[Federal Register Volume 66, Number 163 (Wednesday, August 22, 2001)]
[Notices]
[Pages 44161-44182]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: 01-20885]


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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 July 30, 2001 through August 10, 2001. The 
last biweekly notice was published on August 8, 2001 (66 FR 41609).

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

    The Commission has made a proposed determination that the following 
amendment requests involve no significant hazards consideration. Under 
the Commission's regulations in 10 CFR 50.92, this means that operation 
of the facility in accordance with the proposed amendment would not (1) 
involve a significant increase in the probability or consequences of an 
accident previously evaluated; or (2) create the possibility of a new 
or different kind of accident from any accident previously evaluated; 
or (3) involve a significant reduction in a margin of safety. The basis 
for this proposed determination for each amendment request is shown 
below.
    The Commission is seeking public comments on this proposed 
determination. Any comments received within 30 days after the date of 
publication of this notice will be considered in making any final 
determination.
    Normally, the Commission will not issue the amendment until the 
expiration of the 30-day notice period. However, should circumstances 
change during the notice period such that failure to act in a timely 
way would result, for example, in derating or shutdown of the facility, 
the Commission may issue the license amendment before the expiration of 
the 30-day notice period, provided that its final determination is that 
the amendment involves no significant hazards consideration. The final 
determination will consider all public and State comments received 
before action is taken. Should the Commission take this action, it will 
publish in the Federal Register a notice of issuance and provide for 
opportunity for a hearing after issuance. The Commission expects that 
the need to take this action will occur very infrequently.
    Written comments may be submitted by mail to the Chief, Rules and 
Directives Branch, Division of Administrative Services, Office of 
Administration, U.S. Nuclear Regulatory Commission, Washington, DC 
20555-0001, and should cite the publication date and page number of 
this Federal Register notice. Written comments may also be delivered to 
Room 6D22, Two White Flint North, 11545 Rockville Pike, Rockville, 
Maryland from 7:30 a.m. to 4:15 p.m. Federal workdays. Copies of 
written comments received may be examined at the NRC 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 September 21, 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

[[Page 44162]]

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. 
Publicly available records will be accessible from the Agencywide 
Documents Assess and Management Systems (ADAMS) Public Electronic 
Reading Room on the internet at the NRC Web site, http://www.nrc.gov/NRC/ADAMS/index.html. If you do not have access to ADAMS or if there 
are problems in accessing the documents located in ADAMS, contact the 
NRC Public Document room (PDR) Reference staff at 1-800-397-4209, 304-
415-4737 or by email to [email protected].

AmerGen Energy Company, LLC

[Docket No. 50-461, Clinton Power Station, Unit 1, DeWitt County, 
Illinois]

[Docket No. 50-219, Oyster Creek Generating Station, Ocean County, New 
Jersey]

[Docket No. 50-289, Three Mile Island Nuclear Station, Unit 1, Dauphin 
County, Pennsylvania]

    Date of amendment request: July 9, 2001.
    Description of amendment request: The proposed amendments would 
incorporate TS changes that are being made to provide consistency with 
the changes to 10 CFR 50.59, ``Changes, tests, and experiments,'' as 
published in the Federal Register (FR) Volume 64, beginning on page 
53582 (i.e., 64 FR 53582), dated October 4, 1999. Specifically, the 
changes replace the terms ``safety evaluation'' with ``10 CFR 50.59 
evaluation'' and ``unreviewed safety question'' with ``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:

    1. Does the change involve a significant increase in the 
probability or consequences of an accident previously evaluated?
    The proposed changes reflect revision to 10 CFR 50.59, 
``Changes, tests, and experiments,'' issued as a Final Rule on 
October 4, 1999, and do not impact the operation of any system or 
component assumed in any accident analysis. The proposed changes do 
not change the requirement to perform a 10 CFR 50.59 review when 
required by the Technical Specifications Administrative Controls or 
by a license condition. Due to the administrative nature of these 
proposed changes there will be no direct impact on the consequences 
of any accident previously evaluated. Therefore, these proposed 
changes do not involve a significant increase in the probability or 
consequences of an accident previously evaluated.
    2. Does the change create the possibility of a new or different 
kind of accident from any accident previously evaluated?
    The proposed changes are administrative in nature and do not 
involve a change to the plant design or operation. No new or 
different types of equipment will be installed as a result of these 
changes. The proposed changes make the language in the Technical 
Specifications Administrative Controls and a license condition 
conform to the revised 10 CFR 50.59 rule, dated October 4, 1999. No 
new accident modes or equipment failure modes are created by these 
proposed changes. Therefore, these proposed changes do not create 
the possibility of a new or different kind of accident from any 
accident previously evaluated.
    3. Does the change involve a significant reduction in a margin 
of safety?

[[Page 44163]]

    The proposed changes do not have a direct effect on any safety 
analysis assumptions. The proposed changes are administrative in 
nature and make the Technical Specifications Administrative Controls 
and a license condition language conform to the revised 10 CFR 50.59 
rule, dated October 4, 1999. Changes to the facility that result in 
meeting the criteria of 10 CFR 50.59 will still require NRC approval 
pursuant to 10 CFR 50.59. 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 
requested amendments involve no significant hazards consideration.
    Attorney for licensee: Mr. Edward J. Cullen, Jr., Vice President, 
General Counsel, Exelon Generation Company, LLC, 300 Exelon Way, 
Kennett Square, PA 19348.
    NRC Section Chief: Anthony J. Mendiola.

AmerGen Energy Company, LLC, et al., Docket No. 50-219, Oyster 
Creek Nuclear Generating Station, Ocean County, New Jersey

    Date of amendment request: December 19, 2000.
    Description of amendment request: The proposed amendment would 
revise the Oyster Creek Technical Specification (TS) Section 3.17 Bases 
to remove reference to the current licensing basis control room 
calculated dose consequences and substitute the associated regulatory 
dose limits that apply for control room habitability in accordance with 
General Design Criterion 19 and Section 6.4 of the Standard Review 
Plan. The existing licensing basis control room calculated dose values 
specified in TS Section 3.17 Bases have been reevaluated as a result of 
Oyster Creek Licensee Event Report No. 00-006 dated June 26, 2000. This 
reevaluation has confirmed that the control room habitability dose 
limits continue to be met. However, this reevaluation is based on use 
of the U.S. Nuclear Regulatory Commission (NRC)-approved ARCON96 Code 
methodology for calculation of atmospheric dispersion coefficients (X/
Q) for the control room intakes and updated site meteorological data. 
Incorporation of this new methodology and updated meteorological data 
into the Oyster Creek licensing basis requires prior NRC review and 
approval.
    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.
    Substitution of the applicable regulatory limits for operator 
dose in lieu of the specific analyzed values in Technical 
Specification Section 3.17 Bases is [requested] to be consistent 
with the existing Technical Specification 4.17 Bases. The proposed 
change to utilize ARCON96 methodology and updated meteorological 
data results in control room operator doses that are less than the 
previously analyzed values, and, therefore, remain within the 
allowable limits. The probability of accidents is not affected by 
the computer codes used to assess the consequences of environmental 
releases. The use of updated, more extensive meteorological data 
provides a more accurate atmospheric dispersion coefficient (X/Q) 
value for the Turbine Building release to the control room 
ventilation system air intake.
    Therefore, the proposed change does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    2. The proposed change does not create the possibility of a new 
or different kind of accident from any accident previously 
evaluated.
    The change to incorporate ARCON96 methodology and updated 
meteorological data for assessing the control room operator doses 
from the releases of radioactive material following an accident has 
no [e]ffect on creating a new or different kind of accident. The 
proposed change does not affect the operation or functionality of 
any structures, systems 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. The proposed change does not involve a significant reduction 
in a margin of safety.
    The proposed change involves [a] revision to Technical 
Specification Section 3.17 Bases to substitute applicable regulatory 
limits in lieu of the specific analyzed dose values. The proposed 
change to incorporate ARCON96 methodology and updated meteorological 
data results in a more accurate determination of conservative 
control room air intake X/Q values and the resulting control room 
operator dose. ARCON96 is an NRC approved methodology which provides 
an acceptable level of conservatism. The updated meteorological data 
[are] obtained in accordance with NRC Regulatory Guide 1.23 
requirements.
    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: Kevin P. Gallen, Morgan, Lewis & Bockius, 
LLP, 1800 M Street, NW., Washington, DC 20036-5869.
    NRC Section Chief: Richard P. Correia, Acting.

Connecticut Yankee Atomic Power Company, Docket No. 50-213, Haddam 
Neck Plant, Middlesex County, Connecticut

    Date of amendment request: May 29, 2001.
    Description of amendment request: The proposed amendment would 
revise the Haddam Neck Plant Defueled Physical Security Plan referenced 
in License Condition 2.C(5). The proposed amendment reflects the intent 
of Connecticut Yankee Atomic Power Company (CYAPCO) to transfer all 
spent nuclear fuel and Greater than Class C waste from wet storage in 
the spent fuel pool to dry casks located at an on-site Independent 
Spent Fuel Storage Installation (ISFSI). CYAPCO proposed to make this 
ISFSI Security Plan an attachment to the existing Defueled Physical 
Security Plan. Adding the ISFSI Security Plan as an attachment to the 
Defueled Physical Security Plan would enable CYAPCO to implement the 
ISFSI Security Plan portion prior to commencement of fuel transfer 
operations.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), Connecticut Yankee 
Atomic Power Company (CYAPCO) has provided its analysis of the issue of 
no significant hazards consideration, which is presented below:

    The proposed amendment to the Security Plan provides the basis 
for establishing security functions necessary to implement 
appropriate security/safeguards measures for the CYAPCO Independent 
Spent Fuel Storage Installation (ISFSI). As such, the changes will 
not:
    1. Involve a significant increase in the probability or 
consequences of an accident previously evaluated.
    The proposed amendment to the Security Plan, which incorporates 
ISFSI security functions, does not reduce the ability of the 
Security organization to prevent radiological sabotage and, 
therefore, does not increase the probability or consequences of a 
radiological release previously evaluated. The proposed Security 
Plan changes will not affect any important to safety systems or 
components, their mode of operation or operating strategies. The 
proposed Security Plan changes have no affect on accident initiators 
or mitigation. Therefore, the proposed amendment to the Security 
Plan will not involve a significant increase in the probability or 
consequences of an accident previously evaluated.

[[Page 44164]]

    2. Create the possibility of a new or different kind of accident 
from any accident previously evaluated.
    The proposed amendment of the Security Plan incorporating ISFSI 
security functions does not affect the operation of systems 
important to safety. The Security Plan amendment does not affect any 
of the parameters or conditions that could contribute to the 
initiation of any accident. No new accident scenarios are created as 
a result of Security Plan changes requested to incorporate the ISFSI 
security functions. In addition, the design functions of equipment 
important to safety are not altered as a result of the proposed 
Security Plan changes. Therefore, the proposed Security Plan changes 
will not create the possibility of a new or different accident from 
any previously evaluated.
    3. Involve a significant reduction in the margin of safety. 
Implementation of the proposed amendment to the Security Plan 
incorporating ISFSI security functions will not reduce a margin of 
safety as detailed in the Technical Specifications as there are no 
Technical Specification requirements associated with the physical 
security system. Specifically, the proposed changes to the Security 
Plan do not represent a change in initial conditions, system 
response time, or in any other parameter affecting the course of an 
accident analysis supporting the Basis of any Technical 
Specification. The proposed amendment to the Security Plan does not 
reduce the effectiveness of any security/safeguards measures 
currently in place at CYAPCO. Therefore, the proposed Security Plan 
changes will not involve a significant reduction in the margin of 
safety.
    Based on the considerations noted above, it is concluded that 
the proposed changes will not endanger the public health and safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: Mr. Robert K. Gad, III, Ropes & Gray, One 
International Plaza, Boston, Massachusetts 02110-2624.
    NRC Section Chief: Stephen Dembek.

Connecticut Yankee Atomic Power Company, Docket No. 50-213, Haddam 
Neck Plant, Middlesex County, Connecticut

    Date of amendment request: May 30, 2001.
    Description of amendment request: The proposed amendment would 
correct terminology, clarify the specification for consistency with 
established programs and Standard Technical Specifications, (TSs) and 
reflect current plant conditions. The proposed changes also reflect the 
current organization titles. The licensee also proposed changes to the 
TS Bases for spent fuel pool water level and cooling.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), Connecticut Yankee 
Atomic Power Company (CYAPCO) has provided its analysis of the issue of 
no significant hazards consideration, which is presented below:

    CYAPCO has reviewed the proposed changes to the Operating 
License and the Technical Specifications in accordance with 10 CFR 
50.92 and concluded that the changes do not involve a significant 
hazards consideration (SHC). The basis for this conclusion is that 
the three criteria of 10 CFR 50.92(c) are not compromised. An 
evaluation against these standards is provided below as first a 
summary against the overall change, and also against each of the 
specific proposed changes.
    The proposed changes do not involve an SHC because the changes 
would not:
    1. Involve a significant increase in the probability or 
consequences of an accident previously evaluated.
    In the present plant configuration, the reactor-related 
accidents previously evaluated (i.e., LOCA, MSLB, etc.) are no 
longer possible. The accidents previously evaluated that are still 
applicable to the plant are fuel handling accidents and gaseous and 
liquid radioactive releases. The following events are presently 
considered as bounding of all other events:
     Fuel handling and cask drop accidents in the spent fuel 
building,
     Criticality in the spent fuel pool,
     Loss of spent fuel cooling,
     Resin fire (gaseous release), and
     Rupture of a tank containing radioactive liquid.
    There is no significant increase in the probability of a fuel 
handling accident since refueling operations have ceased, with a 
corresponding decrease in the frequency of fuel movement. The 
radiological consequences of a fuel handling accident, should one 
occur, decrease the longer the spent fuel is allowed to decay. The 
spent fuel inventory of radioactive iodine and noble gases have 
decayed more than 20 half-lives since shutdown and are no longer a 
release concern. The allowed weight over the spent fuel pool is 
still less than that previously approved. Therefore, there has been 
no increase in the probability or consequences of a fuel handling or 
cask drop accident.
    Criticality controls are imposed by specifications 3/4.9.13 and 
3/4.9.14. There have been no technical changes to these 
specifications. Therefore, there has been no increase in the 
probability or consequences of a criticality event.
    Spent fuel cooling is maintained by keeping the pool temperature 
below 150 deg.F. Should normal cooling be lost, the availability of 
an abundant supply of water ensures that sufficient time is 
available to restore cooling. This is controlled by specifications 
3/4.9.11 and 3/4.9.16. There have been no technical changes to these 
specifications. Therefore, there has been no increase in the 
probability or consequences of a loss of cooling event.
    The probability of a gaseous or liquid radioactive release is 
not changed by the proposed revisions. As the plant undergoes 
decommissioning, the previous limiting events are no longer 
applicable, and previous non-limiting events now become limiting. 
These new events have not changed from how they might have occurred 
in the past. The radiological consequences of a gaseous or liquid 
radioactive release are bounded by the fuel handling accident during 
defueled operation and a spent resin fire during processing of resin 
from the reactor coolant system decontamination. The rupture of a 
tank containing radioactive liquid was assessed and found to be 
bounded by these events. With the plant defueled and permanently 
shutdown, the demands on the radwaste systems are lessened since no 
new radioisotopes are being generated by irradiation or fission. 
Therefore, there is no increase in the probability or consequences 
of a gaseous or liquid radioactive release.
    The changes to conform to Section 6.0 to draft NUREG-1625 are of 
an administrative nature, and have been reviewed and found to be 
safe.
    2. Create the possibility of a new or different kind of accident 
from any accident previously evaluated.
    The proposed changes are generally of an administrative nature 
and do not have an effect on the physical plant. The events 
considered bound other potential events and are considered the 
limiting cases for potential gaseous or liquid releases to the 
environment.
    With the plant undergoing decommissioning, the types of 
accidents one might be concerned with involve criticality of the 
spent fuel, or draining of the spent fuel pool. None of the proposed 
changes affect the possibility of such an event. Also, none of the 
proposed changes could lead to a radiological release of a greater 
magnitude than for the events considered, such as might occur with 
the accumulation of a greater quantity of radioactive material in 
one location, or with damage to a greater number of fuel assemblies 
than considered in the fuel handling accident.
    The proposed changes do not affect systems, structures and 
components and have no adverse impact on the storage of fuel nor on 
the processing of radioactive wastes presently at the site. The 
present set of limiting events is a subset of events previously 
considered. Therefore these changes do not create the possibility of 
a new or different kind of accident from any accident previously 
considered.
    3. Involve a significant reduction in a margin of safety.
    The proposed changes do not reduce a margin of safety because 
there is no direct affect on any safety analysis assumptions. 
Changes to the Technical Specifications Bases reflect current plant 
conditions.
    Based on the above evaluation, CYAPCO concludes that the 
activities associated with the above described changes present no 
significant hazards consideration under the standards set forth in 
10 CFR 50.92 and accordingly, a finding by the NRC of no significant 
hazards consideration is justified.


[[Page 44165]]


    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: Mr. Robert K. Gad, III, Ropes & Gray, One 
International Plaza, Boston, Massachusetts 02110-2624.
    NRC Section Chief: Stephen Dembek.

Consolidated Edison Company of New York, Docket No. 50-247, Indian 
Point Nuclear Generating Unit No. 2, Westchester County, New York

    Date of amendment request: July 13, 2001.
    Description of amendment request: The proposed amendment would make 
a one-time change to Technical Specification Surveillance Requirement 
4.4.A.3 to revise the frequency for the containment integrate leak rate 
test (ILRT, Type A test) from at least once per 10 years to once per 15 
years. The change would apply only to the interval following the last 
Type A test that was satisfactorily performed in June 1991 at Indian 
Point Nuclear Generating Unit No. 2 (IP2).
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Operation of the facility in accordance with the proposed 
amendment would not involve a significant increase in the 
probability of occurrence or consequences of an accident previously 
evaluated.
    The change does not affect the ability of the containment to 
mitigate the consequences of an accident. The containment is not an 
accident initiating system or structure. The proposed one time 
change to Type A testing frequency has been determined to be 
adequate as documented in NUREG-1493 [``Performance-Based 
Containment Leak-Test Program,'' September 1995] which determined 
generically that very few potential containment leakage paths are 
not identified by Type B and C tests. The NUREG concluded that 
reducing the Type A (ILRT) testing frequency to one per twenty years 
was found to lead to an imperceptible increase in risk. This generic 
result has been confirmed for IP2 by a plant specific risk impact 
assessment. Past IP2 Type A tests show leakage to be below 
acceptance criteria, indicating a very leak-tight containment, 
without credit for the weld channel and penetration pressurization 
system (WC&PPS). Inspections required by other TS and by the ASME 
[American Society of Mechanical Engineers] code are performed in 
order to identify indications of containment degradation that could 
affect that leak tightness. The WC&PPS monitors the leak tightness 
of liner plate welds in the containment during plant operation as 
required by Technical Specifications. Type B and C testing required 
by TS will identify any containment opening such as valves that 
would otherwise be detected by the Type A tests. The frequency of 
performance of surveillance does not result in any hardware changes 
or the response of equipment in performing its specified function. 
Therefore, operation of the facility in accordance with the proposed 
amendment would not involve a significant increase in the 
probability or consequences of an accident previously evaluated.
    2. Operation of the facility in accordance with the proposed 
amendment would not create the possibility of a new or different 
kind of accident from any accident previously evaluated.
    The proposed change does not introduce nor increase the number 
of failure mechanisms of a new or different type of accident than 
those previously evaluated since there are no physical changes being 
made to the facility. Performance of the testing on the revised 
schedule will not have an adverse affect on the ability of the 
containment to perform its intended function. The proposed change 
does not degrade the reliability of systems, structures, or 
components or create a new accident initiator or precursor. No new 
failure modes are created. Therefore, the change does not create the 
possibility of a new or different kind of accident from any accident 
previously evaluated.
    3. Operation of the facility in accordance with the proposed 
amendment would not involve a significant reduction in the margin of 
safety.
    The one time change to the current frequency for Type A testing 
still provides adequate assurance of containment integrity. The 
NUREG-1493 generic study of the effects of extending containment 
leakage testing found that a 20-year extension in Type A leakage 
testing resulted in an imperceptible increase in risk to the public. 
NUREG -1493 found that, generically, the design containment leakage 
rate contributes about 0.1 percent to the individual risk and that 
the decrease in Type A testing frequency would have a minimal affect 
on this risk since 95% of the potential leakage paths are detected 
by Type B & C testing. The risk impact change of the test frequency 
was small. Online testing of the integrity of liner plate welds 
using the WC&PPS and regular inspections will further reduce the 
risk of a containment leakage path going undetected. There are no 
changes being made to TS safety limits or safety system settings 
that would adversely affect plant safety. Therefore, operation of 
the facility in accordance with the proposed amendment would 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: Brent L. Brandenburg, Esq., 4 Irving Place, 
New York, New York 10003.
    NRC Section Chief: Richard P. Correia, Acting.

Consolidated Edison Company of New York, Docket No. 50-247, Indian 
Point Nuclear Generating Unit No. 2, Westchester County, New York

    Date of amendment request: July 13, 2001.
    Description of amendment request: The proposed amendment would 
revise Technical Specification (TS) 6.12, ``High Radiation Area,'' to 
delete the administrative requirements for the control of access to 
high radiation areas. The control of access to these areas is assured 
by the licensee's radiation protection programs that comply with 10 CFR 
20.1601 by using the alternate methods in NRC Regulatory Guide 8.38, 
``Control of Access to High and Very High Radiation Areas in Nuclear 
Power Plants,'' June 1993.
    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 license amendment involve a significant 
increase in the probability or in the consequences of an accident 
previously evaluated?
    The proposed TS change is administrative in nature. It involves 
deleting specific requirements for complying with a subparagraph of 
10CFR20 for the purpose of controlling access to high radiation 
areas. Accident evaluations do not consider the effects of methods 
of controlling access to high radiation areas. The proposed changes 
do not result in a change to the design or operation of [...] any 
plant structure, system, or component. Therefore any assumptions of 
the operability or performance of any structure, system, or 
component in accident evaluations are unchanged.
    Therefore, there is no increase in the probability or in the 
consequences of an accident previously evaluated.
    2. Does the proposed amendment create the possibility of a new 
or different kind of accident from any accident previously 
evaluated?
    The proposed change is administrative in nature. The methods of 
controlling access to high radiation areas do not affect the design 
or operation of any plant structure, system, or component. 
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 amendment involve a significant reduction 
in a margin of safety?
    The proposed TS change is administrative in nature. It involves 
deleting specific

[[Page 44166]]

requirements for complying with a subparagraph of 10CFR20. However, 
effective compliance with 10CFR20 is mandated by the IP2 [Indian 
Point 2] Facility Operating License Section C. The effectiveness of 
Con Edison compliance with 10CFR20 is not adversely affected by this 
change. In addition, this change does not affect any design function 
for or the operation of any plant structure, system, or component.
    Therefore, the change does not affect any of the safety analyses 
or any margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 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: Brent L. Brandenburg, Esq., 4 Irving Place, 
New York, New York 10003.
    NRC Section Chief: Richard P. Correia, Acting.

Consolidated Edison Company of New York, Docket No. 50-247, Indian 
Point Nuclear Generating Unit No. 2, Westchester County, New York

    Date of amendment request: July 13, 2001.
    Description of amendment request: The proposed amendment would 
revise the Technical Specifications (TSs) to delete TS Tables 3.6-1, 
``Non-Automatic Containment Isolation Valves Open Continuously or 
Intermittently for Plant Operation,'' and 4.4-1, ``Containment 
Isolation Valves.'' The proposed amendment would also revise other TS 
sections that reference these tables. The removal of the tables is in 
accordance with the guidance in NRC Generic Letter (GL) 91-08, 
``Removal of Component Lists from Technical Specifications.''
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Does the proposed license amendment involve a significant 
increase in the probability or in the consequences of an accident 
previously evaluated?
    The proposed changes consist of removal of the containment 
isolation valve component lists from the IP2 [Indian Point 2] TS and 
corresponding editorial changes to support removal of the tables. 
The changes are being made in accordance with the guidance provided 
by the NRC in GL 91-08 and do not alter existing TS requirements or 
those components to which the TS requirements apply. The information 
contained in the Tables being removed is duplicated in the UFSAR 
[Updated Final Safety Analysis Report] and other appropriate plant 
procedures. Any subsequent changes regarding the individual 
components or their operation would be evaluated under the 
requirements of 10CFR50.59. The proposed changes do not involve a 
change to the design or operation of any plant structure, system, or 
component. Nor are the safety analyses affected as a result of the 
changes. Accordingly, the initiators of any accident as well as any 
structure, system or component relied upon for the mitigation of the 
accident are not affected by the proposed changes.
    Therefore, there is no increase in the probability or in the 
consequences of an accident previously evaluated.
    2. Does the proposed amendment create the possibility of a new 
or different kind of accident from any accident previously 
evaluated?
    The proposed changes do not involve a change to the design or 
operation of [...] any plant structure, system or component. The 
proposed changes involve the removal of component lists for 
containment isolation valves from the TS. In accordance with the 
guidance provided by GL 91-08, the conditions, actions, and 
requirements of the TS will apply to those valves that are 
classified as containment isolation valves by the plant licensing 
basis. This includes the testing of Containment Isolation Valves as 
required by 10CFR50 Appendix J and IP2 TS 4.4.D.1.a. Required 
specifications and requirements of the tables remain applicable. 
There are no changes to any parameter used in the accident analyses. 
Therefore, the proposed changes do not create the possibility of a 
new or different kind of accident for any previously evaluated.
    3. Does the proposed amendment involve a significant reduction 
in a margin of safety?
    The proposed changes are in accordance with the guidance 
provided by the NRC in GL 91-08 and NUREG-1431, Standard Technical 
Specifications. The changes will maintain current safety margins 
while reducing the regulatory and administrative burdens to both the 
NRC and IP2. The proposed changes will not result in changes to the 
design or operation of any plant system and do not involve changes 
to any margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 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: Brent L. Brandenburg, Esq., 4 Irving Place, 
New York, New York 10003.
    NRC Section Chief: Richard P. Correia, Acting.

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

    Date of amendment request: July 31, 2001.
    Description of amendment request: The proposed amendment requests 
U.S. Nuclear Regulatory Commission (NRC) approval of Big Rock Point 
Plant's (Big Rock Point) Security Plan, Suitability Training and 
Qualification Plan, and Safeguards Contingency Plan. These plans 
reflect the addition of provisions relating to the loading and storage 
of spent nuclear fuel at the Independent Spent Fuel Storage 
Installation (ISFSI).
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    The proposed change does not:
    1. Involve a significant increase in the probability or 
consequences of an accident previously evaluated.
    The currently approved and implemented Security Plan (Defueled 
Security Plan) is not being changed. The ISFSI Security Plan is 
being added to the scope of the overall security plan for the Big 
Rock Point site. The additions to the overall [Security] Plan have 
been evaluated in accordance with 10 CFR 50.54(p) and 10 CFR 
72.212(b)(4) and it has been determined that the implementation of 
the ISFSI Security Plan would not decrease the effectiveness of the 
Defueled Security Plan, the Defueled Suitability Training and 
Qualification Plan, or the first four categories of the Defueled 
Safeguards Contingency Plan.
    The ISFSI Security Program staffing will be parallel to the 
staffing requirements of the Defueled Security Plan, except that one 
Central Alarm Station [CAS] operator will be employed during the 
period when spent fuel is located in the spent fuel pool in the 
plant and also located in dry fuel storage at the ISFSI facility.
    The operational and physical venues of the Defueled Security 
Plan and the ISFSI Security Plan are separate and distinct, except 
for the utilization of a single CAS operator, and the lines of 
demarcation between the two plans [are] clearly defined and not 
overlapping. The implementation of any of the plans does not 
therefore degrade or inhibit the implementation of the other plan.
    The Defueled Suitability Training and Qualification Plan and the 
Defueled Safeguards Contingency Plan also have not been changed. A 
separate and parallel ISFSI Training and Qualification Plan and 
ISFSI Contingency Plan is included in the ISFSI Security Plan. The 
physical protection systems described in the ISFSI Plans are 
designed to protect against the loss of control of the facility that 
could be sufficient to cause a radiation exposure exceeding the dose 
as described in 10 CFR 72.106.
    Therefore, the ISFSI Plan revisions of the Big Rock Point Plant 
Security Plan, Suitability Training and Qualification Plan and the 
Safeguards Contingency Plan will not increase the probability or the 
consequences of an accident previously evaluated since the 
previously approved Defueled Suitability Training and Qualification 
Plan and the Safeguards Contingency Plan remain unchanged.

[[Page 44167]]

    2. Create the possibility of a new or different kind of accident 
from any accident previously evaluated.
    The ISFSI Security Plan has no impact on the existing Defueled 
Security Plan since they operate in different physical and licensing 
venues. The accidents considered for the Spent Fuel Pool, the venue 
of the Defueled Security Plan, are described in the Big Rock Point 
Updated Final Hazards Summary Report. The accidents considered for 
the ISFSI are contained in the FuelSolutions Final Safety Analysis 
Reports [FSARs] for the W150 Storage Cask, W100 Transfer Cask and 
the W74 Canister under Docket No. 72-1026.
    The ISFSI Security Plan has been crafted to meet or exceed all 
of the assumptions of the FuelSolutions FSARs concerning accident 
analyses and the plan meets or exceeds all of the applicable 
requirements of 10 CFR 73.55 with approved exceptions or approved 
alternative measures. The physical protection systems described in 
the ISFSI Security Plan are designed to protect against the loss of 
control of the facility that could be sufficient to cause a 
radiation exposure exceeding the dose as described in 10 CFR 72.106.
    The proposed action does not affect plant systems, structures or 
components within the venue of the existing Defueled Security Plan. 
The ISFSI additions to the Security Plan, Suitability Training and 
Qualification Plan and the Safeguards Contingency Plan do not create 
the possibility of a new or different kind of accident from any 
accident previously evaluated since the previously approved Defueled 
Security Plan, [Suitability] Training and Qualification Plan and 
Safeguards Contingency Plan remain the same.
    3. Involve a significant reduction in a margin of safety.
    The addition of a separate, parallel ISFSI Security Plan, 
Suitability Training and Qualification Plan, and Safeguards 
Contingency Plan does not alter or reduce the effectiveness of the 
previously approved Defueled Security Plan. The physical protection 
systems described in the ISFSI Plan are designed to protect against 
the loss of control of the facility that could be sufficient to 
cause a radiation exposure exceeding the dose as described in 10 CFR 
72.106. Therefore, the margin of safety will not be reduced as a 
result of the ISFSI addition to the Security Plan, or an ISFSI 
specific addition of a Suitability Training and Qualification Plan 
or an ISFSI specific addition of a Safeguards Contingency Plan.

    The NRC staff has reviewed the licensee's significant hazards 
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: David A. Mikelonis, Esquire, Consumers 
Energy Company, 212 West Michigan Avenue, Jackson, Michigan 49201.
    NRC Section Chief: Robert A. Gramm.

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

    Date of amendment request: July 31, 2001.
    Description of amendment request: The proposed amendment would 
revise and relocate the inservice testing portion of Technical 
Specification (TS) 5.0.5 to TS 6.5.8, and eliminate the inservice 
inspection portion of TS 4.0.5. In addition, other sections of the TSs 
that reference TS 4.0.5 would be revised to be consistent with the 
revisions discussed above.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Will operation of the facility in accordance with this 
proposed change involve a significant increase in the probability or 
consequence of an accident previously evaluated? 
    The proposed change relocates the requirements to test and 
inspect ASME [American Society of Mechanical Engineers] Code [Boiler 
and Pressure Vessel Code] Class 1, 2, and 3 components from TS 4.0.5 
to the administrative section of the TSs and includes modifications 
to the wording to make it consistent with NUREG-1432 [Standard 
Technical Specifications, Combustion Engineering Plants]. This 
change will not reduce the current testing and inspection 
requirements. The performance of a code inservice test is not an 
accident initiator. The proposed change for removing the statement 
for NRC [Nuclear Regulatory Commission] granting written relief for 
[from the] ASME Code does not involve a significant increase in the 
probability or consequences of an accident. Verbally issuing relief 
to the ASME Code by the NRC does not reduce assurance of the health 
and safety of the public since the NRC still reviews the basis for 
the relief on its technical merit and the NRC Staff still obtains 
management approval prior to granting the relief.
    Therefore, this change does not involve a significant increase 
in the probability or consequences of any accident previously 
evaluated.
    2. Will operation of the facility in accordance with this 
proposed change create the possibility of a new or different kind of 
accident from any accident previously evaluated? 
    [The citation at] 10 CFR 50.55a, ``Codes and Standards'' governs 
inservice testing and inspection requirements. The inspection 
requirements contained in 10 CFR 50.55a paragraph (g) are duplicated 
in TS 4.0.5. This duplication is unnecessary and therefore, the 
wording related to the inspection requirements will be deleted in 
the proposed change. No actual change to the inspection or testing 
activities are proposed as the requirements in 10 CFR 50.55a 
continue to govern these. Therefore, the testing and inspection 
requirements will remain the same as those presently required. The 
proposed change is administrative in nature in that it relocates 
testing requirements from one section of the TSs to another and 
modifies the wording to be consistent with NUREG-1432. The removal 
of the requirement to obtain written relief from the NRCc staff will 
not create the possibility of any new or different types of 
accidents. Staff review is still required prior to granting the 
relief.
    Therefore, this change does not create the possibility of a new 
or different kind of accident from any previously evaluated.
    3. Will operation of the facility in accordance with this 
proposed change involve a significant reduction in a margin of 
safety?
    The testing and inspection requirements contained in TS 4.0.5 
are governed by 10 CFR 50.55a, ``Codes and Standards.'' The 10 CFR 
requirements to perform the ASME code testing and inspections will 
not be reduced by the proposed change. The inspection and tests will 
continue to be performed as they are currently. This change moves 
the present requirements from one section of the TSs to another.
    Therefore, this change does not involve a significant reduction 
in the margin of safety.

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

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

    Date of amendment request: July 31, 2001.
    Description of amendment request: The proposed amendment would 
revise the technical specifications (TSs) to allow an extension of the 
three-year inspection interval of the reactor coolant pump flywheel 
voumetric examination to ten years. In addition, the requirement 
discussed above would be moved to the administrative controls section 
of the TSs.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Will operation of the facility in accordance with this 
proposed change involve a significant increase in the

[[Page 44168]]

probability or consequences of an accident previously evaluated?
    Inspections of the reactor coolant pump (RCP) flywheels are 
conducted to detect a flaw in the flywheel prior to it becoming a 
missile that could damage other portions of the facility. The 
fracture mechanics analyses conducted as part fo the NRC [U.S. 
Nuclear Regulatory Commission] approved Topical Report SIR-94-080-A, 
Rev. 1, shows that a conservatively sized pre-existing crack will 
not grow to a flaw size necessary to create flywheel missiles with 
the current or extended life of the facility. This analysis 
conservatively assumes minimum material properties, maximum flywheel 
speed, location of the flaw in the highest stress area, and a number 
of startup and shutdown cycles higher than expected. Since a 
conservative flaw in the RCP flywheels will not grow to the 
allowable flaw size under large break LOCA [loss-of-coolant 
accident] conditions over the life of the plant, reducing the 
inspection frequency of the flywheels will not significantly 
increase the probability or consequences of an accident previously 
evaluated.
    The change to move the survillance requirements for the RCP 
flywheels to the programs section of the technical specifications is 
administrative and has no impact on probability or consequences of 
an accident.
    Therefore, this change does not involve a significant increase 
in the probability or consequences of any accident previously 
evaluated.
    2. Will operation of the facility in accordance with this 
proposed change create the possibility of a new or different kind of 
accident from any accident previously evaluated? 
    The proposed changes will not alter the plant configuration or 
require any new or usual operator actions. They do not alter the way 
any structure, system, or component functions and do not alter the 
manner in which the plant is operated. These changes do not 
introduce any new failure modes.
    Therefore, this change does not create the possibility of a new 
or different kind of accident from any previously evaluated.
    3. Will operation of the facility in accordance with this 
proposed change involve a significant reduction in a margin of 
safety?
    The ANO-2 [Arkansas Nuclear One, Unit 2] flywheels are made of 
either ASTM [American Society for Testing and Materials] A-533, 
Grade B, Class 1 or A-508, Class 5 steel plate material, which is 
pressure vessel quality steel. These materials have high tensile and 
yield strength qualities. The operating temperature of the flywheel 
is not less than 100  deg.F and the RTNDT value is below 
+10  deg.F. Therefore, there is at least 90  deg.F margin below the 
lowest temperature at which operating speed is achieved which is in 
accordance with Regulatory Guide 1.14, Rev. 1, ``Reactor Coolant 
Pump Flywheel Integrity.'' The fracture mechanics analyses conducted 
to support the extension of the inspection frequency from 3 to 10 
years was performed with substantial conservatism built into the 
analyses. Even with this analytical conservatism, the results 
indicate that the flywheels have sufficient margin that there is 
only a negligible potential for gross failure of the flywheels.
    Therefore, this change does not involve a significant reduction 
in the margin of safety.

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

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

    Date of amendment request: July 10, 2001.
    Description of amendment request: Technical Specification (TS) 
Surveillance Requirement (SR) 4.8.1.1.2.e requires certain emergency 
diesel generator (EDG) surveillances be performed during shutdown. The 
proposed change will modify this SR to allow performance of specific 
surveillances during any mode of plant operation. This will provide 
flexibility in the scheduling of testing activities consistent with 
online maintenance activities and improve EDG availability during plant 
shutdown periods.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Will operation of the facility in accordance with this 
proposed change involve a significant increase in the probability or 
consequences of an accident previously evaluated?
    The EDG is designed to operate in the event of a loss of offsite 
power or upon receipt of a SIAS [Safety Injection Actuation Signal]. 
No modifications or design changes are proposed to the EDG in 
conjunction with this proposed TS change. Periodic testing of the 
EDG starting circuitry, lockout relays, capability to reject a load 
and maintain voltage and frequency, ability to run for 24-hours, and 
various other tests prove the EDG is qualified to function upon 
demand. The changes proposed will allow several SRs to be performed 
in modes other than only during shutdown. A review of each of these 
has been performed. The system alignment needed to achieve these 
tests is the same whether the test is performed during shutdown or 
during power operations. When performing SR 4.8.1.1.2.e.1, 2, 4, 6, 
and 9, the EDG is operable and capable of performing its intended 
function, if called upon. When performing SR 4.8.1.1.2.e.10 and 12, 
the EDG that is being tested is inoperable for less than two hours, 
which is well within the allowable outage time. While performing 
these SRs, operations personnel are available to quickly respond to 
align the EDG as needed for an unexpected event. Additionally, the 
equipment covered by these specifications are not accident 
initiators and can not cause an accident.
    Therefore, this change does not involve a significant increase 
in the probability or consequences of any accident previously 
evaluated.
    2. Will operation of the facility in accordance with this 
proposed change create the possibility of a new or different kind of 
accident from any accident previously evaluated?
    The proposed change does not impact any system or component 
which could cause an accident. The proposed change will not alter 
the plant configuration (no system design modifications are 
required) or require any unusual operator actions. The proposed 
change will not alter the way any structure, system, or component 
functions, and will not significantly alter the manner in which the 
plant is operated. A review of the proposed change indicates that 
the required testing will be performed in a similar configuration 
and the interrelationship with other components is the same whether 
the testing is performed at power or during shutdown. The proposed 
change does not introduce any new failure modes. Additionally, the 
response of the plant and the operators following an accident will 
not be significantly different as a result of these changes.
    Therefore, this change does not create the possibility of a new 
or different kind of accident from any previously evaluated.
    3. Will operation of the facility in accordance with this 
proposed change involve a significant reduction in a margin of 
safety?
    The proposed TS change is associated with the surveillance 
requirements for the EDGs. The proposed change allows certain EDG 
surveillance requirements to be performed when the plant is at power 
rather than when shutdown. When performing SR 4.8.1.1.2.e.1, 2, 4, 
6, and 9, the EDG is operable and capable of performing its intended 
function, if called upon. When performing SR 4.8.1.1.2.e.10 and 12, 
the EDG that is being tested is inoperable for less than two hours, 
which is well within the allowable outage time. The proposed change 
will have no adverse effect on plant operation or equipment 
important to safety. The plant response to the design basis 
accidents will not change and the accident mitigation equipment will 
continue to function as assumed in the design basis accident 
analysis.
    Therefore, this change does not involve a significant reduction 
in the margin of safety.

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

[[Page 44169]]

amendment request involves no significant hazards consideration.
    Attorney for licensee: N. S. Reynolds, Esquire, Winston & Strawn 
1400 L Street NW., Washington, DC 20005-3502.
    NRC Section Chief: Robert A. Gramm.

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

    Date of amendment request: July 23, 2001.
    Description of amendment request: This submittal requests a change 
to administrative Technical Specification (TS) 6.15. The change 
postpones the next Type A test performed after May 12, 1991, to no 
later than May 11, 2006, which basically results in an extended 
interval of 15 years for performance of the next Integrated Leak Rate 
Test (ILRT).
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Will operation of the facility in accordance with this 
proposed change involve a significant increase in the probability or 
consequences of an accident previously evaluated?
    [Appendix J of 10 CFR [Part] 50], was amended to incorporate 
provisions for performance-based testing in 1995. The proposed 
amendment to Technical Specification (TS) 6.15 adds a one-time 
extension to the current interval for Type A testing (i.e., the 
integrated leak rate test). The current interval of ten years, based 
on past performance, would be extended on a one-time basis to 15-
years from the date of the last test. The proposed extension to the 
Type A test cannot increase the probability of an accident since 
there are no design or operating changes involved and the test is 
not an accident initiator. The proposed extension of the test 
interval does not involve a significant increase in the consequences 
since research documented in NUREG-1493 has found that, generically, 
fewer than 3% of the potential containment leak paths are not 
identified by Type B and C testing. Waterford 3 [Waterford Steam 
Electric Station, Unit 3], through testing and containment 
inspections, also provides a high degree of assurance that the 
containment will not degrade in a manner detectable only by a Type A 
test. Inspections required by the Maintenance Rule (10 CFR 50.65) 
and by the American Society of Mechanical Engineers Boiler and 
Pressure Vessel Code are performed to identify containment 
degradation that could affect leaktightness.
    Therefore, this change does not involve a significant increase 
in the probability or consequences of any accident previously 
evaluated.
    2. Will operation of the facility in accordance with this 
proposed change create the possibility of a new or different kind of 
accident from any accident previously evaluated?
    The proposed extension to the interval for the Type A test does 
not involve any design or operational changes that could lead to a 
new or different kind of accident from any accidents previously 
evaluated. The test itself is not changing and is just to be 
performed after a longer interval. 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 previously evaluated.
    3. Will operation of the facility in accordance with this 
proposed change involve a significant reduction in a margin of 
safety?
    The generic study of the increase in the Type A test interval, 
NUREG-1493, concluded there is an imperceptible increase in the 
plant risk associated with extending the test interval out to twenty 
years. Further, the extended test interval would have a minimal 
effect on this risk since Type B and C testing detect 97% of 
potential leakage paths. For the requested change in the Waterford 3 
ILRT interval, it was determined that the risk contribution of 
leakage will increase 0.17%. This change is considered very small 
and does not represent a significant reduction in the margin of 
safety.
    Therefore, this change does not involve a significant reduction 
in the margin of safety.

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

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

    Date of amendment request: July 23, 2001.
    Description of amendment request: The proposed change is to delete 
Technical Specifications (TS) 3.9.12, ``Fuel Handling Building 
Ventilation System,'' and TS 3.3.3.1 requirements for the Fuel Storage 
Pool area radiation monitors.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Does Not Involve a Significant Increase in the Probability or 
Consequences of an Accident Previously Evaluated. 
    The FHBVS [Fuel Handling Building Ventilation System] is not 
involved in the initiation of any accidents. The system is not 
credited with providing any supplemental filtration of any releases 
from an accident occurring in the containment building. It was 
designed to provide an accident mitigation function by isolating the 
system and filtering the radioiodines that may be released from a 
damaged fuel assembly in the event of a Fuel Handling Accident 
(FHA). The charcoal adsorber was the primary component that 
supported this filtration function. However, based on a revised 
analysis of the dose consequences of the FHA, it has been 
demonstrated that doses due to the FHA, to both the public and the 
control room operator, remain well within regulatory acceptance 
limits even assuming no credit for either isolation or filtration. 
The charcoal filtration function is not required and need not be 
tested. Thus, there is no required safety function in the event of a 
fuel handling accident provided by either the ventilation system or 
the area radiation monitor.
    Therefore, the proposed changes do not involve a significant 
increase in the probability or consequences of any accident 
previously evaluated.
    2. Does Not Create the Possibility of a New or Different Kind of 
Accident from any Previously Evaluated. 
    The FHBVS is not involved in the initiation of any accidents. It 
was designed to provide an accident mitigation function by isolating 
the system and filtering the radioiodines that may be released from 
a damaged fuel assembly in the event of a Fuel Handling Accident 
(FHA). Recent analyses show that the isolation and filtration 
functions are no longer required. The charcoal adsorber can not 
influence any accident initiators. Further, it has been demonstrated 
that the deletion of the technical specification requirements does 
not impact this conclusion and does not influence any new potential 
accident scenarios in any way.
    Therefore, the proposed changes do not create the possibility of 
a new or different kind of accident from any previously evaluated.
    3. Does Not Involve a Significant Reduction in the Margin of 
Safety.
    The FHBVS was designed to provide an accident mitigation 
function by filtering the radioiodines that may be released from a 
damaged fuel assembly in the event of a Fuel Handling Accident 
(FHA). Charcoal adsorbers had been provided for this function. 
Recent analysis of the FHA in the Fuel Handling Building demonstrate 
that the isolation function and the charcoal adsorber are not 
required to satisfy the margin of safety provided by the Technical 
Specification requirements. Based on a revision to the dose 
consequence analysis of the FHA, it has been determined that doses 
remain well within the regulatory allowable for exposure even 
assuming no credit for charcoal filtration. The margin of safety, as 
defined by SRP [Standard Review Plan] 15.7.4, Rev 1, and General 
Design Criterion 19, has not been significantly reduced.
    Therefore, the proposed changes do not significantly reduce the 
margin [of] safety.


[[Page 44170]]


    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 50-454 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]

[Docket Nos. 50-352 and 50-353, Limerick Generating Station, Units 1 
and 2, Montgomery County, Pennsylvania]

[Docket Nos. STN 50-277 and STN 50-278, Peach Bottom Atomic Power 
Station, Unit Nos. 2 and 3, York County, Pennsylvania]

[Docket Nos. 50-295 and 50-304, Zion Nuclear Power Station, Units 1 and 
2, Lake County, Illinois]

    Date of amendment request: July 9, 2001.
    Description of amendment request: The proposed amendments would 
incorporate Technical Specifications (TS) changes that are being made 
to provide consistency with the changes to 10 CFR 50.59, ``Changes, 
tests, and experiments,'' as published in the Federal Register (FR) 
Volume 64, beginning on page 53582 (i.e., 64 FR 53582), dated October 
4, 1999. Specifically, the changes replace the terms ``safety 
evaluation'' with ``10 CFR 50.59 evaluation'' and ``unreviewed safety 
question'' with ``requires NRC approval pursuant to 10 CFR 50.59.''
    In addition, Exelon proposes to change a condition 3.B of Operating 
License Nos. DPR-44 and DPR-56 for the Peach Bottom Atomic Power 
Station, Units 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. Does the change involve a significant increase in the 
probability or consequences of an accident previously evaluated?
    The proposed changes reflect revision to 10 CFR 50.59, 
``Changes, tests, and experiments,'' issued as a Final Rule on 
October 4, 1999, and do not impact the operation of any system or 
component assumed in any accident analysis. The proposed changes do 
not change the requirement to perform a 10 CFR 50.59 review when 
required by the Technical Specifications Administrative Controls or 
by a license condition. Due to the administrative nature of these 
proposed changes there will be no direct impact on the consequences 
of any accident previously evaluated. Therefore, these proposed 
changes do not involve a significant increase in the probability or 
consequences of an accident previously evaluated.
    2. Does the change create the possibility of a new or different 
kind of accident from any accident previously evaluated?
    The proposed changes are administrative in nature and do not 
involve a change to the plant design or operation. No new or 
different types of equipment will be installed as a result of these 
changes. The proposed changes make the language in the Technical 
Specifications Administrative Controls and a license condition 
conform to the revised 10 CFR 50.59 rule, dated October 4, 1999. No 
new accident modes or equipment failure modes are created by these 
proposed changes. Therefore, these proposed changes do not create 
the possibility of a new or different kind of accident from any 
accident previously evaluated.
    3. Does the change involve a significant reduction in a margin 
of safety?
    The proposed changes do not have a direct effect on any safety 
analysis assumptions. The proposed changes are administrative in 
nature and make the Technical Specifications Administrative Controls 
and a license condition language conform to the revised 10 CFR 50.59 
rule, dated October 4, 1999.
    Changes to the facility that result in meeting the criteria of 
10 CFR 50.59 will still require NRC approval pursuant to 10 CFR 
50.59. 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 
requested amendments involve no significant hazards consideration.
    Attorney for licensee: Mr. Edward J. Cullen, Vice President, 
General Counsel, Exelon Generation Company, LLC, 300 Exelon Way, 
Kennett Square, PA 19348.
    NRC Section Chief: Anthony J. Mendiola.

Exelon Generation Company, LLC, Docket Nos. 50-237 and 50-249, 
Dresden Nuclear Power Station, Units 2 and 3, Grundy County, 
Illinois

    Date of amendment request: September 29, 2000, as supplemented by 
letter dated March 1, 2001 (previously noticed in the Federal Register 
on December 27, 2000, 65 FR 81908).
    Description of amendment request: The March 1, 2001, supplement 
requests an amendment to revise the technical specifications to (1) 
increase the number of required automatic depressurization system (ADS) 
valves from four to five, (2) add surveillance requirements for the 
operability of the additional ADS valve, (3) change a surveillance 
requirement to verify the flow rate of two low-pressure coolant 
injection pumps instead of three pumps, consistent with the accident 
analyses, and (4) remove an allowance to continue operating for 72 
hours if certain combinations of emergency core cooling system systems 
are inoperable. These are additional changes to those that were 
requested in the September 29, 2000, application. The changes to the 
technical specifications support a change in fuel vendors from Siemens 
Power Corporation to General Electric (GE) and a transition to the use 
of GE-14 fuel.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. The proposed TS changes do not involve a significant increase 
in the probability or consequences of an accident previously 
evaluated.
    The proposed changes do not affect the initiators of analyzed 
events or the assumed mitigation of accident or transient events. 
Analyzed events are initiated by the failure of plant structures, 
systems or components. The proposed changes do not impact the 
condition or performance of these structures, systems or components. 
Consequences of analyzed events are the result of the plant being 
operated within assumed parameters at the onset of any events. The 
evaluations supporting the transition to GE fuel revealed that the 
current Technical Specification (TS) Limiting Condition for 
Operation (LCO) and conditions must be revised to place additional 
limitations on equipment to ensure that the plant is operated within 
the assumptions of the safety analyses. With the additional 
limitations, the analyses demonstrate that all of the acceptance 
criteria continue to be met. As a result, the changes do not involve 
a significant increase in the probability of consequences of an 
accident previously evaluated.
    2. The proposed TS changes do not create the possibility of a 
new or different kind of accident from any accident previously 
evaluated.
    The proposed changes do not involve a physical alteration of the 
facility or change the normal facility operation. No new or 
different equipment is being installed and no installed equipment is 
being removed. There is no alteration to the parameters within which 
the plant is normally operated or in the setpoints that initiate 
protective or mitigative actions. Consequently, no new failure modes 
are introduced and the changes therefore do not increase the 
possibility of a new or different kind of accident from any 
previously evaluated.
    3. The proposed TS changes do not involve a significant 
reduction in a margin of safety.

[[Page 44171]]

    Margin of safety is established through the design of the plant 
structures, systems and components, the parameters within which the 
plant is operated, and the establishment of setpoints for the 
actuation of equipment relied upon to respond to an event. The 
proposed changes do not impact the condition or performance of 
structures, systems or components relied upon for accident 
mitigation or any safety analysis assumptions. The changes reflect a 
reduction in redundancy in the capability of the Automatic 
Depressurization System (ADS)[.] However, the proposed changes 
impose more restrictive requirements on operation to ensure that all 
of the accident analyses continue to meet acceptance criteria. 
Therefore the proposed changes do not involve a significant 
reduction in 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. Edward J. Cullen, Vice President, 
General Counsel, Exelon Generation Company, LLC, 300 Exelon Way, 
Kennett Square, PA 19348.
    NRC Section Chief: Anthony J. Mendiola.

Exelon Energy Company, LLC, Docket Nos. 50-352 and 50-353, Limerick 
Generating Station, Units 1 and 2, Montgomery County, Pennsylvania

    Date of amendment request: May 23, 2001.
    Description of amendment request: Exelon proposed changes that 
would delete Action Statement b. associated with Limiting Condition for 
Operation 3.4.2 regarding operations with a stuck open safety/relief 
valve.
    Basis for proposed no significant hazards consideration 
determination: As required by Section 50.91 (a) of Title 10 of the Code 
of Federal Regulations (CFR), the licensee has provided its analysis of 
the issue of no significant hazards consideration. The NRC staff has 
review the licensee's analysis against the standards of 10 CFR 
50.92(c). The NRC staff review is presented below:

    1. The proposed TS change does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    The proposed TS change deletes Action Statement b. associated 
with Limiting Condition for Operation (LCO) 3.4.2 concerning plant 
operations with stuck open safety/relief valves. The operator action 
described in the LCO represents detailed methods of responding to an 
event, and therefore, if eliminated, would not result in increasing 
the probability of the event nor act as an additional initiator of 
an event. Therefore, this action can be eliminated, and will not 
involve a significant increase in the probability of an accident 
previously evaluated.
    As discussed in Section 15.1.4 (``Inadvertent Main Steam Relief 
Valve Opening''), of the Limerick Generating Station, Units 1 and 2, 
Updated Final Safety Analysis Report (UFSAR), a main steam relief 
valve is postulated to inadvertently open. While this transient does 
not result in fuel failure, it does result in the discharge of 
normal coolant activity to the suppression pool via relief valve 
operation. Because this activity is contained within the primary 
containment, there is no exposure to operating personnel or 
uncontrolled release of radioactivity to the environment. Therefore, 
this change does not increase the consequences of an accident 
previously evaluated.
    The requirement to scram the reactor within 2 minutes of 
identifying a stuck open safety/relief valve was not incorporated 
into the BWR Standard Technical Specifications (NUREG-1433, 
``Standard Technical Specifications General Electric Plants, BWR/
4,'' Revision 1, dated April 1995).
    2. The proposed TS change does not create the possibility of a 
new or different kind of accident from any accident previously 
evaluated.
    The proposed TS change deletes Action Statement b. associated 
with Limiting Condition for Operation 3.4.2 concerning safety/relief 
valves. This change does not change the design or configuration of 
the plant. The safety/relief valves are accident mitigators. Section 
15.1.4 (``Inadvertent Main Steam Relief Valve Opening''), of the 
Limerick Generating Station, Units 1 and 2, Updated Final Safety 
Analysis Report (UFSAR), postulates an inadvertent opening of a main 
steam relief valve. This change will not alter the assumptions or 
results of this analysis. No new operation or failure modes are 
created, nor is a system-level failure mode created that is 
different than those that already exist. Therefore, this change does 
not create the possibility of a new or different kind of accident 
from any accident previously evaluated.

    3. The proposed TS change does not involve a significant 
reduction in a margin of safety.
    The proposed change does not involve a significant reduction in 
a margin of safety, nor does it affect any analytical limits. There 
are no changes to accident or transient core thermal hydraulic 
conditions, or fuel or reactor coolant boundary design limits, as a 
result of the proposed change. The proposed change will not alter 
the assumptions or results of the analysis contained in the UFSAR. 
Therefore, the proposed change does not involve a significant 
reduction in a margin of safety.

    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. Edward Cullen, Vice President & General 
Counsel, Exelon Generation Company, LLC, 300 Exelon Way, Kennett 
Square, PA 19348.
    NRC Section Chief: James W. Clifford.

Exelon Energy Company, LLC, Docket Nos. 50-352 and 50-353, Limerick 
Generating Station, Units 1 and 2, Montgomery County, Pennsylvania

    Date of amendment request: June 26, 2001.
     Description of amendment request: The proposed amendment would 
revise the Limerick Generating Station (LGS) Units 1 and 2, Technical 
Specifications (TSs) 3/4.3.3, Actions 36 and 37 of Table 3.3.3-1, and 
the associated TS Bases. The change to Action 36 clarifies equipment 
affected by inoperable components. The change to Action 37 takes 
advantage of the inherent overlap of the degraded voltage relays' 
characteristics such that inoperable relays that define a channel can 
be taken out of service without placing its associated source breaker 
in the trip position.
    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. The NRC staff has reviewed the licensee's analysis 
against the standards of 10 CFR 50.92(c). The NRC staff's review is 
presented below:

    1. The proposed TS amendment does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    The changes to Action 36 detail what equipment is impacted by an 
inoperable bus under voltage relay. Making these changes assures 
that the appropriate equipment is considered inoperable. Identifying 
the impacted equipment for an inoperable under voltage relay does 
not increase the probability or the consequences of an accident 
previously evaluated.
    Action 37 presently requires placing an inoperable channel 
(relay) in the tripped condition which results in making the 
associated offsite source circuit breaker unavailable to that bus. 
Changing Action 37 to place a relay in the bypass condition, rather 
than the tripped condition, permits the offsite source of power to 
still be available to the bus in the event of an inoperable degraded 
voltage relay. The change to Action 37 takes advantage of the 
inherent overlap of the degraded voltage relays' characteristics 
such that inoperable relays that define the channel can be taken out 
of service without placing its associated source breaker in the trip 
position. The change to Action 37 does not adversely impact the 
availability or reliability of the offsite power system. Therefore, 
the proposed changes to Action 37 do not increase the probability or

[[Page 44172]]

consequences of an accident previously evaluated.
    2. The proposed TS amendment does not create the possibility of 
a new or different kind of accident from any accident previously 
evaluated.
    The changes to Action 36 detail what equipment is impacted by an 
inoperable bus under voltage relay and does not involve physical 
changes to the plant that would create the possibility of a new or 
different kind of accident from any accident previously evaluated.
    The proposed changes to Action 37 take advantage of the overlap 
of the degraded voltage relays by providing actions to be taken when 
an individual relay within a channel is inoperable. Changing Action 
37 does not make any physical changes to the plant. Therefore, the 
changes to Action 37 do not create the possibility of a new or 
different kind of accident from any accident previously evaluated.
    3. The proposed TS amendment does not involve a significant 
reduction in a margin of safety.
    The changes to Action 36 and Action 37 do not affect the 
availability or operation of mitigation systems. Therefore, there is 
no impact on event analysis that would affect the resultant analyses 
or reduce a margin of safety.
    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. Edward Cullen, Vice President & General 
Counsel, Exelon Generation Company, LLC, 300 Exelon Way, Kennett 
Square, PA 19348.
    NRC Section Chief: James W. Clifford.

Exelon Generation Company, LLC, Docket Nos. 50-254 and 50-265, Quad 
Cities Nuclear Power Station, Units 1 and 2, Rock Island County, 
Illinois

    Date of amendment request: September 29, 2000, as supplemented by 
letter dated March 1, 2001 (previously noticed in the Federal Register 
on December 27, 2000, 65 FR 81912).
    Description of amendment request: The March 1, 2001, supplement 
requests an amendment to revise the technical specifications to 
increase the number of required automatic depressurization system (ADS) 
valves from four to five, to add surveillance requirements for the 
operability of the additional ADS valve, and to remove an allowance to 
continue operating for 72 hours if certain combinations of emergency 
core cooling systems are inoperable. These are additional changes to 
those that were requested in the September 29, 2000, application. The 
changes to the technical specifications support a change in fuel 
vendors from Siemens Power Corporation to General Electric (GE) and a 
transition to the use of GE-14 fuel.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. The proposed TS changes do not involve a significant increase 
in the probability or consequences of an accident previously 
evaluated.
    The proposed changes involve more restrictive limitations on 
operation. These changes do not affect the initiators of analyzed 
events or the assumed mitigation of accident or transient events. 
Analyzed events are initiated by the failure of plant structures, 
systems or components. The proposed changes do not impact the 
condition or performance of these structures, systems or components. 
Consequences of analyzed events are the result of the plant being 
operated within assumed parameters at the onset of any events. The 
evaluations supporting the transition to GE fuel revealed that the 
current Technical Specification (TS) Limiting Condition for 
Operation (LCO) and conditions must be revised to place additional 
limitations on equipment to ensure that the plant is operated within 
the assumptions of the safety analyses. With the additional 
limitations, the analyses demonstrate that all of the acceptance 
criteria continue to be met. As a result, the changes do not involve 
a significant increase in the probability of consequences of an 
accident previously evaluated.
    2. The proposed TS changes do not create the possibility of a 
new or different kind of accident from any accident previously 
evaluated.
    The proposed changes do not involve a physical alteration of the 
facility or change the manner in which the facility is operated. No 
new or different equipment is being installed and no installed 
equipment is being removed. There is no alteration to the parameters 
within which the plant is normally operated or in the setpoints that 
initiate protective or mitigative actions. Consequently, no new 
failure modes are introduced and the changes therefore do not 
increase the possibility of a new or different kind of accident from 
any previously evaluated.
    3. The proposed TS changes do not involve a significant 
reduction in a margin of safety.
    Margin of safety is established through the design of the plant 
structures, systems and components, the parameters within which the 
plant is operated, and the establishment of setpoints for the 
actuation of equipment relied upon to respond to an event. The 
proposed changes do not impact the condition or performance of 
structures, systems or components relied upon for accident 
mitigation or any safety analysis assumptions. The changes reflect a 
reduction in redundancy in the capability of the Automatic 
Depressurization System (ADS). However, the proposed changes impose 
more restrictive requirements on operation to ensure that all of the 
accident analyses continue to meet acceptance criteria. Therefore 
the proposed changes do not involve a significant reduction in 
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. Edward J. Cullen, Vice President, 
General Counsel, Exelon Generation Company, LLC, 300 Exelon Way, 
Kennett Square, PA 19348.
    NRC Section Chief: Anthony J. Mendiola.

FirstEnergy Nuclear Operating Company, et al., Docket No. 50-412, 
Beaver Valley Power Station, Unit 2, Beaver County, Pennsylvania

    Date of amendment request: March 28, 2001.
    Description of amendment request: The proposed amendment would 
revise the Technical Specification (TS) requirements to credit the 
soluble boron in the fuel storage pool analyses. This amendment would 
revise the index, modify TS 3.9.14, ``Fuel Storage--Spent Fuel Storage 
Pool,'' add TS 3.9.15, ``Fuel Storage Pool Boron Concentration,'' 
modify applicable Bases and revise Design Feature Section 5.3.1.1, 
``Criticality.'' TS 3.9.14 would be modified by separating this 
specification into two specifications to support crediting soluble 
boron in the fuel storage pool. The revised TS 3.9.14 would provide 
controls for fuel assembly enrichment and burnup in the spent fuel pool 
and also include an increase in the maximum enrichment from 4.85 weight 
percent (w/o) to 5.0 w/o. A new TS 3.9.15 would provide control for 
soluble boron requirements in the spent fuel pool. Separating this 
specification into two specifications follows the general guidance 
provided in the improved standard TS (ISTS) of NUREG-1431.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Does the change involve a significant increase in the 
probability or consequences of an accident previously evaluated?
    Because of the Boraflex deterioration that has been observed, 
the spent fuel racks have been reanalyzed neglecting the presence of 
Boraflex to allow storage of Westinghouse 17x17 fuel assemblies with 
nominal

[[Page 44173]]

enrichments up to 5.0 weight percent (w/o) using credit for 
checkerboarding, burnup and soluble boron. The proposed changes will 
not have a significant impact on the safety of the plant or on the 
spent fuel storage pool and are consistent with the NRC approved 
changes identified for other plants (i.e., Prairie Island Units 1 
and 2, Vogtle Units 1 and 2). Criteria set forth in Table 3.9-1 
provide qualification requirements for fuel assembly storage to 
ensure the NRC acceptance criteria and accident analysis assumptions 
are satisfied. Increasing the enrichment from 4.85 w/o up to and 
including 5.0 w/o U-235 [uranium 235] has minor effects on the 
radiological source terms and subsequently the potential releases, 
both normal and accidental, are not significantly affected.
    The proposed Technical Specification changes credit the use of 
soluble boron in the spent fuel pool criticality analyses. These 
criticality analyses were performed using the NRC approved 
methodology developed by the Westinghouse Owners Group (WOG) and 
described in WCAP-14416-NP-A, Revision 1, ``Westinghouse Spent Fuel 
Rack Criticality Analysis Methodology,'' November 1996. The analysis 
includes evaluations that factor in the axial burnup bias correction 
and utilizing identified conservatisms in the analysis demonstrate 
that Keff remains less than or equal to the design limits.
    The proposed changes do not involve a change to plant equipment 
and do not affect the performance of plant equipment used to 
mitigate an accident. They do not affect the operation of the spent 
fuel pool cooling system or any other system and are consistent with 
applicable analyses including [those associated with postulated] 
fuel handling accidents. They will not affect the ability of any 
system to perform its design function; therefore, the proposed 
changes do not involve a significant increase in the probability or 
consequences of an accident previously evaluated.
    2. Does the change create the possibility of a new or different 
kind of accident from any accident previously evaluated?
    There are no hardware changes associated with this license 
amendment nor are there any changes in the method by which any 
safety-related plant system performs its safety function. No new 
accident scenarios, transient precursors, failure mechanisms or 
limiting single failures are introduced as a result of the proposed 
changes. The proposed changes do not introduce any adverse effects 
or challenges to any safety-related systems.
    The potential criticality accidents have been reanalyzed to 
demonstrate that the pool remains subcritical. Soluble boron has 
been maintained in the fuel storage pool water since its initial 
operation. The possibility of a fuel storage pool dilution is not 
affected by the proposed changes to the Technical Specifications. 
Therefore, implementation of Technical Specification controls for 
the soluble boron will not create the possibility of a new or 
different kind of accidental pool dilution.
    With credit for soluble boron now a major factor in controlling 
subcriticality, an evaluation of fuel storage pool dilution events 
was completed. This evaluation concluded that no credible events 
would result in a reduction of the criticality margin below the 5% 
margin recommended by the NRC. In addition, the No Soluble Boron 95/
95 probability/confidence level criticality analysis assures that 
dilution to 0 ppm [parts per million] will not result in 
criticality.
    The proposed Technical Specification changes ensure the 
maintenance of the fuel pool boron concentration and storage 
configuration. Therefore, the proposed changes will not create the 
possibility of any new or different kind of accident from any 
accident previously evaluated.
    3. Does the change involve a significant reduction in a margin 
of safety?
    The proposed changes do not affect the acceptance criteria for 
any analyzed event nor impact any plant safety analyses since the 
analysis assumptions are not changed. The safety limits assumed in 
the accident analyses and the design function of the equipment 
required to mitigate the consequences of any postulated accidents 
will not be changed since the proposed changes do not affect 
equipment required to mitigate design basis accidents described in 
the Updated Final Safety Analysis Report. The Technical 
Specifications continue to assure that applicable operating 
parameters are maintained within required limits.
    The proposed changes to the fuel storage pool boron 
concentration and storage requirements will provide adequate margin 
to assure that the fuel storage array will always remain subcritical 
by the 5% margin recommended by the NRC. These limits are based on a 
criticality analysis performed in accordance with NRC approved 
Westinghouse fuel storage rack criticality analysis methodology.
    While criticality analysis utilized credit for soluble boron, 
the storage configurations have been defined using Keff 
calculations to ensure that the spent fuel rack Keff will 
be less than 1.0 with no soluble boron. Soluble boron credit is used 
to offset off-normal conditions (such as a misplaced assembly) and 
to provide subcritical margin such that the fuel storage pool 
Keff is maintained less than or equal to 0.95.
    The spent fuel pool boron dilution analysis concludes that an 
unplanned or inadvertent event which would result in dilution of the 
spent fuel pool boron concentration from 2000 ppm to 450 ppm is not 
a credible event. This conclusion is based on the substantial volume 
of unborated water required to dilute the pool and the fact that a 
large dilution event would be readily detected by plant personnel 
via alarms, flooding in the fuel handling building or detected 
during normal operator rounds through the spent fuel pool area.
    The margin of safety depends upon maintenance of specific 
operating parameters within design limits. The Technical 
Specifications continue to require that these limits be maintained 
and provide appropriate remedial actions if a limit is exceeded. The 
maintenance of these limits continues to be assured through 
performance of surveillances. Therefore, the plant will be 
maintained within the analyzed limits and the proposed changes will 
not involve a significant reduction in a margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: Mary O'Reilly, FirstEnergy Nuclear Operating 
Company, FirstEnergy Corporation, 76 South Main Street, Akron, OH 
44308.
    NRC Section Chief: Richard P. Correia, Acting.

Florida Power Corporation, et al., Docket No. 50-302, Crystal River 
Unit No. 3 Nuclear Generating Plant, Citrus County, Florida

    Date of amendment request: July 24, 2001.
    Description of amendment request: The proposed amendment would 
accommodate future changes in plant design, including increased levels 
of Once-Through Steam Generator tube plugging. The changes are 
categorized into two sets. The first set of changes relocate parameters 
from the Improved Technical Specifications (ITS) to the cycle-specific 
Core Operating Limits Report (COLR). These parameters are the Variable 
Low Pressure Trip equation specified in ITS Table 3.3.1-1, and Reactor 
Coolant System (RCS) pressure limit within Surveillance Requirement 
(SR) 3.4.1.1. The second set of changes is directly related to tube 
plugging equivalent to up to 20% of all tubes, and addresses its 
impact. These changes include the revision of the hot leg maximum 
temperature limit, and the revision of the RCS minimum flow limits for 
four- and three-reactor coolant pump operation. The RCS limits 
associated with 20% plugging will be maintained in the ITS, however, 
cycle-specific values for these limits will be relocated to the COLR. 
The hot leg temperature and RCS flow limit values within SR 3.4.1.2 and 
3.4.1.3 ``RCS Pressure, Temperature, and Flow DNB [departure from 
nucleate boiling] Limits,'' will be relocated to reflect their location 
in the COLR. For both sets of changes, ITS 5.6.2.18(a) will be modified 
to reflect the relocation of cycle-specific values from the ITS to the 
COLR.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    (1) Does not involve a significant increase in the probability 
or consequences of an accident previously analyzed.

[[Page 44174]]

    The proposed change relocates several Reactor Coolant System 
(RCS) parameters from the ITS to the Core Operating Limits Report 
(COLR). The purpose for this relocation is to permit the values of 
these parameters to be changed under the 10 CFR 50.59 change process 
for cycle-specific analyses. In addition, these changes will allow 
increased Once-Through Steam Generator (OTSG) tube plugging. The 
increased plugging limit is in accordance with the analysis and will 
support continued proper maintenance of the OTSGs. The increased 
OTSG plugging will result in a small decrease in RCS flow and 
primary to secondary heat transfer. The difference in heat transfer 
results in small changes to primary and secondary operational 
parameters but will not result in any challenges to plant equipment. 
The change in RCS parameters will have no impact on the probability 
of accident initiators or precursors. Increased OTSG plugging will 
slightly reduce mass release to the containment following some loss 
of primary coolant accidents. Previously analyzed accidents were 
reevaluated considering the proposed changes and were found to be 
within established limits. Therefore, the change will not 
significantly increase the probability or consequences of an 
accident previously evaluated.
    (2) Does not create the possibility of a new or different kind 
of accident from any accident previously analyzed.
    The proposed changes do not introduce any new operating methods 
or configurations. The revised RCS parameters have been analyzed and 
have been determined to be within established limits. No new failure 
modes or limiting single failures were identified. All safety and 
design criteria continue to be met. Therefore, the proposed change 
will not create the possibility of a new or different kind of 
accident from any accident previously evaluated.
    (3) Does not involve a significant reduction in the margin of 
safety.
    The proposed changes affect RCS parameters, which are inputs to 
the plant's safety limits. The changes have been evaluated and the 
resultant plant analysis and configuration remain within the 
existing safety limits. The safety limits themselves are not being 
altered. The accident analysis was reevaluated and it has been 
determined that there is no significant impact on the fuel cladding, 
reactor coolant system or the containment structure. Therefore, this 
change does not involve a significant reduction in the margin of 
safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 50.92(c) are 
satisfied. Therefore the NRC staff proposes to determine if the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: R. Alexander Glenn, Associate General 
Counsel (MAC-BT15A), Florida Power Corporation, P.O. Box 14042, St. 
Petersburg, Florida 33733-4042.
    NRC Section Chief, Acting: Kahtan N. Jabbour.

Indiana Michigan Power Company, Docket Nos. 50-315 and 50-316, 
Donald C. Cook Nuclear Plant, Units 1 and 2, Berrien County, 
Michigan

    Date of amendment requests: May 17, 2001.
    Description of amendment requests: The proposed amendments would 
revise Technical Specification (TS) 3/4.9.3, ``Decay Time,'' to allow 
the start of a core offload at 100 hours after reactor subcriticality 
between September 15 and June 15, and 148 hours after reactor 
subcriticality between June 16 and September 14. The difference in the 
required decay times is dependent on the time of year due to the lake 
temperature assumed in the spent fuel pool cooling analysis. In 
addition, the proposed license amendment would make format changes to 
the TS pages.
    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. The NRC staff has reviewed the licensee's analysis 
against the standards of 10 CFR 50.92(c). The NRC staff's review is 
presented below: According to 10 CFR 50.92(c), a proposed amendment to 
an operating license involves no significant hazards consideration if 
operation of the facility in accordance with the proposed amendment 
would not:
    1. involve a significant increase in the probability of occurrence 
or consequences of an accident previously evaluated;
    2. create the possibility of a new or different kind of accident 
from any previously analyzed; or
    3. involve a significant reduction in a margin of safety.
    The determinations that the criteria set forth in 10 CFR 50.92 are 
met for this amendment request is indicated below:

    1. Does the change involve a significant increase in the 
probability of occurrence or consequences of an accident previously 
evaluated?
    The proposed license amendment would allow fuel assemblies to be 
removed from the reactor core and be stored in the spent fuel pool 
in less time after subcriticality than currently allowed by the TSs. 
Decreasing the decay time of the fuel affects the isotopic make-up 
of the fuel to be offloaded as well as the amount of decay heat that 
is present from the fuel at the time of offload. The proposed 
changes do not involve a significant increase in the probability of 
occurrence of an accident previously evaluated. The accident 
previously evaluated that is associated with the proposed license 
amendment is the fuel handling accident. Allowing the fuel to be 
offloaded as early as 100 hours after subcriticality does not impact 
the manner in which the fuel is offloaded. The accident initiator is 
the dropping of the fuel assembly. Since earlier offload does not 
effect fuel handling, there is no increase in the probability of 
occurrence of a fuel handling accident. The time frame in which the 
fuel assemblies are moved has been evaluated against the 10 CFR Part 
20 and 10 CFR Part 100 dose limits for members of the public and 
licensee personnel and 10 CFR 50.67 control room dose limits. All 
dose limits are met with the reduced core offload times.
    The proposed changes do not involve a significant increase in 
the consequences of an accident previously evaluated. The accident 
previously evaluated that is associated with fuel movement is the 
fuel handling accident. Thus, there is no significant increase in 
consequences.
    The TS page format changes are administrative in nature and have 
no impact on any accident previously evaluated. Thus, the 
probability of occurrence of an accident previously evaluated is not 
changed.
    Therefore, the proposed license amendment does not increase the 
probability of occurrence or the consequences of accidents 
previously evaluated are not increased.
    2. Create the possibility of a new or different kind of accident 
from any accident previously evaluated.
    The proposed license amendment would allow core offload to occur 
in less time after subcriticality, which affects the isotopic make-
up of the fuel to be offloaded as well as the amount of decay heat 
that is present from the fuel at the time of offload. The isotopic 
makeup of the fuel assemblies and the amount of decay heat produced 
by the fuel assemblies do not currently initiate any accident. A 
change in the isotopic makeup of the fuel at the time of core 
offload or an increase in the decay heat produced by the fuel being 
offloaded will not cause the initiation of any accident. There is no 
change to the manner in which fuel is being handled or in the 
equipment used to offload or store the fuel. The effects of the 
additional decay heat load have been analyzed. The analysis 
demonstrated that the existing spent fuel pool cooling system and 
all associated systems under worst-case circumstances would maintain 
the integrity of the spent fuel pool and the proposed method of 
offload does not create a new or different kind of accident from any 
accident previously evaluated.
    The TS page format changes are administrative in nature and have 
no impact on the operation of either unit. Therefore, the proposed 
changes do not create the possibility of a new or different kind of 
accident from any accident previously evaluated.
    Therefore, the proposed license amendment does not create the 
possibility of a new or different kind of accident from any accident 
previously evaluated.
    3. Does the change involve a significant reduction in a margin 
of safety?
    The margin of safety pertinent to the proposed changes is the 
dose consequences

[[Page 44175]]

resulting from a fuel handling accident. The shorter decay time 
prior to fuel movement has been evaluated against the 10 CFR Part 
100 in the current licensing basis and all limits continue to be 
met. In addition, the integrity of the spent fuel pool has been 
demonstrated with the additional decay heat load. As stated above, 
the changes in isotopic makeup and additional heat load do not 
impact any safety settings and do not cause any safety limit to not 
be met. In addition, the integrity of the spent fuel pool is 
maintained.
    The proposed format changes do not affect plant operation, and, 
therefore, do not involve a significant reduction in a margin of 
safety.
    Therefore, the proposed changes do not involve a significant 
reduction in a margin of safety.


    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.
    Attorney for licensee: David W. Jenkins, Esq., 500 Circle Drive, 
Buchanan, MI 49107.
    NRC Section Chief: Claudia M. Craig.

Indiana Michigan Power Company, Docket Nos. 50-315 and 50-316, 
Donald C. Cook Nuclear Plant, Units 1 and 2, Berrien County, 
Michigan

    Date of amendment requests: July 17, 2001.
    Description of amendment requests: The proposed amendments would 
revise Technical Specification Surveillance Requirement 4.0.3 to 
provide a delay period following discovery of a missed surveillance 
prior to declaring that the Limiting Condition for Operation has not 
been met. The proposed delay period would be 24 hours from the time of 
discovery of the missed surveillance or the limit of the specified 
surveillance interval, whichever is less. The proposed changes are 
consistent with the intent of Generic Letter 87-09, ``Sections 3.0 and 
4.0 of the Standard Technical Specifications on the Applicability of 
Limiting Conditions for Operation and Surveillance Requirements.'' 
Indiana Michigan Power Company is submitting this request to reduce the 
potential for unnecessary plant system and equipment manipulations.
    The proposed license amendment also includes format changes that 
improve appearance and are not intended to introduce other changes.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Does the change involve a significant increase in the 
probability of occurrence or consequences of an accident previously 
evaluated?
    It is overly conservative to assume components are inoperable 
when a surveillance requirement has not been performed. The 24-hour 
delay period to perform a missed surveillance does not involve a 
significant increase in the probability of occurrence of an accident 
previously evaluated because it allows time to perform the 
surveillance without requiring other plant manipulations such as a 
plant shutdown. If a plant shutdown is required before a missed 
surveillance is completed, it is likely that the surveillance would 
be conducted when the plant is being shut down because completion of 
a missed surveillance would terminate the shutdown requirement. A 
forced plant shutdown or other forced actions prior to completion of 
the missed surveillance increases risk to the plant, as it requires 
the manipulation of additional equipment. Delaying a surveillance 
test on a component cannot cause a failure of the component, nor 
would it significantly affect accident initiators or precursors. 
Therefore, there is no significant increase in the probability of 
occurrence of an accident previously evaluated.
    Since this change does not affect plant design, operation, or 
the manner in which testing is performed, there is no effect on the 
consequences of an accident previously evaluated.
    The T/S page format changes are administrative in nature and 
have no impact on plant operation.
    Thus, the proposed change does not involve a significant 
increase in the probability of occurrence or consequences of an 
accident previously evaluated.
    2. Does the change create the possibility of a new or different 
kind of accident from any accident previously evaluated?
    The proposed change does not affect plant design, operation, or 
the manner in which testing is performed. Delaying a surveillance 
test on a component cannot cause a failure of the component. As 
such, the proposed delay period will not cause any equipment 
malfunctions or introduce any changes to the way in which components 
operate. The T/S page format changes are administrative in nature 
and have no impact on plant operation. Therefore, the proposed 
changes do not increase the possibility of a new or different kind 
of accident from any accident previously evaluated.
    3. Does the change involve a significant reduction in a margin 
of safety?
    The margin of safety is neither described or prescribed for this 
specification. The proposed change simply provides additional time 
to perform a surveillance and verify that the operability of 
equipment is in conformance with the T/S requirements.
    The T/S page format changes are administrative in nature and 
have no impact on plant operation.
    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 requests involve no significant hazards consideration.
    Attorney for licensee: David W. Jenkins, Esq., 500 Circle Drive, 
Buchanan, MI 49107.
    NRC Section Chief: Claudia M. Craig.

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: June 29, 2001.
    Description of amendment requests: The licensee requests to revise 
Technical Specifications (TSs) 3.7.10, ``Emergency Chilled Water 
(ECW)'' and 3.7.11, ``Control Room Emergency Air Cleanup System 
(CREACUS)'' and the associated TSs Bases. The proposed change would 
revise the Allowed Outage Time (AOT) for a single inoperable train of 
both the ECW and CREACUS from 7 days to 14 days.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    The Commission has provided standards for determining whether a 
significant hazards consideration exists as stated in 10 CFR 50.92. 
A proposed amendment to an operating license for a facility involves 
no significant hazards consideration if operation of the facility in 
accordance with a 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. A discussion of these standards as they relate to this 
amendment request follows:
    (1) Will operation of the facility in accordance with this 
proposed change involve a significant increase in the probability or 
consequences of an accident previously evaluated?
    Response: No.
    This proposed change is to revise the Allowed Outage Time (AOT) 
for a single inoperable train of the Emergency Chilled Water (ECW) 
and Control Room Emergency Air Cleanup System (CREACUS) systems from 
7 days to 14 days. The proposed change does not involve a change in 
the design configuration, or operation of the plant.
    Therefore, this change does not involve a significant increase 
in the probability or consequences of an accident previously 
evaluated.

[[Page 44176]]

    (2) Will operation of the facility in accordance with this 
proposed change create the possibility of a new or different kind of 
accident from any accident previously evaluated?
    Response: No.
    This proposed change does not involve a change in the design, 
configuration, or method of operation of the plant.
    Therefore, this proposed change will not create the possibility 
of a new or different kind of accident from any accident that has 
been previously evaluated.
    (3) Will operation of the facility in accordance with this 
proposed change involve a significant reduction in a margin of 
safety?
    Response: No.
    The proposed change does not affect the limiting conditions for 
operation or their bases that are used in the deterministic analyses 
to establish the margin of safety. Probabilistic risk analysis was 
used to evaluate these changes.
    Therefore, there will be no significant reduction in a margin of 
safety as a result of this change.
    Based on the responses to these three criteria, Southern 
California Edison (SCE) has concluded that the proposed amendment 
involves no significant hazards consideration.

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

STP Nuclear Operating Company, Docket Nos. 50-498 and 50-499, South 
Texas Project, Units 1 and 2, Matagorda County, Texas

    Date of amendment request: May 9, 2001.
    Description of amendment request: Proposed amendments revise 
Technical Specification 3.5.1, ``Emergency Core Cooling Systems--
Accumulators,'' to extend the allowed outage time allowed for an 
inoperable accumulator to 24 hours.
    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:

    STPNOC has evaluated whether or not a significant hazards 
consideration is involved with the proposed amendment by focusing on 
the three standards set forth in 10 CFR 50.92, ``Issuance of 
amendment,'' as discussed below.
    1. Does the proposed change involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    Response: No.
    The proposed changes involve no significant increase in the 
probability of an accident previously evaluated because the 
accumulator has no role as an accident initiator.
    The proposed extension to the allowed outage time has no 
significant effect on the availability of the accumulator to perform 
its design function and has no effect on the configuration or 
accident response of the accumulator. The proposed change involves 
no changes to the accident analyses. Consequently, the proposed 
extended allowed outage time involves no significant increase in the 
consequences of an accident previously evaluated.
    The proposed changes to eliminate the surveillance requirements 
also have no significant effect on the availability of the 
accumulator to perform its design function and have no effect on the 
configuration or accident response of the accumulator. The changes 
to the surveillance requirements involve no change to the accident 
analyses. Consequently, the changes to the surveillance requirements 
involve no significant increase in the consequences of an accident 
previously evaluated.
    The proposed changes in the structure of the specification to be 
more consistent with ITS are administrative and have no technical 
impact. Consequently, they involve no significant increase in the 
probability or consequences of an accident previously evaluated.
    The correction of the typographical error is an administrative 
change which has no operational significance.
    2. Does the proposed change create the probability of a new or 
different kind of accident from any accident previously evaluated?
    Response: No.
    The proposed change does not involve the installation or 
operation of any new or different kinds of equipment, nor does it 
involve a new or different mode of operation. The proposed changes 
do not result in systems operating in a manner different from 
existing procedures and practices. Therefore, the proposed change 
does not create the possibility of a new or different kind of 
accident from any accident previously evaluated.
    The proposed changes in the structure of the specification to be 
more consistent with ITS are administrative and have no technical 
impact. Consequently, they do not create the possibility of a new or 
different kind of accident from any accident previously evaluated.
    The correction of the typographical error is an administrative 
change which has no operational significance.
    3. Does the proposed change involve a significant reduction in a 
margin of safety?
    Response: No.
    The proposed change will allow plant operation in a 
configuration outside the design basis for up to 24 hours before 
being required to begin shutdown. The impact of this on plant risk 
was evaluated and found to be very small. That is, increasing the 
time the accumulators will be unavailable to respond to large LOCA 
event, assuming design basis accumulator success criteria is 
necessary to mitigate the event, has a very small impact on plant 
risk. The analyses quantitatively demonstrate the change does not 
involve a significant reduction in the margin of safety.
    The proposed change removes the 18 month test to verify that the 
accumulator isolation valves automatically open when a simulated or 
actual P-11 interlock setpoint is exceeded, or when an SI signal is 
received. The valves are verified open every 24 hours and the power 
is verified removed every 31 days in accordance with the TS. Should 
the valves be inadvertently closed, the normal testing would 
adequately identify the condition. If the condition is recognized, 
the failure would be addressed by plant administrative controls that 
would immediately result in the appropriate Actions being taken for 
all affected systems. Based on the existence of other measures which 
adequately address the reason for the current requirement, this 
change does not involve a significant reduction in a margin of 
safety.
    The proposed change removes the requirement from the Technical 
Specifications to perform surveillances on the accumulator 
instrumentation. The TS does not specifically require this 
instrumentation to be used to meet the required pressure and level 
verification surveillances. The verification of accumulator level 
and pressure may be determined by either installed instrumentation 
or temporary test equipment. Therefore, the change does not involve 
a significant reduction in a margin of safety.
    The proposed changes in the structure of the specification to be 
more consistent with ITS are administrative and have no technical 
impact. Consequently, they do not involve a significant reduction in 
the margin of safety.
    The correction of the typographical error is an administrative 
change which has no operational significance.
    Based upon the analysis provided herein, the proposed amendments 
will not increase the probability or consequences of an accident 
previously evaluated, create the possibility of a new or different 
kind of accident previously evaluated, or involve a reduction in a 
margin of safety. Therefore, the proposed amendments meet the 
requirements of 10 CFR 50.92 and do not involve a significant 
hazards consideration.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the 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: Alvin H. Gutterman, Esq., Morgan, Lewis & 
Bockius, 1800 M Street, NW., Washington, DC 20036-5869.
    NRC Section Chief: Robert A. Gramm.

[[Page 44177]]

STP Nuclear Operating Company, Docket Nos. 50-498 and 50-499, South 
Texas Project, Units 1 and 2, Matagorda County, Texas

    Date of amendment request: May 30, 2001.
    Description of amendment request: Proposed amendments would permit 
relaxation of the allowed outage times and bypass test times for 
limiting conditions for operations under Technical Specifications 3.31, 
``Reactor Trip System Instrumentation,'' and 3.3.2, ``Engineered Safety 
Features Actuation System Instrumentation.''
    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 change involve a significant increase in the 
probability or consequences of an accident previously evaluated?
    Response: No.
    The reactor protection and engineered safety features functions 
are not initiators of any design basis accident or event and 
therefore the proposed changes do not increase the probability of 
any accident previously evaluated. The proposed changes to the 
allowed outage and bypass test times have an insignificant impact on 
plant safety based on the calculated core damage frequency increase 
being approximately 1.OE-06. Therefore, the proposed changes do not 
result in a significant increase in the consequences of an accident 
previously evaluated.
    2. Will the change create the possibility of a new or different 
kind of accident from any accident previously evaluated?
    Response: No.
    The proposed changes do not result in a change in the manner in 
which the Reactor Trip System (RTS) and Engineered Safety Features 
Actuation System (ESFAS) provide plant protection. The existing RTS 
and ESFAS actuation setpoints will be unaffected by these proposed 
changes. The changes to the allowed outage and bypass test times do 
not change any existing accident scenarios nor create any new or 
different accident scenarios. Therefore, this request does not 
create the possibility of a new or different kind of accident from 
any accident previously evaluated.
    3. Will the change involve a significant reduction in a margin 
of safety?
    Response: No.
    The proposed changes do not alter the manner in which safety 
limits, limiting safety system settings or limiting conditions for 
operation are determined. The impact of increased allowed outage 
times and bypass test times should result in an overall improvement 
in safety by reducing the potential for spurious reactor trips and 
spurious actuation of safety equipment. The longer allowed outage 
times and bypass test times will provide additional time before 
being required to place the associated channel in trip. With the 
channel in trip, the logic required to cause a reactor trip or 
safety system actuation is reduced to 1-out-or-2 (for 2-out-of-3 
logic) and 1-out-of-3 (for 2-out-or-4 logic). With one channel 
tripped, the potential for a spurious actuation is increased. 
Placing a channel in bypass for additional time does reduce the 
availability of signals to initiate component actuation for event 
mitigation when required, but as shown in WCAP-14333, the impact on 
safety is small due to the availability of other signals or operator 
action to trip the reactor or cause component actuation. Therefore, 
these proposed changes should reduce the potential for inadvertent 
reactor trips and inadvertent equipment actuations due to human 
error or spurious actuation, and will not involve a significant 
reduction in the margin of safety.
    Based upon the analysis provided herein, the proposed amendments 
will not increase the probability or consequences of an accident 
previously evaluated, create the possibility of a new or different 
kind of accident from any accident previously evaluated, or involve 
a reduction in a margin of safety. Therefore, the proposed 
amendments meet the requirements of 10 CFR 50.92 and do not involve 
a significant hazards consideration.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the 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: Alvin H. Gutterman, Esq., Morgan, Lewis & 
Bockius, 1800 M Street, NW., Washington, DC 20036-5869.
    NRC Section Chief: Robert A. Gramm.

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

    Date of amendment request: August 6, 2001 (TS 01-05).
    Description of amendment request: The proposed amendment would 
revise the Technical Specification (TS) surveillance requirements for 
containment isolation valves (CIVs) to be verified closed. More 
specifically, valves in high radiation areas may be verified by 
administrative means. In addition, valves which are locked sealed or 
otherwise secured do not need to be reverified closed and are 
eliminated from the scope of the surveillance.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    A. The proposed amendment does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    The proposed changes to the surveillance requirements (SR) for 
verification of valve position continues to assure the operability 
of these valves such that the containment isolation function assumed 
in the safety analyses is maintained. Since these proposed revisions 
will continue to support the required safety functions without 
modification of the plant features, the probability of an accident 
is not increased.
    The provisions proposed in this change request will continue to 
maintain an acceptable level of protection for the health and safety 
of the public and will not impact the potential for the offsite 
release of radioactive products. The overall effect of the proposed 
change will result in specifications that have equivalent 
requirements compared to existing specifications for CIV operability 
and will not increase the consequences of an accident.
    B. The proposed amendment does not create the possibility of a 
new or different kind of accident from any accident previously 
evaluated.
    The proposed revisions are not the result of changes to plant 
equipment, system design, testing methods, or operating practices. 
The modified requirements will allow the use of administrative means 
for verification of valve closure for those CIVs located in high 
radiation areas and eliminate the requirement to verify close those 
valves that are locked, sealed, or otherwise secured. The 
specifications for CIVs serve to provide controls for maintaining 
the containment pressure boundary. TVA's proposed changes does not 
contribute to the generation of postulated accidents. Since the 
function of the CIVs and their associated systems remains unchanged, 
and the effects do not contribute to accident generation, the 
proposed changes will not create the possibility of a new or 
different kind of accident.
    C. The proposed amendment does not involve a significant 
reduction in a margin of safety.
    The proposed change involves upgrading the CIV TS surveillance 
requirement to be consistent with the S[Standard]TS. The proposed 
change has been developed considering the importance of the CIVs in 
limiting the consequences of a design basis event and the concerns 
for the plant's ability to perform required operational support 
functions with the necessary systems isolated. The proposed change 
allows for alternative protection to assure the isolation function 
of the valves remain available.
    Since the proposed revision does not alter the intent or 
application of the current TS requirements, and the function of the 
CIVs and their associated systems remains unchanged, the proposed 
change will continue to provide controls for maintaining the 
containment pressure boundary.

    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

[[Page 44178]]

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.

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. 
Publicly available records will be accessible from the Agencywide 
Documents Access and Management Systems (ADAMS) Public Electronic 
Reading Room on the internet at the NRC web site, http://www.nrc.gov/NRC/ADAMS/index.html. If you do not have access to ADAMS or if there 
are problems in accessing the documents located in ADAMS, contact the 
NRC Public Document Room (PDR) Reference staff at 1-800-397-4209, 301-
415-4737 or by email to [email protected].

Carolina Power & Light Company, Docket No. 50-261, H. B. Robinson 
Steam Electric Plant, Unit No. 2, Darlington County, South Carolina

    Date of application for amendment: June 5, 2000, as supplemented 
August 4, 2000, and July 6, 2001.
    Brief description of amendment: This amendment revises Technical 
Specification (TS) 3.7.8 to establish Required Actions and Completion 
Times in the event that the service water system exceeds the maximum 
allowed TS temperature of 97 degrees F.
    Date of issuance: August 9, 2001.
    Effective date: August 9, 2001.
    Amendment No. 191.
    Facility Operating License No. DPR-23. Amendment revises the 
Technical Specifications.
    Date of initial notice in Federal Register: August 9, 2000 (65 FR 
48745). The August 4, 2000, and July 6, 2001, supplements contained 
clarifying information only, and did not change the initial no 
significant hazards consideration determination or expand the scope of 
the initial application.
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated August 9, 2001.
    No significant hazards consideration comments received: No.

Carolina Power & Light Company, et al., Docket No. 50-400, Shearon 
Harris Nuclear Power Plant, Unit 1, Wake and Chatham Counties, 
North Carolina

    Date of application for amendment: May 18, 2001
    Brief description of amendment: The amendment revises Technical 
Specification (TS) 3/4.9.4 ``Containment Building Penetrations'' and 
the associated Bases to permit containment building penetrations to 
remain open, under administrative controls, during core alterations or 
the movement of irradiated fuel within the containment. Specifically, 
the amendment: (1) Incorporates an alternate source term methodology in 
the fuel handling accident analysis; (2) revises TS 3.9.4 to remove 
portions of a note restricting the applicability of administrative 
controls with respect to containment penetrations; and (3) includes the 
use of administrative controls on the equipment hatch and other 
penetrations that provide access from containment atmosphere to outside 
atmosphere.
    Date of issuance: July 30, 2001.
    Effective date: July 30, 2001.
    Amendment No.: 104.
    Facility Operating License No. NPF-63: Amendment revises the 
Technical Specifications.
    Date of initial notice in Federal Register: June 27, 2001 (66 FR 
34280).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated July 30, 2001.
    No significant hazards consideration comments received: No.

Consolidated Edison Company of New York, Docket No. 50-247, Indian 
Point Nuclear Generating Unit No. 2, Westchester County, New York

    Date of application for amendment: May 10, 2001:
     Brief description of amendment: The amendment removes Technical 
Specification surveillance requirement 4.6.A.4 that requires each 
emergency diesel generator (EDG) to be given a thorough inspection at 
least annually following the manufacturer's recommendations. The 
requirement for the EDG inspection will be relocated to the Updated 
Final Safety Analysis Report and will be in accordance with the 
licensee-controlled maintenance program. The inspection period required 
by the maintenance program will also be changed to specify that it will 
be ``in accordance with the manufacturer's recommendations.''
    Date of issuance: July 30, 2001.
    Effective date: As of the date of issuance to be implemented within 
60 days.
    Amendment No.: 218.
    Facility Operating License No. DPR-26: Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: June 12, 2001 (66 FR 
31704).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated July 30, 2001.
    No significant hazards consideration comments received: No.

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

    Date of application for amendments: March 1, 2001.
    Brief description of amendments: The amendments revised the 
Technical Specifications (TS) to permit implementation of 10 CFR part 
50, Appendix J, Option B and to reference Regulatory Guide 1.163, 
``Performance-Based Containment Leak Test Program,'' dated September 
1995, which specifies a method acceptable to the NRC for complying with 
Option B. These changes relate only to Type B and Type C (local) 
leakage rate testing. In addition, the amendments revised Surveillance 
Requirement 3.6.3.8 by deleting the requirement for soap bubble testing 
of welded penetrations that are

[[Page 44179]]

not individually testable and clarified the Bases for TS 3.6.2 
pertaining to the containment air lock door.
    Date of issuance: July 31, 2001.
    Effective date: As of the date of issuance and shall be implemented 
within 30 days from the date of issuance.
    Amendment Nos.: 192/184.
    Facility Operating License Nos. NPF-35 and NPF-52: Amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: April 19, 2001 (66 FR 
22028).
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated July 31, 2001.
    No significant hazards consideration comments received: No.

Entergy Nuclear Operations, Inc., Docket No. 50-333, James A. 
FitzPatrick Nuclear Power Plant, Oswego County, New York

    Date of application for amendment: May 11, 2001.
    Brief description of amendment: The amendment extends, on a one-
time-basis, the Limiting Condition for Operation allowable out-of-
service time for the residual heat removal service water (RHRSW) system 
from 7 days to 11 days. The applicability of this change is limited to 
the one-time-only installation of the modification to the ``B'' RHRSW 
strainer.
    Date of issuance: July 27, 2001.
    Effective date: As of the date of issuance to be implemented within 
30 days.
    Amendment No.: 271.
    Facility Operating License No. DPR-59: Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: June 27, 2001 (66 FR 
34282).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated July 27, 2001.
    No significant hazards consideration comments received: No.

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 application for amendments: February 9, 2001 as 
supplemented by letters dated May 18, 2001 and June 26, 2001.
    Brief description of amendments: The one-time amendments revise 
Braidwood Unit 1 Technical Specifications (TS), section 5.5.9.d.2, 
``Steam Generator Tube Surveillance Program, Inspection Frequencies,'' 
for the Braidwood Station, Unit 1, fall 2001 refueling outage to allow 
a 40 month inspection interval after its first (post-replacement) 
inservice inspection, resulting in a C-1 classification, rather than 
after two consecutive inspections.
    Date of issuance: August 9, 2001.
    Effective date: As of the date of issuance and shall be implemented 
within 30 days.
    Amendment Nos.: 117 and 117.
    Facility Operating License Nos. NPF-72 and NPF-77: The amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: May 2, 2001 (66 FR 
22030). The May 18, 2001 and June 26, 2001, supplemental letters 
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 amendments is contained in a Safety 
Evaluation dated August 9, 2001.
    No significant hazards consideration comments received: No.

FirstEnergy Nuclear Operating Company, et al., Docket No. 50-334, 
Beaver Valley Power Station, Unit No. 1 (BVPS-1), Beaver County, 
Pennsylvania

    Date of application for amendment: December 21, 2000.
    Brief description of amendment: The amendment revised Technical 
Specification 3/4.3.1, ``Reactor Trip System Instrumentation,'' and 
associated bases to reflect the deletion of the steam/feedwater flow 
mismatch and low steam generator water level reactor trip function.
    Date of issuance: August 8, 2001.
    Effective date: As of the day of issuance and shall be implemented 
by the first entry into MODE 2 following the BVPS-1 Refueling Outage 
14.
    Amendment No: 240.
    Facility Operating License No. DPR-66: Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: January 24, 2001 (66 FR 
7680).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated August 8, 2001.
    No significant hazards consideration comments received: No.

FirstEnergy Nuclear Operating Company, et al., Docket Nos. 50-334 
and 50-412, Beaver Valley Power Station, Unit Nos. 1 and 2, Beaver 
County, Pennsylvania

    Date of application for amendments: December 27, 2000, as 
supplemented on March 28, April 12, June 9, June 13, and June 29 (3), 
2001. The addition of a Technical Specification (TS) Bases control 
program was requested on March 28, 2001.
    Brief description of amendments: These amendments allow: (1) 
Revisions to reactor trip and engineered safety feature actuation 
setpoints and allowable values, (2) implementation of the revised 
thermal design procedure, (3) relocations of TS requirements to the 
core operating limits report, (4) relocation of TS requirements to the 
licensee requirements manual, (5) miscellaneous editorial changes. In 
addition, License Condition 2.(C).(3) regarding less than 3-loop 
operation was deleted.
    Date of issuance: July 20, 2001.
    Effective date: Immediately and to be implemented within 120 days.
    Amendment Nos.: 239 and 120.
    Facility Operating License Nos. DPR-66 and NPF-73: Amendments 
revised the Technical Specifications and License.
    Date of initial notice in Federal Register: April 18, 2001 (66 FR 
20002) for the December 27, 2000, amendment request. A portion of a 
March 28, 2001, amendment request was also issued in this amendment. 
The date of the initial notice for the March 28, 2001, amendment 
request was June 20, 2001 (66 FR 33111).
    The March 28, April 12, June 9, June 13, and June 29 (3), 2001, 
letters provided clarifying information that did not change the initial 
proposed no significant hazards consideration determination and did not 
expand the scope of the original Federal Register notice.
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated July 20, 2001.
    No significant hazards consideration comments received: No.

Florida Power and Light Company, Docket Nos. 50-250 and 50-251, 
Turkey Point Plant, Units 3 and 4, Miami-Dade County, Florida

    Date of application for amendments: March 12, 2001 as supplemented 
June 26, 2001.
    Brief description of amendments: The amendments to the emergency 
diesel generators (EDG) Technical Specifications (TS) revised the 72-
hour allowed outage time specified in TS 3.8.1.1, Actions b and f, and 
Tss 3.4.3 and 3.5.2 to allow 14 days to restore an inoperable EDG to 
operable status. In addition, the amendments deleted TS Surveillance 
Requirement 4.8.1.1.2.g.1 and allowed its relocation to a licensee-
controlled maintenance program that will be incorporated by reference 
into the Updated Final Safety Analysis Report.
    Date of issuance: August 8, 2001.

[[Page 44180]]

    Effective date: As of the date of issuance and shall be implemented 
within 60 days of issuance.
    Amendment Nos: 215 and 209.
    Facility Operating License Nos. DPR-31 and DPR-41: Amendments 
revised the TS.
    Date of initial notice in Federal Register: April 18, 2001 (66 FR 
20005). The supplemental submittal of June 26, 2001, provided 
clarifying information that did not change the scope of the original 
request or change the initial proposed no significant hazards 
consideration.
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated August 8, 2001.
    No significant hazards consideration comments received: No.

Indiana Michigan Power Company, Docket Nos. 50-315 and 50-316, 
Donald C. Cook Nuclear Plant, Units 1 and 2, Berrien County, 
Michigan

    Date of application for amendments: August 18, 2000.
    Brief description of amendments: The amendments would change 
Technical Specification (TS) 3/4.7.4, ``Essential Service Water (ESW) 
System,'' and the associated Bases to add requirements that would 
support cross-connection to the opposite unit. The proposed amendment 
would also delete a provision for a 60-day allowed outage time when an 
ESW flowpath is not available to support the opposite unit's shutdown 
functions. Administrative and editorial changes are also made to 
provide consistency between units, correct typographical errors, 
improve readability, and improve page layout.
    Date of issuance: August 3, 2001.
    Effective date: As of the date of issuance and shall be implemented 
within 60 days.
    Amendment Nos.: 253 and 235.
    Facility Operating License Nos. DPR-58 and DPR-74: Amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: September 20, 2000 (65 
FR 56951) The Commission's related evaluation of the amendments is 
contained in a Safety Evaluation dated August 3, 2001.
    No significant hazards consideration comments received: No.

Indiana Michigan Power Company, Docket Nos. 50-315 and 50-316, 
Donald C. Cook Nuclear Plant, Units 1 and 2, Berrien County, 
Michigan

    Date of application for amendments: July 17, 2001.
    Brief description of amendments: The amendments revise Technical 
Specification 3.3.1.1, Table 3.3-1, Action 2a, to increase the amount 
of time allowed to place an inoperable power range neutron flux channel 
in the tripped condition from one hour to six hours.
    Date of issuance: August 8, 2001.
    Effective date: As of the date of issuance and shall be implemented 
within 3 days.
    Amendment Nos.: 254 and 236.
    Facility Operating License Nos. DPR-58 and DPR-74: Amendments 
revised the Technical Specifications.
    Public comments requested as to proposed no significant hazards 
consideration: Yes (66 FR 38753, dated July 25, 2001). The notice 
provided an opportunity to submit comments on the Commission's proposed 
NSHC determination. No comments have been received. The notice also 
provided for an opportunity to request a hearing by August 24, 2001, 
but indicated that if the Commission makes a final NSHC determination, 
any such hearing would take place after issuance of the amendment.
    The Commission's related evaluation of the amendments, finding of 
exigent circumstances, state consultation, and final NSHC determination 
are contained in a Safety Evaluation dated August 8, 2001.
    Attorney for licensee: David W. Jenkins, Esq., Indiana Michigan 
Power Company, Nuclear Generation Group, One Cook Place, Bridgman, MI 
49106.
    NRC Section Chief: Claudia M. Craig.

Niagara Mohawk Power Corporation, Docket No. 50-220, Nine Mile 
Point Nuclear Station Unit No. 1, Oswego County, New York

    Date of application for amendment: March 6, 2001.
    Brief description of amendment: The amendment revises the Technical 
Specifications to change the standard by which the licensee tests 
charcoal used in engineered safeguards features systems to American 
Society for Testing and Materials D3803-1989. These revisions are made 
in accordance with Generic Letter 99-02, ``Laboratory Testing of 
Nuclear-Grade Activated Charcoal.''
    Date of issuance: July 30, 2001.
    Effective date: As of the date of issuance and shall be implemented 
within 30 days.
    Amendment No.: 171.
    Facility Operating License No. NPF-69: Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: April 4, 2001 (66 FR 
17968).
    The staff's related evaluation of the amendment is contained in a 
Safety Evaluation dated
    No significant hazards consideration comments received: No.

Niagara Mohawk Power Corporation, Docket No. 50-410, Nine Mile 
Point Nuclear Station Unit No. 2, Oswego County, New York

    Date of application for amendment: March 29, 2001.
    Brief description of amendment: The amendment revises the Technical 
Specifications, Section 3.7.2, ``Control Room Envelope Filtration 
(CREF) System,'' to establish actions to be taken for an inoperable 
CREF system due to a degraded control room envelope boundary.
    Date of issuance: August 7, 2001.
    Effective date: As of the date of issuance and shall be implemented 
within 60 days.
    Amendment No.: 97.
    Facility Operating License No. NPF-69: Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: May 30, 2001 (66 FR 
29360).
    The staff's related evaluation of the amendment is contained in a 
Safety Evaluation dated August 7, 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 March 23, April 9, and June 27, 2001.
    Brief description of amendment: The amendment revises the licensing 
basis to utilize the full scope of an alternative radiological source 
term for accidents as described in NUREG-1465, ``Accident Source Terms 
for Light-Water Nuclear Power Plants,'' and revises the Technical 
Specifications implementing various assumptions in the alternative 
source term analyses.
    Date of issuance: July 31, 2001.
    Effective date: As of the date of issuance and shall be implemented 
within 30 days.
    Amendment No.: 240.
    Facility Operating License No. DPR-49: The amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: March 6, 2001 (66 FR 
13598).
    The March 6, 2001, notice provided an opportunity for a hearing and 
petition for leave to intervene. No requests for hearing or petition 
for leave to intervene were received. Subsequently, the staff 
determined that the licensing action was eligible for categorical 
exclusion from environmental review. The amendment request was noticed 
on June 27, 2001

[[Page 44181]]

(66 FR 34285), with the staff's proposed no significant hazards 
consideration determination. The June 27, 2001, supplement contained 
corrected TS pages and did not change the initial no significant 
hazards consideration determination and did not expand the scope of the 
original Federal Register notice.
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated July 31, 2001.
    No significant hazards consideration comments received: No.

Nuclear Management Company, LLC, Docket No. 50-263, Monticello 
Nuclear Generating Plant, Wright County, Minnesota

    Date of application for amendment: May 2, 2001, as supplemented 
June 22 and July 27, 2001.
    Brief description of amendment: The amendment (1) relocates 
requirements of the American Society of Mechanical Engineers (ASME) 
Boiler and Pressure Vessel Code (the Code), Section XI, Inservice 
Testing (IST) Program currently contained in Technical Specification 
(TS) Surveillance Requirement (TSSR) 4.15.B to TS Administrative 
Control Section 6.8, ``Programs and Manuals,'' (2) makes conforming 
changes to several SRs to reflect the change in reference from TSSR 
4.15.B to the licensee-controlled IST Program, (3) rewords TSSRs 
4.5.A.3 and 4.5.D.1 to be consistent with NUREG-1433, (4) incorporates 
TS Task Force (TSTF) initiative TSTF-279 into TS Administrative Control 
Section 6.8, and (5) revises TSSRs 4.6.H.1, 4.6.H.3, and Table 4.6.1 to 
change the inspection and functional testing interval extensions 
reference from plus-or-minus 25 percent to plus 25 percent.
    Date of issuance: August 1, 2001.
    Effective date: As of the date of issuance and shall be implemented 
within 45 days.
    Amendment No.: 122.
    Facility Operating License No. DPR-22. Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: May 30, 2001 (66 FR 
29360).
    The June 22, 2001, supplement provided clarifying information to 
the application and added a table defining IST testing frequencies to 
the proposed TS 6.8.G in order to be consistent with NUREG-1433. The 
July 27, 2001, supplement provided updated TS pages to reflect 
amendments issued subsequent to the application. The supplements were 
within the scope of the original Federal Register notice and did not 
change the staff's initial proposed no significant hazards 
considerations determination. The Commission's related evaluation of 
the amendment is contained in a Safety Evaluation dated August 1, 2001.
    No significant hazards consideration comments received: No.

Nuclear Management Company, LLC, Docket No. 50-255, Palisades 
Plant, Van Buren County, Michigan

    Date of application for amendment: April 2, 2001.
    Brief description of amendment: Removes from the Technical 
Specifications all requirements for, and references to, the term 
``Assembly Radial Peaking Factor.''
    Date of issuance: August 1, 2001.
    Effective date: As of the date of issuance and shall be implemented 
within 60 days.
    Amendment No.: 205.
    Facility Operating License No. DPR-20. Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: May 2, 2001 (66 FR 
22027).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated August 1, 2001.
    No significant hazards consideration comments received: No.

Portland General Electric Company, et al., Docket No. 50-344, 
Trojan Nuclear Plant, Columbia County, Oregon

    Date of application for amendment: March 6, 2001, as supplemented 
by letter dated June 6, 2001.
    Brief description of amendment: The amendment revises Section 5.0, 
``Administrative Controls,'' of the Permanently Defueled Technical 
Specifications by eliminating the position of Senior Vice President, 
Power Supply, and assigning those duties to the Trojan Site Executive; 
and dividing the position and duties of the Trojan Site Executive and 
Plant General Manager between two separate positions: (1) Trojan Site 
Executive, and (2) General Manager, Trojan. The amendment also revises 
the language used in Section 5.0 of the Permanently Defueled Technical 
Specifications to conform with the language of revised 10 CFR 50.59 by 
replacing phrases which included the wording ``unreviewed safety 
question'' and ``safety evaluation'' with wording that will continue to 
conform to the requirements of revised 10 CFR 50.59.
    Date of issuance: July 31, 2001.
    Effective date: This license amendment is effective as of the date 
of issuance.
    Amendment No.: 207.
    Facility Operating License No. NPF-1: The amendment changes the 
Permanently Defueled Technical Specifications.
    Date of initial notice in Federal Register: April 4, 2001 (66 FR 
17962).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated July 31, 2001.
    No significant hazards consideration comments received: No.

South Carolina Electric & Gas Company, South Carolina Public 
Service Authority, Docket No. 50-395, Virgil C. Summer Nuclear 
Station, Unit No. 1, Fairfield County, South Carolina

    Date of application for amendment: September 14, 2000, as 
supplemented April 24 and May 24, 2001.
    Brief description of amendment: The amendment removes the 
references to the Independent Safety Engineering Group.
    Date of issuance: July 30, 2001.
    Effective date: July 30, 2001.
    Amendment No.: 151.
    Facility Operating License No. NPF-12: Amendment revises the 
Technical Specifications.
    Date of initial notice in Federal Register: November 1, 2000 (65 FR 
65349). The April 24 and May 24, 2001, supplements contained clarifying 
information only and did not change the initial no significant hazards 
consideration determination or expand the scope of the initial 
application.
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated July 30, 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: May 30, 2001 (ULNRC-04481).
    Brief description of amendment: The amendment removes the phrase 
``and the charging flow control valve full open'' from Limiting 
Condition for Operation 3.5.5, Required Action A.1, and Surveillance 
Requirement 3.5.5.1 for the reactor coolant pump seal injection flow in 
the technical specifications.
    Date of issuance: August 7, 2001.
    Effective date: August 7, 2001, and shall be implemented within 60 
days from the date of issuance.
    Amendment No.: 146.
    Facility Operating License No. NPF-30: The amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: June 27, 2001 (66 FR 
34289)
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated August 7, 2001.

[[Page 44182]]

    No significant hazards consideration comments received: No.

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

    Date of amendment request: March 22, 2001.
    Brief description of amendment: The amendment changed the 
penetration values in Technical Specification (TS) 5.5.11.c for 
laboratory testing of the charcoal adsorber for the control room 
ventilation system from 2 percent to 2.5 percent and the auxiliary/fuel 
building emergency exhaust system from 2 percent to 5 percent. The 
amendment also deleted the ``'' sign associated with the 
temperature for the laboratory test of a sample of the charcoal 
adsorber.
    Date of issuance: August 7, 2001.
    Effective date: August 7, 2001, and shall be implemented within 60 
days from the date of issuance.
    Amendment No.: 139.
    Facility Operating License No. NPF-42. The amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: May 16, 2001 (66 FR 
27178).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated August 7, 2001.
    No Significant Hazards Consideration comments received: No.

    Dated at Rockville, Maryland, this 14th day of August 2001.

    For the Nuclear Regulatory Commission.
John A. Zwolinski,
Director, Division of Licensing Project Management, Office of Nuclear 
Reactor Regulation.
[FR Doc. 01-20885 Filed 8-21-01; 8:45 am]
BILLING CODE 7590-01-P