[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.

    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.

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:
    <bullet> 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.
    <bullet> 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.
    <bullet> 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.
    <bullet> 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.
    <bullet> 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:
    <bullet> 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,'' 
(F<INF>R</INF><SUP>A</SUP>). 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 
(F<INF>R</INF><SUP>T</SUP>) would be corrected to read: 
``F<INF>R</INF><SUP>T</SUP> 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 (F<INF>R</INF><SUP>T</SUP>);'' the wording would state 
``F<INF>R</INF><SUP>T</SUP> shall be within the limits specified in the 
[Core Operating Limits Report] COLR;'' Condition A would state 
``F<INF>R</INF><SUP>T</SUP> not within limits specified in the COLR;'' 
Required Action A.1 would state ``Restore F<INF>R</INF><SUP>T</SUP> to 
within limits;'' and Surveillance Requirement (SR) 3.2.2.1 would state 
``Verify F<INF>R</INF><SUP>T</SUP> is within limits specified in the 
COLR.'' In LCO 3.2.3, Required Action A.1 would state: Verify 
F<INF>R</INF><SUP>T</SUP> is within the limits of LCO 3.2.2, ``Total 
Radial Peaking Factor (F<INF>R</INF><SUP>T</SUP>)'.'' 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 (F<INF>R</INF><SUP>A</SUP>) has been 
used in the past safety analyses and radiological consequence 
analyses. These analyses utilized the assumption that 
F<INF>R</INF><SUP>A</SUP> 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 (F<INF>R</INF><SUP>T</SUP>) limit 
(which remains unchanged), without exceeding the specified Safety 
Limits. The radiological consequence events that previously used the 
F<INF>R</INF><SUP>A</SUP> limit have been re-analyzed using the 
slightly higher F<INF>R</INF><SUP>T</SUP> 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 F<INF>R</INF><SUP>A</SUP>

[[Page 22028]]

assumptions, continue to meet the Safety Limits, and because the 
dose consequences of all analyzed events, which were also performed 
without F<INF>R</INF><SUP>A</SUP> 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, F<INF>R</INF><SUP>A</SUP>, (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 F<INF>R</INF><SUP>A</SUP> 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 F<INF>R</INF><SUP>A</SUP> assumptions, continue to meet the 
Safety Limits, and the dose consequence of all analyzed events, 
which were also performed without F<INF>R</INF><SUP>A</SUP> 
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.

    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.

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 <plus-minus>1 second to 5 seconds <plus-minus>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 <greek-D>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 <greek-D>T.
    (3) Revise Note 2 on the overpower <greek-D>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 
f<INF>1</INF>(<greek-D>I) term, which is in the over temperature 
<greek-D>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 
f<INF>1</INF>(<greek-D>I) term, which is in the over temperature 
<greek-D>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 
f<INF>1</INF>(<greek-D>I) term, which is in the over temperature 
<greek-D>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