET 01-0008).
Description of amendment request: The amendment would make the
following changes to the technical specifications (TSs):
(1) Revise Safety Limit 2.1.1 by replacing Figure 2.1.1-1,
``Reactor Core Safety Limits,'' with a reference to limits being
specified in the Core Operating Limits Report (COLR) and by adding two
reactor core safety limits on departure from nucleate boiling ratio
(DNBR) and peak fuel centerline temperature.
(2) Revise Note 1 on the over temperature T in Table
3.3.1-1 of TS 3.3.1, ``Reactor Trip System Instrumentation,'' by
replacing values of parameters with a reference to the values being
specified in the COLR and correcting the expression for one term in the
inequality for over temperature T.
(3) Revise Note 2 on the overpower T in Table 3.3.1-1 by
replacing values of parameters with a reference to the values being
specified in the COLR.
(4) Replace the limits for the reactor coolant system (RCS)
pressure and average temperature with a reference to the limits being
specified in the COLR for Limiting Condition for Operation (LCO) 3.4.1
and Surveillance Requirements (SRs) 3.4.1.1 and 3.4.1.2.
(5) Add the phrase ``and greater than or equal to the limit
specified in the COLR'' to the RCS total flow rate in LCO 3.4.1 and SRs
3.4.1.3 and 3.4.1.4.
(6) Move items a. and b. to the left in the Note to the
applicability in LCO 3.4.1.
(7) Revise TS Section 5.6.5 by adding TS 3.3.1 on over temperature
and overpower ``T trip setpoints and TS 3.4.1 on RCS pressure,
temperature, and flow limits to the existing list of core operating
limits for each reload cycle that are documented in the COLR and
revising the list of topical reports in the COLR that represent the
analytical methods approved by the Commission to determine core
operating limits.
The proposed changes remove cycle-specific parameter limits and
relocate them to the COLR, but they (1) do not change any of the
limits, (2) add more specific requirements regarding DNBR limit and
peak fuel centerline temperature limit to the TSs, (3) revise the list
of topical reports in the list of NRC-approved analytical methods, (4)
correct one term of an expression, and (5) move terms in a Note to the
mode applicability for an LCO.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:

1. The proposed change does not involve a significant increase
in the probability or consequences of an accident previously
evaluated.
The proposed changes are programmatic and administrative in
nature which do not physically alter safety related systems, nor
affect the way in which safety related systems perform their
functions. More specific requirements regarding the safety limits
(i.e., DNBR limit and peak fuel centerline temperature limit) are
being imposed in TS 2.1.1, ``Reactor Core Safety Limits,'' which
replace the Reactor Core Safety Limits figure and are consistent
with the values stated in the USAR [Updated Safety Analysis Report].
The proposed changes remove the cycle-specific parameter limits from
TS 3.4.1 and relocate them to the COLR which do not change plant
design or affect system operating parameters. In addition, the
minimum limit for RCS total flow rate is being retained in TS 3.4.1
to assure that a lower flow rate than reviewed by the NRC will not
be used. The proposed changes do not, by themselves, alter any of
the parameter limits. The removal of the cycle-specific parameter
limits from the TS does not eliminate existing requirements to
comply with the parameter limits. The existing TS Section 5.6.5b,
COLR Reporting Requirements, continues to ensure that the analytical
methods used to determine the core operating limits meet NRC
reviewed and approved methodologies. The existing TS Section 5.6.5c,
COLR Reporting Requirements, continues to ensure that applicable
limits of the safety analyses are met.
The proposed changes to reference only the Topical Report number
and title do not alter the use of the analytical methods used to
determine core operating limits that have been reviewed and approved
by the NRC. This method of referencing Topical Reports would allow
the use of current Topical Reports to support limits in the COLR
without having to submit an amendment to [the TS of] the operating
license. Implementation of revisions to Topical Reports would still
be reviewed in accordance with 10 CFR 50.59 and where required
receive NRC review and approval.
Although the relocation of the cycle-specific parameter limits
to the COLR would allow revision of the affected parameter limits
without prior NRC approval, there is no significant effect on the
probability or consequences of an accident previously evaluated.
Future changes to the COLR parameter limits could result in event
consequences which are either slightly less or slightly more severe
than the consequences for the same event using the present parameter
limits. The differences would not be significant and would be
bounded by the existing requirement of TS Section 5.6.5c to meet the
applicable limits of the safety analyses.
The cycle-specific parameter limits being transferred from the
TS to the COLR will continue to be controlled under existing
programs and procedures. The USAR accident analyses will continue to
be examined with respect to changes in the cycle-dependent
parameters obtained using NRC reviewed and approved reload design
methodologies, ensuring that the transient evaluation of new reload
designs are bounded by previously accepted analyses. This
examination will continue to be performed pursuant to 10 CFR 50.59
requirements ensuring that future reload designs will not involve a
significant increase in the probability or consequences of an
accident previously evaluated. Additionally, the proposed changes do
not allow for an increase in plant power levels, do not increase the
production, nor alter the flow path or method of disposal of
radioactive waste or byproducts. Therefore, the proposed changes do
not change the types or increase the amounts of any effluents
released offsite.
[The proposed changes to the expression of the
f1(I) term, which is in the over temperature
T inequality, clarifies and corrects the term. Moving the
terms in a Note to the LCO mode applicability is an administrative
action.]
Therefore, the proposed changes do not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
2. The proposed change does not create the possibility of a new
or different kind of accident from any accident previously
evaluated.
[The proposed changes are programmatic and administrative in
nature which do not physically alter safety related systems, nor
affect the way in which safety related systems perform their
functions.]
The proposed changes that retain the minimum limit for RCS total
flow rate in the TS, and that relocate certain cycle-specific
parameter limits from the TS to the COLR, thus removing the
requirement for prior NRC approval of revisions to those parameters,
do not involve a physical change to the plant. No new equipment is
being introduced, and installed equipment is not being operated in a
new or different manner. There are no

[[Page 22037]]

changes being made to the parameters within which the plant is
operated, other than their relocation to the COLR. There are no
setpoints affected by the proposed changes at which protective or
mitigative actions are initiated. The proposed changes will not
alter the manner in which equipment operation is initiated, nor will
the function demands on credited equipment be changed. No alteration
in the procedures which ensure the plant remains within analytical
limits is being proposed, and no change is being made to the
procedures relied upon to respond to an off-normal event. As such,
no new failure modes are being introduced.
The proposed changes to reference only the Topical Report number
and title do not alter the use of the analytical methods used to
determine core operating limits that have been reviewed and approved
by the NRC. This method of referencing Topical Reports would allow
the use of current Topical Reports to support limits in the COLR
without having to submit an amendment to [the TS of] the operating
license. Implementation of revisions to Topical Reports would still
be reviewed in accordance with 10 CFR 50.59 and where required
receive NRC review and approval.
Relocation of cycle-specific parameter limits has no influence
or impact on, nor does it contribute in any way to the possibility
of a new or different kind of accident. The relocated cycle-specific
parameter limits will continue to be calculated using the NRC
reviewed and approved methodology. The proposed changes do not alter
assumptions made in the safety analysis and operation within the
core operating limits will continue.
[The proposed changes to the expression of the
f1(I) term, which is in the over temperature
T inequality, clarifies and corrects the term.]
Therefore, the proposed changes do 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.
The margin of safety is established through equipment design,
operating parameters, and the setpoints at which automatic actions
are initiated. The proposed changes [are programmatic and
administrative in nature and] do not physically alter safety related
systems, nor does it [a]ffect the way in which safety-related
systems perform their functions. The setpoints at which protective
actions are initiated are not altered by the proposed changes.
Therefore, sufficient equipment remains available to actuate upon
demand for the purpose of mitigating an analyzed event. As the
proposed changes to relocate cycle-specific parameter limits to the
COLR will not affect plant design or system operating parameters,
there is no detrimental impact on any equipment design parameter,
and the plant will continue to operate within prescribed limits.
The development of cycle-specific parameter limits for future
reload designs will continue to conform to NRC reviewed and approved
methodologies, and will be performed pursuant to 10 CFR 50.59 to
assure that plant operation [is] within cycle-specific parameter
limits.
The proposed changes to reference only the Topical Report number
and title do not alter the use of the analytical methods used to
determine core operating limits that have been reviewed and approved
by the NRC. This method of referencing Topical Reports would allow
the use of [the] current Topical Reports to support limits in the
COLR without having to submit an amendment to [the TS of] the
operating license. Implementation of revisions to Topical Reports
would still be reviewed in accordance with 10 CFR 50.59 and where
required receive NRC review and approval.
[The proposed changes to the expression of the
f1(I) term, which is in the over temperature
T inequality, clarifies and corrects the term. Moving the
terms in a Note to the LCO mode applicability is an administrative
action.]
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 request involves no significant hazards consideration.
Attorney for licensee: Jay Silberg, Esq., Shaw, Pittman, Potts and
Trowbridge, 2300 N Street, NW., Washington, DC 20037.
NRC Section Chief: Stephen Dembek.

Notice of Issuance of Amendments to Facility Operating Licenses

During the period since publication of the last biweekly notice,
the Commission has issued the following amendments. The Commission has
determined for each of these amendments that the application complies
with the standards and requirements of the Atomic Energy Act of 1954,
as amended (the Act), and the Commission's rules and regulations. The
Commission has made appropriate findings as required by the Act and the
Commission's rules and regulations in 10 CFR Chapter I, which are set
forth in the license amendment.
Notice of Consideration of Issuance of Amendment to Facility
Operating License, Proposed No Significant Hazards Consideration
Determination, and Opportunity for A Hearing in connection with these
actions was published in the Federal Register as indicated.
Unless otherwise indicated, the Commission has determined that
these amendments satisfy the criteria for categorical exclusion in
accordance with 10 CFR 51.22. Therefore, pursuant to 10 CFR 51.22(b),
no environmental impact statement or environmental assessment need be
prepared for these amendments. If the Commission has prepared an
environmental assessment under the special circumstances provision in
10 CFR 51.12(b) and has made a determination based on that assessment,
it is so indicated.
For further details with respect to the action see (1) the
applications for amendment, (2) the amendment, and (3) the Commission's
related letter, Safety Evaluation and/or Environmental Assessment as
indicated. All of these items are available for public inspection at
the Commission's Public Document Room, located at One White Flint
North, 11555 Rockville Pike (first floor), Rockville, Maryland 20852.
Publicly available records will be accessible and 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 Nos.
1, 2, and 3, Maricopa County, Arizona

Date of application for amendments: February 28, 2001.
Brief description of amendments: The amendments revise the
definitions of engineered safety feature response time and reactor
protection system response time in Technical Specification (TS) 1.1,
``Definitions,'' to add the following statement: ``In lieu of
measurement, response time may be verified for selected components
provided that the components and methodology for verification have been
previously reviewed and approved by the [Nuclear Regulatory Commission
] NRC.'' Approval of the amendments will allow either an allocated
sensor response time or a measured sensor response time for the
identified Reactor Protection System and Engineered Safety Features
Actuation System pressure sensors when performing response time
testing.
Date of issuance: April 19, 2001.
Effective date: April 19, 2001, and shall be implemented within 45
days of the date of issuance.
Amendment Nos.: Unit 1-135, Unit 2-135, Unit 3-135.
Facility Operating License Nos. NPF-41, NPF-51, and NPF-74: The
amendments revised the Technical Specifications.
Date of initial notice in Federal Register: March 20, 2001 (66 FR
15766).
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated April 19, 2001.
No significant hazards consideration comments received: No. Calvert
Cliffs Nuclear Power Plant, Inc., Docket Nos. 50-317 and 50-318,
Calvert.

[[Page 22038]]

Cliffs Nuclear Power Plant, Unit Nos. 1 and 2, Calvert County, Maryland

Date of application for amendments: December 21, 2000, as
supplemented on February 12, 2001, and March 5, 2001.
Brief description of amendments: The amendments revise Technical
Specification 5.2.2.e by removing the reference to the Nuclear
Regulatory Commission Policy Statement on working hours.
Date of issuance: April 5, 2001.
Effective date: As of the date of issuance to be implemented within
30 days.
Amendment Nos.: 245 and 219.
Renewed Facility Operating License Nos. DPR-53 and DPR-69:
Amendments revised the Technical Specifications.
Date of initial notice in Federal Register: February 7, 2001 (66 FR
9380).
The February 12, 2001, and March 5, 2001, letters provided
clarifying information that did not change the initial proposed no
significant hazards consideration determination.
The Commission's related evaluation of these amendments is
contained in a Safety Evaluation dated April 5, 2001.
No significant hazards consideration comments received: No.

Calvert Cliffs Nuclear Power Plant, Inc., Docket No. 50-318, Calvert
Cliffs Nuclear Power Plant, Unit No. 2, Calvert County, Maryland

Date of application for amendment: September 14, 2000, as
supplemented on December 21, 2000.
Brief description of amendment: The amendment permits operation of
Calvert Cliffs Unit 2 with a core containing a lead fuel (test)
assembly that includes fuel rods with advanced zirconium alloy
cladding.
Date of issuance: April 5, 2001.
Effective date: As of the date of issuance to be implemented within
30 days.
Amendment No.: 220.
Renewed License No. DPR-69: Amendment revised the Technical
Specifications.
Date of initial notice in Federal Register: January 10, 2001 (66 FR
2012).
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated April 5, 2001.
No significant hazards consideration comments received: No.

Commonwealth Edison Company, Docket Nos. STN 50-454 and STN 50-455,
Byron Station, Unit Nos. 1 and 2, Ogle County, Illinois

Docket Nos. STN 50-456 and STN 50-457, Braidwood Station, Unit Nos.
1 and 2, Will County, Illinois.
Date of application for amendments: June 19, 2000, as supplemented
by letters dated March 16, 2001, and April 4, 2001.
Brief description of amendments: The proposed amendments revised
the technical specifications to remove their applicability related to
the Boron Dilution Protection System (BDPS) after the next refueling
outage for each unit. During the refueling outages, modifications are
scheduled to be made which will permit mitigation of a boron dilution
event without the use of the BDPS.
Date of issuance: April 6, 2001.
Effective date: Immediately, to be implemented upon completion of
the modifications scheduled to be completed after cycle 9 for Byron,
Unit 2, and Braidwood, Units 1 and 2, and after cycle 11 for Byron,
Unit 1.
Amendment Nos.: 117 and 111.
Facility Operating License Nos. NPF-37, NPF-66, NPF-72 and NPF-77:
The amendments revised the Technical Specifications.
Date of initial notice in Federal Register: September 6, 2000 (65
FR 54084).
Since the proposed additional changes provided in this supplement
are more restrictive than the originally proposed changes, it does not
change the previous determination of no significant hazards
consideration determination.
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated April 6, 2001.
No significant hazards consideration comments received: No.

Consumers Energy Company, Docket No. 50-255, Palisades Plant, Van Buren
County, Michigan

Date of application for amendment: February 12, 2001.
Brief description of amendment: The amendment changes Technical
Specification Section 5.6.5b, ``Reporting Requirements--Core Operating
Limits Report (COLR),'' to add a report pertaining to statistical
setpoint methodology to the list of approved methodology references.
Date of issuance: April 9, 2001.
Effective date: As of the date of issuance, to be implemented
within 30 days.
Amendment No.: 195.
Facility Operating License No. DPR-20. Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: March 7, 2001 (66 FR
13801).
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated April 9, 2001.
No significant hazards consideration comments received: No.

Duke Energy Corporation, Docket Nos. 50-369 and 50-370, McGuire Nuclear
Station, Units 1 and 2, Mecklenburg County, North Carolina

Date of application for amendments: August 22, 2000, as
supplemented by letter dated November 7, 2000.
Brief description of amendments: The amendments revise the
Technical Specifications (TS) of each unit to restore a time limit for
an allowable condition for the occurrence of an inoperable refueling
water storage tank level transmitter in TS 3.3.2.
Date of issuance: April 12, 2001.
Effective date: As of the date of issuance and shall be implemented
within 30 days from the date of issuance.
Amendment Nos.: 198 and 179.
Facility Operating License Nos. NPF-9 and NPF-17: Amendments
revised the Technical Specifications.
Date of initial notice in Federal Register: November 1, 2000 (65 FR
65341).
The supplement dated November 7, 2000, provided clarifying
information that did not change the scope of the August 22, 2000,
application nor the initial proposed no significant hazards
consideration determination.
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated April 12, 2001.
No significant hazards consideration comments received: No.

Entergy Nuclear Operations, Inc., Docket No. 50-286, Indian Point
Nuclear Generating Unit No. 3, Westchester County, New York

Date of application for amendment: September 6, 2000, as
supplemented on January 18, and April 2, 2001.
Brief description of amendment: The amendment revises Technical
Specification 5.5.15 to allow a one time change in the 10 CFR part 50,
Appendix J, Type A test interval from the required 10 years to a test
interval of 15 years.
Date of issuance: April 17, 2001.
Effective date: April 17, 2001.
Amendment No.: 206.
Facility Operating License No. DPR-64:
Date of initial notice in Federal Register: January 24, 2000 (66 FR
7665).
The January 18, and April 2, 2001, submittals contained clarifying
information only, and did not change the initial no significant hazards
consideration determination.
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated April 17, 2001.

[[Page 22039]]

No significant hazards consideration comments received: No.

Entergy Nuclear Generation Company, Docket No. 50-293, Pilgrim Nuclear
Power Station, Plymouth County, Massachusetts

Date of application for amendment: February 16, 2001.
Brief description of amendment: This amendment substitutes a
surveillance interval of ``Once/Operating Cycle'' for the current
surveillance interval of ``Each Refueling Outage,'' for the following
instruments in Technical Specification Table 4.2.F: Containment High
Radiation Monitor, Reactor Building Vent Radiation Monitor, Main Stack
Vent Radiation Monitor, and Turbine Building Vent Radiation Monitor.
Date of issuance: April 9, 2001.
Effective date: As of the date of issuance, and shall be
implemented within 30 days.
Amendment No.: 189.
Facility Operating License No. DPR-35: Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: March 7, 2001 (66 FR
13802).
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated April 9, 2001.
No significant hazards consideration comments received: No.

Entergy Nuclear Generation Company, Docket No. 50-293, Pilgrim Nuclear
Power Station, Plymouth County, Massachusetts

Date of application for amendment: November 22, 2000, as
supplemented on January 30 and February 2, 2001.
Brief description of amendment: This amendment changes the
pressure-temperature limit curves of Figures 3.6.1, 3.6.2, and 3.6.3 of
the Technical Specifications (TS) over operation between 20, 32, and 48
Effective Full Power Years. However, these curves will only apply for
the remainder of operating cycles 13 and 14. The Bases section has been
modified to reflect these TS changes.
Date of issuance: April 13, 2001.
Effective date: As of the date of issuance, and shall be
implemented within 30 days.
Amendment No.: 190.
Facility Operating License No. DPR-35: Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: December 27, 2000 (65
FR 81915).
The January 30 and February 2, 2001, letters provided 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 April 13, 2001.
No significant hazards consideration comments received: No.

Entergy Operations, Inc., Docket No. 50-368, Arkansas Nuclear One, Unit
No. 2, Pope County, Arkansas

Date of application for amendment: August 10, 2000, as supplemented
by letter dated March 22, 2001.
Brief description of amendment: The amendment revised Technical
Specification 3/4.9.4, ``Refueling Operations, Containment Building
Penetrations,'' by deleting the requirements for the containment purge
and exhaust system and by revising the closure requirements for
containment building penetrations to require that containment
penetrations are capable of being closed during the handling of
irradiated fuel within the containment.
Date of issuance: April 18, 2001.
Effective date: As of the date of issuance to be implemented within
60 days from the date of issuance.
Amendment No.: 230.
Facility Operating License No. NPF-6: Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: September 20, 2000 (65
FR 56950).
The March 22, 2001, supplemental letter provided clarifying
information that was within the scope of the original Federal Register
notice and did not change the staff's initial no significant hazards
consideration determination.
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated April 18, 2001.
No significant hazards consideration comments received: No.

Exelon Generation Company, LLC, Docket No. 50-353, Limerick Generating
Station, Unit 2, Montgomery County, Pennsylvania

Date of application for amendment: February 1, 2001, as
supplemented March 6 and 23, 2001.
Brief description of amendment: This amendment revises the minimum
critical power ratio safety limits for operating cycle 7.
Date of issuance: April 12, 2001.
Effective date: As of date of issuance, and shall be implemented
within 30 days.
Amendment No.: 114.
Facility Operating License No. NPF-85. This amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: February 21, 2001 (66
FR 11061).
The March 6 and 23, 2001, letters provided 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 April 12, 2001.
No significant hazards consideration comments received: No.

Florida Power and Light Company, Docket No. 50-335, St. Lucie Plant,
Unit No. 1, St. Lucie County, Florida

Date of application for amendment: December 4, 2000, as
supplemented February 9, 2001.
Brief description of amendment: This amendment changes the
licensing bases to incorporate a revised analysis of the Main Steam
Line Break inside containment.
Date of Issuance: April 20, 2001.
Effective Date: April 20, 2001.
Amendment No.: 175.
Facility Operating License No. DPR-67: Amendment revised the
Updated Final Safety Analysis Report.
Date of initial notice in Federal Register: February 7, 2001 (66 FR
9383).
The February 9, 2001, Supplement did not affect the original
proposed no significant hazards determination, or expand the scope of
the request as noticed in the Federal Register.
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated April 20, 2001.
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 18, 2000.
Description of amendment request: The amendment deletes Technical
Specifications Section 6.7.6.e, ``Post-Accident Sampling,'' for
Seabrook Station, Unit No. 1 and thereby eliminates the requirements to
have and maintain the post-accident sampling system.
Date of issuance: April 17, 2001.
Effective date: As of its date of issuance, and shall be
implemented within 60 days.
Amendment No.: 78.
Facility Operating License No. NPF-86: Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: January 24, 2000 (66 FR
7683).

[[Page 22040]]

The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated April 17, 2001.
No significant hazards consideration comments received: No.

Nuclear Management Company, LLC, Docket No. 50-331, Duane Arnold Energy
Center, Linn County, Iowa

Date of application for amendment: October 19, 2000, as
supplemented November 16, 2000, and April 9, 2001, and as limited in
scope by letter dated March 23, 2001.
Brief description of amendment: The amendment revises the Technical
Specifications regarding operability requirements during core
alterations and while moving irradiated fuel assemblies within the
secondary containment. The amendment also provides for a change in
design and licensing bases for a selective application of the alternate
radiological source term in accordance with 10 CFR 50.67, ``Accident
Source Term,'' and revised meteorology dispersion values, both being
limited to a design-basis fuel handling accident.
Date of issuance: April 16, 2001.
Effective date: As of the date of issuance and shall be implemented
within 60 days.
Amendment No.: 237.
Facility Operating License No. DPR-49: The amendment revised the
Technical Specifications and the licensing and design bases regarding a
design-basis fuel handling accident.
Date of initial notice in Federal Register: March 6, 2001 (66 FR
13598).
NMC's letters dated March 23 and April 9, 2001, are within the
scope of the changes proposed in NMC's letter of October 19, 2000, that
was noticed in the Federal Register on March 6, 2001.
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated April 16, 2001.
No significant hazards consideration comments received: No.

Nuclear Management Company, LLC, Docket Nos. 50-282 and 50-306, Prairie
Island Nuclear Generating Plant, Units 1 and 2, Goodhue County,
Minnesota

Date of application for amendments: April 17, 2000, as supplemented
February 2, 2001.
Brief description of amendments: The amendments change the
Technical Specifications (TSs) for removal of boric acid storage tanks
from the safety injection (SI) system. These changes accomplish two
objectives: (1) Eliminate high concentration boric acid from the SI
system and (2) align this specific TS section with the standard TSs.
Date of issuance: April 16, 2001.
Effective date: As of the date of issuance and shall be implemented
within 90 days.
Amendment Nos.: 156 and 147.
Facility Operating License Nos. DPR-42 and DPR-60: Amendments
revised the Technical Specifications.
Date of initial notice in Federal Register: March 7, 2001 (66 FR
13806).
The February 2, 2001, supplement provided clarifying information
that was within the scope of the original application and did not
change the staff's initial proposed no significant hazards
consideration determination.
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated April 16, 2001.
No significant hazards consideration comments received: No.

PPL Susquehanna, LLC, Docket Nos. 50-387 and 50-388, Susquehanna Steam
Electric Station, Units 1 and 2, Luzerne County, Pennsylvania

Date of application for amendments: October 4, 2000, as
supplemented March 12, April 2, and April 5, 2001.
Brief description of amendments: The amendments revise the
surveillance test requirements for excess flow check valves (EFCVs) to
allow testing of a representative sample at 24-month intervals such
that each EFCV is tested at least once every 10 years.
Date of issuance: April 11, 2001.
Effective date: As of date of issuance to be implemented within 30
days.
Amendment Nos.: 193 and 168.
Facility Operating License Nos. NPF-14 and NPF-22. The amendments
revised the Technical Specifications.
Date of initial notice in Federal Register: January 10, 2001 (66 FR
2021).
The March 12, April 2, and April 5, letters provided clarifying
information that did not change the initial proposed no significant
hazards consideration determination or expand the amendment beyond the
scope of the initial notice.
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated April 11, 2001.
No significant hazards consideration comments received: No.

Union Electric Company, Docket No. 50-483, Callaway Plant, Unit 1,
Callaway County, Missouri

Date of application for amendment: January 18, 2001, as
supplemented by letter dated February 21, 2001.
Brief description of amendment: The amendment authorizes the
installation of new engineered safety feature transformers as an
improvement. This amendment will allow the installation and use of the
new transformers equipped with automatic load tap changers and an
update to the Final Safety Analysis Report (FSAR) to reflect their
installation.
Date of issuance: April 6, 2001.
Effective date: April 6, 2001, and shall be implemented in the next
periodic update to the FSAR in accordance with 10 CFR 50.71(e).
Amendment No.: 143.
Facility Operating License No. NPF-30: The amendment revised the
FSAR.
Date of initial notice in Federal Register: February 21, 2001 (66
FR 11063).
The February 21, 2001, supplemental letter 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 April 6, 2001.
No significant hazards consideration comments received: No.

Union Electric Company, Docket No. 50-483, Callaway Plant, Unit 1,
Callaway County, Missouri

Date of application for amendment: January 18, 2001.
Brief description of amendment: The amendment deletes Section
5.5.3, ``Post Accident Sampling,'' of the Technical Specifications for
the Callaway Plant and thereby eliminates the requirements to have and
maintain the post-accident sampling system (PASS).
Date of issuance: April 6, 2001.
Effective date: April 6, 2001, to be implemented within 60 days of
the date of issuance.
Amendment No.: 144.
Facility Operating License No. NPF-30: The amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: March 7, 2001 (66 FR
13808).
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated April 6, 2001.
No significant hazards consideration comments received: No.

Dated at Rockville, Maryland this 25th day of April 2001.

For The Nuclear Regulatory Commission.
John A. Zwolinski,
Director, Division of Licensing Project Management, Office of Nuclear
Reactor Regulation
[FR Doc. 01-10822 Filed 5-1-01; 8:45 am]
BILLING CODE 7590-01-P