[Federal Register Volume 66, Number 211 (Wednesday, October 31, 2001)]
[Notices]
[Pages 55007-55031]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: 01-27261]


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NUCLEAR REGULATORY COMMISSION


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

    Note: The publication date for this notice will change from 
every other Wednesday to every other Tuesday, effective January 8, 
2002. The notice will contain the same information and will continue 
to be published biweekly.

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 October 9, 2001 through October 19, 2001. 
The last biweekly notice was published on October 17, 2001 (66 FR 
52794).

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

[[Page 55008]]

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. The filing of requests for a hearing and 
petitions for leave to intervene is discussed below.
    By November 30, 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 NRC's Public Document Room, located at One White Flint 
North, 11555 Rockville Pike (first floor), Rockville, Maryland 20852. 
Publicly available records will be accessible electronically from the 
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 a request for a hearing or 
petition for leave to intervene is filed by the above date, the 
Commission or an Atomic Safety and Licensing Board, designated by the 
Commission or by the Chairman of the Atomic Safety and Licensing Board 
Panel, will rule on the request and/or petition; and the Secretary or 
the designated Atomic Safety and Licensing Board will issue a notice of 
a hearing or an appropriate order.
    As required by 10 CFR 2.714, a petition for leave to intervene 
shall set forth with particularity the interest of the petitioner in 
the proceeding, and how that interest may be affected by the results of 
the proceeding. The petition should specifically explain the reasons 
why intervention should be permitted with particular reference to the 
following factors: (1) The nature of the petitioner's right under the 
Act to be made a party to the proceeding; (2) the nature and extent of 
the petitioner's property, financial, or other interest in the 
proceeding; and (3) the possible effect of any order which may be 
entered in the proceeding on the petitioner's interest. The petition 
should also identify the specific aspect(s) of the subject matter of 
the proceeding as to which petitioner wishes to intervene. Any person 
who has filed a petition for leave to intervene or who has been 
admitted as a party may amend the petition without requesting leave of 
the Board up to 15 days prior to the first prehearing conference 
scheduled in the proceeding, but such an amended petition must satisfy 
the specificity requirements described above.
    Not later than 15 days prior to the first prehearing conference 
scheduled in the proceeding, a petitioner shall file a supplement to 
the petition to intervene which must include a list of the contentions 
which are sought to be litigated in the matter. Each contention must 
consist of a specific statement of the issue of law or fact to be 
raised or controverted. In addition, the petitioner shall provide a 
brief explanation of the bases of the contention and a concise 
statement of the alleged facts or expert opinion which support the 
contention and on which the petitioner intends to rely in proving the 
contention at the hearing. The petitioner must also provide references 
to those specific sources and documents of which the petitioner is 
aware and on which the petitioner intends to rely to establish those 
facts or expert opinion. Petitioner must provide sufficient information 
to show that a genuine dispute exists with the applicant on a material 
issue of law or fact. Contentions shall be limited to matters within 
the scope of the amendment under consideration. The contention must be 
one which, if proven, would entitle the petitioner to relief. A 
petitioner who fails to file such a supplement which satisfies these 
requirements with respect to at least one contention will not be 
permitted to participate as a party.
    Those permitted to intervene become parties to the proceeding, 
subject to any limitations in the order granting leave to intervene, 
and have the opportunity to participate fully in the conduct of the 
hearing, including the opportunity to present evidence and cross-
examine witnesses.
    If a hearing is requested, the Commission will make a final 
determination on the issue of no significant hazards consideration. The 
final determination will serve to decide when the hearing is held.
    If the final determination is that the amendment request involves 
no significant hazards consideration, the Commission may issue the 
amendment and make it immediately effective, notwithstanding the 
request for a hearing. Any hearing held would take place after issuance 
of the amendment.
    If the final determination is that the amendment request involves a 
significant hazards consideration, any hearing held would take place 
before the issuance of any amendment.
    A request for a hearing or a petition for leave to intervene must 
be filed with the Secretary of the Commission, U.S. Nuclear Regulatory 
Commission, Washington, DC 20555-0001, Attention: Rulemaking and 
Adjudications Branch, or may be delivered to the Commission's Public 
Document Room, located at One White Flint North, 11555 Rockville Pike 
(first floor), Rockville, Maryland 20852, by the above date. A copy of 
the petition should also be sent to the Office of the General Counsel, 
U.S. Nuclear Regulatory Commission, Washington, DC 20555-0001, and to 
the attorney for the licensee.
    Nontimely filings of petitions for leave to intervene, amended 
petitions, supplemental petitions and/or requests for a hearing will 
not be entertained absent a determination by the Commission, the 
presiding officer or the Atomic Safety and Licensing Board that the 
petition and/or request should be granted based upon a balancing of 
factors specified in 10 CFR 2.714(a)(1)(i)-(v) and 2.714(d).
    For further details with respect to this action, see the 
application for amendment which is available for

[[Page 55009]]

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 Nos. 50-289, Three Mile Island Nuclear Station, Unit 1, Dauphin 
County, Pennsylvania

    Date of amendment request: August 1, 2001.
    Description of amendment request: The requested changes to the 
technical specifications (TSs) propose to revise requirements that have 
been superceded based on licensed operator training programs being 
accredited by the Institute for Nuclear Power Operations (INPO), 
promulgation of the revised 10 CFR part 55, Operators' Licenses, and 
adoption of a systems approach to training as required by 10 CFR 
50.120, Training and qualification of nuclear power plant personnel. 
The same changes were requested by Exelon Generation Company, LLC 
(Exelon) for Braidwood Station, Units 1 and 2; Byron Station, Units 1 
and 2; Dresden Nuclear Power Station, Units 2 and 3; LaSalle County 
Station, Units 1 and 2; Limerick Generating Station, Units 1 and 2; 
Peach Bottom Atomic Power Station, Units 2 and 3; and Quad Cities 
Nuclear Power Station, Units 1 and 2. The proposed no significant 
hazards consideration for those plants is published elsewhere in the 
Federal Register under Exelon Generation Company, LLC.
    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 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 change does not involve a significant increase in 
the probability or consequences of an accident previously evaluated.
    There will be no changes to the procedures by which the operators 
operate the plants. There will be no changes to the systems, 
structures, or components in the plants.
    Based on the above, these proposed changes do not involve a 
significant increase in the probability or consequences of an accident 
previously evaluated.
    2. The proposed change does not create the possibility of a new or 
different kind of accident previously evaluated.
    There will be no changes to the procedures by which the operators 
operate the plants. There will be no changes to the systems, 
structures, or components in the plants.
    Therefore, the proposed changes will not create the possibility of 
a new or different kind of accident previously evaluated.
    3. The proposed change does not involve a significant reduction in 
a margin of safety.
    There will be no change in the plants' systems, structures, or 
components, nor in the way in which they will be operated as a result 
of the proposed changes. Therefore, the proposed changes will 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 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 Chiefs: Anthony J. Mendiola, Lakshminaras Raghavan.

Dominion Nuclear Connecticut, Inc., Docket No. 50-336, Millstone 
Nuclear Power Station, Unit No. 2, New London County, Connecticut

    Date of amendment request: July 31, 2001.
    Description of amendment request: The proposed amendment deletes 
requirements from the Technical Specifications (and, as applicable, 
other elements of the licensing bases) to maintain a Post Accident 
Sampling System (PASS). Licensees were generally required to implement 
PASS upgrades as described in NUREG-0737, ``Clarification of TMI [Three 
Mile Island] Action Plan Requirements,'' and Regulatory Guide 1.97, 
``Instrumentation for Light-Water-Cooled Nuclear Power Plants to Assess 
Plant and Environs Conditions During and Following an Accident.'' 
Implementation of these upgrades was an outcome of the lessons learned 
from the accident that occurred at TMI, Unit 2. Requirements related to 
PASS were imposed by Order for many facilities and were added to or 
included in the technical specifications (TS) for nuclear power 
reactors currently licensed to operate. Lessons learned and 
improvements implemented over the last 20 years have shown that the 
information obtained from PASS can be readily obtained through other 
means or is of little use in the assessment and mitigation of accident 
conditions.
    The NRC staff issued a notice of opportunity for comment in the 
Federal Register on August 11, 2000 (65 FR 49271) on possible 
amendments to eliminate PASS, including a model safety evaluation and 
model no significant hazards consideration (NSHC) determination, using 
the consolidated line item improvement process. The NRC staff 
subsequently issued a notice of availability of the models for 
referencing in license amendment applications in the Federal Register 
on October 31, 2000 (65 FR 65018). The licensee affirmed the 
applicability of the following NSHC determination in its application 
dated July 31, 2001.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), an analysis of the issue 
of no significant hazards consideration is presented below:

Criterion 1--The Proposed Change Does Not Involve a Significant 
Increase in the Probability or Consequences of an Accident 
Previously Evaluated.

    The PASS was originally designed to perform many sampling and 
analysis functions. These functions were designed and intended to be 
used in post accident situations and were put into place as a result 
of the TMI-2 accident. The specific intent of the PASS was to 
provide a system that has the capability to obtain and analyze 
samples of plant fluids containing potentially high levels of 
radioactivity, without exceeding plant personnel radiation exposure 
limits. Analytical results of these samples would be used largely 
for verification purposes in aiding the plant staff in assessing the 
extent of core damage and subsequent offsite radiological dose 
projections. The system was not intended to and does not serve a 
function for preventing accidents and its elimination would not 
affect the probability of accidents previously evaluated.
    In the 20 years since the TMI-2 accident and the consequential 
promulgation of post accident sampling requirements, operating 
experience has demonstrated that a PASS provides little actual 
benefit to post accident mitigation. Past experience has indicated 
that there exists in-plant instrumentation and methodologies 
available in lieu of a PASS for collecting and assimilating 
information

[[Page 55010]]

needed to assess core damage following an accident. Furthermore, the 
implementation of Severe Accident Management Guidance (SAMG) 
emphasizes accident management strategies based on in-plant 
instruments. These strategies provide guidance to the plant staff 
for mitigation and recovery from a severe accident. Based on current 
severe accident management strategies and guidelines, it is 
determined that the PASS provides little benefit to the plant staff 
in coping with an accident.
    The regulatory requirements for the PASS can be eliminated 
without degrading the plant emergency response. The emergency 
response, in this sense, refers to the methodologies used in 
ascertaining the condition of the reactor core, mitigating the 
consequences of an accident, assessing and projecting offsite 
releases of radioactivity, and establishing protective action 
recommendations to be communicated to offsite authorities. The 
elimination of the PASS will not prevent an accident management 
strategy that meets the initial intent of the post-TMI-2 accident 
guidance through the use of the SAMGs, the emergency plan (EP), the 
emergency operating procedures (EOP), and site survey monitoring 
that support modification of emergency plan protective action 
recommendations (PARs).
    Therefore, the elimination of PASS requirements from Technical 
Specifications (TS) (and other elements of the licensing bases) does 
not involve a significant increase in the consequences of any 
accident previously evaluated.

Criterion 2--The Proposed Change Does Not Create the Possibility of 
a New or Different Kind of Accident from any Previously Evaluated.

    The elimination of PASS related requirements will not result in 
any failure mode not previously analyzed. The PASS was intended to 
allow for verification of the extent of reactor core damage and also 
to provide an input to offsite dose projection calculations. The 
PASS is not considered an accident precursor, nor does its existence 
or elimination have any adverse impact on the pre-accident state of 
the reactor core or post accident confinement of radionuclides 
within the containment building.
    Therefore, this change does not create the possibility of a new 
or different kind of accident from any previously evaluated.

Criterion 3--The Proposed Change Does Not Involve a Significant 
Reduction in the Margin of Safety.

    The elimination of the PASS, in light of existing plant 
equipment, instrumentation, procedures, and programs that provide 
effective mitigation of and recovery from reactor accidents, results 
in a neutral impact to the margin of safety. Methodologies that are 
not reliant on PASS are designed to provide rapid assessment of 
current reactor core conditions and the direction of degradation 
while effectively responding to the event in order to mitigate the 
consequences of the accident. The use of a PASS is redundant and 
does not provide quick recognition of core events or rapid response 
to events in progress. The intent of the requirements established as 
a result of the TMI-2 accident can be adequately met without 
reliance on a PASS.
    Therefore, this change does not involve a significant reduction 
in the margin of safety. Based upon the reasoning presented above 
and the previous discussion of the amendment request, the requested 
change does not involve a significant hazards consideration.

    The NRC staff proposes to determine that the amendment request 
involves no significant hazards consideration.
    Attorney for licensee: Lillian M. Cuoco, Senior Nuclear Counsel, 
Dominion Nuclear Connecticut, Inc., Rope Ferry Road, Waterford, CT 
06385.
    NRC Section Chief: James W. Clifford.

Dominion Nuclear Connecticut, Inc., Docket No. 50-336, Millstone 
Nuclear Power Station, Unit No. 2, New London County, Connecticut

    Date of amendment request: August 27, 2001.
    Description of amendment request: The proposed amendment would 
revise the Technical Specification action and surveillance requirements 
associated with the containment air lock and expand the current 
guidance provided to address inoperable air lock components.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Involve a significant increase in the probability or 
consequences of an accident previously evaluated.
    The proposed Technical Specification changes to revise the 
action and surveillance requirements associated with the containment 
air lock will not cause an accident to occur and will not result in 
any change in the operation of the associated accident mitigation 
equipment. The containment air lock is not an accident initiator. 
The proposed changes will not revise the operability requirements 
(e.g., leakage limits) for the containment air lock. Proper 
operation of the containment air lock will still be verified. As a 
result, the design basis accidents will remain the same postulated 
events described in the Millstone Unit No. 2 Final Safety Analysis 
Report, and the consequences of the design basis accidents will 
remain the same. Therefore, the proposed changes will not increase 
the probability or consequences of an accident previously evaluated.
    2. Create the possibility of a new or different kind of accident 
from any accident previously evaluated.
    The proposed changes to the Technical Specifications do not 
impact any system or component that could cause an accident. The 
proposed changes will not alter the plant configuration (no new or 
different type of equipment will be installed) or require any 
unusual operator actions. The proposed changes will not alter the 
way any structure, system, or component functions, and will not 
significantly alter the manner in which the plant is operated. The 
response of the plant and the operators following an accident will 
not be different. In addition, the proposed changes do not introduce 
any new failure modes. Therefore, the proposed changes will not 
create the possibility of a new or different kind of accident from 
any accident previously analyzed.
    3. Involve a significant reduction in a margin of safety.
    The proposed Technical Specification changes to revise the 
action and surveillance requirements associated with the containment 
air lock will not cause an accident to occur and will not result in 
any change in the operation of the associated accident mitigation 
equipment. The operability requirements for the containment air lock 
have not been changed. The containment air lock will continue to 
function as assumed in the safety analysis. In addition, the 
proposed changes will not adversely affect equipment design or 
operation, and there are no changes being made to the Technical 
Specification required safety limits or safety system settings that 
would adversely affect plant safety. Therefore, the proposed changes 
will not result in a reduction in a margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: Lillian M. Cuoco, Senior Nuclear Counsel, 
Dominion Nuclear Connecticut, Inc., Rope Ferry Road, Waterford, CT 
06385.
    NRC Section Chief: James W. Clifford.

Dominion Nuclear Connecticut, Inc., Docket No. 50-336, Millstone 
Nuclear Power Station, Unit No. 2, New London County, Connecticut

    Date of amendment request: August 28, 2001.
    Description of amendment request: The proposed amendment would 
revise the Technical Specifications by removing the surveillance 
requirement that verifies the automatic opening features of the safety 
injection tank outlet isolation valves.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Involve a significant increase in the probability or 
consequences of an accident previously evaluated.
    The proposed Technical Specification change to remove the 
surveillance

[[Page 55011]]

requirement that verifies the automatic opening features of the 
safety injection tank outlet isolation valves will not cause an 
accident to occur since the safety injection tanks and associated 
isolation valves are not accident initiators. In addition, the 
proposed change will not alter the operation of the associated 
accident mitigation equipment. The operability requirement for the 
safety injection tank outlet isolation valves to be deenergized open 
when the safety injection tanks are required to be operable will not 
be affected, and outlet isolation valve position will still be 
verified periodically. As a result, the design basis accidents will 
remain the same postulated events described in the Millstone Unit 
No. 2 Final Safety Analysis Report, and the consequences of the 
design basis accidents will remain the same. Therefore, the proposed 
change will not increase the probability or consequences of an 
accident previously evaluated.
    2. Create the possibility of a new or different kind of accident 
from any accident previously evaluated.
    The proposed Technical Specification change does not impact any 
system or component that could cause an accident. The proposed 
change will not alter the plant configuration (no new or different 
type of equipment will be installed) 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. The response of the plant and 
the operators following an accident will not be different. In 
addition, the proposed change does not introduce any new failure 
modes. Therefore, the proposed change will not create the 
possibility of a new or different kind of accident from any accident 
previously analyzed.
    3. Involve a significant reduction in a margin of safety.
    The proposed Technical Specification change to remove the 
surveillance requirement that verifies the automatic opening 
features of the safety injection tank outlet isolation valves will 
not cause an accident to occur and will not result in any change in 
the operation of the associated accident mitigation equipment. The 
proposed change will not revise the operability requirement for the 
safety injection tank outlet isolation valves to be deenergized open 
when the safety injection tanks are required to be operable. The 
safety injection tanks will continue to be able to mitigate the 
design basis accidents as assumed in the safety analysis. In 
addition, the proposed change will not adversely affect equipment 
design or operation, and there are no changes being made to the 
Technical Specification required safety limits or safety system 
settings that would adversely affect plant safety. Therefore, the 
proposed change will not result in a reduction in a margin of 
safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: Lillian M. Cuoco, Senior Nuclear Counsel, 
Dominion Nuclear Connecticut, Inc., Rope Ferry Road, Waterford, CT 
06385.
    NRC Section Chief: James W. Clifford.

Dominion Nuclear Connecticut Inc., et al., Docket Nos. 50-336 and 50-
423, Millstone Nuclear Power Station, Unit Nos. 2 and 3, New London 
County, Connecticut

    Date of amendment request: June 4, 2001.
    Description of amendment request: The proposed amendment modifies 
the Millstone Nuclear Power Station, Unit No. 2 (MP2) and Unit No. 3 
(MP3) Technical Specifications (TSs) to relocate selected MP2 and MP3 
technical specifications related to the reactor coolant system to the 
respective Technical Requirements Manual (TRM), with the exception of 
MP3 Technical Specification section 4.4.10, which will be relocated to 
section 6 of MP3's TS. The Bases of the affected TSs will be modified 
to address the proposed 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. The NRC staff reviewed the licensee's analysis against 
the standards of 10 CFR 50.92(c). The NRC staff's analysis, which is 
based on the representation made by the licensee in the June 4, 2001, 
application, is presented below:
    1. Involve a significant increase in the probability or 
consequences of an accident previously evaluated.
    The proposed requirements remain the same except that the 
requirements will be relocated to the TRM. Since the proposed 
requirements are the same, this proposed change will not increase the 
probability or consequences of an accident previously evaluated.
    2. Create the possibility of a new or different kind of accident 
from any accident previously evaluated.
    Since the requirements remain the same, these proposed changes do 
not alter the way any system, structure, or component functions and do 
not alter the manner in which the plant is operated. The proposed 
changes do not introduce any new failure modes. Therefore, the proposed 
changes will not create the possibility of a new or different kind of 
accident from any accident previously evaluated.
    3. Involve a significant reduction in a margin of safety.
    Since the proposed changes are solely to relocate the existing 
requirements, it does not affect plant operation in any way. Therefore, 
the proposed change will not result in a reduction in a margin of 
safety.
    Based on the NRC staff's analysis, 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: Lillian M. Cuoco, Senior Nuclear Counsel, 
Dominion Nuclear Connecticut, Inc., Hartford, CT 06141-5127.
    NRC Section Chief: James W. Clifford.

Dominion Nuclear Connecticut Inc., et al., Docket No. 50-423, Millstone 
Nuclear Power Station, Unit No. 3, New London County, Connecticut

    Date of amendment request: July 31, 2001.
    Description of amendment request: The proposed amendment deletes 
requirements from the Technical Specifications (and, as applicable, 
other elements of the licensing bases) to maintain a Post Accident 
Sampling System (PASS). Licensees were generally required to implement 
PASS upgrades as described in NUREG-0737, ``Clarification of TMI [Three 
Mile Island] Action Plan Requirements,'' and Regulatory Guide 1.97, 
``Instrumentation for Light-Water-Cooled Nuclear Power Plants to Assess 
Plant and Environs Conditions During and Following an Accident.'' 
Implementation of these upgrades was an outcome of the lessons learned 
from the accident that occurred at TMI, Unit 2. Requirements related to 
PASS were imposed by Order for many facilities and were added to or 
included in the technical specifications (TS) for nuclear power 
reactors currently licensed to operate. Lessons learned and 
improvements implemented over the last 20 years have shown that the 
information obtained from PASS can be readily obtained through other 
means or is of little use in the assessment and mitigation of accident 
conditions.
    The NRC staff issued a notice of opportunity for comment in the 
Federal Register on August 11, 2000 (65 FR 49271) on possible 
amendments to eliminate PASS, including a model safety evaluation and 
model no significant hazards consideration (NSHC) determination, using 
the consolidated line item improvement process. The NRC staff 
subsequently issued a notice of availability of the models for 
referencing in license amendment applications in the Federal Register 
on October 31, 2000 (65 FR 65018). The licensee affirmed the 
applicability of the following NSHC

[[Page 55012]]

determination in its application dated July 31, 2001.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), an analysis of the issue 
of no significant hazards consideration is presented below:

Criterion 1--The Proposed Change Does Not Involve a Significant 
Increase in the Probability or Consequences of an Accident 
Previously Evaluated.

    The PASS was originally designed to perform many sampling and 
analysis functions. These functions were designed and intended to be 
used in post accident situations and were put into place as a result 
of the TMI-2 accident. The specific intent of the PASS was to 
provide a system that has the capability to obtain and analyze 
samples of plant fluids containing potentially high levels of 
radioactivity, without exceeding plant personnel radiation exposure 
limits. Analytical results of these samples would be used largely 
for verification purposes in aiding the plant staff in assessing the 
extent of core damage and subsequent offsite radiological dose 
projections. The system was not intended to and does not serve a 
function for preventing accidents and its elimination would not 
affect the probability of accidents previously evaluated.
    In the 20 years since the TMI-2 accident and the consequential 
promulgation of post accident sampling requirements, operating 
experience has demonstrated that a PASS provides little actual 
benefit to post accident mitigation. Past experience has indicated 
that there exists in-plant instrumentation and methodologies 
available in lieu of a PASS for collecting and assimilating 
information needed to assess core damage following an accident. 
Furthermore, the implementation of Severe Accident Management 
Guidance (SAMG) emphasizes accident management strategies based on 
in-plant instruments. These strategies provide guidance to the plant 
staff for mitigation and recovery from a severe accident. Based on 
current severe accident management strategies and guidelines, it is 
determined that the PASS provides little benefit to the plant staff 
in coping with an accident.
    The regulatory requirements for the PASS can be eliminated 
without degrading the plant emergency response. The emergency 
response, in this sense, refers to the methodologies used in 
ascertaining the condition of the reactor core, mitigating the 
consequences of an accident, assessing and projecting offsite 
releases of radioactivity, and establishing protective action 
recommendations to be communicated to offsite authorities. The 
elimination of the PASS will not prevent an accident management 
strategy that meets the initial intent of the post-TMI-2 accident 
guidance through the use of the SAMGs, the emergency plan (EP), the 
emergency operating procedures (EOP), and site survey monitoring 
that support modification of emergency plan protective action 
recommendations (PARs).
    Therefore, the elimination of PASS requirements from Technical 
Specifications (TS) (and other elements of the licensing bases) does 
not involve a significant increase in the consequences of any 
accident previously evaluated.

Criterion 2--The Proposed Change Does Not Create the Possibility of 
a New or Different Kind of Accident from any Previously Evaluated.

    The elimination of PASS related requirements will not result in 
any failure mode not previously analyzed. The PASS was intended to 
allow for verification of the extent of reactor core damage and also 
to provide an input to offsite dose projection calculations. The 
PASS is not considered an accident precursor, nor does its existence 
or elimination have any adverse impact on the pre-accident state of 
the reactor core or post accident confinement of radionuclides 
within the containment building.
    Therefore, this change does not create the possibility of a new 
or different kind of accident from any previously evaluated.

    Criterion 3--The Proposed Change Does Not Involve a Significant 
Reduction in the Margin of Safety.

    The elimination of the PASS, in light of existing plant 
equipment, instrumentation, procedures, and programs that provide 
effective mitigation of and recovery from reactor accidents, results 
in a neutral impact to the margin of safety. Methodologies that are 
not reliant on PASS are designed to provide rapid assessment of 
current reactor core conditions and the direction of degradation 
while effectively responding to the event in order to mitigate the 
consequences of the accident. The use of a PASS is redundant and 
does not provide quick recognition of core events or rapid response 
to events in progress. The intent of the requirements established as 
a result of the TMI-2 accident can be adequately met without 
reliance on a PASS.
    Therefore, this change does not involve a significant reduction 
in the margin of safety.
    Based upon the reasoning presented above and the previous 
discussion of the amendment request, the requested change does not 
involve a significant hazards consideration.

    The NRC staff proposes to determine that the amendment request 
involves no significant hazards consideration.
    Attorney for licensee: Lillian M. Cuoco, Senior Nuclear Counsel, 
Dominion Nuclear Connecticut, Inc., Rope Ferry Road, Waterford, CT 
06385.
    NRC Section Chief: James W. Clifford.

Entergy Nuclear Operations, Docket No. 50-247, Indian Point Nuclear 
Generating Unit No. 2, Westchester County, New York

    Date of amendment request: September 20, 2001.
    Description of amendment request: The proposed amendment would 
revise Technical Specification (TS) 3.8, ``Refueling, Fuel Storage and 
Operations with the Reactor Vessel Head Bolts Less Than Fully 
Tensioned,'' TS Table 4.1-2, ``Frequencies for Sampling Tests,'' and TS 
Section 5.4, ``Fuel Storage,'' to allow the credit for soluble boron in 
the criticality analysis of the spent fuel pit (SFP). The revisions 
also incorporate changes to the SFP rack layout by dividing it into 
sub-regions and specifying requirements for fuel assembly burnup and 
soluble boron concentration for various loading configurations in these 
sub-regions.
    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.
    Current TS contain minimum requirements for the SFP boron 
concentration. The actual boron concentration in the SFP has been 
maintained at a higher value. The proposed changes to the TS 
establish new boron concentration requirements for the SFP water 
that are consistent with the new criticality analysis. Since soluble 
boron has already been maintained in the SFP water and is currently 
required by the TS, the implementation of this new requirement will 
have no effect on the normal SFP operations and maintenance.
    The presence of an increased requirement for soluble boron in 
the SFP water does not increase the probability of a fuel assembly 
drop accident in the SFP. The handling of the fuel assemblies in the 
SFP has always been performed in borated water. The criticality 
analysis shows the consequences of a fuel assembly drop accident in 
the SFP are not affected when considering the presence of soluble 
boron since the rack keff remains  0.95.
    Fuel assembly placement will continue to be controlled in 
accordance with approved fuel handling procedures and will be in 
accordance with TS spent fuel rack storage configuration 
limitations. The proposed SFP storage configuration limitations will 
be more complex but will be similar to those previously approved. 
Therefore, the new limitations will not significantly increase the 
probability of accident occurrence. There is no increase in the 
consequences of the accidental misloading of spent fuel assemblies 
into the spent fuel racks since the criticality analysis 
demonstrates that the SFP keff will remain  
0.95 following an accidental misloading.
    There is no increase in the probability of the loss of normal 
cooling to the spent fuel pit water when considering the presence of 
soluble boron in the pit water for subcriticality control since a 
high concentration of soluble boron has always been maintained in 
the SFP water.
    Soluble boron requirements for mitigating reactivity effects due 
to increased pool temperatures are adequately met by the proposed 
increase in minimum TS soluble boron concentration. A negligible 
increase in

[[Page 55013]]

the probability of a criticality accident due to increased pool 
temperature exists with the proposed TS changes, as the minimum 
soluble boron concentration will not change. The positive reactivity 
introduced as a result of the higher TS boron concentration effect 
on moderator reactivity coefficient will be sufficiently mitigated 
by the substantial margin to the amount actually required to 
maintain keff  0.95.
    Decreased fuel temperatures will increase the water density in 
the SFP, therefore increasing the thermal neutron flux, possibly 
causing an increase in reactivity. This density increase will 
increase the differential worth of the soluble boron but the excess 
soluble boron in the SFP is more than sufficient to offset any 
reactivity increase introduced by a temperature decrease.
    Therefore, the proposed changes do 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.
    Spent fuel handling accidents are not new or different types of 
accidents, they have been analyzed for the UFSAR [Updated Final 
Safety Analysis Report] and in Criticality Analysis reports 
associated with License Amendment 150 up to the nominal 5.0 \w\/o 
\235\ U [weight percent uranium-235] that is assumed for the 
proposed change.
    A dilution of the SFP soluble boron has always been a 
possibility. However the boron dilution event previously had no 
consequences since boron was not previously credited. With the 
proposed TS, credit is taken for soluble boron. So a boron dilution 
has been evaluated as a possible new accident. The evaluation 
concluded a boron dilution accident was not credible, that processes 
were in place to detect and mitigate the possible events, and that, 
even if the SFP boron concentration was diluted to zero, criticality 
would not occur. Therefore, there would be no additional hazards if 
this request were approved.
    There is no other change in the plant configuration or equipment 
design.
    Therefore, the proposed change does not create a new accident 
initiator or precursor, or 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 [a] margin of 
safety.
    The TS changes proposed by this LAR [license amendment request] 
and the resulting spent fuel storage operation limits will provide 
adequate safety margin to ensure that the stored fuel assembly array 
will always remain subcritical. These limits are based on the plant 
specific criticality analysis and boron dilution analysis. The 
proposed TS changes rely upon known and predictable reactivity 
effects to ensure required criticality margins in the SFP.
    While the criticality analysis utilizes credit for soluble 
boron, storage configurations have been defined using 95/95 
keff calculations to ensure the spent fuel rack 
keff  will be 1.0 with no soluble boron. Soluble boron 
credit is used to offset uncertainties, tolerances, and off-normal 
conditions and to provide subcritical margin such that the SFP 
keff  is maintained 0.95.
    The loss of substantial amounts of soluble boron from the SFP, 
which could lead to keff  exceeding 0.95, has been 
evaluated and shown to be not credible. An evaluation has been 
performed that shows that the dilution of the SFP boron 
concentration from 2000 ppm [parts per million] to 786 ppm is not 
credible. Also the spent fuel rack keff  will remain 1.0 
with the SFP flooded with unborated water. These safety analyses 
demonstrate a level of safety comparable to the conservative 
criticality analysis approved for License Amendment 150 and show 
that the requirements of 10CFR50.68 are met.
    The reactivity credit for additional poisons in the spent and 
fresh fuel assemblies increases the margin of safety in the SFP. No 
credit is taken for Boraflex in certain regions, when in reality 
some residual Boraflex does remain in these regions. In regions that 
do take credit for Boraflex, the amount of credit is conservative. 
These conservatisms add an increased safety margin. Predictions of 
the effective neutron multiplication factors have shown that, under 
the worst of scenarios, the SFP remains subcritical when 
conservative credit for future expected loss of Boraflex poison 
plates is considered.
    The analysis show that the level of safety required by 
10CFR50.68 is achieved for the IP2 [Indian Point 2] SFP with the 
proposed TS.
    Therefore, operation of the facility in accordance with the 
proposed amendment would not involve a significant reduction in [a] 
margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: Mr. John Fulton, Assistant General Counsel, 
Entergy Nuclear Generating Co., Pilgrim Station, 600 Rocky Hill Road, 
Plymouth, MA 02360.
    NRC Section Chief: L. Raghavan, Acting.

Entergy Nuclear Operations, Docket No. 50-247, Indian Point Nuclear 
Generating Unit No. 2, Westchester County, New York 

    Date of amendment request: September 20, 2001.
    Description of amendment request: The proposed amendment deletes 
requirements from the Technical Specifications (TSs) (and, as 
applicable, other elements of the licensing bases) to maintain a Post 
Accident Sampling System (PASS). Licensees were generally required to 
implement PASS upgrades as described in NUREG-0737, ``Clarification of 
TMI [Three Mile Island] Action Plan Requirements,'' and Regulatory 
Guide 1.97, ``Instrumentation for Light-Water-Cooled Nuclear Power 
Plants to Assess Plant and Environs Conditions During and Following an 
Accident.'' Implementation of these upgrades was an outcome of the 
lessons learned from the accident that occurred at TMI, Unit 2. 
Requirements related to PASS were imposed by Order for many facilities 
and were added to or included in the TSs for nuclear power reactors 
currently licensed to operate. Lessons learned and improvements 
implemented over the last 20 years have shown that the information 
obtained from PASS can be readily obtained through other means or is of 
little use in the assessment and mitigation of accident conditions.
    The NRC staff issued a notice of opportunity for comment in the 
Federal Register on August 11, 2000 (65 FR 49271) on possible 
amendments to eliminate PASS, including a model safety evaluation and 
model no significant hazards consideration (NSHC) determination, using 
the consolidated line item improvement process. The NRC staff 
subsequently issued a notice of availability of the models for 
referencing in license amendment applications in the Federal Register 
on October 31, 2000 (65 FR 65018). The licensee affirmed the 
applicability of the following NSHC determination in its application 
dated September 20, 2001.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), an analysis of the issue 
of no significant hazards consideration is presented below:

Criterion 1--The Proposed Change Does Not Involve a Significant 
Increase in the Probability or Consequences of an Accident 
Previously Evaluated.

    The PASS was originally designed to perform many sampling and 
analysis functions. These functions were designed and intended to be 
used in post accident situations and were put into place as a result 
of the TMI-2 [Three Mile Island 2] accident. The specific intent of 
the PASS was to provide a system that has the capability to obtain 
and analyze samples of plant fluids containing potentially high 
levels of radioactivity, without exceeding plant personnel radiation 
exposure limits. Analytical results of these samples would be used 
largely for verification purposes in aiding the plant staff in 
assessing the extent of core damage and subsequent offsite 
radiological dose projections. The system was not intended to and 
does not serve a function for preventing accidents and its 
elimination would not affect the probability of accidents previously 
evaluated.

[[Page 55014]]

    In the 20 years since the TMI-2 accident and the consequential 
promulgation of post accident sampling requirements, operating 
experience has demonstrated that a PASS provides little actual 
benefit to post accident mitigation. Past experience has indicated 
that there exists in-plant instrumentation and methodologies 
available in lieu of a PASS for collecting and assimilating 
information needed to assess core damage following an accident. 
Furthermore, the implementation of Severe Accident Management 
Guidance (SAMG) emphasizes accident management strategies based on 
in-plant instruments. These strategies provide guidance to the plant 
staff for mitigation and recovery from a severe accident. Based on 
current severe accident management strategies and guidelines, it is 
determined that the PASS provides little benefit to the plant staff 
in coping with an accident.
    The regulatory requirements for the PASS can be eliminated 
without degrading the plant emergency response. The emergency 
response, in this sense, refers to the methodologies used in 
ascertaining the condition of the reactor core, mitigating the 
consequences of an accident, assessing and projecting offsite 
releases of radioactivity, and establishing protective action 
recommendations to be communicated to offsite authorities. The 
elimination of the PASS will not prevent an accident management 
strategy that meets the initial intent of the post-TMI-2 accident 
guidance through the use of the SAMGs, the emergency plan (EP), the 
emergency operating procedures (EOP), and site survey monitoring 
that support modification of emergency plan protective action 
recommendations (PARs).
    Therefore, the elimination of PASS requirements from Technical 
Specifications (TS) (and other elements of the licensing bases) does 
not involve a significant increase in the consequences of any 
accident previously evaluated.

Criterion 2--The Proposed Change Does Not Create the Possibility of 
a New or Different Kind of Accident from any Previously Evaluated.

    The elimination of PASS related requirements will not result in 
any failure mode not previously analyzed. The PASS was intended to 
allow for verification of the extent of reactor core damage and also 
to provide an input to offsite dose projection calculations. The 
PASS is not considered an accident precursor, nor does its existence 
or elimination have any adverse impact on the pre-accident state of 
the reactor core or post accident confinement of radionuclides 
within the containment building.
    Therefore, this change does not create the possibility of a new 
or different kind of accident from any previously evaluated.

Criterion 3--The Proposed Change Does Not Involve a Significant 
Reduction in [a] Margin of Safety.

    The elimination of the PASS, in light of existing plant 
equipment, instrumentation, procedures, and programs that provide 
effective mitigation of and recovery from reactor accidents, results 
in a neutral impact to the margin of safety. Methodologies that are 
not reliant on PASS are designed to provide rapid assessment of 
current reactor core conditions and the direction of degradation 
while effectively responding to the event in order to mitigate the 
consequences of the accident. The use of a PASS is redundant and 
does not provide quick recognition of core events or rapid response 
to events in progress. The intent of the requirements established as 
a result of the TMI-2 accident can be adequately met without 
reliance on a PASS.
    Therefore, this change does not involve a significant reduction 
in [a] margin of safety.
    Based upon the reasoning presented above and the previous 
discussion of the amendment request, the requested change does not 
involve a significant hazards consideration.

    The NRC staff proposes to determine that the amendment request 
involves no significant hazards consideration.
    Attorney for licensee: Mr. John Fulton, Assistant General Counsel, 
Entergy Nuclear Operations, Inc., 440 Hamilton Avenue, White Plains, NY 
10601.
    NRC Section Chief: L. Raghavan, Acting.

Entergy Nuclear Operations, Docket No. 50-247, Indian Point Nuclear 
Generating Unit No. 2, Westchester County, New York 

    Date of amendment request: September 20, 2001.
    Description of amendment request: The proposed amendment would 
allow the one-time extension of the intervals for selected Technical 
Specification (TS) surveillance requirements (SRs) to enable the tests 
to be performed during the next refueling outage starting no later than 
November 19, 2002. Specifically, the surveillance interval would be 
extended for certain SRs associated with the volume control tank (VCT), 
residual heat removal system (RHR) , emergency diesel generators 
(EDGs), and shock suppressors (snubbers). In addition, the proposed 
amendment would: (1) Correct the channel functional test interval in 
Items 3 and 4 of TS Table 4.10-4 and Items 4 and 5 of Table 4-10-4, (2) 
delete alternate inspection requirements for the steam generator 
snubbers, and (3) remove the reference to a prior one-time extension of 
checks, calibrations and tests for certain instrument channels in TS 
Table 4.1-1 that is no longer applicable.
    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 consequences of an accident 
previously evaluated?
    There is no change to the design, function, or capability of any 
plant structure, system, or component as a result of the proposed 
surveillance interval extensions. Hence there is no change in the 
probability of occurrence of an accident previously evaluated.
    The proposed surveillance interval extensions do not affect the 
ability of any plant structure, system, or component to mitigate the 
consequences of any accident previously evaluated. The surveillance 
interval extensions do not alter or prevent the ability of the 
affected structures, systems, and components to perform their 
intended functions.

[VCT]

    A statistical analysis of channel uncertainty for a proposed 31 
month operating cycle has been performed. It confirms that the 
channel drift for the proposed 31 month interval is bounded by the 
existing drift allowance used in the current uncertainty 
calculations. Therefore, there is no expected decrease in 
reliability for the VCT level channel for the proposed 31 month 
operating cycle. Since there is no expected decrease in the 
reliability of the VCT level channels, the design safety functions 
of the VCT are not affected.

[RHR]

    Since the past test data supports the integrity of the system 
and an extended standby period is not expected to affect any 
potential leak path, there is a reasonable expectation that the RHR 
and Safety Injection systems will continue to perform their intended 
safety functions without excessive leakage. It is concluded that a 
one-time extension of less than one month for the leakage test 
surveillance intervals will have minimal impact on the system 
reliability.

[EDG]

    The identified anomalies with valve and filter operation for EDG 
23 were evaluated and corrected and are not indicative of any 
inability of the machine to meet performance requirements. The 
anomalous adjustment affecting movement of the fuel control lever 
arm for EDG 22 was properly evaluated and eliminated as evidenced by 
subsequent successful testing. Therefore, the historical data 
together with the positive verification of the adequacy of 
corrective actions for previous test failures demonstrate that the 
EDGs have met the required performance criteria. Therefore the 
ability of the EDGs to mitigate accidents is not affected by this 
proposed change.
    Failure of an EDG cannot, of itself, initiate an accident.

[Snubbers]

    The TS functional testing program requires a sampling program 
that provides a 95% confidence level that 90-100% of the snubbers 
operate within acceptance limits. For each snubber failing the 
functional test an additional sample lot must be selected and tested 
to assure that the required confidence level is maintained. The past 
functional test history with very few functional test failures 
provides assurance that an extension in the surveillance will not

[[Page 55015]]

result in increased snubber failures. In all cases, the functional 
test failures were thoroughly analyzed and appropriate action was 
taken to prevent recurrence. Subsequent testing resulted in all 
snubbers meeting their design requirements.
    The operability of snubbers is not affected by the deletion of 
the allowance to separately group steam generator snubbers for the 
purposes of determining inspection intervals.
    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) Does the proposed license amendment create the possibility 
of a new or different kind of accident from any accident previously 
evaluated?
    The proposed changes do not involve any physical design change 
or operational change to any plant system, structure or component. 
Thus a new failure mode is not introduced. Therefore, the proposed 
changes do not create a new accident initiator or precursor, or 
create the possibility of a new or different kind of accident from 
any accident previously evaluated.

[VCT]

    The proposed change does not involve the addition of any new or 
different type of equipment, nor does it involve operating equipment 
required for safe operation of the facility in a manner that is 
different from that addressed in the Updated Final Safety Analysis 
Report (UFSAR). The proposed change in the surveillance interval has 
been evaluated to have a negligible effect on the reliability of the 
existing instruments.

[RHR]

    The proposed change does not involve the addition of any new or 
different type of equipment. Nor does it involve operating equipment 
required for safe operation of the facility in a manner that is 
different from that addressed in the UFSAR.

[EDG]

    The proposed change does not involve the addition of any new or 
different type of equipment, nor does it involve operating equipment 
required for safe operation of the facility in a manner that is 
different from that addressed in the UFSAR. Also, the increased 
surveillance interval (one-time only) will not adversely affect the 
reliability of the EDGs.

[Snubbers]

    The proposed license amendment does not create the possibility 
of a new or different kind of accident from any previously 
evaluated. The proposed change does not involve the addition of any 
new or different type of equipment, nor does it involve operating 
equipment required for safe operation of the facility in a manner 
that is different from that addressed in the UFSAR. Also, the 
increased surveillance interval (one-time only) will not adversely 
affect the snubbers.
    (3) Does the proposed amendment involve a significant reduction 
in a margin of safety?
    As a result of these proposed surveillance interval extensions, 
there are no changes to IP2's design or to the IP2 TS safety limits, 
limiting safety system settings, or limiting conditions for 
operation. The only change is a change to the surveillance testing 
frequency for affected structures, systems, and components.
    The proposed surveillance interval extensions have been 
evaluated to not significantly degrade the reliability of any 
existing system, structure, or component. Therefore, testing in 
accordance with the proposed test intervals continues to ensure that 
the necessary quality of affected structures, systems, and 
components is maintained, that IP2 operation will be within safety 
limits, and that the IP2 limiting conditions for operation will be 
met.
    The proposed surveillance interval extensions do not adversely 
affect the ability of any IP2 structures, systems, or components to 
function when required to mitigate any accident or licensing basis 
event.

[VCT]

    The proposed change in surveillance interval resulting from an 
increased operating cycle will not result in a channel statistical 
allowance that impacts any TS limit or any UFSAR requirement. 
Protective functions will continue to occur so that safety analysis 
limits are not exceeded.
    Based on past rest results, the one-time extension of nine days 
does not involve a significant reduction in a margin of safety.

[RHR]

    There is minimal risk that a surveillance interval extension of 
less than one month will increase leakage in the piping systems 
under review beyond the TS limits or that the system performance 
will be influenced. Past test data indicate that there was no impact 
on the margin imposed by the TS.

[EDG]

    The functional test history indicates the functional test 
failures were the result of actions independent of actual EDG load 
performance. Apart from these anomalous actions, the record does not 
indicate a potential for failure to meet performance criteria. In 
all cases, the functional test failures were thoroughly analyzed and 
appropriate actions were taken to prevent recurrence.
    Subsequent testing resulted in the EDG meeting its design 
requirements.
    There is no reduction of margin indicated by the surveillance 
testing. The proposed change for a one-time extension of the test 
interval does not adversely affect the performance of any safety 
related system, component or structure and does not result in 
increased severity of any of the accidents considered in the UFSAR. 
Surveillance test results indicate no trend toward margin reduction.

[Snubbers]

    The objective of the functional test is to provide a 95% 
confidence level that 90-100% of the snubbers operate within the 
specified acceptance limits. The review of past test history 
indicates that this objective was met at the time of the testing. 
There are no identified trends that would suggest that the same 
success rate would not be maintained over the requested extension 
period. The proposed license amendment does not involve a 
significant reduction in a margin of safety. The proposed change for 
a one-time extension of the test interval does not adversely affect 
the performance of any safety related system, component or structure 
and does not result in increased severity of any of the accidents 
considered in the UFSAR.
    Therefore, the one-time extension of less than one month for the 
functional tests does not involve a significant reduction in a 
margin of safety.
    The proposed deletion of the allowance to separately group steam 
generator snubbers for the purpose of determining inspection 
intervals does not affect the effectiveness of the surveillance 
requirements. The steam generator snubbers will still be inspected 
at the interval required by the TS.
    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: Mr. John Fulton, Assistant General Counsel, 
Entergy Nuclear Generating Co., Pilgrim Station, 600 Rocky Hill Road, 
Plymouth, MA 02360.
    NRC Section Chief: L. Raghavan, Acting.

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

    Date of amendment request: October 2, 2001.
    Description of amendment request: The proposed amendment would 
revise Technical Specification (TS) Table 3.3-4, ``Engineered Safety 
Feature Actuation System Instrumentation Trip Values,'' Functional Unit 
7.b, ``Loss of Power, 460 volt Emergency Bus Undervoltage,'' by 
changing the referenced bus from the 460 volt (V) bus to the 480 V bus, 
by removing the trip setpoint, and by slightly increasing the range of 
allowable values for the degraded voltage setting and its associated 
time delay.
    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:

Criterion 1--Does Not Involve a Significant Increase in the 
Probability or Consequences of an Accident Previously Evaluated.

    The two degraded voltage protection relays that are provided on 
each of the 480 V safety buses act to mitigate the consequences of 
an

[[Page 55016]]

accident by detecting a sustained undervoltage condition, isolating 
the safety buses from offsite power, and starting the associated 
emergency diesel generator (EDG). This safety function is unchanged 
by the proposed allowable voltage setting revisions. The revised 
settings for the degraded voltage protection relays will continue to 
provide the safety function of protecting the associated Class 1E 
equipment from the effects of a low voltage condition. The time 
delays remain within those assumed in the ANO-2 [Arkansas Nuclear 
One, Unit 2] safety analyses. Additionally, the revised allowable 
voltage settings will not result in any unnecessary isolation from 
the off-site power sources. The relocation of trip setpoint values 
to station surveillance procedures allows operational flexibility to 
account for additional margins, drifts, or uncertainties while 
ensuring that the relays are set to actuate within the acceptable 
range of allowable values denoted in the TSs. Since the proposed 
change does not adversely impact the mitigating function of the 
relays, the consequences of an accident previously evaluated remains 
unchanged.
    The ANO-2 technical specifications will continue to require the 
480 V bus degraded voltage functions to be surveillance tested at 
their present frequency without changing the modes in which the 
surveillance is required or the modes of applicability for these 
components. The technical specifications will continue to require 
the same actions as currently exist for the inoperability of one or 
more of the 480 V bus degraded voltage relays.
    Therefore, this change does not involve a significant increase 
in the probability or consequences of any accident previously 
evaluated.

Criterion 2--Does Not Create the Possibility of a New or Different 
Kind of Accident from any Previously Evaluated.

    The proposed change introduces no new modes of plant operation 
or new plant configuration that could lead to a new or different 
kind of accident from any previously evaluated being introduced. The 
480 V bus degraded voltage relays are required to operate upon 
detection of a sustained undervoltage condition to protect the Class 
1E components from damage from low voltage by initiating transfer of 
the 4160 V safety bus power source to the EDG. This safety function 
remains unchanged by the proposed allowable voltage setting 
revisions, and the proposed values continue to provide the required 
actions consistent with the ANO-2 safety analysis.
    Therefore, this change does not create the possibility of a new 
or different kind of accident from any previously evaluated.

Criterion 3--Does Not Involve a Significant Reduction in the Margin 
of Safety.

    The two degraded voltage relays located on each 480 V safety bus 
are provided to detect sustained undervoltage, isolate the safety 
buses, and start the EDGs. This safety function remains unchanged by 
the proposed revisions to the allowable values. The proposed changes 
to the allowable values for the degraded voltage relays incorporate 
channel uncertainties and calibration tolerances, while fully 
meeting their required safety functions of degraded voltage 
protection without resulting in undesired tripping of the offsite 
power source.
    The slightly higher range of allowable values for the degraded 
voltage settings allows enhanced protection of the Class 1E 
components, but does not result in undesired tripping of the offsite 
power source for the analyzed grid minimum normal condition. In 
addition, the slight increase in the range of allowable values for 
the degraded voltage time delay remains well within the assumption 
of the 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 
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: October 2, 2001.
    Description of amendment request: The proposed amendment would 
relocate the technical specification (TS) requirement that the reactor 
core be subcritical for a minimum of 175 hours prior to discharge of 
more than 70 assemblies to the spent fuel pool (SFP), to the technical 
requirements manual (TRM).
    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 accident of concern related to the proposed change is the 
fuel handling accident. This accident assumes a dropped fuel 
assembly. One of the assumptions made in the analysis is that fuel 
movement is delayed at least 100 hours after shutdown to allow for 
radioactive decay of the fission product inventory. TS 3.9.3.a 
provides this restriction. The analysis does not assume any further 
delay in fuel movement following the initial 100-hour decay period. 
The relocation of TS 3.9.3.b will not impact this assumption.
    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 relocates TS 3.9.3.b to the TRM. There are 
no changes to the design or operation of the facility proposed. 
Thus, there are no new or different kinds of accidents created. SFP 
cooling capability and the heat load generated by the movement of 
fuel into the SFP will continue to be evaluated under 10 CFR 50.59. 
The SFP cooling system includes two cooling pumps and one heat 
exchanger. In addition, several systems are available for makeup 
when needed. Under postulated accident conditions, when no pool 
cooling systems are operational, the maximum temperature at the 
inlet to the cells is assumed to be equal to the saturation 
temperature at atmospheric pressure or 212F [Fahrenheit] (allowed to 
boil). The proposed change does not increase the possibility of a 
complete loss of pool cooling.
    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 change relocates TS 3.9.3.b to the TRM. Following 
relocation, any future changes to TRM 3.9.3.b will be assessed under 
the guidance of 10 CFR 50.59. The ANO [Arkansas Nuclear One] 50.59 
process will provide an evaluation to ensure heat loads transferred 
will be within the cooling capacity of the service water system. 
Analyses will continue to demonstrate that even in the event of a 
loss of SFP cooling, the maximum temperature in the pool is such 
that design limits associated with assuring the integrity of the 
fuel cladding are satisfied.
    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: October 2, 2001.
    Description of amendment request: The proposed amendment would 
change the technical specification definitions of response time for the 
reactor trip system (RTS) and for engineered safety features (ESFs) to 
allow use of either an allocated or a

[[Page 55017]]

measured response time for select sensors in these two systems.
    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?
    Response time testing is not an initiator of any previously 
evaluated accident. The proposed change to the definition of RTS and 
ESF response time allows substitution of an allocated response time 
for selected sensors in lieu of measuring the sensor response time. 
The allocated response times adequately represent the response time 
of the components such that the safety systems utilizing these 
components will continue to perform their accident mitigation 
function as assumed in the safety analysis. Response time testing 
for the non-sensor portions of the channels will continue to use a 
series of sequential or overlapping test measurements.
    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 involve a physical change to the 
plant. Modifications will not be made to existing components nor 
will any new or different types of equipment be installed. The 
proposed change modifies the definitions for RTS and ESF response 
time and allows the substitution of an allocated response time in 
lieu of measured sensor response time for selected sensors. The 
response time assumed in the accident analysis for the non-sensor 
portions of the channels will continue to be verified using a series 
of sequential or overlapping test measurements. Appropriate actions 
will be taken to ensure overall channel response time remains within 
the times specified in the accident analysis.
    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 change modifies the definitions of RTS and ESF 
response time to allow a substitution of an allocated response time 
for selected sensors in lieu of measuring the response time. The 
allocated time adequately represents the actual measured time for 
the associated sensors. The overall response time of each channel 
will continue to be measured using a series of sequential, 
overlapping or entire channel measurements to ensure the components 
actuated by each channel perform their accident mitigation function 
within the response time assumed in the safety 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 
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. 50-382, Waterford Steam Electric 
Station, Unit 3, St. Charles Parish, Louisiana

    Date of amendment request: September 21, 2001.
    Description of amendment request: Entergy Operations, Inc. 
(Entergy, the licensee) is requesting approval of changes to the 
Waterford Steam Electric Station, Unit 3, Operating License and 
Technical Specifications associated with an increase in the licensed 
power level. The changes involve a proposed increase in the power level 
from 3,390 Megawatts thermal (MWt) to 3,441 MWt. These changes result 
from increased feedwater flow measurement accuracy to be achieved by 
utilizing high accuracy ultrasonic flow measurement 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 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 comprehensive analytical efforts performed to support the 
proposed change included a review of the Nuclear Steam Supply System 
(NSSS) systems and components that could be affected by this change. 
All systems and components will function as designed, and the 
applicable performance requirements have been evaluated and found to 
be acceptable.
    The primary loop components (reactor vessel, reactor internals, 
control element drive mechanisms, loop piping and supports, reactor 
coolant pumps, steam generators, and pressurizer) continue to comply 
with their applicable structural limits and will continue to perform 
their intended design functions. Thus, there is no increase in the 
probability of a structural failure of these components. The Leak 
Before Break analysis conclusions remain valid, and thus the 
limiting break sizes determined in this analysis remain bounding. 
All of the NSSS will still perform the intended design functions 
during normal and accident conditions. The auxiliary systems and 
components continue to meet their applicable structural limits and 
will continue to perform their intended design functions. Thus, 
there is no increase in the probability of a structural failure of 
these components. All of the NSSS and Balance of Plant (BOP) 
interface systems will continue to perform their intended design 
functions. The main steam safety valves (MSSVs) will provide 
adequate relief capacity to maintain the steam generator pressures 
within design limits. The atmospheric dump valves and steam bypass 
valves meet design sizing requirements at the uprated power level. 
The current Loss of Coolant Accident (LOCA) hydraulic forcing 
functions are still bounding for the proposed 1.5 percent increase 
in power.
    Because the integrity of the plant will not be affected by 
operation at the uprated condition, it is concluded that all 
structures, systems, and components required to mitigate a transient 
remain capable of fulfilling their intended functions. The reduced 
uncertainty in the flow input to the power calorimetric measurement 
allows the current safety analyses to be used, without change, to 
support operation at a core power of 3,441 megawatts thermal (MWt). 
As such, all Updated Final Safety Analysis Report (UFSAR) Chapter 15 
accident analyses continue to demonstrate compliance with the 
relevant event acceptance criteria. Those analyses performed to 
assess the effects of mass and energy releases remain valid. The 
source terms used to assess radiological consequences have been 
reviewed and determined to either bound operation at the 1.5 percent 
uprated condition, or new analyses were performed to verify all 
acceptance criteria continue to be met.
    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?
    No new accident scenarios, failure mechanisms, or single 
failures are introduced as a result of the proposed changes. The new 
installation of the LEFM [leading edge flow meter] CheckPlus system 
has been analyzed, and failures of this system will have no effect 
on any safety-related system or any systems, structures or 
components required for transient mitigation. All systems, 
structures, and components previously required for the mitigation of 
a transient remain capable of fulfilling their intended design 
functions. The proposed changes have no adverse effects on any 
safety-related system or component and do not challenge the 
performance or integrity of any safety related system.
    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

[[Page 55018]]

involve a significant reduction in a margin of safety?
    Operation at the uprated power condition does not involve a 
significant reduction in a margin of safety. Analyses of the primary 
fission product barriers have concluded that all relevant design 
criteria remain satisfied, both from the standpoint of the integrity 
of the primary fission product barrier and from the standpoint of 
compliance with the required acceptance criteria.
    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.

Exelon Generation Company, LLC, Docket Nos. STN 50-456 and STN 50-457, 
Braidwood Station, Units 1 and 2, Will County, Illinois
Docket Nos. STN 50-454 and STN 50-455, Byron Station, Units 1 and 2, 
Ogle County, Illinois
Docket Nos. 50-237 and 50-249, Dresden Nuclear Power Station, Units 2 
and 3, Grundy County, Illinois
Docket Nos. 50-373 and 50-374, LaSalle County Station, Units 1 and 2, 
LaSalle 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, Units 2 and 3, York County, Pennsylvania
Docket Nos. 50-254 and 50-265, Quad Cities Nuclear Power Station, Units 
1 and 2, Rock Island County, Illinois
Date of amendment request: August 1, 2001.
    Description of amendment request: The requested changes to the 
technical specifications (TSs) propose to revise requirements that have 
been superceded based on licensed operator training programs being 
accredited by the Institute for Nuclear Power Operations (INPO), 
promulgation of the revised 10 CFR part 55, Operators' Licenses, and 
adoption of a systems approach to training as required by 10 CFR 
50.120, Training and qualification of nuclear power plant personnel. 
The same changes were requested by AmerGen Energy Company, LLC 
(AmerGen) for the Clinton Power Station, Oyster Creek, and Three Mile 
Island, Unit 1. The proposed no significant hazards consideration for 
those plants is published elsewhere in the Federal Register under 
AmerGen.
    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 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 change does not involve a significant increase 
in the probability or consequences of an accident previously 
evaluated.
    There will be no changes to the procedures by which the 
operators operate the plants. There will be no changes to the 
systems, structures, or components in the plants.
    Based on the above, these proposed changes do not involve a 
significant increase in the probability or consequences of an 
accident previously evaluated.
    2. The proposed change does not create the possibility of a new 
or different kind of accident previously evaluated.
    There will be no changes to the procedures by which the 
operators operate the plants. There will be no changes to the 
systems, structures, or components in the plants.
    Therefore, the proposed changes will not create the possibility 
of a new or different kind of accident previously evaluated.
    3. The proposed change does not involve a significant reduction 
in a margin of safety.
    There will be no change in the plants' systems, structures, or 
components, nor in the way in which they will be operated as a 
result of the proposed changes. Therefore, the proposed changes will 
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 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 Chiefs: Anthony J. Mendiola, James W. Clifford.

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

    Date of amendment request: September 21, 2001.
    Description of amendment request: The proposed amendment deletes 
requirements from the Technical Specifications (and, as applicable, 
other elements of the licensing bases) to maintain a Post Accident 
Sampling System (PASS). Licensees were generally required to implement 
PASS upgrades as described in NUREG-0737, ``Clarification of TMI [Three 
Mile Island] Action Plan Requirements,'' and Regulatory Guide 1.97, 
``Instrumentation for Light-Water-Cooled Nuclear Power Plants to Assess 
Plant and Environs Conditions During and Following an Accident.'' 
Implementation of these upgrades was an outcome of the lessons learned 
from the accident that occurred at TMI, Unit 2. Requirements related to 
PASS were imposed by Order for many facilities and were added to or 
included in the technical specifications (TS) for nuclear power 
reactors currently licensed to operate. Lessons learned and 
improvements implemented over the last 20 years have shown that the 
information obtained from PASS can be readily obtained through other 
means or is of little use in the assessment and mitigation of accident 
conditions.
    The NRC staff issued a notice of opportunity for comment in the 
Federal Register on August 11, 2000 (65 FR 49271) on possible 
amendments to eliminate PASS, including a model safety evaluation and 
model no significant hazards consideration (NSHC) determination, using 
the consolidated line item improvement process. The NRC staff 
subsequently issued a notice of availability of the models for 
referencing in license amendment applications in the Federal Register 
on October 31, 2000 (65 FR 65018). The licensee affirmed the 
applicability of the following NSHC determination in its application 
dated September 21, 2001.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), an analysis of the issue 
of no significant hazards consideration is presented below:

Criterion 1--The Proposed Change Does Not Involve a Significant 
Increase in the Probability or Consequences of an Accident 
Previously Evaluated.

    The PASS was originally designed to perform many sampling and 
analysis functions. These functions were designed and intended to be 
used in post accident situations and were put into place as a result 
of the TMI-2 accident. The specific intent of the PASS was to 
provide a system that has the capability to obtain and analyze 
samples of plant fluids containing potentially high levels of 
radioactivity, without exceeding plant personnel radiation exposure 
limits. Analytical results of these samples would be used largely 
for verification purposes in aiding the plant staff in assessing the 
extent of core damage and subsequent offsite

[[Page 55019]]

radiological dose projections. The system was not intended to and 
does not serve a function for preventing accidents and its 
elimination would not affect the probability of accidents previously 
evaluated.
    In the 20 years since the TMI-2 accident and the consequential 
promulgation of post accident sampling requirements, operating 
experience has demonstrated that a PASS provides little actual 
benefit to post accident mitigation. Past experience has indicated 
that there exists in-plant instrumentation and methodologies 
available in lieu of a PASS for collecting and assimilating 
information needed to assess core damage following an accident. 
Furthermore, the implementation of Severe Accident Management 
Guidance (SAMG) emphasizes accident management strategies based on 
in-plant instruments. These strategies provide guidance to the plant 
staff for mitigation and recovery from a severe accident. Based on 
current severe accident management strategies and guidelines, it is 
determined that the PASS provides little benefit to the plant staff 
in coping with an accident.
    The regulatory requirements for the PASS can be eliminated 
without degrading the plant emergency response. The emergency 
response, in this sense, refers to the methodologies used in 
ascertaining the condition of the reactor core, mitigating the 
consequences of an accident, assessing and projecting offsite 
releases of radioactivity, and establishing protective action 
recommendations to be communicated to offsite authorities. The 
elimination of the PASS will not prevent an accident management 
strategy that meets the initial intent of the post-TMI-2 accident 
guidance through the use of the SAMGs, the emergency plan (EP), the 
emergency operating procedures (EOP), and site survey monitoring 
that support modification of emergency plan protective action 
recommendations (PARs).
    Therefore, the elimination of PASS requirements from Technical 
Specifications (TS) (and other elements of the licensing bases) does 
not involve a significant increase in the consequences of any 
accident previously evaluated.

    Criterion 2--The Proposed Change Does Not Create the Possibility 
of a New or Different Kind of Accident From Any Previously 
Evaluated.

    The elimination of PASS related requirements will not result in 
any failure mode not previously analyzed. The PASS was intended to 
allow for verification of the extent of reactor core damage and also 
to provide an input to offsite dose projection calculations. The 
PASS is not considered an accident precursor, nor does its existence 
or elimination have any adverse impact on the pre-accident state of 
the reactor core or post accident confinement of radionuclides 
within the containment building.
    Therefore, this change does not create the possibility of a new 
or different kind of accident from any previously evaluated.

    Criterion 3--The Proposed Change Does Not Involve a Significant 
Reduction in the Margin of Safety

    The elimination of the PASS, in light of existing plant 
equipment, instrumentation, procedures, and programs that provide 
effective mitigation of and recovery from reactor accidents, results 
in a neutral impact to the margin of safety. Methodologies that are 
not reliant on PASS are designed to provide rapid assessment of 
current reactor core conditions and the direction of degradation 
while effectively responding to the event in order to mitigate the 
consequences of the accident. The use of a PASS is redundant and 
does not provide quick recognition of core events or rapid response 
to events in progress. The intent of the requirements established as 
a result of the TMI-2 accident can be adequately met without 
reliance on a PASS.
    Therefore, this change does not involve a significant reduction 
in the margin of safety.
    Based upon the reasoning presented above and the previous 
discussion of the amendment request, the requested change does not 
involve a significant hazards consideration.

    The NRC staff proposes to determine that the amendment request 
involves 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: June 28, 2001.
    Description of amendment request: The proposed amendment would 
revise the technical specification (TS) 3.1.1.4 upper limit for the 
moderator temperature coefficient (MTC) from 0  x  10-4 
change in reactivity per degree Fahrenheit (``k/k/ deg.F) to +0.2  x  
10-4 ``k/k/ deg.F for power levels up to 70 percent of rated 
thermal power (RTP), and ramping linearly to 0  x  10-4 ``k/
k/ deg.F from 70 percent to 100 percent RTP. The proposed change is 
needed to address future core designs with higher energy requirements, 
associated with plant operation at higher capacity factors.
    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?
    No. The proposed change from a[n] MTC of 0  x  10-4 
``k/k/ deg.F to a positive moderator temperature coefficient (PMTC) 
of +0.2  x  10-4 ``k/k/ deg.F does not introduce an 
initiator of any design basis accident or event. The proposed change 
does not adversely affect accident initiators or precursors nor 
alter the configuration of the facility or the manner in which the 
plant is maintained. Thus, the proposed change does not involve a 
significant increase in the probability of an accident previously 
evaluated.
    The proposed change to a PMTC does not alter or prevent the 
ability of structures, systems, and components (SSCs) from 
performing their intended function to mitigate the consequences of 
an initiating event within the assumed acceptance limits. The 
proposed change is consistent with the safety analysis assumptions 
and resultant consequences. Accident analyses affected by the 
proposed change have been reanalyzed and all applicable acceptance 
criteria have been met. Thus, the proposed change does not involve a 
significant increase in the consequences of an accident previously 
evaluated.
    Therefore, the proposed amendment does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    2. Does the change create the possibility of a new or different 
kind of accident from any accident previously evaluated?
    No. The change to a PMTC does not involve a physical alteration 
of the plant (i.e., no new or different type of equipment will be 
installed), subsequently no new or different failure modes or 
limiting single failures are created. The plant will not be operated 
in a different manner due to the proposed change. All SSCs will 
continue to function as currently designed. Thus, the proposed 
change does not create any new or different accident scenarios.
    Therefore, the proposed 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?
    No. The proposed change to a PMTC does not involve revisions to 
any safety limits or safety system settings that would adversely 
impact plant safety. The proposed amendment does not alter the 
functional capabilities assumed in a safety analysis for any SSCs 
important to the mitigation and control of design bases accident 
conditions within the facility.
    All of the applicable acceptance criteria (i.e., preventing 
reactor coolant system [RCS] or main steam system 
overpressurization, maintaining the minimum departure from nucleate 
boiling ratio [DNBR], preventing core uncovery, preventing fuel 
temperatures from exceeding their limit, preventing clad damage, and 
limiting the number of fuel rods that enter a departure from 
nucleate boiling [DNB] condition) for each of the analyses affected 
by the proposed change continue to be met. The conclusions of the 
UFSAR remain valid. Thus, since the operating parameters and system 
performance will remain within design requirements and safety 
analysis assumptions, safety margin is maintained.
    Therefore, the proposed amendment 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

[[Page 55020]]

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: Lakshminaras Raghavan, (Acting).

GPU Nuclear Inc., Docket No. 50-320, Three Mile Island Nuclear 
Generating Station, Unit 2, Dauphin County, Pennsylvania

    Date of amendment request: June 21, 2001.
    Description of amendment request: The proposed technical 
specifications change request (TSCR) No. 81 is to revise Three Mile 
Island Nuclear Generating Station, Unit 2 (TMI-2) Technical 
Specification (TS) Administrative Controls section that will provide 
consistency with the changes to the revised 50.59 rule of Title 10 of 
the Code of Federal Regulations (10 CFR) Regulations, as published in 
the Federal Register on October 4, 1999 (64 FR 53582).
    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:

    Does the change involve a significant increase in the 
probability or consequences of an accident previously evaluated?
    The proposed changes reflect the revised 50.59 rule, issued as a 
Final Rule in October 4, 1999, and do not impact the operation of 
any system or component assumed in any accident analysis. The 
proposed change does not change the requirement to perform a 50.59 
review when required by the Technical Specification Administrative 
Controls. Based on the administrative nature of this change there 
will be no direct impact on the radiological source term. Therefore, 
these changes will not involve a significant increase in the 
probability or consequences of an accident previously evaluated.
    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 this 
change. The proposed change is administrative in nature and makes 
the language in the Technical Specification Administrative Controls 
conform to the Final Rule, dated October 4, 1999, related to the 10 
CFR 50.59 rule. No new accident mode or equipment failure modes are 
created by these changes. Therefore, these proposed changes do not 
create the possibility of a new or different kind of accident from 
any accident previously evaluated.
    Does the change involve a significant reduction in a margin of 
safety?
    The proposed change does not impact or have a direct effect on 
any safety analysis assumptions. The proposed change is 
administrative in nature and makes the TS Administrative Control 
language conform to the Final Rule, dated October 4, 1999, related 
to the 10 CFR 50.59 rule. 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 
amendment request involves no significant hazards consideration.
    Attorney for licensee: Ernest L. Blake, Jr., Esquire, Shaw, 
Pittman, Potts & Trowbridge, 2300 N Street, NW., Washington, DC 20037.
    NRC Section Chief: Robert A. Gramm.

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

    Date of amendment request: October 5, 2001.
    Description of amendment request: The licensee proposed to amend 
the Technical Specifications (TSs) to change the licensing basis 
requirement for establishing containment hydrogen monitoring ``within 
30 minutes'' to ``within 3 hours'' of initiating emergency core cooling 
following a loss-of-coolant accident (LOCA).
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration. The NRC staff has reviewed the licensee's analysis 
against the three standards of 10 CFR 50.92(c). The licensee's analysis 
is presented below:

    1. The operation of Nine Mile Point Unit 2 in accordance with 
the proposed amendment will not involve a significant increase in 
the probability or consequences of an accident previously evaluated.
    The Updated Safety Analysis Report (USAR) Chapter 15 accident 
analyses do not require or take credit for hydrogen monitoring to be 
established shortly after a loss of coolant accident (LOCA). Post-
LOCA hydrogen production occurs over a long period of time, and an 
extension from 30 minutes to 3 hours for establishing hydrogen 
monitoring will have a positive impact on the ability of the 
operators to concentrate on their more immediate actions while 
having no negative impact on containment integrity or the long-term 
assessment efforts. Therefore, the proposed license amendment will 
not involve a significant increase in the probability or 
consequences of an accident previously evaluated.
    2. The operation of Nine Mile Point Unit 2 in accordance with 
the proposed amendment will not create the possibility of a new or 
different kind of accident from any accident previously evaluated.
    Control room operators use the containment hydrogen monitors 
following a LOCA to establish hydrogen control measures should it 
become necessary. The proposed license amendment would not eliminate 
the requirement to establish hydrogen monitoring, but would allow it 
to be delayed until those actions required to mitigate the accident 
and verify proper operation of essential safety equipment have been 
completed. The proposed extension maintains the requirement to 
establish hydrogen monitoring well before calculated conditions 
inside the containment indicate any need to initiate hydrogen 
control measures. Therefore, the proposed license amendment will not 
create a new or different kind of accident from any accident 
previously evaluated.
    3. The operation of Nine Mile Point Unit 2 in accordance with 
the proposed amendment will not involve a significant reduction in a 
margin of safety.
    The need to establish hydrogen control measures will not be 
present within the first 3 hours following a LOCA since there will 
not be significant hydrogen accumulation. By extending the time 
allowed to establish containment hydrogen monitoring, the operators 
can remain focused on the actions necessary to mitigate the accident 
before directing their attention to hydrogen control measures and 
other long-term actions. The proposed extension maintains the 
requirement to establish hydrogen monitoring well before calculated 
conditions inside the containment indicate any need to initiate 
hydrogen control measures. Therefore, the proposed license amendment 
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 
requested amendment involves no significant hazards consideration.
    Attorney for licensee: Mark J. Wetterhahn, Esquire, Winston & 
Strawn, 1400 L Street, NW., Washington, DC 20005-3502.
    NRC Section Chief: L. Raghavan, Acting.

Pacific Gas and Electric Company, Docket Nos. 50-275 and 50-323, Diablo 
Canyon Nuclear Power Plant (DCPP), Units 1 and 2, San Luis Obispo 
County, California

    Date of amendment requests: September 13, 2001.
    Description of amendment requests: The proposed license amendments

[[Page 55021]]

would revise Technical Specification (TS) 3.7.16, ``Spent Fuel Pool 
Boron Concentration,'' TS 3.7.17, ``Spent Fuel Assembly Storage--Region 
1/Region 2,'' and TS 4.3, ``Fuel Storage'' for DCPP Units 1 and 2, to 
allow the use of credit for soluble boron in the spent fuel pool 
criticality analysis.
    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.
    Response: No.
    There is no increase in the probability of a fuel assembly drop 
accident in the spent fuel pool (SFP) when considering the presence 
of soluble boron in the SFP water for criticality control. The 
handling of the fuel assemblies does not change as a result of 
crediting soluble boron in the SFP.
    There is no increase in the probability of the accidental 
misloading of a fuel assembly into the SFP racks when considering 
the presence of soluble boron in the SFP water for criticality 
control. Fuel assembly placement will continue to be controlled 
pursuant to approved fuel handling procedures and will be in 
accordance with the Technical Specification (TS) SFP storage 
configuration limitations.
    There is no increase in the consequences of an accidental drop 
or accidental misloading of a fuel assembly into the SFP racks 
because the criticality analysis demonstrates that the pool will 
remain subcritical following either event even if the pool contains 
a boron concentration less than that currently specified in the TS. 
The current TS limitation will ensure that an adequate SFP boron 
concentration will be maintained.
    There is no increase in the probability of the loss of normal 
cooling to the SFP water considering the presence of soluble boron 
in the pool water for subcriticality control since a high 
concentration of soluble boron has always been maintained in the SFP 
water.
    There is no increase in the consequences of a loss of normal SFP 
cooling because the 2,000 ppm boron concentration required by TS 
provides significant negative reactivity to provide subcritical 
margin such that the SFP keff is maintained less than or 
equal to 0.95 up to boiling (212 deg.F).
    Therefore, the proposed change does not involve a significant 
increase in the probability or consequence 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.
    Response: No.
    Spent fuel handling accidents are not new or different types of 
accidents; they have been analyzed in Section 15.5.22 of the Updated 
Final Safety Analysis Report (UFSAR).
    Criticality accidents in the SFP are not new or different types 
of accidents; they have been analyzed in the UFSAR and in the 
Criticality Analysis reports associated with the specific license 
amendments for fuel enrichments up to 5.0 weight percent U-235.
    Because soluble boron has always been required in the SFP water, 
and is currently required by TS, credit for soluble boron will have 
no effect on normal pool operation and maintenance. Crediting 
soluble boron in the SFP criticality analysis will only result in 
increased sampling to verify the boron concentration. This increased 
sampling frequency will not create the possibility of a new or 
different kind of accident.
    The SFP dilution analysis demonstrates that a dilution which 
could increase the rack keff to greater than 0.95 is not 
a credible event. Therefore, crediting soluble boron in the SFP 
criticality analysis will not result in the possibility of a new 
kind of accident.
    Revised specifications continue to specify the requirements for 
SFP storage configurations. The only significant changes relate to 
the criteria for determining the storage configuration. Because the 
proposed SFP storage configuration limitations will be similar to 
those currently contained in the TS, the new limitations will not 
have any significant effect on normal SFP operations and maintenance 
and will not create the possibility of a new or different kind of 
accident. A SFP loading verification will continue to be performed 
to ensure that the SFP loading configuration meets the specified 
requirements.
    Therefore, the proposed change does not create the possibility 
of a new or different kind of accident from any previously 
evaluated.
    3. The proposed change does not involve a significant reduction 
in a margin of safety.
    Response: No.
    The TS changes proposed by this license amendment request and 
the resulting spent fuel storage limitations will provide an 
adequate safety margin to ensure that the stored fuel assembly array 
will always remain subcritical. Those limits are based on a plant 
specific criticality analysis performed for the Diablo Canyon Units 
1 and 2 SFPs that includes technically supported margins.
    While the criticality analysis utilized credit for soluble 
boron, storage configurations have been defined to ensure that the 
spent fuel rack keff will be less than 1.0 with no 
soluble boron with a 95 percent probability at a 95 percent 
confidence level. Soluble boron credit is used to offset 
uncertainties, tolerances and off-normal conditions, and to provide 
subcritical margin such that the SFP keff is maintained 
less than or equal to 0.95. Since keff is less than or 
equal to 0.95, the current margin of safety is maintained.
    A substantial reduction in the SFP soluble boron concentration 
that could lead to exceeding a keff of 0.95 has been 
evaluated and shown not to be credible.
    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: Christopher J. Warner, Esq., Pacific Gas and 
Electric Company, P.O. Box 7442, San Francisco, California 94120.
    NRC Section Chief: Stephen Dembek.

Pacific Gas and Electric Company, Docket Nos. 50-275 and 50-323, Diablo 
Canyon Nuclear Power Plant (DCPP), Unit Nos. 1 and 2, San Luis Obispo 
County, California

    Date of amendment requests: September 13, 2001.
    Description of amendment requests: The proposed license amendments 
would modify Technical Specification (TS) 5.5.9, ``Steam Generator Tube 
Surveillance Program,'' to allow extension of steam generator tube W 
star alternate repair criteria for DCPP Units 1 and 2, from Cycles 10 
and 11 to Cycles 12 and 13.
    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.
    Of the various accidents previously evaluated, the extension of 
the steam generator (SG) tube W star (W*) alternate repair criteria 
(ARC) through Cycles 12 and 13 only affects the steam generator tube 
rupture (SGTR) accident evaluation and the postulated steam line 
break (SLB) accident evaluation. Loss-of-coolant accident (LOCA) 
conditions cause a compressive axial load to act on the tube. 
Therefore, since the LOCA tends to force the tube into the tubesheet 
rather than pull it out, it is not a factor in this evaluation.
    For the SGTR accident, the required structural margins of the SG 
tubes will be maintained by the presence of the tubesheet. Tube 
rupture is precluded for cracks in the Westinghouse explosive tube 
expansion (WEXTEX) region due to the constraint provided by the 
tubesheet. Therefore, Regulatory Guide (RG) 1.121, ``Bases for 
Plugging Degraded PWR Steam Generator Tubes,'' margins against burst 
are maintained for both normal and postulated accident conditions.
    WCAP-14797, Revision 1, defines a length, W*, of degradation 
free expanded tubing that provides the necessary resistance to tube 
pullout due to the pressure induced forces (with applicable safety 
factors applied). The W* length supplies the necessary resistive 
force to preclude pullout loads under both normal operating and 
accident conditions. The contact pressure results from the WEXTEX 
expansion process, thermal

[[Page 55022]]

expansion mismatch between the tube and tubesheet and from the 
differential pressure between the primary and secondary side. The 
proposed changes do not affect other systems, structures, 
components, or operational features. Therefore, the proposed change 
results in no significant increase in the probability of the 
occurrence of an SGTR or SLB accident.
    The consequences of an SGTR accident are affected by the 
primary-to-secondary leakage flow during the accident. Primary-to-
secondary leakage flow through a postulated broken tube is not 
affected by the proposed changes since the tubesheet enhances the 
tube integrity in the region of the WEXTEX expansion by precluding 
tube deformation beyond its initial expanded outside diameter. The 
resistance to both tube rupture and collapse is strengthened by the 
tubesheet in that region. At normal operating pressures, leakage 
from primary water stress corrosion cracking (PWSCC) in the W* 
length is limited by both the tube-to-tubesheet crevice and the 
limited crack opening permitted by the tubesheet constraint. No 
leakage has been observed in any in-situ test of W* indications 
identified to date. Consequently, negligible normal operating 
leakage is expected from cracks within the tubesheet region.
    SLB leakage is limited by leakage flow restrictions resulting 
from the crack and tube-to-tubesheet contact pressures that provide 
a restricted leakage path above the indications and also limit the 
degree of crack face opening compared to free span indications. The 
total leakage, that is, the combined leakage for all such tubes, 
plus the combined leakage developed by any other ARC, are maintained 
below the maximum allowable SLB leak rate limit, such that off-site 
doses are maintained less than 10 CFR 100 guideline values.
    Therefore, based on the above evaluation, the proposed changes 
do not involve a significant increase in the probability or 
consequences of an accident previously evaluated.
    2. The proposed change does not create the possibility of a new 
or different kind of accident from any accident previously 
evaluated.
    The proposed changes do not introduce any changes or mechanisms 
that create the possibility of a new or different kind of accident. 
Tube bundle integrity is expected to be maintained for all plant 
conditions upon continued implementation of the W* ARC.
    Axial indications left in service shall have the upper crack tip 
below the top of the tubesheet (TTS) by at least the value of the 
nondestructive examination (NDE) uncertainty and crack growth 
allowance, such that at the end of the subsequent operating cycle 
the entire crack remains below the tubesheet secondary face, thereby 
minimizing the potential for free span cracking and demonstrating 
that an acceptable level of risk is maintained for tubes returned to 
service under W* ARC. This repair criteria is in addition to 
ensuring that the upper crack tip is located below the bottom of the 
WEXTEX transition by at least the NDE measurement uncertainty. 
Condition monitoring will verify that all tubes returned to service 
under W* ARC remain below the TTS, including an allowance for NDE 
uncertainty.
    These changes do not introduce any new equipment or any change 
to existing equipment. No new effects on existing equipment are 
created nor are any new malfunctions introduced.
    Therefore, based on the above evaluation, the proposed changes 
do 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 changes maintain the required structural margins of 
the SG tubes for both normal and accident conditions. RG 1.121 is 
used as the basis in the development of the W* ARC for determining 
that SG tube integrity considerations are maintained within 
acceptable limits. RG 1.121 describes a method acceptable to the NRC 
staff for meeting General Design Criteria 14, 15, 31, and 32 by 
reducing the probability and consequences of an SGTR. RG 1.121 
concludes that by determining the limiting safe conditions of tube 
wall degradation beyond which tubes with unacceptable cracking, as 
established by inservice inspection, should be removed from service 
or repaired, the probability and consequences of a SGTR are reduced. 
This RG uses safety factors on loads for tube-burst that are 
consistent with the requirements of Section III of the ASME Code.
    For primarily axially oriented cracking located within the 
tubesheet, tube-burst is precluded due to the presence of the 
tubesheet. WCAP-14797, Revision 1, defines a length, W*, of 
degradation free expanded tubing that provides the necessary 
resistance to tube pullout due to the pressure induced forces (with 
applicable safety factors applied). Application of the W* ARC will 
preclude unacceptable primary-to-secondary leakage during all plant 
conditions. The methodology for determining leakage provides for 
large margins between calculated and actual leakage values in the W* 
ARC.
    Plugging of the SG tubes reduces the reactor coolant flow margin 
for core cooling. Continued implementation of W* ARC will result in 
maintaining the margin of flow that may have otherwise been reduced 
by tube plugging.
    Based on the above, it is concluded that the proposed changes do 
not result in a significant reduction of margin with respect to 
plant safety as defined in the Final Safety Analysis Report Update 
or Bases of the plant Technical Specifications.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment requests involve no significant hazards consideration.
    Attorney for licensee: Christopher J. Warner, Esq., Pacific Gas and 
Electric Company, P.O. Box 7442, San Francisco, California 94120.
    NRC Section Chief: Stephen Dembek.

PSEG Nuclear LLC, Docket Nos. 50-272 and 50-311, Salem Nuclear 
Generating Station, Unit Nos. 1 and 2, Salem County, New Jersey

    Date of amendment request: August 17, 2001.
    Description of amendment request: The proposed amendments delete 
requirements from the Technical Specifications (and, as applicable, 
other elements of the licensing bases) to maintain a Post Accident 
Sampling System (PASS). Licensees were generally required to implement 
PASS upgrades as described in NUREG-0737, ``Clarification of TMI [Three 
Mile Island] Action Plan Requirements,'' and Regulatory Guide 1.97, 
``Instrumentation for Light-Water-Cooled Nuclear Power Plants to Assess 
Plant and Environs Conditions During and Following an Accident.'' 
Implementation of these upgrades was an outcome of the lessons learned 
from the accident that occurred at TMI, Unit 2. Requirements related to 
PASS were imposed by Order for many facilities and were added to or 
included in the technical specifications (TS) for nuclear power 
reactors currently licensed to operate. Lessons learned and 
improvements implemented over the last 20 years have shown that the 
information obtained from PASS can be readily obtained through other 
means or is of little use in the assessment and mitigation of accident 
conditions.
    The NRC staff issued a notice of opportunity for comment in the 
Federal Register on August 11, 2000 (65 FR 49271) on possible 
amendments to eliminate PASS, including a model safety evaluation and 
model no significant hazards consideration (NSHC) determination, using 
the consolidated line item improvement process. The NRC staff 
subsequently issued a notice of availability of the models for 
referencing in license amendment applications in the Federal Register 
on October 31, 2000 (65 FR 65018). The licensee affirmed the 
applicability of the following NSHC determination in its application 
dated August 17, 2001.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), an analysis of the issue 
of no significant hazards consideration is presented below:

Criterion 1--The Proposed Change Does Not Involve a Significant 
Increase in the Probability or Consequences of an Accident 
Previously Evaluated.

    The PASS was originally designed to perform many sampling and 
analysis

[[Page 55023]]

functions. These functions were designed and intended to be used in 
post accident situations and were put into place as a result of the 
TMI-2 accident. The specific intent of the PASS was to provide a 
system that has the capability to obtain and analyze samples of 
plant fluids containing potentially high levels of radioactivity, 
without exceeding plant personnel radiation exposure limits. 
Analytical results of these samples would be used largely for 
verification purposes in aiding the plant staff in assessing the 
extent of core damage and subsequent offsite radiological dose 
projections. The system was not intended to and does not serve a 
function for preventing accidents and its elimination would not 
affect the probability of accidents previously evaluated.
    In the 20 years since the TMI-2 accident and the consequential 
promulgation of post accident sampling requirements, operating 
experience has demonstrated that a PASS provides little actual 
benefit to post accident mitigation. Past experience has indicated 
that there exists in-plant instrumentation and methodologies 
available in lieu of a PASS for collecting and assimilating 
information needed to assess core damage following an accident. 
Furthermore, the implementation of Severe Accident Management 
Guidance (SAMG) emphasizes accident management strategies based on 
in-plant instruments. These strategies provide guidance to the plant 
staff for mitigation and recovery from a severe accident. Based on 
current severe accident management strategies and guidelines, it is 
determined that the PASS provides little benefit to the plant staff 
in coping with an accident.
    The regulatory requirements for the PASS can be eliminated 
without degrading the plant emergency response. The emergency 
response, in this sense, refers to the methodologies used in 
ascertaining the condition of the reactor core, mitigating the 
consequences of an accident, assessing and projecting offsite 
releases of radioactivity, and establishing protective action 
recommendations to be communicated to offsite authorities. The 
elimination of the PASS will not prevent an accident management 
strategy that meets the initial intent of the post-TMI-2 accident 
guidance through the use of the SAMGs, the emergency plan (EP), the 
emergency operating procedures (EOP), and site survey monitoring 
that support modification of emergency plan protective action 
recommendations (PARs).
    Therefore, the elimination of PASS requirements from Technical 
Specifications (TS) (and other elements of the licensing bases) does 
not involve a significant increase in the consequences of any 
accident previously evaluated.

Criterion 2--The Proposed Change Does Not Create the Possibility of 
a New or Different Kind of Accident from any Previously Evaluated.

    The elimination of PASS related requirements will not result in 
any failure mode not previously analyzed. The PASS was intended to 
allow for verification of the extent of reactor core damage and also 
to provide an input to offsite dose projection calculations. The 
PASS is not considered an accident precursor, nor does its existence 
or elimination have any adverse impact on the pre-accident state of 
the reactor core or post accident confinement of radionuclides 
within the containment building.
    Therefore, this change does not create the possibility of a new 
or different kind of accident from any previously evaluated.

Criterion 3--The Proposed Change Does Not Involve a Significant 
Reduction in the Margin of Safety.

    The elimination of the PASS, in light of existing plant 
equipment, instrumentation, procedures, and programs that provide 
effective mitigation of and recovery from reactor accidents, results 
in a neutral impact to the margin of safety. Methodologies that are 
not reliant on PASS are designed to provide rapid assessment of 
current reactor core conditions and the direction of degradation 
while effectively responding to the event in order to mitigate the 
consequences of the accident. The use of a PASS is redundant and 
does not provide quick recognition of core events or rapid response 
to events in progress. The intent of the requirements established as 
a result of the TMI-2 accident can be adequately met without 
reliance on a PASS.
    Therefore, this change does not involve a significant reduction 
in the margin of safety.
    Based upon the reasoning presented above and the previous 
discussion of the amendment requests, the requested changes do not 
involve a significant hazards consideration.

    The NRC staff proposes to determine that the amendment requests 
involve no significant hazards consideration.
    Attorney for licensee: Jeffrie J. Keenan, Esquire, Nuclear Business 
Unit--N21, P.O. Box 236, Hancocks Bridge, NJ 08038.
    NRC Section Chief: James W. Clifford.

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

    Date of amendment request: October 1, 2001.
    Description of amendment request: The proposed amendment deletes 
requirements from the Technical Specifications (TS) (and, as 
applicable, other elements of the licensing bases) to maintain a Post 
Accident Sampling System (PASS). Licensees were generally required to 
implement PASS upgrades as described in NUREG-0737, ``Clarification of 
TMI [Three Mile Island] Action Plan Requirements,'' and Regulatory 
Guide 1.97, ``Instrumentation for Light-Water-Cooled Nuclear Power 
Plants to Assess Plant and Environs Conditions During and Following an 
Accident.'' Implementation of these upgrades was an outcome of the 
lessons learned from the accident that occurred at TMI, Unit 2. 
Requirements related to PASS were imposed by Order for many facilities 
and were added to or included in the TS for nuclear power reactors 
currently licensed to operate. Lessons learned and improvements 
implemented over the last 20 years have shown that the information 
obtained from PASS can be readily obtained through other means or is of 
little use in the assessment and mitigation of accident conditions.
    The NRC staff issued a notice of opportunity for comment in the 
Federal Register on August 11, 2000 (65 FR 49271) on possible 
amendments to eliminate PASS, including a model safety evaluation and 
model no significant hazards consideration (NSHC) determination, using 
the consolidated line item improvement process. The NRC staff 
subsequently issued a notice of availability of the models for 
referencing in license amendment applications in the Federal Register 
on October 31, 2000 (65 FR 65018). The licensee affirmed the 
applicability of the following NSHC determination in its application 
dated October 1, 2001.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), an analysis of the issue 
of no significant hazards consideration is presented below:

Criterion 1--The Proposed Change Does Not Involve a Significant 
Increase in the Probability or Consequences of an Accident 
Previously Evaluated.

    The PASS was originally designed to perform many sampling and 
analysis functions. These functions were designed and intended to be 
used in post accident situations and were put into place as a result 
of the TMI-2 accident. The specific intent of the PASS was to 
provide a system that has the capability to obtain and analyze 
samples of plant fluids containing potentially high levels of 
radioactivity, without exceeding plant personnel radiation exposure 
limits. Analytical results of these samples would be used largely 
for verification purposes in aiding the plant staff in assessing the 
extent of core damage and subsequent offsite radiological dose 
projections. The system was not intended to and does not serve a 
function for preventing accidents and its elimination would not 
affect the probability of accidents previously evaluated.
    In the 20 years since the TMI-2 accident and the consequential 
promulgation of post accident sampling requirements, operating 
experience has demonstrated that a PASS provides little actual 
benefit to post accident mitigation. Past experience has indicated 
that there exists in-plant instrumentation and methodologies 
available in lieu of a PASS for collecting and assimilating 
information needed to assess core damage following an accident. 
Furthermore, the implementation of Severe Accident Management 
Guidance (SAMG) emphasizes accident management

[[Page 55024]]

strategies based on in-plant instruments. These strategies provide 
guidance to the plant staff for mitigation and recovery from a 
severe accident. Based on current severe accident management 
strategies and guidelines, it is determined that the PASS provides 
little benefit to the plant staff in coping with an accident.
    The regulatory requirements for the PASS can be eliminated 
without degrading the plant emergency response. The emergency 
response, in this sense, refers to the methodologies used in 
ascertaining the condition of the reactor core, mitigating the 
consequences of an accident, assessing and projecting offsite 
releases of radioactivity, and establishing protective action 
recommendations to be communicated to offsite authorities. The 
elimination of the PASS will not prevent an accident management 
strategy that meets the initial intent of the post-TMI-2 accident 
guidance through the use of the SAMGs, the emergency plan (EP), the 
emergency operating procedures (EOP), and site survey monitoring 
that support modification of emergency plan protective action 
recommendations (PARs).
    Therefore, the elimination of PASS requirements from Technical 
Specifications (TS) (and other elements of the licensing bases) does 
not involve a significant increase in the consequences of any 
accident previously evaluated.

Criterion 2--The Proposed Change Does Not Create the Possibility of 
a New or Different Kind of Accident from any Previously Evaluated.

    The elimination of PASS related requirements will not result in 
any failure mode not previously analyzed. The PASS was intended to 
allow for verification of the extent of reactor core damage and also 
to provide an input to offsite dose projection calculations. The 
PASS is not considered an accident precursor, nor does its existence 
or elimination have any adverse impact on the pre-accident state of 
the reactor core or post accident confinement of radionuclides 
within the containment building.
    Therefore, this change does not create the possibility of a new 
or different kind of accident from any previously evaluated.

Criterion 3--The Proposed Change Does Not Involve a Significant 
Reduction in the Margin of Safety.

    The elimination of the PASS, in light of existing plant 
equipment, instrumentation, procedures, and programs that provide 
effective mitigation of and recovery from reactor accidents, results 
in a neutral impact to the margin of safety. Methodologies that are 
not reliant on PASS are designed to provide rapid assessment of 
current reactor core conditions and the direction of degradation 
while effectively responding to the event in order to mitigate the 
consequences of the accident. The use of a PASS is redundant and 
does not provide quick recognition of core events or rapid response 
to events in progress. The intent of the requirements established as 
a result of the TMI-2 accident can be adequately met without 
reliance on a PASS.
    Therefore, this change does not involve a significant reduction 
in the margin of safety.
    Based upon the reasoning presented above and the previous 
discussion of the amendment request, the requested change does not 
involve a significant hazards consideration.

    The NRC staff proposes to determine that the amendment request 
involves no significant hazards consideration.
    Attorney for licensee: Thomas G. Eppink, South Carolina Electric & 
Gas Company, Post Office Box 764, Columbia, South Carolina 29218.
    NRC Section Chief: Richard J. Laufer, Acting.

Southern Nuclear Operating Company, Inc, Docket Nos. 50-348 and 50-364, 
Joseph M. Farley Nuclear Plant, Units 1 and 2, Houston County, Alabama

    Date of amendment request: May 3, 2001.
    Description of amendment request: The proposed amendments would 
relocate cycle-specific reactor coolant system Technical Specifications 
parameters limits to the Core Operating Limits Report. Also, a 
reference to the Refueling Operations Boron Concentration is added to 
TS 5.6.5 to correct an omission.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. The proposed change does not involve a significant increase 
in the probability or consequences of an accident previously 
evaluated.
    The proposed amendment is a programmatic and administrative 
change that does not physically alter plant systems, nor does it 
impact the performance of their functions. No new equipment is added 
nor is installed equipment being changed or operated in a different 
manner. Because the design of the facility and system operating 
parameters are not being changed, the proposed amendment does not 
involve an increase in the probability or consequences of any 
accident previously evaluated.
    The cycle-specific limits in the Core Operating Limits Report 
(COLR) will continue to be controlled by the Farley Nuclear Plant 
(FNP) programs and procedures. Each accident analysis addressed in 
the Final Safety Analysis Report (FSAR) will be examined with 
respect to changes in the cycle dependent parameters, which are 
obtained from the use of Nuclear Regulatory Commission (NRC) 
approved reload design methodologies, to ensure that the transient 
evaluation of new reloads are bounded by previously accepted 
analyses. This examination, which will be conducted per the 
requirements of 10 CFR 50.59, will ensure that future reloads will 
not involve a significant increase in the probability or 
consequences of an accident previously evaluated. The safety limits 
imposed in Technical Specification (TS) 2.1 are consistent with the 
values stated in the FNP FSAR.
    This change does not involve an increase in the probability or 
consequences of any 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.
    Relocation of the cycle-specific parameters has no influence or 
impact on, nor does it contribute in any way to the probability or 
consequences of an accident. No plant equipment, function or plant 
operation will be altered as a result of this proposed change. The 
cycle-specific parameters are calculated using the NRC approved 
methods and submitted to the NRC to allow the staff to continue to 
trend the values of these limits. The TS will continue to require 
operation within the core operating limits and appropriate actions 
will be required if these limits are exceeded. The safety limits are 
maintained in the COLR and appropriate actions will be required if 
these limits are exceeded. In addition, the minimum limit for 
Reactor Coolant System flow will be retained in the TS. The safety 
limits imposed in TS 2.1 are consistent with the values stated in 
the FNP FSAR.
    This proposed amendment 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 margin of safety is not affected by the removal of cycle-
specific core operating limits from the TS. The margin of safety 
presently provided by current TS limits remains unchanged. 
Appropriate measures exist to control the values of these cycle-
specific limits. The proposed amendment continues to require 
operation within the core limits as obtained from NRC approved 
reload design methodologies and the actions to be taken if a limit 
is exceeded remain unchanged.
    The development of the limits for future reloads will continue 
to conform to those methods described in NRC approved documentation. 
In addition, each future reload will involve a 10 CFR 50.59 
evaluation to assure that operation of the unit within the cycle-
specific limits will not involve a significant reduction in the 
margin of safety.
    The proposed changes to relocate cycle specific parameter limits 
to the COLRs will not affect plant design or system operating 
parameters, there is no detrimental impact on any equipment design 
parameters, and the plant will continue to operate within prescribed 
limits. The safety limits imposed in TS 2.1 are consistent with the 
values stated in the FNP FSAR.
    This proposed change does not involve a significant reduction in 
the margin of safety.

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

[[Page 55025]]

amendment request involves no significant hazards consideration.
    Attorney for licensee: M. Stanford Blanton, Esq., Balch and 
Bingham, Post Office Box 306, 1710 Sixth Avenue North, Birmingham, 
Alabama 35201.
    NRC Section Chief: Richard J. Laufer, Acting.

Southern Nuclear Operating Company, Inc, Docket Nos. 50-348 and 50-364, 
Joseph M. Farley Nuclear Plant, Units 1 and 2, Houston County, Alabama

    Date of amendment request: June 5, 2001.
    Description of amendment request: The proposed amendments would 
modify Technical Specifications (TS) Surveillance Requirement 3.4.14.1 
to clarify that the frequency does not apply to Reactor Coolant System 
Pressure Isolation Valves in the Residual Heat Removal System flow 
path. Also, related TS Bases and editorial changes are part of this TS 
change.
    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 changes do not involve a significant increase in 
the probability or consequences of an accident previously evaluated.
    The proposed change to Surveillance Requirement (SR) 3.4.14.1 
clarifies that the requirement to test Reactor Coolant System (RCS) 
Pressure Isolation Valves (PIVs) following valve actuation due to 
automatic or manual action or flow through the valve does not apply 
to PIVs in the Residual Heat Removal (RHR) flow path. This resolves 
a source of potential confusion and ensures that the testing 
requirements are implemented consistent with the historical 
licensing basis for Farley and the Improved Technical Specification 
conversion NRC Safety Evaluation Report. The valves will continue to 
be tested for back leakage every 18 months. The proposed change does 
not affect the consequences of a previously analyzed accident since 
the magnitude and duration of analyzed events are not impacted by 
this change. Thus, the consequences of a previously evaluated 
accident are unchanged.
    Therefore, the proposed TS change does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    2. The proposed changes do not create the possibility of a new 
or different kind of accident from any accident previously 
evaluated.
    The proposed change involves no change to the physical plant. It 
allows for a clarification to the testing requirements to ensure 
that the historical licensing basis for Farley is maintained. These 
valves are tested every 18 months to ensure that the back leakage is 
within acceptable limits. This testing will continue. These changes 
do not impact the function of the valves.
    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 changes do not involve a significant reduction 
in a margin of safety.
    The physical plant is unaffected by this change. The proposed 
change does not impact accident offsite dose, containment pressure 
or temperature, emergency core cooling system (ECCS) or reactor 
protection system (RPS) settings or any other parameter that could 
affect a margin of safety. The clarification of the testing 
requirements ensures that future testing is consistent with the 
historical licensing basis for Farley.
    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: M. Stanford Blanton, Esq., Balch and 
Bingham, Post Office Box 306, 1710 Sixth Avenue North, Birmingham, 
Alabama 35201.
    NRC Section Chief: Richard J. Laufer, Acting.

Virginia Electric and Power Company, Docket Nos. 50-338 and 50-339, 
North Anna Power Station, Units No. 1 and No. 2, Louisa County, 
Virginia

    Date of amendment request: September 10, 2001.
    Description of amendment request: The proposed amendment deletes 
requirements from the Technical Specifications (and, as applicable, 
other elements of the licensing bases) to maintain a Post Accident 
Sampling System (PASS). Licensees were generally required to implement 
PASS upgrades as described in NUREG-0737, ``Clarification of TMI [Three 
Mile Island] Action Plan Requirements,'' and Regulatory Guide 1.97, 
``Instrumentation for Light-Water-Cooled Nuclear Power Plants to Assess 
Plant and Environs Conditions During and Following an Accident.'' 
Implementation of these upgrades was an outcome of the lessons learned 
from the accident that occurred at TMI, Unit 2. Requirements related to 
PASS were imposed by Order for many facilities and were added to or 
included in the technical specifications (TS) for nuclear power 
reactors currently licensed to operate. Lessons learned and 
improvements implemented over the last 20 years have shown that the 
information obtained from PASS can be readily obtained through other 
means or is of little use in the assessment and mitigation of accident 
conditions.
    The NRC staff issued a notice of opportunity for comment in the 
Federal Register on August 11, 2000 (65 FR 49271) on possible 
amendments to eliminate PASS, including a model safety evaluation and 
model no significant hazards consideration (NSHC) determination, using 
the consolidated line item improvement process. The NRC staff 
subsequently issued a notice of availability of the models for 
referencing in license amendment applications in the Federal Register 
on October 31, 2000 (65 FR 65018). The licensee affirmed the 
applicability of the following NSHC determination in its application 
dated September 10, 2001.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), an analysis of the issue 
of no significant hazards consideration is presented below:

Criterion 1--The Proposed Change Does Not Involve a Significant 
Increase in the Probability or Consequences of an Accident 
Previously Evaluated.

    The PASS was originally designed to perform many sampling and 
analysis functions. These functions were designed and intended to be 
used in post accident situations and were put into place as a result 
of the TMI-2 accident. The specific intent of the PASS was to 
provide a system that has the capability to obtain and analyze 
samples of plant fluids containing potentially high levels of 
radioactivity, without exceeding plant personnel radiation exposure 
limits. Analytical results of these samples would be used largely 
for verification purposes in aiding the plant staff in assessing the 
extent of core damage and subsequent offsite radiological dose 
projections. The system was not intended to and does not serve a 
function for preventing accidents and its elimination would not 
affect the probability of accidents previously evaluated.
    In the 20 years since the TMI-2 accident and the consequential 
promulgation of post accident sampling requirements, operating 
experience has demonstrated that a PASS provides little actual 
benefit to post accident mitigation. Past experience has indicated 
that there exists in-plant instrumentation and methodologies 
available in lieu of a PASS for collecting and assimilating 
information needed to assess core damage following an accident. 
Furthermore, the implementation of Severe Accident Management 
Guidance (SAMG) emphasizes accident management strategies based on 
in-plant instruments. These strategies provide guidance to the plant 
staff for mitigation and recovery from a severe accident. Based on 
current severe accident management strategies and guidelines, it is 
determined that the PASS provides little benefit to the plant staff 
in coping with an accident.

[[Page 55026]]

    The regulatory requirements for the PASS can be eliminated 
without degrading the plant emergency response. The emergency 
response, in this sense, refers to the methodologies used in 
ascertaining the condition of the reactor core, mitigating the 
consequences of an accident, assessing and projecting offsite 
releases of radioactivity, and establishing protective action 
recommendations to be communicated to offsite authorities. The 
elimination of the PASS will not prevent an accident management 
strategy that meets the initial intent of the post-TMI-2 accident 
guidance through the use of the SAMGs, the emergency plan (EP), the 
emergency operating procedures (EOP), and site survey monitoring 
that support modification of emergency plan protective action 
recommendations (PARs).
    Therefore, the elimination of PASS requirements from Technical 
Specifications (TS) (and other elements of the licensing bases) does 
not involve a significant increase in the consequences of any 
accident previously evaluated.

Criterion 2--The Proposed Change Does Not Create the Possibility of 
a New or Different Kind of Accident from any Previously Evaluated.

    The elimination of PASS related requirements will not result in 
any failure mode not previously analyzed. The PASS was intended to 
allow for verification of the extent of reactor core damage and also 
to provide an input to offsite dose projection calculations. The 
PASS is not considered an accident precursor, nor does its existence 
or elimination have any adverse impact on the pre-accident state of 
the reactor core or post accident confinement of radionuclides 
within the containment building.
    Therefore, this change does not create the possibility of a new 
or different kind of accident from any previously evaluated.

Criterion 3--The Proposed Change Does Not Involve a Significant 
Reduction in the Margin of Safety.

    The elimination of the PASS, in light of existing plant 
equipment, instrumentation, procedures, and programs that provide 
effective mitigation of and recovery from reactor accidents, results 
in a neutral impact to the margin of safety. Methodologies that are 
not reliant on PASS are designed to provide rapid assessment of 
current reactor core conditions and the direction of degradation 
while effectively responding to the event in order to mitigate the 
consequences of the accident. The use of a PASS is redundant and 
does not provide quick recognition of core events or rapid response 
to events in progress. The intent of the requirements established as 
a result of the TMI-2 accident can be adequately met without 
reliance on a PASS.
    Therefore, this change does not involve a significant reduction 
in the margin of safety.
    Based upon the reasoning presented above and the previous 
discussion of the amendment request, the requested change does not 
involve a significant hazards consideration.

    The NRC staff proposes to determine that the amendment request 
involves no significant hazards consideration.
    Attorney for licensee: Mr. Donald P. Irwin, Esq., Hunton and 
Williams, Riverfront Plaza, East Tower, 951 E. Byrd Street, Richmond, 
Virginia 23219.
    NRC Section Chief: Richard J. Laufer, Acting.

Virginia Electric and Power Company, Docket Nos. 50-280 and 50-281, 
Surry Power Station, Unit Nos. 1 and 2, Surry County, Virginia

    Date of amendment request: September 10, 2001.
    Description of amendment request: The proposed amendment deletes 
requirements from the Technical Specifications (and, as applicable, 
other elements of the licensing bases) to maintain a Post Accident 
Sampling System (PASS). Licensees were generally required to implement 
PASS upgrades as described in NUREG-0737, ``Clarification of TMI [Three 
Mile Island] Action Plan Requirements,'' and Regulatory Guide 1.97, 
``Instrumentation for Light-Water-Cooled Nuclear Power Plants to Assess 
Plant and Environs Conditions During and Following an Accident.'' 
Implementation of these upgrades was an outcome of the lessons learned 
from the accident that occurred at TMI, Unit 2. Requirements related to 
PASS were imposed by Order for many facilities and were added to or 
included in the technical specifications (TS) for nuclear power 
reactors currently licensed to operate. Lessons learned and 
improvements implemented over the last 20 years have shown that the 
information obtained from PASS can be readily obtained through other 
means or is of little use in the assessment and mitigation of accident 
conditions.
    The NRC staff issued a notice of opportunity for comment in the 
Federal Register on August 11, 2000 (65 FR 49271) on possible 
amendments to eliminate PASS, including a model safety evaluation and 
model no significant hazards consideration (NSHC) determination, using 
the consolidated line item improvement process. The NRC staff 
subsequently issued a notice of availability of the models for 
referencing in license amendment applications in the Federal Register 
on October 31, 2000 (65 FR 65018). The licensee affirmed the 
applicability of the following NSHC determination in its application 
dated September 10, 2001.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), an analysis of the issue 
of no significant hazards consideration is presented below:
Criterion 1--The Proposed Change Does Not Involve a Significant 
Increase in the Probability or Consequences of an Accident 
Previously Evaluated.

    The PASS was originally designed to perform many sampling and 
analysis functions. These functions were designed and intended to be 
used in post accident situations and were put into place as a result 
of the TMI-2 accident. The specific intent of the PASS was to 
provide a system that has the capability to obtain and analyze 
samples of plant fluids containing potentially high levels of 
radioactivity, without exceeding plant personnel radiation exposure 
limits. Analytical results of these samples would be used largely 
for verification purposes in aiding the plant staff in assessing the 
extent of core damage and subsequent offsite radiological dose 
projections. The system was not intended to and does not serve a 
function for preventing accidents and its elimination would not 
affect the probability of accidents previously evaluated.
    In the 20 years since the TMI-2 accident and the consequential 
promulgation of post accident sampling requirements, operating 
experience has demonstrated that a PASS provides little actual 
benefit to post accident mitigation. Past experience has indicated 
that there exists in-plant instrumentation and methodologies 
available in lieu of a PASS for collecting and assimilating 
information needed to assess core damage following an accident. 
Furthermore, the implementation of Severe Accident Management 
Guidance (SAMG) emphasizes accident management strategies based on 
in-plant instruments. These strategies provide guidance to the plant 
staff for mitigation and recovery from a severe accident. Based on 
current severe accident management strategies and guidelines, it is 
determined that the PASS provides little benefit to the plant staff 
in coping with an accident.
    The regulatory requirements for the PASS can be eliminated 
without degrading the plant emergency response. The emergency 
response, in this sense, refers to the methodologies used in 
ascertaining the condition of the reactor core, mitigating the 
consequences of an accident, assessing and projecting offsite 
releases of radioactivity, and establishing protective action 
recommendations to be communicated to offsite authorities. The 
elimination of the PASS will not prevent an accident management 
strategy that meets the initial intent of the post-TMI-2 accident 
guidance through the use of the SAMGs, the emergency plan (EP), the 
emergency operating procedures (EOP), and site survey monitoring 
that support modification of emergency plan protective action 
recommendations (PARs).
    Therefore, the elimination of PASS requirements from Technical 
Specifications (TS) (and other elements of the licensing bases) does 
not involve a significant increase in the consequences of any 
accident previously evaluated.

Criterion 2--The Proposed Change Does Not Create the Possibility of 
a New or Different Kind of Accident from any

[[Page 55027]]

Previously Evaluated.

    The elimination of PASS related requirements will not result in 
any failure mode not previously analyzed. The PASS was intended to 
allow for verification of the extent of reactor core damage and also 
to provide an input to offsite dose projection calculations. The 
PASS is not considered an accident precursor, nor does its existence 
or elimination have any adverse impact on the pre-accident state of 
the reactor core or post accident confinement of radionuclides 
within the containment building.
    Therefore, this change does not create the possibility of a new 
or different kind of accident from any previously evaluated.

Criterion 3--The Proposed Change Does Not Involve a Significant 
Reduction in the Margin of Safety.

    The elimination of the PASS, in light of existing plant 
equipment, instrumentation, procedures, and programs that provide 
effective mitigation of and recovery from reactor accidents, results 
in a neutral impact to the margin of safety. Methodologies that are 
not reliant on PASS are designed to provide rapid assessment of 
current reactor core conditions and the direction of degradation 
while effectively responding to the event in order to mitigate the 
consequences of the accident. The use of a PASS is redundant and 
does not provide quick recognition of core events or rapid response 
to events in progress. The intent of the requirements established as 
a result of the TMI-2 accident can be adequately met without 
reliance on a PASS.
    Therefore, this change does not involve a significant reduction 
in the margin of safety.
    Based upon the reasoning presented above and the previous 
discussion of the amendment request, the requested change does not 
involve a significant hazards consideration.

    The NRC staff proposes to determine that the amendment request 
involves no significant hazards consideration.
    Attorney for licensee: Donald P. Irwin, Esq., Hunton and Williams, 
Riverfront Plaza, East Tower, 951 E. Byrd Street, Richmond, Virginia 
23219.
    NRC Section Chief: Richard J. Laufer, Acting.

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].

Arizona Public Service Company, et al., Docket Nos. STN 50-528, STN 50-
529, and STN 50-530, Palo Verde Nuclear Generating Station, Units Nos. 
1, 2, and 3, Maricopa County, Arizona

    Date of application for amendments: April 1, 2001, as supplemented 
by letter dated July 26, 2001.
    Brief description of amendments: The amendment revises Technical 
Specifications (TS) section 5.0, ``Administrative Controls,'' by (1) 
clarifying new diesel fuel oil limits for water and sediment, (2) 
revising guidance on changes to TS Bases consistent with changes to 10 
CFR 50.59, (3) adding clarification to the requirements for the Safety 
Function Determination Program, (4) adding the CENTS computer code to 
the list of analytical methods used to determine core operating limits, 
and (5) revising the Core Operating Limits Report list of references to 
approved topical reports.
    Date of issuance: October 15, 2001.
    Effective date: October 15, 2001, and shall be implemented within 
60 days of the date of issuance.
    Amendment Nos.: Unit 1-137, Unit 2-137, Unit 3-137.
    Facility Operating License Nos. NPF-41, NPF-51, and NPF-74: The 
amendments revised the Technical Specifications.
    Date of initial notice in Federal Register: May 2, 2001 (66 FR 
22022).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated October 15, 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: October 4, 2000, as supplemented 
March 8, 2001, March 27, April 26, May 14, May 18, June 4, June 11, 
June 26, June 29, July 3, July 16 (2 letters), July 17, August 17, and 
September 20, 2001.
    Brief description of amendment: The license amendment revises the 
Harris Nuclear Plant Technical Specifications to support the 
replacement of the current Westinghouse Model D4 steam generators with 
Westinghouse Model Delta 75 replacement steam generators and revises 
the accident analyses to adopt the alternate source term (AST) 
methodology, using the guidance of Nuclear Regulatory Commission (NRC) 
Regulatory Guide 1.183.
    Date of issuance: October 12, 2001.
    Effective date: October 12, 2001.
    Amendment No.: 107.
    Facility Operating License No. NPF-63: Amendment revises the 
Technical Specifications.
    Date of initial notice in Federal Register: For the steam generator 
replacement amendment request, the initial notice is dated November 1, 
2000 (65 FR 65338). The March 8, 2001, March 27, April 26, May 14, May 
18, June 4, June 11, June 26, June 29, July 3, July 16 (2 letters), and 
September 20, 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 
initial notice for the adoption of the AST methodology, using the 
guidance of NRC Regulatory Guide 1.183, was published on August 8, 2001 
(66 FR 41612). The August 17, 2001, supplement contained clarifying 
information only, and did not change the initial no significant hazards

[[Page 55028]]

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 October 12, 2001.
    No significant hazards consideration comments received: No.

Entergy Nuclear Indian Point 2 and Entergy Nuclear Operations, Docket 
Nos. 50-003 and 50-247, Indian Point Nuclear Generating Unit Nos. 1 and 
2, Westchester County, New York

    Date of application for amendment: July 13, 2001.
    Brief description of amendment: The amendments revise Technical 
Specification (TS) 4.1.8, ``High Radiation Area,'' for Indian Point 
Unit 1 and TS 6.12, ``High Radiation Area,'' for Indian Point Unit 2 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.
    Date of issuance: October 10, 2001.
    Effective date: As of the date of issuance to be implemented within 
60 days.
    Amendment No.: 51 and 221.
    Facility Operating License Nos. DPR-5 and DPR-26: Amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: September 5, 2001 (66 
FR 46477).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated October 10, 2001.
    No significant hazards consideration comments received: No.

Entergy Nuclear Operations, Docket Nos. 50-003 and 50-247, Indian Point 
Nuclear Generating Unit Nos. 1 and 2, Westchester County, New York

    Date of application for amendments: December 12, 2000, as 
supplemented on April 12, April 16, May 24, June 6, and June 8, 2001.
    Brief description of amendments: The conforming amendments 
reflected the transfer of the licenses, formerly held by Consolidated 
Edison Company of New York, Inc., to Entergy Nuclear Indian Point 2, 
LLC, as the owner of Indian Point 1 and 2, and to Entergy Nuclear 
Operations, Inc., as the entity authorized to maintain Indian Point 1 
and operate Indian Point 2. The amendments were approved pursuant to 
Section 50.90 of Title 10 of the Code of Federal Regulations.
    Date of issuance: September 6, 2001.
    Effective date: As of the date of issuance to be implemented within 
30 days.
    Amendment Nos.: 50 (Indian Point 1) and 220 (Indian Point 2).
    Facility Operating License Nos. DPR-05 and DPR-26: Amendments 
revised the Licenses and Technical Specifications.
    Date of initial notice in Federal Register: January 29, 2001 (66 FR 
8122).
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated August 27, 2001.
    No significant hazards consideration comments received: Not 
applicable.

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

    Date of application for amendment: June 12, 2001, as supplemented 
by letters dated July 31, September 19 and September 25, 2001.
    Brief description of amendment: The amendment revised Technical 
Specification (TS) 3.8.1.1 to provide a one-time extension of the 
allowed outage time (AOT) for an inoperable emergency diesel generator 
(EDG) from three days to ten days. In addition, the amendment revised 
TS 3.4.4 to make the action associated with an inoperable emergency 
power supply to the pressurizer heaters consistent with the proposed 
EDG AOT.
    Date of issuance: October 15, 2001.
    Effective date: As of the date of issuance to be implemented within 
60 days from the date of issuance.
    Amendment No.: 234.
    Facility Operating License No. NPF-6: Amendment revised the TSs.
    Date of initial notice in Federal Register: July 11, 2001 (66 FR 
36341).
    The July 31, September 19 and September 25, 2001, supplemental 
letters provided clarifying information and revised TSs that were 
within the scope of the original Federal Register notice and did not 
change the staff's initial no significant hazards consideration 
determination.
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated October 15, 2001.
    No significant hazards consideration comments received: No.

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

    Date of amendment request: January 8, 2001.
    Brief description of amendment: The change revises the lower limit 
of the allowable containment internal pressure in Technical 
Specification (TS) 3.6.1.4, ``Containment Systems--Internal Pressure,'' 
from 14.375 pounds per square inch, absolute (psia) to 14.275 psia.
    Date of issuance: October 10, 2001.
    Effective date: As of the date of issuance and shall be implemented 
60 days from the date of issuance.
    Amendment No.: 174.
    Facility Operating License No. NPF-38: The amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: February 21, 2001 (66 
FR 11058).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated October 10, 2001.
    No significant hazards consideration comments received: No.

Exelon Generation Company, LLC, PSEG Nuclear LLC, and Atlantic City 
Electric Company, Docket Nos. 50-277 and 50-278, Peach Bottom Atomic 
Power Station, Units 2 and 3, York County, Pennsylvania

    Date of application for amendments: October 10, 2000, as 
supplemented October 9, 2001.
    Brief description of amendments: The amendments revised the 
licenses for Peach Bottom Units 2 and 3 to remove Atlantic City 
Electric Company as a licensee, in conjunction with the transfer of the 
minority ownership interests of Atlantic City Electric Company to the 
majority owners, Exelon Generation Company, LLC, and PSEG Nuclear LLC.
    Date of issuance: October 18, 2001.
    Effective date: As of date of issuance, and shall be implemented 
within 30 days of issuance.
    Amendments Nos.: 241 and 245.
    Facility Operating License Nos. DPR-44 and DPR-56: The amendments 
revised the License.
    Date of initial notice in Federal Register: November 27, 2000 (65 
FR 70740).
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated December 27, 2000.

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

    Date of application for amendment: March 28, 2001, as supplemented 
July 19, and October 2, 2001.

[[Page 55029]]

    Brief description of amendment: The amendment revised the Improved 
Technical Specifications 3.7.18, ``Control Complex Cooling System'' to 
allow a one-time increase in the completion time for restoring an 
inoperable Control Complex Cooling System train from 7 to 35 days.
    Date of issuance: October 16, 2001.
    Effective date: As of the date of issuance and shall be implemented 
within 30 days of issuance.
    Amendment No.: 200.
    Facility Operating License No. DPR-72: Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: April 18, 2001 (66 FR 
20006). The supplemental letters provided clarifying information that 
did not change the initial proposed no significant hazards 
consideration determination.
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated October 16, 2001.
    No significant hazards consideration comments received: No.

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

    Date of application for amendment: February 21, 2000, as 
supplemented June 27, 2001.
    Brief description of amendment: The amendments revise various 
administrative actions, requirements, and responsibilities contained in 
Improved Technical Specifications (ITS) 2.0, ``Safety Limits,'' and ITS 
5.0, ``Administrative Controls,'' to reflect the recent CR-3 Nuclear 
Operations reorganization and the amended requirements of 10 CFR 50.72, 
10 CFR 50.73 and 10 CFR 50.59.
    Date of issuance: October 18, 2001.
    Effective date: As of the date of issuance and shall be implemented 
within 30 days of issuance.
    Amendment No.: 201.
    Facility Operating License No. DPR-72: Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: March 21, 2001 (66 FR 
15926). The supplemental letter provided clarifying information that 
did not change the initial proposed no significant hazards 
consideration determination.
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated October 20, 2001.
    No significant hazards consideration comments received: No.

    Florida Power and Light Company, et al., Docket Nos. 50-335 and 50-
389, St. Lucie Plant, Unit Nos. 1 and 2, St. Lucie County, Florida

    Date of application for amendments: April 17, 2001.
    Brief description of amendments: Minor changes and corrections to a 
Unit 1 license condition and to Technical Specifications of both Unit 1 
and 2 to correct administrative errors, or to incorporate changes 
justified by previous submittals, or to correct logic errors, or to 
delete obsolete terminology and provide conforming changes to reflect 
the revisions to 10 CFR 50.59.
    Date of Issuance: October 18, 2001.
    Effective Date: As of the date of issuance and shall be implemented 
within 60 days of issuance.
    Amendment Nos.: 177 and 119.
    Facility Operating License Nos. DPR-67 and NPF-16: Amendments 
revised the Unit 1 Operating License and the Technical Specifications 
of both units.
    Date of initial notice in Federal Register: May 30, 2001 (66 FR 
29357).
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated October 18, 2001.
    No significant hazards consideration comments received: No.

Nebraska Public Power District, Docket No. 50-298, Cooper Nuclear 
Station, Nemaha County, Nebraska

    Date of amendment request: February 28, 2001.
    Brief description of amendment: The amendment revises the technical 
specifications to reflect changes in the standard by which the licensee 
will test charcoal used in engineered safety feature systems. The 
requested changes satisfy the requirements of NRC Generic Letter 99-02.
    Date of issuance: October 16, 2001.
    Effective date: As of the date of issuance and shall be implemented 
within 30 days from the date of issuance.
    Amendment No.: 186.
    Facility Operating License No. DPR-46: Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: June 12, 2001 (66 FR 
31710).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated October 16, 2001.
    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: February 27, 2001, as 
supplemented on September 6, 2001.
    Brief description of amendment: The amendment revises surveillance 
requirements associated with Technical Specifications Section 3.3.8.2, 
``Reactor Protection System (RPS) Electric Power Monitoring--Logic,'' 
and Section 3.3.8.3, ``Reactor Protection System (RPS) Electric Power 
Monitoring--Scram Solenoids.'' Specifically, the overvoltage allowable 
values and associated channel calibration frequency interval are 
changed.
    Date of issuance: October 17, 2001.
    Effective date: As of the date of issuance and shall be implemented 
by March 15, 2002.
    Amendment No.: 99.
    Facility Operating License No. NPF-69: Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: March 21, 2001 (66 FR 
15928).
    The staff's related evaluation of the amendment is contained in a 
Safety Evaluation dated October 17, 2001.
    No significant hazards consideration comments received: No.

PPL Susquehanna, LLC, Docket No. 50-388, Susquehanna Steam Electric 
Station, Unit 2, Luzerne County, Pennsylvania
    Date of application for amendment: November 16, 2000.
    Brief description of amendment: This amendment deleted a note in TS 
Surveillance Requirement 3.6.1.1.1 which extended the leak rate testing 
surveillance interval on the 2S299A and 2S299B spectacle flange o-rings 
until the Unit 2 10th refueling outage or a prior Unit 2 outage 
requiring entry into Mode 4. The note, added in Amendment No. 160 to 
Facility Operating License No. NPF-22 which was issued on May 8, 2000, 
was necessitated because of a Notice of Enforcement Discretion 
documented in a letter dated April 11, 2000. This note is no longer 
required to be included in TS 3.6.1.1.1 because the surveillance test 
was conducted during the Unit 2 forced outage in August of 2000.
    Date of issuance: October 9, 2001.
    Effective date: As of the date of issuance and shall be implemented 
within 30 days.
    Amendment No.: 171.
    Facility Operating License No. NPF-22: This amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: July 11, 2001 (66 FR 
36343).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated October 9, 2001.

[[Page 55030]]

    No significant hazards consideration comments received: No.

PSEG Nuclear LLC, Docket No. 50-354, Hope Creek Generating Station, 
Salem County, New Jersey

    Date of application for amendment: December 20, 1999, as 
supplemented on February 11, February 25, and October 10, 2000.
    Brief description of amendment: The amendment revises the license 
to reflect changes related to the transfer of the license for the Hope 
Creek Generating Station, to the extent held by Atlantic City Electric 
Company to PSEG Nuclear LLC.
    Date of issuance: October 18, 2001.
    Effective date: As of the date of issuance, and shall be 
implemented within 30 days.
    Amendment No.: 135.
    Facility Operating License No. NPF-57: This amendment revised the 
License.
    Date of initial notice in Federal Register: February 18, 2000 (65 
FR 8453).
    The letters dated February 11, February 25, and October 10, 2000, 
provided clarifying information that did not change the initial 
proposed no significant hazards consideration determination.
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated April 21, 2000.
    No significant hazards consideration comments received: No.

Public Service Electric & Gas Company, Docket Nos. 50-272 and 50-311, 
Salem Nuclear Generating Station, Unit Nos. 1 and 2, Salem County, New 
Jersey

    Date of application for amendments: December 20, 1999, as 
supplemented February 11, February 25, and October 10, 2000.
    Brief description of amendments: The amendments revised Facility 
Operating Licenses DPR-70 and DPR-75 to reflect changes related to the 
transfer of the license for the Salem Nuclear Generating Station, Unit 
Nos. 1 and 2, to the extent held by the Atlantic City Electric Company, 
to PSEG Nuclear Limited Liability Company.
    Date of issuance: October 18, 2001.
    Effective date: As of the date of issuance, and shall be 
implemented within 30 days.
    Amendment Nos.: 246 and 227.
    Facility Operating License Nos. DPR-70 and DPR-75: The amendments 
revised the License.
    Date of initial notice in Federal Register: February 18, 2000 (65 
FR 8452). The February 11, February 25, and October 10, 2000, 
supplements did not expand the scope of the original application with 
respect to both the proposed transfer action and the proposed amendment 
action as initially noticed in the Federal Register. No hearing 
requests or comments were received. In addition, the submittals did not 
affect the applicability of the Commission's generic no significant 
hazards consideration determination set forth in 10 CFR 2.1315.
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated December 21, 2000.

Southern Nuclear Operating Company, Inc., Georgia Power Company, 
Oglethorpe Power Corporation, Municipal Electric Authority of Georgia, 
City of Dalton, Georgia, Docket No. 50-366, Edwin I. Hatch Nuclear 
Plant, Unit 2, Appling County, Georgia

    Date of application for amendment: May 23, 2001.
    Brief description of amendments: The amendment revises the Safety 
Limit Minimum Critical Power Ratio to reflect the results of a cycle-
specific calculation that was performed using NRC-approved methodology.
    Date of issuance: October 12, 2001.
    Effective date: As of the date of issuance and shall be implemented 
within 30 days from the date of issuance.
    Amendment No.: 167.
    Facility Operating License No. NPF-5: Amendment revises the 
Technical Specifications.
    Date of initial notice in Federal Register: June 12, 2001 (66 FR 
31714).
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated October 12, 2001.
    No significant hazards consideration comments received: No.

TXU Electric, Docket Nos. 50-445 and 50-446, Comanche Peak Steam 
Electric Station, Unit Nos. 1 and 2, Somervell County, Texas 

    Date of amendment request: April 25, 2001, as supplemented by 
letters dated July 31 and August 23, 2001.
    Brief description of amendments: The amendments change the 
Technical Specifications (TS) to allow a one-time only change to TS 
3.8.1, ``AC Sources--Operating,'' Action A.3, by extending the required 
Completion Time for restoration of an inoperable offsite circuit from 
72 hours to 21 days.
    Date of issuance: October 9, 2001.
    Effective date: As of the date of issuance and shall be implemented 
no later than February 28, 2002.
    Amendment Nos.: 88/88.
    Facility Operating License Nos. NPF-87 and NPF-89: The amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: September 5, 2001 (66 
FR 46482).
    The supplemental letter dated August 23, 2001, provided clarifying 
information that did not change the Nuclear Regulatory Commission (the 
Commission) staff's proposed no significant hazards consideration 
determination.
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated October 9, 2001.
    No significant hazards consideration comments received: No.

Vermont Yankee Nuclear Power Corporation, Docket No. 50-271, Vermont 
Yankee Nuclear Power Station, Vernon, Vermont 

    Date of application for amendment: August 14, 2001, as supplemented 
on August 21, 2001.
    Brief description of amendment: The proposed amendment would extend 
the allowed outage time (AOT) for the High Pressure Coolant Injection 
(HPCI) and Reactor Core Isolation Cooling systems from 7 days to 14 
days. Requirements were added to immediately ensure the availability of 
alternate means of high pressure coolant makeup. Also, clarifying 
changes were made to Technical Specifications (TSs) 3.5.E.2 and 3.5.G.2 
by reformatting the TSs to make the nomenclature consistent regarding 
HPCI and the Automatic Depressurization System as being systems, not 
subsystems.
    Date of Issuance: October 18, 2001.
    Effective date: As of the date of issuance, and shall be 
implemented within 30 days.
    Amendment No.: 205.
    Facility Operating License No. DPR-28: Amendment revised the 
Technical Specifications.
    Date of initial notice in  Federal Register: September 18, 2001 (66 
FR 48152).
    The Commission's related evaluation of this amendment is contained 
in a Safety Evaluation dated October 18, 2001.
    No significant hazards consideration comments received: No.


    Note: The publication date for this notice will change from 
every other Wednesday to every other Tuesday, effective January 8, 
2002. The notice will contain the same information and will continue 
to be published biweekly.



    Dated at Rockville, Maryland, this 22nd day of October 2001.


[[Page 55031]]


    For the Nuclear Regulatory Commission.
John A. Zwolinski,
Director, Division of Licensing Project Management, Office of Nuclear 
Reactor Regulation.
[FR Doc. 01-27261 Filed 10-30-01; 8:45 am]
BILLING CODE 7590-01-P