[Federal Register Volume 66, Number 239 (Wednesday, December 12, 2001)]
[Notices]
[Pages 64284-64312]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: 01-30455]


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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

[[Page 64285]]

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 November 19, 2001 through November 30, 2001. 
The last biweekly notice was published on November 28, 2001 (66 FR 
59498).

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

    The Commission has made a proposed determination that the following 
amendment requests involve no significant hazards consideration. Under 
the Commission's regulations in 10 CFR 50.92, this means that operation 
of the facility in accordance with the proposed amendment would not (1) 
Involve a significant increase in the probability or consequences of an 
accident previously evaluated; or (2) create the possibility of a new 
or different kind of accident from any accident previously evaluated; 
or (3) involve a significant reduction in a margin of safety. The basis 
for this proposed determination for each amendment request is shown 
below.
    The Commission is seeking public comments on this proposed 
determination. Any comments received within 30 days after the date of 
publication of this notice will be considered in making any final 
determination.
    Normally, the Commission will not issue the amendment until the 
expiration of the 30-day notice period. However, should circumstances 
change during the notice period such that failure to act in a timely 
way would result, for example, in derating or shutdown of the facility, 
the Commission may issue the license amendment before the expiration of 
the 30-day notice period, provided that its final determination is that 
the amendment involves no significant hazards consideration. The final 
determination will consider all public and State comments received 
before action is taken. Should the Commission take this action, it will 
publish in the Federal Register a notice of issuance and provide for 
opportunity for a hearing after issuance. The Commission expects that 
the need to take this action will occur very infrequently.
    Written comments may be submitted by mail to the Chief, Rules and 
Directives Branch, Division of Administrative Services, Office of 
Administration, U.S. Nuclear Regulatory Commission, Washington, DC 
20555-0001, and should cite the publication date and page number of 
this Federal Register notice. Written comments may also be delivered to 
Room 6D22, Two White Flint North, 11545 Rockville Pike, Rockville, 
Maryland from 7:30 a.m. to 4:15 p.m. Federal workdays. Copies of 
written comments received may be examined at the NRC Public Document 
Room, located at One White Flint North, 11555 Rockville Pike (first 
floor), Rockville, Maryland. The filing of requests for a hearing and 
petitions for leave to intervene is discussed below.
    By January 11, 2002, 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.

[[Page 64286]]

    If the final determination is that the amendment request involves a 
significant hazards consideration, any hearing held would take place 
before the issuance of any amendment.
    A request for a hearing or a petition for leave to intervene must 
be filed with the Secretary of the Commission, U.S. Nuclear Regulatory 
Commission, Washington, DC 20555-0001, Attention: Rulemaking and 
Adjudications Branch, or may be delivered to the Commission's Public 
Document Room, located at One White Flint North, 11555 Rockville Pike 
(first floor), Rockville, Maryland 20852, by the above date. A copy of 
the petition should also be sent to the Office of the General Counsel, 
U.S. Nuclear Regulatory Commission, Washington, DC 20555-0001, and to 
the attorney for the licensee.
    Nontimely filings of petitions for leave to intervene, amended 
petitions, supplemental petitions and/or requests for a hearing will 
not be entertained absent a determination by the Commission, the 
presiding officer or the Atomic Safety and Licensing Board that the 
petition and/or request should be granted based upon a balancing of 
factors specified in 10 CFR 2.714(a)(1)(i)-(v) and 2.714(d).
    For further details with respect to this action, see the 
application for amendment which is available for public inspection at 
the Commission's Public Document Room, located at One White Flint 
North, 11555 Rockville Pike (first floor), Rockville, Maryland. 
Publicly available records will be accessible from the Agencywide 
Documents Assess and Management Systems (ADAMS) Public Electronic 
Reading Room on the internet at the NRC Web site, http://www.nrc.gov/NRC/ADAMS/index.html. If you do not have access to ADAMS or if there 
are problems in accessing the documents located in ADAMS, contact the 
NRC Public Document room (PDR) Reference staff at 1-800-397-4209, 304-
415-4737 or by email to [email protected].

AmerGen Energy Company, LLC, Docket No. 50-461, Clinton Power Station, 
Unit 1, DeWitt County, Illinois

    Date of amendment request: July 5, 2001.
    Description of amendment request: The proposed amendment would 
relax Technical Specification (TS) operability requirements for primary 
containment systems, secondary containment systems, and the standby gas 
treatment system during the movement of irradiated fuel and during core 
alterations.
    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 equipment affected by the proposed changes are mitigative in 
nature, and relied upon after an accident has been initiated. 
Application of the Alternative Source Term (AST) does not involve a 
change to the plant design. While the operation of the primary and 
secondary containment systems do change as a result of these 
proposed changes, these systems are not accident initiators. 
Application of the AST does not initiate a design basis accident. 
Similarly, application of the AST does not affect the design or 
operation for any equipment or systems involved in the mitigation of 
accidents. The proposed changes to the Technical Specifications 
(TS), while they revise certain performance requirements, do not 
involve any physical modifications to the plant. As a result, the 
proposed changes do not affect any of the parameters or conditions 
that could contribute to the initiation of any accidents. As such, 
removal of operability requirements during the specified conditions 
will not significantly increase the probability of occurrence for an 
accident previously analyzed.
    The AST changes do not affect the design and operation of the 
facility. Rather, once the accident has been postulated the new 
source term is an input to the evaluation of the consequences. The 
implementation of the AST has been evaluated in revisions to the 
analyses of the worst case Fuel Handling Accident (FHA) at Clinton 
Power Station (CPS). Based on the results of the analyses, it has 
been demonstrated that, with the proposed changes, the dose 
consequences of the worst case FHA remain a small fraction of the 
regulatory guidance provided by the NRC for the AST in RG 
[regulatory guide] 1.183, ``Alternative Radiological Source Terms 
for Evaluating Design Basis Accidents at Nuclear Power Reactors,'' 
dated July 2000. Since the primary containment systems, secondary 
containment systems and the Standby Gas Treatment (SGT) are not 
assumed to be operable in the FHA, the consequences of eliminating 
the requirements that these systems be operable during the handling 
of irradiated fuel in both primary and secondary containment or 
during core alterations will not increase significantly.
    In summary, the proposed changes do 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?
    No new equipment is introduced, and no installed equipment is 
operated in a new or different manner. There is no change to the 
predicted accident response of any plant structure, system or 
component. The proposed change in availability of mitigative 
equipment has been evaluated in accordance with the guidance in RG 
1.183 and does not produce different or more limiting accident 
progression or results. As such, no new accident modes or equipment 
failure modes are created by these proposed changes.
    Therefore, these proposed changes do not create the possibility 
of a new or different kind of accident from any accident previously 
evaluated.
    Does the change involve a significant reduction in a margin of 
safety?
    The proposed changes involve a selective application of the AST 
for the FHA consistent with the guidance provided in RG 1.183. The 
existing analyses demonstrated that the dose consequences associated 
with the FHA were within the applicable NRC specified limits. For 
offsite dose, the margin to safety for the FHA using the 10 CFR 100, 
``Reactor Site Criteria,'' limits was maintained by the existing 
analysis. For the Control Room dose, the margin of safety using the 
10 CFR 50, ``Domestic Licensing of Production and Utilization 
Facilities,'' Appendix A, ``General Design Criteria for Nuclear 
Power Plants,'' General Design Criteria 19, ``Control room,'' dose 
limits was conservatively maintained by the existing analyses. The 
results of the FHA analysis revised in support of this submittal 
however, are subject to revised acceptance criteria. The revised 
dose consequences of the limiting design basis FHA are within the 
acceptance criteria found in RG 1.183 and 10 CFR 50.67, ``Domestic 
Licensing of Production and Utilization Facilities, Accident Source 
Term.'' The proposed changes ensure that the doses at the exclusion 
area boundary (EAB), low population zone (LPZ), and control room 
remain a small fraction of the new regulatory limits in RG 1.183 and 
10 CFR 50.67.
    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: Robert Helfrich, Mid-West Regional Operating 
Group, Exelon Generation Company, LLC, 4300 Winfield Road, Warrenville, 
IL 60555.
    NRC Section Chief: Anthony J. Mendiola.

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

    Date of amendment request: October 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,

[[Page 64287]]

``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 October 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: William D. Johnson, Vice President and 
Corporate Secretary, Carolina Power & Light Company, Post Office Box 
1551, Raleigh, North Carolina 27602.
    NRC Section Chief: Richard P. Correia.

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 amendment request: October 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

[[Page 64288]]

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 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: William D. Johnson, Vice President and 
Corporate Secretary, Carolina Power & Light Company, Post Office Box 
1551, Raleigh, North Carolina 27602.
    NRC Section Chief: Richard P. Correia.

Detroit Edison Company, Docket No. 50-341, Fermi 2, Monroe County, 
Michigan

    Date of amendment request: May 24, 2001.
    Description of amendment request: The proposed amendment would 
delete License Condition 2.C.(11), which is no longer applicable to the 
facility. License Condition 2.C.(11) requires inspection of the low-
pressure turbine discs during the second refueling outage, including 
volumetric examination of the disc base using ultrasonic techniques, 
and specifies that the frequency of subsequent inspections shall be in 
accordance with the turbine manufacturer's recommendations. The 
amendment request states that the license condition is no longer 
applicable for the following reasons: (1) the initial inspection was 
completed during the second refueling outage as required; and (2) 
during fifth refueling outage, the low-pressure turbine rotors were 
replaced with monoblock designed rotors that do not utilize shrunk-on 
discs, and therefore the subsequent inspections specified in License 
Condition 2.C.(11) for shrunk-on discs would be meaningless with the 
new rotor design. The licensee's inspection and maintenance program for 
the new low-pressure turbine is based on the current turbine 
manufacturer's recommendations for the monoblock design.
    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 removes Fermi 2 Operating License 
Condition 2.C.(11) which details the inspection frequency of the 
low-pressure (LP) turbine discs. The inspection frequency was 
recommended because the original turbine rotor design involved a 
shrunk-on disc configuration. The inspection attributes applied 
specifically to this disc design and were intended to enhance design 
reliability. In 1996, however, the LP turbine steam path consisting 
of rotors, buckets (blades), diaphragms and steam flow guides, all 
manufactured by English Electric Co., were replaced with General 
Electric (GE) components. In particular, the GE design does not 
utilize shrunk-on discs; it includes rotors of monoblock 
construction, thus negating the applicability of License Condition 
2.C.(11). There are no relevant aspects of the

[[Page 64289]]

previously recommended inspections that apply to the new monoblock 
construction.
    Section 3.5.1.2.1 of the Fermi 2 UFSAR [Updated Final Safety 
Analysis Report] addresses the potential for missiles generated from 
rotating equipment including those generated from a low-pressure 
turbine rotor segment. Section 10.2.3 of the UFSAR states that 
following the low-pressure turbine rotor replacement during RFO05, 
``there will no longer be a design basis turbine missile at Fermi 
2.'' Section 3.5.1.2.2 further states, ``The new low-pressure rotors 
are of monoblock construction. The monoblock rotors have higher 
speed capability than the maximum attainable speed of the turbine 
generator units. Per General Electric, the supplier of the new 
rotors, the probability of missiles being generated is well below 10 
to the -8 power.'' There are no other postulated accidents that were 
directly attributable to the English Electric Company shrunk-on disc 
design; therefore, the removal of License Condition 2.C.(11) does 
not increase the probability of occurrence or the 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.
    The proposed change removes License Condition 2.C.(11) because 
it is no longer applicable to the design of the low-pressure turbine 
currently installed at the facility. Therefore, removal of the 
license condition affects neither the design nor the operation of 
the plant. It cannot create a new failure mode, nor can its removal 
create the possibility of a new or different kind of accident than 
any accident previously evaluated.
    3. The change does not involve a significant reduction in the 
margin of safety.
    License Condition 2.C.(11) is not applicable to the facility 
because the low-pressure turbine rotor was replaced with a design 
which does not include shrunk-on turbine discs. This rotor 
replacement eliminated the potential for a design basis accident 
resulting from the turbine missiles at Fermi 2, which was the 
accident scenario that the inspections referenced in License 
Condition 2.C.(11) were intended to prevent. Since the license 
condition no longer applies to the current facility design, and the 
potential design basis accident associated with the license 
condition no longer exists, the removal of the license condition 
will not reduce any margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 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: Peter Marquardt, Legal Department, 688 WCB, 
Detroit Edison Company, 2000 2nd Avenue, Detroit, Michigan 48226-1279.
    NRC Acting Section Chief: William D. Reckley, Acting.

Detroit Edison Company, Docket No. 50-341, Fermi 2, Monroe County, 
Michigan

    Date of amendment request: November 11, 2001.
    Description of amendment request: A change is proposed to Technical 
Specification 3.0.3 to allow a longer period of time to perform a 
missed surveillance. The time is extended from the current limit of `` 
* * * up to 24 hours or up to the limit of the specified Frequency, 
whichever is less'' to ``* * *up to 24 hours or up to the limit of the 
specified Frequency, whichever is greater.'' In addition, the following 
requirement would be added to the specification: ``A risk evaluation 
shall be performed for any Surveillance delayed greater than 24 hours 
and the risk impact shall be managed.''
    The Nuclear Regulatory Commission (NRC) staff issued a notice of 
opportunity for comment in the Federal Register on June 14, 2001 (66 FR 
32400), on possible amendments concerning missed surveillances, 
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 September 28, 2001 (66 FR 
49714). The licensee affirmed the applicability of the following NSHC 
determination in its application dated November 11, 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 proposed change relaxes the time allowed to perform a missed 
surveillance. The time between surveillances is not an initiator of 
any accident previously evaluated. Consequently, the probability of 
an accident previously evaluated is not significantly increased. The 
equipment being tested is still required to be operable and capable 
of performing the accident mitigation functions assumed in the 
accident analysis. As a result, the consequences of any accident 
previously evaluated are not significantly affected. Any reduction 
in confidence that a standby system might fail to perform its safety 
function due to a missed surveillance is small and would not, in the 
absence of other unrelated failures, lead to an increase in 
consequences beyond those estimated by existing analyses. The 
addition of a requirement to assess and manage the risk introduced 
by the missed surveillance will further minimize possible concerns. 
Therefore, this change does not involve a significant increase in 
the probability or consequences of an 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 proposed change does not involve a physical alteration of 
the plant (no new or different type of equipment will be installed) 
or a change in the methods governing normal plant operation. A 
missed surveillance will not, in and of itself, introduce new 
failure modes or effects and any increased chance that a standby 
system might fail to perform its safety function due to a missed 
surveillance would not, in the absence of other unrelated failures, 
lead to an accident beyond those previously evaluated. The addition 
of a requirement to assess and manage the risk introduced by the 
missed surveillance will further minimize possible concerns. Thus, 
this change does not create the possibility of a new or different 
kind of accident from any accident previously evaluated.

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

    The extended time allowed to perform a missed surveillance does 
not result in a significant reduction in the margin of safety. As 
supported by the historical data, the likely outcome of any 
surveillance is verification that the LCO [Limiting Condition for 
Operation] is met. Failure to perform a surveillance within the 
prescribed frequency does not cause equipment to become inoperable. 
The only effect of the additional time allowed to perform a missed 
surveillance on the margin of safety is the extension of the time 
until inoperable equipment is discovered to be inoperable by the 
missed surveillance. However, given the rare occurrence of 
inoperable equipment, and the rare occurrence of a missed 
surveillance, a missed surveillance on inoperable equipment would be 
very unlikely. This must be balanced against the real risk of 
manipulating the plant equipment or condition to perform the missed 
surveillance. In addition, parallel trains and alternate equipment 
are typically available to perform the safety function of the 
equipment not tested. Thus, there is confidence that the equipment 
can perform its assumed safety function.
    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: Peter Marquardt, Legal Department, 688 WCB, 
Detroit Edison Company, 2000 2nd Avenue, Detroit, Michigan 48226-1279.

[[Page 64290]]

    NRC Section Chief: William D. Reckley, Acting.

Dominion Nuclear Connecticut, Inc., et al., Docket Nos. 50-245, 50-336, 
and 50-423, Millstone Nuclear Power Station, Unit Nos. 1, 2, and 3, New 
London County, Connecticut

    Date of amendment request: November 8, 2001.
    Description of amendment request: The proposed amendments would 
incorporate administrative and editorial changes into the Millstone 
Unit No. 1 Permanently Defueled Technical Specifications (PDTS) and 
into the Millstone Unit Nos. 2 and 3 Technical Specifications (TSs). 
Specifically, the proposed changes would: (1) Relocate redundant design 
features information already included in other licensing basis (LB) 
documents (e.g., the Final Safety Analysis Report (FSAR)), from Section 
5.0, ``Design Features,'' of the Unit Nos. 2 and 3 TS, to other LB 
documents, consistent with the improved Standard Technical 
Specifications (STSs) for the respective unit design; (2) revise TS 
5.6.2, ``Technical Specifications Bases Control Program,'' in the Unit 
No. 1 PDTS to incorporate the 10 CFR 50.59 rule change; (3) add a new 
TS (TS 6.22 for Unit No. 2 and TS 6.17 for Unit No. 3), to incorporate 
a TS bases control program within the Unit Nos. 2 and 3 TS; (4) add a 
new TS (TS 6.18, ``Component Cyclic or Transient Limits''), to the Unit 
No. 3 TS to define the program for tracking cyclic (or transient) 
limits. These limits are proposed to be relocated from where they are 
listed in TS 5.7, ``Component Cyclic or Transient Limit,'' in the Unit 
No. 3 TS, to the FSAR; (5) revise the Unit No. 1 PDTS and the Unit Nos. 
2 and 3 TS related to Radiological Environmental Monitoring Program 
(REMP) procedure processing to: (a) remove reference to an organization 
affiliated with Northeast Utilities (NU), the Production Operations 
Services Laboratory, which is no longer applicable following the change 
in ownership from NU to Dominion Nuclear Connecticut (DNC); (b) replace 
the reference to the Radiological Assessment Branch (a Millstone DNC 
organization) with the ``organization responsible for the REMP'' for 
review/approval of changes to the REMP to avoid future TS changes due 
to a change in organizational titles; (c) correct an inconsistency 
within the Unit No. 1 PDTS which implies that REMP procedures are 
processed under the general procedure processing specification (i.e., 
TS 5.5.1), in addition to the specific specifications for processing 
REMP procedure changes (i.e., Specifications 5.5.6 and 5.5.7); and (6) 
correct miscellaneous editorial issues and achieve consistency between 
the TSs for each unit. These changes include: (a) Changes to and 
corrections in titles; (b) correct references to the quality assurance 
program, and (c) change titles to utilize the term radiation protection 
rather than health physics.
    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 changes related to Section 5, ``Design Features,'' 
of the Unit Nos. 2 or 3 TS either relocates or deletes certain 
details from the Technical Specifications and relocates them to the 
respective unit's updated FSAR or other plant controlled documents. 
The FSAR and other plant controlled documents will be maintained in 
accordance with 10 CFR 50.59. The proposed changes to Section 6, 
``Administrative Controls,'' adds new administrative specifications 
consistent with the guidance of the improved STS, corrects 
inconsistencies, or represents changes in nomenclature, and will 
correct editorial issues and achieve consistency within the 
individual TS and between individual TS. The changes are purely 
administrative or editorial and do not alter any regulatory 
requirements or have an impact on the acceptance criteria for any 
design basis accident described in the respective Unit Nos. 2 or 3 
FSAR or the Unit No. 1 Defueled Safety Analysis Report (DSAR).
    These changes have no impact on plant equipment operation. Since 
the changes are solely an administrative or editorial change to the 
TS, they cannot affect the likelihood or consequences of accidents. 
Therefore, these 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 have no impact on plant operation. Since 
the proposed changes are solely an administrative or editorial 
change to the TS, they do not affect plant operation in any way. The 
proposed changes do not involve a physical alteration of the plant 
or change the plant configuration (no new or different type of 
equipment will be installed). The proposed changes do not require 
any new or unusual operator actions. The changes do not alter the 
way any structure, system, or component functions and do not alter 
the manner in which the plant is operated. The 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 administrative or 
editorial changes to the TS, they do not affect plant operation in 
any way. The proposed changes to the respective unit's technical 
specifications will standardize terminology, remove extraneous 
information and make minor format changes that will not result in 
any technical changes to current requirements.
    The proposed changes do not impact any acceptance criteria for 
the design basis accidents described in the respective Unit Nos. 2 
or 3 FSAR or the Unit No. 1 DSAR and do not impact the consequences 
of accidents previously evaluated. 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.

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

    Date of amendment request: May 25, 2001.
    Description of amendment request: The amendments would revise 
Technical Specifications (TS) Definitions for ENGINEERED SAFETY FEATURE 
(ESF) RESPONSE TIME and REACTOR TRIP SYSTEM (RTS) RESPONSE TIME to 
provide for verification of response time for selected components 
provided that the components and the methodology for verification have 
been previously reviewed and approved by the Nuclear Regulatory 
Commission. The associated Bases will also be revised. The licensee has 
referenced previously approved WCAP-13632-P-A, Revision 2, 
``Elimination of Pressure Sensor Response Time Testing Requirements,'' 
and WCAP-14036-P-A Revision 1, ``Elimination of Periodic Protection 
Channel Response Time Tests'' as the justifications for proposing these 
changes.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:


[[Page 64291]]


    Conformance of the proposed amendments to the standards for a 
determination of no significant hazards as defined in 10 CFR 50.92 
is shown in the following:
    (1) The proposed license amendments do not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    This change to the TS does not result in a condition where the 
design, material, and construction standards that were applicable 
prior to the change are altered. The same RTS and ESFAS 
instrumentation is being used; the time response allocations/
modeling assumptions in the UFSAR Chapter 15 analyses are still the 
same; only the method of verifying time response is changed. The 
proposed change will not modify any system interface and could not 
increase the likelihood of an accident since these events are 
independent of this change. The proposed activity will not change, 
degrade, or prevent actions or alter any assumptions previously made 
in evaluating the radiological consequences of an accident described 
in the UFSAR. Therefore, the proposed amendments do not result in 
any increase in the probability or consequences of an accident 
previously evaluated.
    (2) The proposed license amendments do not create the 
possibility of a new or different kind of accident from any accident 
previously evaluated.
    This change does not alter the performance of the reactor 
protection system (RPS) or the engineered safety features actuation 
system (ESFAS). All RPS and ESFAS channels will still have response 
time verified by test before placing the channel in operational 
service and after any maintenance that could affect response time. 
Changing the method of periodically verifying instrument response 
for certain RPS and ESFAS channels (assuring equipment operability) 
from time response testing to calibration and channel checks will 
not create any new accident initiators or scenarios. Periodic 
surveillance of these instruments will detect significant 
degradation in the channel characteristic. Implementation of the 
proposed amendments does not create the possibility of a new or 
different kind of accident from any accident previously evaluated.
    (3) The proposed license amendments do not involve a significant 
reduction in a margin of safety.
    This change does not affect the total system response time 
assumed in the safety analysis. The periodic system response time 
verification method is modified to allow use of actual test data or 
engineering data. The method of verification still provides 
assurance that the total system response is within that defined in 
the safety analysis, since calibration tests will detect any 
degradation which might significantly affect channel response time. 
Based on the above, it is concluded that the proposed license 
amendment request does not result in a reduction in a margin with 
respect to plant safety.
    Based on the preceding analysis, it is concluded that 
elimination of periodic [response time testing] RTT is acceptable 
and the proposed license amendments do not involve a significant 
hazards consideration finding as defined in 10 CFR 50.92.

    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: Ms. Lisa F. Vaughn , Legal Department 
(PB05E), Duke Energy Corporation, 422 South Church Street, Charlotte, 
North Carolina 28201-1006.
    NRC Section Chief: Richard J. Laufer, Acting.

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

    Date of amendment request: August 6, 2001.
    Description of amendment request: The amendments would revise 
Technical Specifications (TS) 3.3.2 for engineered safety feature 
actuation system instrumentation, TS 3.3.6 for containment purge and 
exhaust isolation instrumentation. The amendments would also revise the 
appropriate bases, and the bases for Containment Isolation Valves (TS 
3.6.3). Specifically, the proposed amendments would modify the TS 
requirements so that they exclude the Containment Purge Ventilation 
System and the Hydrogen Purge System, thereby applying the requirements 
to only the Containment Air Release and Addition System. At Catawba, 
the containment isolation valves for the Containment Purge Ventilation 
System and the Hydrogen Purge System are sealed closed in the modes of 
applicability (Modes 1, 2, 3, and 4) according to TS requirements.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    The following discussion is a summary of the evaluation of the 
changes contained in this proposed amendment against the 10 CFR 
50.92(c) requirements to demonstrate that all three standards are 
satisfied. A no significant hazards consideration is indicated if 
operation of the facility in accordance with the proposed amendment 
would not:
    1. Involve a significant increase in the probability or 
consequences of an accident previously evaluated, or
    2. Create the possibility of a new or different kind of accident 
from any accident previously evaluated, or
    3. Involve a significant reduction in a margin of safety.

First Standard

    Implementation of this amendment would not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated. Neither the Containment Purge Ventilation 
System, the Hydrogen Purge System, nor the Containment Air Release 
and Addition System is capable of by itself initiating any accident. 
The safety related portions of these systems, which are responsible 
for maintaining containment isolation during accident conditions, 
will continue to function as designed, and in accordance with all 
applicable TS. The design and operation of the systems are not being 
modified by this proposed amendment. Therefore, there will be no 
impact on any accident probabilities or consequences.

Second Standard

    Implementation of this amendment would not create the 
possibility of a new or different kind of accident from any accident 
previously evaluated. No new accident causal mechanisms are created 
as a result of NRC approval of this amendment request. No changes 
are being made to the plant which will introduce any new accident 
causal mechanisms. This amendment request does not impact any plant 
systems that are accident initiators and does not impact any safety 
analyses.

Third Standard

    Implementation of this amendment would not involve a significant 
reduction in a margin of safety. Margin of safety is related to the 
confidence in the ability of the fission product barriers to perform 
their design functions during and following an accident situation. 
These barriers include the fuel cladding, the reactor coolant 
system, and the containment system. The performance of these fission 
product barriers will not be impacted by implementation of this 
proposed amendment. It has already been shown that the performance 
of all containment isolation functions in response to accident 
conditions will not be impacted by this proposed amendment. There is 
no risk significance to this proposed amendment, as no reduction in 
system or component availability will be incurred. No safety margins 
will be impacted.
    Based upon the preceding discussion, Duke Energy has concluded 
that the proposed amendment does not involve a significant hazards 
consideration.

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

[[Page 64292]]

Entergy Gulf States, Inc., and Entergy Operations, Inc., Docket No. 50-
458, River Bend Station, Unit 1, West Feliciana Parish, Louisiana

    Date of amendment request: September 24, 2001.
    Description of amendment request: The amendment request proposes to 
extend the allowed outage time for a Division I or Division II 
Emergency Diesel Generator (EDG) from 72 hours to 14 days. The proposed 
changes are intended to provide flexibility in scheduling EDG 
maintenance activities, reduce refueling outage duration, and improve 
EDG availability during plant shutdowns.
    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 proposed Technical Specification (TS) changes do not affect 
the design, operational characteristics, function, or reliability of 
the EDGs. The EDGs are not the initiators of previously evaluated 
accidents. The EDGs are designed to mitigate the consequences of 
previously evaluated accidents including a loss of offsite power. 
Extending the allowed outage time (AOT) for a single EDG would not 
significantly affect the previously evaluated accidents since the 
remaining EDGs supporting the redundant ESF [Engineered Safety 
Feature] systems would continue to perform the accident mitigating 
functions as designed.
    The duration of a TS AOT is determined considering that there is 
a minimal possibility that an accident will occur while a component 
is removed from service. A risk-informed assessment was performed 
which concluded that the increase in plant risk is small and 
consistent with the USNRC [United States Nuclear Regulatory 
Commission (NRC)] ``Safety Goals for the Operations of Nuclear Power 
Plants; Policy Statement,'' Federal Register, Vol. 51, p.30028 (51 
FR 30028), August 4, 1986, as further described by NRC Regulatory 
Guide 1.177.
    The current TS requirements establish controls to ensure that 
redundant systems relying on the remaining EDGs are Operable. In 
addition to these requirements, administrative controls will be 
established to provide assurance that the AOT extension is not 
applied during adverse weather conditions that could potentially 
affect offsite power availability.
    Both the RBS [River Bend Station, Unit 1] risk-based analysis 
and the deterministic evaluation support the increased AOT. 
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 TS changes do not involve a change in the design, 
configuration, or method of operation of the plant that could create 
the possibility of a new or different kind of accident. The proposed 
change extends the AOT currently allowed by the TS.
    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 extended AOT is not in conflict with any of the 
approved codes and standards applicable to the onsite AC 
[Alternating Current] power sources. The proposed changes do deviate 
from the recommendations of Regulatory Guide (RG) 1.93. An extension 
of the 72 hour AOT recommended in the RG to 14 days is demonstrated 
herein to be acceptable and has been approved for several other 
licensees. Assuming there are no additional failures of redundant 
equipment during the time that the EDG is removed from service, the 
intended safety functions would still be met.
    The proposed AOT change does not affect any of the assumptions 
or inputs to the safety analyses of the FSAR [Final Safety 
Assessment Report] and does not erode the decrease in severe 
accident risk achieved with the issuance of the Station Blackout 
(SBO) Rule, 10 CFR 50.63 ``Loss of All Alternating Current Power''. 
RBS is classified as a four-hour coping plant with 0.95 EDG 
reliability (see U[pdated] FSAR Appendix 15C). The assumptions used 
in the SBO [Station Blackout] analysis regarding reliability of the 
EDGs are unaffected by the proposed TS changes since preventive 
maintenance and testing will continue to be performed to maintain 
reliability assumptions.
    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: Mark Wetterhahn, Esq., Winston & Strawn, 
1400 L Street, NW., Washington, DC 20005.
    NRC Section Chief: Robert A. Gramm.

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

    Date of amendment request: October 23, 2001.
    Description of amendment request: The proposed amendment would 
revise Technical Specification (TS) 5.5.10, ``Ventilation Filter 
Testing Program,'' to adopt the requirements of the American Society 
for Testing and Materials Standard (ASTM) D3803-1989, ``Standard Test 
Method for Nuclear-Grade Activated Carbon.'' The proposed TS revisions 
are in response to Nuclear Regulatory Commission (NRC) Generic Letter 
99-02, ``Laboratory Testing of Nuclear-Grade Activated Charcoal.'' The 
proposed amendment would also revise the differential pressure criteria 
for the test of the filter system for the Control Room Ventilation 
System.
    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?
    The proposed license amendment adopts the new test method and 
acceptance criteria of ASTM D3803-1989, with the exceptions 
identified, for activated charcoal filters and changes the allowable 
pressure differential for Control Room ventilation. The changes 
require laboratory performance testing of adsorber carbon that 
yields a more accurate result than the testing currently required by 
the TS and requires a more stringent limit on the Control Room 
ventilation pressure differential. The proposed change to delete 
non-conservative TS requirements for testing of adsorber carbon and 
limiting the Control Room ventilation differential pressure are not 
plant accident initiators as described in the Final Safety Analysis 
Report (FSAR). The proposed amendment does not change the function 
of any structure, system or component (SSC). The function of the 
ventilation systems is filtration of radiological releases during 
postulated accidents. The proposed changes will provide greater 
assurance that this function is provided. The revised TS 
requirements are for laboratory tests and pressure differential 
measurements that are currently in place and change only the TS 
testing requirements. They will not result in any changes to the 
efficiency assumed in accident analysis. The changes do not alter, 
degrade or prevent actions described or assumed in an accident 
described in the FSAR. Therefore, the proposed amendment does not 
change the possibility of an accident previously evaluated or 
significantly increase the 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 license amendment adopts the new test method and 
acceptance criteria of ASTM D3803-1989, with the exceptions

[[Page 64293]]

identified, for activated charcoal filters and changes the allowable 
pressure differential for Control Room ventilation. The change does 
not involve any modifications to the plant, will not require changes 
to how the plant is operated nor will it affect the operation of the 
plant. The changes require laboratory performance testing of 
adsorber carbon that yields a more accurate result than the testing 
currently required by the TS and requires a more stringent limit on 
the Control Room ventilation pressure differential. The proposed 
changes to delete non-conservative TS requirements for testing of 
adsorber carbon and limiting the Control Room ventilation 
differential pressure are not plant accident initiators as described 
in the Final Safety Analysis Report (FSAR). The proposed amendment 
does not change the function of any structure, system or component 
(SSC). The function of the ventilation systems is filtration of 
radiological releases during postulated accidents. The proposed 
changes will provide greater assurance that this function is 
provided. The revised TS requirements are for laboratory tests and 
pressure differential measurements that are currently in place and 
change only the TS testing requirements. They will not result in any 
changes to the efficiency assumed in accident analysis. The changes 
do not alter, degrade or prevent actions described or assumed in an 
accident described in the FSAR. 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 proposed license amendment involve a significant 
reduction in a margin of safety?
    The proposed license amendment adopts the new test method and 
acceptance criteria of ASTM D3803-1989, with the exceptions 
identified, for activated charcoal filters and changes the allowable 
pressure differential for Control Room ventilation. The proposed 
license amendment does not reduce the margin of safety but enhances 
by requiring more accurate testing and a more conservative pressure 
differential. The proposed test change will require the use of a 
current and improved ASTM standard to ensure that the carbon ability 
to adsorb radioactive material will remain at or above the 
capability credited in our accident analysis. The proposed 
differential pressure limit will assure that the system provides 
sufficient flow though the charcoal to meet accident analyses.

    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 Operations, Inc., 440 Hamilton Avenue, White Plains, NY 
10601.
    NRC Section Chief: L. Raghavan (Acting).

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

    Date of amendment request: October 23, 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 Nuclear Regulatory Commission (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 23, 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

[[Page 64294]]

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

    Date of amendment request: September 28, 2001.
    Description of amendment request: The licensee proposes to revise a 
single Anticipated Transient Without Scram (ATWS) Recirculation Pump 
Trip Reactor Pressure High setpoint to replace the current conditional 
setpoints which are based upon the number of Safety Relief Valves out 
of service.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Involve a significant increase in the probability or 
consequences of an accident previously evaluated because: a change 
in the ATWS high RPV [reactor pressure vessel] pressure RWR [ATWS 
Reactor Pressure High Recirculation Pump] pump trip setpoint does 
not affect initiation of any accident. Operation in accordance with 
the revised setpoint ensures the consequences of previously analyzed 
accidents are not changed.
    2. Create the possibility of a new or different kind of accident 
from any accident previously evaluated because: RPV pressure 
following an ATWS with PRFO [Pressure Regulating Valve Open] event 
(worst case transient for RPV pressurization) remains within 
acceptable limits with the revised setpoint. Therefore, changing the 
setpoint will not lead to a new or different kind of accident.
    3. Involve a significant reduction in a margin of safety 
because: the analyses performed to determine the revised ATWS high 
pressure RWR pump trip setpoint assure maintenance of the same 
margin of safety as presently exists for limiting RPV pressure 
following an ATWS with PRFO (limiting transient). The current 
analyses actually shows an improved margin over the results of the 
previous analyses (References 2 and 3), which were performed using 
an earlier computer code.

    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. David E. Blabey, 1633 Broadway, New 
York, New York 10019.
    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 30, 2001.
    Description of amendment request: The license amendment request 
proposes changes to Arkansas Nuclear One, Unit 2 (ANO-2) Technical 
Specification (TS) 3.4.9, ``Pressure/Temperature Limits,'' and TS 
3.4.12, ``Low Temperature Overpressure Protection (LTOP) System.'' The 
primary changes are to update the existing pressure/temperature (P/T) 
limits from 21 to 32 effective full power years (EFPYs) and to include 
additional restrictions in the LTOP TSs.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

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

    The probability of occurrence of an accident previously 
evaluated for ANO-2 is not altered by the proposed amendment to the 
technical specifications (TSs). The accidents remain the same as 
currently analyzed in the ANO-2 Safety Analysis Report (SAR) as a 
result of changes to the P/T limits as well as those for LTOP. The 
new P/T and LTOP limits were based on NRC [Nuclear Regulatory 
Commission] accepted methodologies along with ASME [American Society 
of Mechanical Engineers] Code [Boiler and Pressure Vessel Code] 
alternatives. The proposed changes do not impact the integrity of 
the reactor coolant pressure boundary (RCPB) (i.e. there is no 
change to the operating pressure, materials, loadings, etc.) as a 
result of this change. In addition, there is no increase in the 
potential for the occurrence of a loss of coolant accident. The 
probability of any design basis accident is not affected by this 
change, nor are the consequences of any design basis accident (DBA) 
affected by this proposed change. The proposed P/T limit curves and 
the LTOP limits are not considered to be an initiator or contributor 
to any accident currently evaluated in the ANO-2 SAR. These new 
limits ensure the long term integrity of the RCPB.
    Fracture toughness test data are obtained from material 
specimens contained in capsules that are periodically withdrawn from 
the reactor vessel. These data permit determination of the 
conditions under which the vessel can be operated with adequate 
safety margins against non-ductile fracture throughout its service 
life. A new reactor vessel specimen capsule was withdrawn at the 
most recent refueling outage and was analyzed to predict the 
fracture toughness requirements using projected neutron fluence 
calculations. For each analyzed transient and steady state 
condition, the allowable pressure is determined as a function of 
reactor coolant temperature considering postulated flaws in the 
reactor vessel beltline, inlet nozzle, outlet nozzle, and closure 
head.
    The predicted radiation induced RTNDT [shift 
in reference temperature for nil-ductility transition] was 
calculated using the respective reactor vessel beltline materials 
copper and nickel contents and the neutron fluence applicable to 32 
EFPY including an estimated increase in flux due to a proposed power 
uprate. The RTNDT [reference temperature for 
nil-ductility transition] and, in turn, the operating limits for 
ANO-2 were adjusted to account for the effects of irradiation on the 
fracture toughness of the reactor vessel materials. Therefore, new 
operating limits are established which are represented in the 
revised operating curves for heatup/criticality, cooldown and 
inservice hydrostatic testing contained in the technical 
specifications.
    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 changes to the P/T and LTOP limits will not create 
a new accident scenario. The requirements to have P/T and LTOP 
protection are part of the licensing basis of ANO-2. The proposed 
changes

[[Page 64295]]

reflect the change in vessel material properties acknowledged and 
managed by regulation and the best data available in response to NRC 
Generic Letter 92-01, Revision 1. The approach used meets NRC and 
ASME regulations and guidelines. The calculational methodology for 
fluence is based on an NRC approved Framatome ANP approach. 
Therefore, the adjusted reference temperatures for fracture 
toughness are consistent with that previously provided to the NRC. 
The data analysis for the vessel specimen removed at 2R14 
(approximately 15.7 EFPY of exposure) confirms that the vessel 
materials are responding as predicted.
    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 existing P/T curves and LTOP limits in the technical 
specifications are reaching their expiration period for the number 
of years at effective full power operation. The revision of the P/T 
limits and curves will ensure that ANO-2 continues to operate within 
the operating margins allowed by 10 CFR 50.60 and the ASME Code. The 
material properties used in the analysis are based on results 
established through CE [Combustion Engineering] material reports for 
copper and nickel content. The application of ASME Code Case N-641 
presents alternative procedures for calculating P/T and LTOP 
temperatures and pressures in lieu of that established for ASME 
Section XI, Appendix G-2215. This Code alternative allows certain 
assumptions to be conservatively reduced. However, the procedures 
allowed by Code Case N-641 still provide significant conservatism 
and ensure an adequate margin of safety in the development of P/T 
operating and pressure test limits to prevent non-ductile fractures.
    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.

Indiana Michigan Power Company, Docket No. 50-315, Donald C. Cook 
Nuclear Plant, Unit 1, Berrien County, Michigan

    Date of amendment request: November 19, 2001.
    Description of amendment request: The proposed amendment would to 
eliminate restrictions imposed by technical specification (TS) 3.0.4 
for the Remote Shutdown 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. Does the proposed change involve a significant increase in 
the probability of occurrence or consequences of an accident 
previously evaluated?
    Response: No.

Probability of Occurrence of an Accident Previously Evaluated

    The Remote Shutdown Instrumentation system ensures that 
sufficient capability is available to permit shutdown and 
maintenance of Hot Standby of the plant from locations outside of 
the control room. The proposed change allows Unit 1 to ascend in 
mode without meeting the LCO [limiting condition for operation] for 
TS 3.3.3.5. The proposed change does not impact the ability to 
comply with the allowed outage time (AOT) described in TS 3.3.3.5. 
As such, the proposed change does not affect any accident initiators 
or precursors, since the AOT for TS 3.3.3.5 will continue to be met. 
The proposed change is also consistent with the Unit 2 TS. 
Therefore, the probability of occurrence of an accident previously 
evaluated is not significantly increased.
    The format changes do not impact any accident initiators or 
precursors. Thus, the probability of occurrence of an accident 
previously evaluated is not significantly increased.

Consequences of an Accident Previously Evaluated

    The proposed change to allow Unit 1 to ascend in mode without 
meeting the LCO for TS 3.3.3.5, while continuing to meet the action 
statement, will not significantly impact the Remote Shutdown 
Instrumentation systems' capability of performing its design 
function. The Remote Shutdown Instrumentation ensures that 
sufficient capability is available to permit shutdown and 
maintenance of Hot Standby of the plant from locations outside of 
the control room. The proposed change does not impact the ability to 
comply with AOT described in TS 3.3.3.5. The proposed change is also 
consistent with the Unit 2 TS. Thus, there will be no increase in 
offsite doses, and the consequences of an accident previously 
analyzed are not increased.
    The format changes do not impact the function of the Remote 
Shutdown Instrumentation. Thus, there will be no increase in offsite 
doses, and the consequences of an accident previously analyzed are 
not significantly increased.
    2. Does the proposed change create the possibility of a new or 
different kind of accident from any accident previously evaluated?
    Response: No.
    The Remote Shutdown Instrumentation system ensures that 
sufficient capability is available to permit shutdown and 
maintenance of Hot Standby of the plant from locations outside of 
the control room. Allowing Unit 1 to ascend in mode without meeting 
the LCO for TS 3.3.3.5, while continuing to meet the action 
statement, does not change the function of the Remote Shutdown 
Instrumentation system or create the possibility of a new or 
different type of accident. The proposed change does not impact the 
ability to comply with the AOT described in TS 3.3.3.5. The proposed 
change is also consistent with the Unit 2 TS. Therefore, the 
proposed change does not create the possibility of a new or 
different kind of accident from any previously evaluated.
    The format changes do not create the possibility of a new or 
different kind of accident from any previously evaluated.
    3. Does the proposed change involve a significant reduction in a 
margin of safety?
    Response: No.
    The proposed change does not impact the Remote Shutdown 
Instrumentation system's capability of performing its design 
function, nor does the proposed change impact the operational 
characteristics of the Remote Shutdown Instrumentation system. The 
Remote Shutdown Instrumentation ensures that sufficient capability 
is available to permit shutdown and maintenance of Hot Standby of 
the plant from locations outside of the control room. Allowing Unit 
1 to ascend in mode without meeting the LCO for TS 3.3.3.5, while 
continuing to meet the action statement, does not impact CNP's 
accident analysis. The proposed change is also consistent with the 
Unit 2 TS. Therefore, the proposed change does not involve a 
significant reduction in the margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment requests involve no significant hazards consideration.
    Attorney for licensee: David W. Jenkins, Esq., 500 Circle Drive, 
Buchanan, MI 49107.
    NRC Section Chief: William D. Reckley, Acting.

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

    Date of amendment requests: October 12, 2001.
    Description of amendment requests: The proposed amendments would 
delete 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

[[Page 64296]]

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 October 12, 2001.
    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--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.

    Therefore, the NRC staff proposes to determine that the amendment 
requests involve no significant hazards consideration.
    Attorney for licensee: David W. Jenkins, Esq., 500 Circle Drive, 
Buchanan, MI 49107.
    NRC Section Chief: William D. Reckley, Acting.

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

    Date of amendment requests: November 1, 2001.
    Description of amendment requests: The proposed amendments would 
revise technical specification (TS) surveillance requirements (SR) 
4.8.2.3.2.c.2 and 4.8.2.5.2.c.2 and associated TS bases concerning the 
safety-related batteries to make them more consistent with the 
Westinghouse Standard TSs.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Does the change involve a significant increase in the 
probability of occurrence or consequences of an accident previously 
evaluated?

Probability of Occurrence of an Accident Previously Evaluated

    The proposed change to SRs 4.8.2.3.2.c.2 and 4.8.2.5.2.c.2 to 
add a requirement to remove visible corrosion and to delete the 
requirement that the battery be free of corrosion does not affect 
any accident initiators or precursors. The batteries perform a 
mitigating function following a loss of AC power, and the presence 
of corrosion will not adversely impact components whose failure 
would initiate an accident. Thus, the probability of occurrence of 
an accident previously evaluated is not significantly increased.
    The proposed change to the TS 3/4.8 bases provides clarification 
and does not affect any accident initiators or precursors. Thus, the 
probability of occurrence of an accident previously evaluated is not 
significantly increased.
    The proposed change to SRs 4.8.2.3.2.c.3 and 4.8.2.5.2.c.3 
increases the battery charger current required during surveillance 
testing. The required value is within the capability of

[[Page 64297]]

the battery charger. Thus, the battery charger is not degraded by 
this change, and the change does not affect any accident initiators 
or precursors. Thus, the probability of occurrence of an accident 
previously evaluated is not significantly increased.
    The proposed changes to SR 4.8.2.3.2.d delete the requirement 
that the battery terminal voltage be maintained greater than or 
equal to 210 volts during the battery service test, and delete the 
description of the composite load profile. The removal of the 
requirement and the description from the SR do not affect any 
accident initiators or precursors. Thus, the probability of 
occurrence of an accident previously evaluated is not significantly 
increased.
    The deletion of Tables 4.8-2 and 4.8-3, the incorporation of the 
words ``this page intentionally left blank,'' and the deletion of 
the SR 4.8.2.3.2.d and SR 4.8.2.5.2.d references to the tables do 
not impact battery operation as the tables summarize information 
used as calculation inputs. These changes do not affect any accident 
initiators or precursors. Thus, the probability of occurrence of an 
accident previously evaluated is not significantly increased.
    The proposed changes to SR 4.8.2.5.2.d to delete the requirement 
that the battery terminal voltage be maintained greater than or 
equal to 210 volts during the battery service test, and to add the 
term ``design duty cycle'' does not affect any accident initiators 
or precursors. Thus, the probability of occurrence of an accident 
previously evaluated is not significantly increased.
    The editorial change does not impact any accident initiators or 
precursors. Thus, the probability of occurrence of an accident 
previously evaluated is not significantly increased.

Consequences of an Accident Previously Evaluated

    The batteries and their associated chargers provide power to 
emergency equipment that is used in the mitigation of accidents. The 
batteries provide power to this equipment following a loss of AC 
power until the battery chargers are powered by the emergency diesel 
generators.
    The proposed change to SRs 4.8.2.3.2.c.2 and 4.8.2.5.2.c.2 to 
add a requirement to remove visible corrosion and to delete the 
requirement that the battery connections be free of corrosion does 
not impact a battery's capability to power its safety-related loads 
as the presence of corrosion at the terminal connections does not 
indicate that the battery is unable to perform its function. Rather, 
it is the impact of the corrosion on the connections that is of 
concern. This concern will be addressed by performing a resistance 
check to verify that battery performance is acceptable. Therefore, 
this change does not result in an increase in offsite doses. Thus, 
the consequences of an accident previously analyzed are not 
increased.
    The proposed change to the TS 3/4.8 bases provides clarification 
and does not impact the battery's capability to power its safety-
related loads. Thus, the consequences of an accident previously 
analyzed are not increased.
    The proposed change to SRs 4.8.2.3.2.c.3 and 4.8.2.5.2.c.3 to 
increase the required battery charger current ensures that the 
battery charger has sufficient capacity to provide power to 
emergency equipment while simultaneously recharging batteries that 
were discharged following a loss of AC power. This ensures that 
emergency equipment connected to the battery will continue to 
operate as designed, and offsite doses will not be increased. Thus, 
the consequences of an accident previously analyzed are not 
increased.
    The proposed changes to SR 4.8.2.3.2.d delete the requirement 
that the battery terminal voltage be maintained greater than or 
equal to 210 volts during the battery service test, and delete the 
description of the composite load profile. However, the SR will 
still require that the service test demonstrate that the battery 
capacity is adequate to supply emergency loads. The voltage 
requirements for the batteries are determined by battery-system 
specific calculations, and the calculation results are incorporated 
into the test procedures. This assures that the equipment connected 
to the battery will continue to operate as designed, and offsite 
doses will not be increased. Thus, the consequences of an accident 
previously analyzed are not increased.
    The deletion of Tables 4.8-2 and 4.8-3, the addition of the 
words ``this page intentionally left blank,'' and the deletion of 
the SR 4.8.2.3.2.d and SR 4.8.2.5.2.d references to the tables do 
not impact battery operation as the tables summarize information 
used as calculation inputs. The batteries are tested to a load 
profile that is developed on the basis of the battery loads for a 
loss of AC power, and the testing assures that the batteries are 
capable of performing their safety function. Thus, these changes 
will not impact battery capability, will not result in an increase 
in offsite doses, and the consequences of an accident previously 
analyzed are not increased.
    The proposed changes to SR 4.8.2.5.2.d to delete the requirement 
that the battery terminal voltage be maintained greater than or 
equal to 210 volts during the battery service test, and to add the 
term ``design duty cycle'' requires that the battery be tested in 
accordance with a load profile developed on the basis of the battery 
loads for a loss of AC power. The testing of the battery assures 
that it is capable of performing its safety function. Thus, the 
capability of the battery is not impacted, there will be no increase 
in offsite doses, and the consequences of an accident previously 
analyzed are not increased.
    The editorial change does not impact battery capability. Thus, 
there will be no increase in offsite doses, and the consequences of 
an accident previously analyzed are not increased.
    Therefore, the probability of occurrence or the consequences of 
accidents previously evaluated are not increased.
    2. Does the change create the possibility of a new or different 
kind of accident from any accident previously evaluated?
    The batteries perform a mitigating function by providing power 
to emergency equipment following a loss of AC power.
    The proposed change to SRs 4.8.2.3.2.c.2 and 4.8.2.5.2.c.2 adds 
a requirement to remove visible corrosion and deletes the 
requirement that the battery terminals be free of corrosion. The 
presence of corrosion on the battery terminals does not introduce a 
mechanism that would cause a plant transient, and I&M will ensure 
that the corrosion does not impact the battery's function. Thus, the 
possibility of a new or different kind of accident is not created.
    The proposed change to the TS 3/4.8 bases provides clarification 
and does not introduce a mechanism that would cause a plant 
transient. Thus, the possibility of a new or different kind of 
accident is not created.
    The proposed change to SRs 4.8.2.3.2.c.3 and 4.8.2.5.2.c.3 
increases the acceptance criterion for battery charger current to 
reflect the present demand on the battery charger when it is 
simultaneously supplying power to emergency equipment and charging a 
discharged battery. The increase in the acceptance criterion is 
within the capability of the battery charger, and no failure 
mechanisms are introduced by this change. Thus, the change does not 
create the possibility of a new or different kind of accident from 
any previously evaluated.
    The proposed changes to SR 4.8.2.3.2.d to delete the requirement 
that the battery terminal voltage be maintained greater than or 
equal to 210 volts during a battery service test, and to delete the 
load profile description do not directly impact any emergency 
equipment as the SR continues to require that the battery service 
test demonstrate that the battery is capable of supplying power to 
connected equipment, and this change does not introduce any battery 
failure mechanisms. Thus, the change does not create the possibility 
of a new or different kind of accident from any previously 
evaluated.
    The deletion of Tables 4.8-2 and 4.8-3, the incorporation of the 
words ``this page intentionally left blank,'' and the deletion of 
the SR 4.8.2.3.2.d and SR 4.8.2.5.2.d references to the tables do 
not impact battery operation as the tables summarize information 
used as calculation inputs. Thus, the changes do not create the 
possibility of a new or different kind of accident from any 
previously evaluated.
    The proposed changes to SR 4.8.2.5.2.d to delete the requirement 
that the battery terminal voltage be maintained greater than 210 
volts during a battery service test, and to add the term ``design 
duty cycle'' do not introduce any battery failure mechanisms as they 
do not alter the battery's physical characteristics or the battery 
testing requirements. Additionally, the term ``design duty cycle'' 
more accurately reflects the use of a simulated load for the battery 
test. Thus, the change does not create the possibility of a new or 
different kind of accident from any previously evaluated.
    The editorial change does not create the possibility of a new or 
different kind of accident from any previously evaluated.
    Therefore, the proposed changes do not create the possibility of 
a new or different kind of accident from any accident previously 
evaluated.
    3. Does the change involve a significant reduction in a margin 
of safety?
    The proposed changes do not impact the functional requirements 
of either the

[[Page 64298]]

batteries or the battery chargers, nor do the changes impact the 
operational characteristics of the equipment that is connected to 
the battery. The batteries will continue to be subjected to a system 
test to verify that the battery capacity is adequate, and the 
battery chargers will be tested to verify that they are capable of 
meeting their rated capacity. These tests will demonstrate that the 
batteries and the battery chargers are capable of performing their 
mitigation function for analyzed accidents.
    Therefore, the proposed changes do not involve a significant 
reduction in a margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment requests involve no significant hazards consideration.
    Attorney for licensee: David W. Jenkins, Esq., 500 Circle Drive, 
Buchanan, MI 49107.
    NRC Section Chief: William D. Reckley, Acting.

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

    Date of amendment requests: November 16, 2001.
    Description of amendment requests: The proposed amendments would 
revise technical specification (TS) Table 3.3-4, ``Engineered Safety 
Feature Actuation System Instrumentation Trip Setpoints.'' The proposed 
changes are part of a planned design change to replace the existing 4kV 
offsite power transformers, loss of voltage relays, and degraded 
voltage relays with components of an improved design to increase the 
reliability of offsite power for safety-related equipment.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Does the change involve a significant increase in the 
probability of occurrence or consequences of an accident previously 
evaluated?

Probability of Occurrence of an Accident Previously Evaluated

    The proposed changes to the degraded voltage and loss of voltage 
setpoints and time delay affect when an emergency bus that is 
experiencing low or degraded voltage will trip from offsite power 
and shift to an emergency diesel generator. While the setpoints that 
initiate this action will be modified, the function remains the 
same. The setpoints have been analyzed to ensure spurious trips will 
be avoided. The proposed changes will not significantly affect any 
accident initiators or precursors. The format changes are intended 
to improve readability, consistency with NUREG-1431, Revision 2, and 
appearance. In addition, they do not alter any requirements. The 
bases change provides explanatory information only. Thus, the 
probability of occurrence of an accident previously evaluated is not 
significantly increased.

Consequences of an Accident Previously Evaluated

    The proposed changes to the degraded voltage and loss of voltage 
setpoints and time delay affect when an emergency bus that is 
experiencing low or degraded voltage will trip from offsite power 
and shift to an emergency diesel generator. While the setpoints that 
initiate this action will be modified, they are bounded by the 
current safety analysis. The function of the plant equipment remains 
the same. The proposed changes improve the reliability of safety-
related equipment to operate as designed. The format changes are 
intended to improve readability, consistency with NUREG-1431, 
Revision 2, and appearance. In addition, they do not alter any 
requirements. The bases change provides explanatory information 
only. Thus, the consequences of an accident previously analyzed are 
not significantly increased.
    2. Does the change create the possibility of a new or different 
kind of accident from any accident previously evaluated?
    The proposed changes to the degraded voltage and loss of voltage 
setpoints and time delay do not affect existing or introduce any new 
accident precursors or modes of operation. The relays will continue 
to detect undervoltage conditions and transfer safety loads to the 
emergency diesel generators at a voltage level adequate to ensure 
proper safety equipment performance and to prevent equipment damage. 
The function of the relays remains the same. The format changes are 
intended to improve readability, consistency with NUREG-1431, 
Revision 2, and appearance. In addition, they do not alter any 
requirements. The bases change provides explanatory information 
only. Thus, the proposed changes do not create the possibility of a 
new or different kind of accident from any accident previously 
evaluated.
    3. Does the change involve a significant reduction in a margin 
of safety?
    The proposed changes will allow all safety-related loads to have 
sufficient voltage to perform their intended safety function while 
ensuring spurious trips are avoided. Thus, the results of the 
accident analyses will not be affected as the input assumptions are 
protected. The format changes are intended to improve readability, 
consistency with NUREG-1431, Revision 2, and appearance. In 
addition, they do not alter any requirements. The bases change 
provides explanatory information only. Thus, the proposed changes do 
not involve a significant reduction in a margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment requests involve no significant hazards consideration.
    Attorney for licensee: David W. Jenkins, Esq., 500 Circle Drive, 
Buchanan, MI 49107.
    NRC Section Chief: William D. Reckley, Acting Section Chief.

North Atlantic Energy Service Corporation, Docket No. 50-443, Seabrook 
Station, Unit No. 1, Rockingham County, New Hampshire

    Date of amendment request: August 2, 2001, as supplemented November 
2, 2001.
    Description of amendment request: The amendment would change the 
Seabrook Station Technical Specification (TS) 6.15 to permit a one-time 
exception to the 10-year frequency for the Integrated Leakage Rate Test 
(ILRT). This exception would permit the existing ILRT frequency to be 
extended from 10 years to 15 years from the last test completed on 
October 30, 1992.
    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 the Seabrook Station Technical 
Specifications does not involve a significant increase in the 
probability or consequences of an accident previously analyzed. The 
proposed revision to TS 6.15 adds a one-time extension to the 
current interval for the ILRT test. It is proposed that the current 
test interval be extended from ten-years to fifteen-years from the 
date of the last ILRT performed on October 30, 1992. The proposed 
extension cannot increase the probability of an accident previously 
evaluated since the test interval extension does not involve 
modification of the plant, nor a operation of the plant that could 
initiate an accident. The proposed extension of the ILRT does not 
involve a significant increase in the consequences of an accident. 
The increase in risk is very small because ILRTs identify only a few 
potential leakage paths that cannot be identified by local leakage 
rate [Type B and C] testing, and the leaks that have been found by 
ILRTs have been only marginally above existing requirements. An 
analysis of the 144 ILRT results including 23 failures, found that 
no ILRT failures were due to a containment liner breach. NUREG-1493 
[``Performance-Based Containment Leak Test Program''] concluded that 
reducing the ILRT testing frequency to one per twenty years would 
lead to an imperceptible increase in risk.
    Therefore, it is concluded that the proposed change to TS 6.15 
does not involve

[[Page 64299]]

a significant increase in the probability or consequence of an 
accident previously evaluated.
    2. The proposed changes do not create the possibility of a new 
or different kind of accident from any previously evaluated.
    The proposed change to Technical Specification 6.15 does not 
create the possibility of a new or different kind of accident from 
any previously evaluated. The proposed change adds a one-time 
extension to the current Integrated Leakage Rate Test frequency of 
ten-years to fifteen-years from the date of the last test. The 
proposed change cannot create the possibility of a new or different 
type of accident since there are no physical changes being made to 
the plant. Additionally, there are no changes to the operation of 
the plant that could introduce a new failure mode creating an 
accident.
    3. The proposed changes do not involve a significant reduction 
in the margin of safety.
    The proposed change does not involve a significant reduction in 
the margin of safety. The proposed revision to TS 6.15 adds a one-
time extension to the current interval for the ILRT test. It is 
proposed that the current test interval be extended from ten-years 
to fifteen-years from the date of the last ILRT performed on October 
30, 1992. A reduction in the ILRT frequency was found to lead to an 
imperceptible decrease in the margin of safety. The estimated 
increase in risk is very small because ILRTs identify only a few 
potential leakage paths that cannot be identified by local leakage 
rate [Type B and C] testing, and the leaks that have been found by 
ILRTs have been only marginally above existing requirements. A 
Seabrook Station specific risk evaluation is consistent with the 
generic conclusions identified in NUREG-1493.
    Based on the above evaluation, North Atlantic concludes that the 
proposed change to TS 6.15 does not constitute a significant hazard.

    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, Esq., Senior Nuclear 
Counsel, Northeast Utilities Service Company, P.O. Box 270, Hartford, 
CT 06141-0270.
    NRC Section Chief: James W. Clifford.

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

    Date of amendment request: October 22, 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 Nuclear Regulatory Commission (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 22, 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

[[Page 64300]]

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: Bradley D. Jackson, Esq., Foley and Lardner, 
P.O. Box 1497, Madison, WI 53701-1497.
    NRC Section Chief: William D. Reckley, Acting.

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

    Date of amendment request: October 16, 2001.
    Description of amendment request: The proposed amendment would 
revise the Susquehanna Steam Electric Station (SSES), Units 1 and 2, 
Technical Specifications (TSs). The licensee proposed to revise 
selected sections of the administrative controls chapter of the TSs 
consistent with Nuclear Regulatory Commission (NRC) approved Technical 
Specification Task Force (TSTF) generic changes to NUREG-1433, 
``Standard Technical Specifications for General Electric Plants (BWR/
4),'' Revision 1 (STS). The licensee also proposed editorial and 
administrative changes to the affected sections.
    The licensee categorized the proposed changes as either 
``Administrative Changes'' or ``Less Restrictive Changes--Removed 
Detail.'' The licensee categorized proposed changes consistent with the 
approved versions of TSTF-273, TSTF-299, TSTF-308, TSTF-348, and TSTF-
364 as ``Administrative Changes.'' An administrative change involves 
editorial restructuring of the current requirements, or modification of 
wording that does not affect the technical content of the current TSs. 
Administrative changes are not intended to add, delete, or relocate any 
technical requirements of the current TSs. The licensee categorized 
proposed changes consistent with the approved versions of TSTF-279 and 
TSTF-363 as ``Less Restrictive Changes--Removed Detail.'' The proposed 
changes involve moving details out of the TSs and into the TS Bases, 
the updated Final Safety Analysis Report (UFSAR), the Technical 
Requirements Manual (TRM), or other documents for which changes are 
subject to regulatory control. The removal of this information is 
considered to be less restrictive because it is no longer controlled by 
the TS change process.
    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:

Administrative Changes

    1. Does the proposed change involve a significant increase in 
the probability of occurrence or consequences of an accident 
previously evaluated?
    The proposed change involves reformatting, renumbering, and 
rewording the existing [technical specification] TS. The 
reformatting, renumbering, and rewording process involves no 
technical changes to the existing TS. As such, this change is 
administrative in nature and does not affect the initiators of 
analyzed events or assumed mitigation of accidents or transient 
events. Therefore, this change does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    2. Does the proposed change create the possibility of a new or 
different kind of accident from any accident previously evaluated?
    The proposed change does not involve a physical alteration of 
the plant (no new or different type of equipment will be installed) 
or changes in methods governing normal plant operation. The proposed 
change will not impose any new or eliminate any old requirements. 
Thus, this change does not create the possibility of a new or 
different kind of accident from any accident previously evaluated.
    3. Does the proposed change involve a significant reduction in a 
margin of safety?
    The proposed change will not reduce a margin of safety because 
it has no effect on any safety analyses assumptions. Therefore, the 
change does not involve a significant reduction in a margin of 
safety.

Less Restrictive Changes--Removed Detail

    1. Does the proposed change involve a significant increase in 
the probability of occurrence or consequences of an accident 
previously evaluated?
    The proposed change relocates certain details from the TS to 
other documents under regulatory control. The TS Bases, [updated 
final safety analysis report] UFSAR, and [Technical Requirements 
Manual] TRM will be maintained in accordance with 10 CFR 50.59. In 
addition to 10 CFR 50.59 provisions, the TS Bases are subject to the 
change control provisions in the Administrative Controls Chapter of 
the TS. The UFSAR is subject to the change control provisions of 10 
CFR 50.71(e). Other documents are subject to controls imposed by TS 
or regulations. Since any changes to these documents will be 
evaluated, no significant increase in the probability or 
consequences of an accident previously evaluated will be allowed. 
Therefore, this change does not involve a significant increase in 
the probability or consequences of an accident previously evaluated.
    2. Does the proposed change create the possibility of a new or 
different kind of accident from any accident previously evaluated?
    The proposed change does not involve a physical alteration of 
the plant (no new or different type of equipment will be installed) 
or changes in methods governing normal plant operation. The proposed 
change will not impose any new or eliminate any old requirements, 
and adequate control of the information will be maintained. Thus, 
this change does not create the possibility of a new or different 
kind of accident from any accident previously evaluated.
    3. Does the proposed change involve a significant reduction in a 
margin of safety?
    The proposed change will not reduce a margin of safety because 
it has no effect on any safety analyses assumptions. In addition, 
the details to be moved from the TS to other documents are the same 
as the existing TS. Since any future changes to these details will 
be evaluated, no significant reduction in a margin of safety will be 
allowed. A significant reduction in a margin of safety is not 
associated with the elimination of the 10 CFR 50.92 requirement for 
NRC review and approval of future changes to the relocated details. 
The proposed change is consistent with NUREG 1433, issued by the NRC 
staff, revising the TS to reflect the approved level of detail, 
which indicates that there is no 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: Bryan A. Snapp, Esquire, Assoc. General 
Counsel, PPL Services Corporation, 2 North Ninth St., GENTW3, 
Allentown, PA 18101-1179.
    NRC Section Chief: L. Raghavan, Acting.

Rochester Gas and Electric Corporation, Docket No. 50-244, R. E. Ginna 
Nuclear Power Plant, Wayne County, New York

    Date of amendment request: October 25, 2001.
    Description of amendment request: The proposed amendment deletes 
requirements from the Technical Specifications (TSs) (and, as 
applicable,

[[Page 64301]]

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 an 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 25, 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 [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: Nicholas S. Reynolds, Winston & Strawn, 1400 
L Street, NW., Washington, DC 20005.
    NRC Section Chief: L. Raghavan, Acting.

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: August 20, 2001.
    Description of amendment request: The licensee is proposing to 
revise Virgil C. Summer Nuclear Station (VCSNS) Technical 
Specifications (TS) to add a footnote to Table 3.3-3 regarding the 
Steam Line Isolation and Engineered Safety Feature Actuation System 
(ESFAS) functions. This revision will allow VCSNS to exclude ESFAS 
steam line isolation instrumentation operability in Mode 3 when the 
main steam isolation valves, along with associated bypass valves, are 
closed and disabled, and ease the restriction of Specification 3.0.4 
when performing reactor coolant system (RCS) resistance temperature 
device (RTD) cross calibrations at temperatures below the ESFAS P-12 
Interlock for Low-Low Tavg. This request is consistent in 
part with the improved Standard Technical Specifications (ITS).
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Does the proposed change involve a significant increase in 
the probability or consequences of an accident previously evaluated?

[[Page 64302]]

    [The] proposed changes involve upgrading the VCSNS TS to more 
closely agree with ITS and does not result in any hardware changes. 
The proposed change revises the applicability for the initiating 
functions of the main steam isolation function such that when a main 
steam line isolation valve is closed and the isolation function is 
accomplished, the automatic initiation of this function is no longer 
required to be operable. The ESFAS is not assumed to be an initiator 
of any analyzed event. The role of the ESFAS is in mitigating and 
thereby limiting the consequences of accidents. The proposed change 
continues to adequately ensure the operability of the ESFAS main 
steam line isolation function when the lines are unisolated and 
thereby ensures the protection provided by the function remains 
operable when required. The relaxation of the P-12 Function during 
RCS RTD cross calibration allows all associated narrow range 
temperature channels to remain in test, with test circuitry 
installed, during the transition between Modes 4 and 3. Surveillance 
performance is administratively controlled by plant procedures which 
assure testing is conducted below the ESFAS P-12 interlock setpoint 
of 552  deg.F and that TS limits for mode operability are not 
exceeded. Therefore, the results of the analyses described in the 
FSAR [Final Safety Analysis Report] remain bounding. Additionally, 
the proposed change does not impose any new safety analyses limits 
or alter the plant's ability to detect or mitigate events. 
Therefore, this change does not involve a significant increase in 
the probability or consequences of an accident previously evaluated.
    2. Does the proposed change create the possibility of a new or 
different kind of accident from any accident previously evaluated?
    The proposed changes involve upgrading the ESFAS area of the 
VCSNS TS to more closely agree with ITS and to support RCS RTD cross 
calibration. The changes do not necessitate a physical alteration of 
the plant (no new or different type of equipment will be installed) 
or changes in parameters governing normal plant operation. Thus, 
this change does not create the possibility of a new or different 
kind of accident from any accident previously evaluated.
    3. Does this change involve a significant reduction in margin of 
safety?
    The proposed change, which upgrades the ESFAS area of the VCSNS 
TS to be more consistent with ITS and supports RCS RTD cross 
calibration, does not have an adverse impact on any design basis 
safety analysis. In combination with administrative controls, 
required safety functions will continue to be accomplished in 
accordance with safety analysis assumptions. As such, the results of 
the analyses described in the FSAR remain bounding [, thus] assuring 
the proposed change does not involve a significant reduction in 
margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: Randolph R. Mahan, South Carolina Electric & 
Gas Company, Post Office Box 764, Columbia, South Carolina 29218.
    NRC Section Chief: Richard Laufer, Acting.

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

    Date of amendment request: October 31, 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 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.

[[Page 64303]]

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: General Counsel, Tennessee Valley Authority, 
400 West Summit Hill Drive, ET 11A Knoxville, Tennessee 37902.
    NRC Section Chief: Richard P. Correia.

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

    Date of amendment request: November 15, 2001 (TS-01-08).
    Description of amendment request: The proposed amendment would 
increase the full core thermal power rating by 1.3 percent from 3411 
MWt to 3455 MWt, based on planned installation of the improved Caldon, 
Incorporated (Caldon) Leading Edge Flow Meter, LEFMTM (LEFM) 
feedwater flow measurement instrumentation. This change affects 
Operating License Condition 2.C.(1) and Definition 1.26 for Rated 
Thermal Power. In addition, changes are necessary to the reactor power 
limits of Technical Specification (TS) Table 3.7.1 with an inoperable 
main steam safety valve for both units and, for Unit 2 only, the 
interval for which the pressure and temperature curves and the low 
temperature over pressure protection curves (TS Figures 3.4-2, 3.4-3, 
and 3.4-4) are valid. A change to the Bases for TS Section 3/4.7.1.1 is 
also included to address the changes in main steam safety valve 
capabilities.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    A. The proposed amendment does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    The comprehensive analytical efforts performed to support the 
proposed change included a review of the nuclear steam supply 
systems (NSSSs) 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 rod drive mechanism, loop piping and supports, reactor 
coolant pump, steam generator 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 rod 
control cluster assembly (RCCA) drop time remains within the current 
limits assumed in the accident analyses. Thus, there is no increase 
in the consequences of the accidents which credit RCCA drop. Several 
steam generator tubes may need to be plugged to preclude the 
potential for U-bend fatigue if the plant operates below certain 
steam pressure values. As long as these provisions are maintained, 
there is no increase in the probability of an steam generator tube 
rupture event. The leak before break analysis conclusions remain 
valid and thus the limiting break sizes determined in this analysis 
remain bounding.
    All of the NSSS systems will continue to perform their intended 
design functions during normal and accident conditions. The 
pressurizer spray flow remains above its design value. Thus, the 
control system design analyses that credit the spray flow do not 
need to be modified for changes in this flow. The auxiliary systems 
and components continue to comply with 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/or balance of plant (BOP) 
interface systems will continue to perform their intended design 
functions. The steam generator safety valves will provide adequate 
relief capacity to maintain the steam generators within design 
limits. The steam dump system will still relieve 40 percent of the 
maximum full load steam flow. The current loss-of-coolant accident 
(LOCA) hydraulic forcing functions are still bounding. Thus, there 
is no significant increase in the probability of an accident 
previously evaluated.
    The fuel has been completely analyzed to determine the effect of 
the 1.3 percent power uprate. The fuel assembly and fuel rod 
integrity have been evaluated. The change results in negligible 
changes to the hydraulic lift forces and the existing holddown 
margins remain acceptable. The increase in corrosion of the fuel 
assembly structural Zircaloy-4 components due to a slight increase 
in temperature is small, thus acceptable structural margin for 
normal operating, faulted, and handling conditions exist. The fuel 
assembly and fuel rod flow-induced vibration (FIV) performance 
remains acceptable. The existing fuel assembly faulted condition 
loading and analysis remain applicable and acceptable. The fuel rod 
strain, creep collapse, and corrosion performance were evaluated at 
the higher power level with acceptable results.
    The fuel cycle design was evaluated and there was no significant 
effect caused by the 1.3 percent power uprate. The operational 
analysis of the core was evaluated for the change and found to 
remain applicable with acceptable results.
    The thermal-hydraulic analysis was evaluated and found to remain 
applicable. The safety analysis addressed all Condition II, III, and 
IV events with the conclusion that current analyses remain 
applicable or bounding. The radiological consequences were evaluated 
and determined to be bounded by current analyses.
    Additionally, the current licensing basis steamline break and 
LOCA mass and energy releases that are used to determine the peak 
containment pressure and temperature limits continue to remain 
bounding with the increase in power. Thus, there is no significant 
increase in the consequences of an accident previously evaluated.
    The heatup and cooldown curves for Unit 2 are now applicable for 
14.5 EFPY [effective full-power year] instead of 16 EFPY. The heatup 
and cooldown curves define limits that still ensure the prevention 
of nonductile failure for the SQN Units 1 and 2 reactor coolant 
system (RCS). The design-basis events that were protected have not 
changed. This modification does not alter any assumptions previously 
made in the radiological consequence evaluations nor affect 
mitigation of the radiological consequences of an accident described 
in the Updated Final Safety Analysis Report. Therefore, the proposed 
changes will not significantly increase the probability or 
consequences of an accident previously evaluated.
    The revised requirements for inoperable MSSVs [main steam safety 
valves] provide limits for the power range high flux trip

[[Page 64304]]

setpoint that ensure adequate relief capability for postulated 
accidents. This change does not alter any plant systems, components, 
or operating methods. Since the plant will continue to operate in 
the same manner with the same protective features, this change will 
not increase the possibility of an accident. The revised setpoint is 
a conservative change that provides additional margin considering 
the effect of the proposed power uprate. Since the revised setpoint 
continues to provide an equivalent level of safety function, this 
change will not significantly increase the consequences of an 
accident and the offsite dose impact will not be significantly 
increased.
    B. The proposed amendment does not 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. All systems, 
structures, and components previously required for the mitigation of 
an event remain capable of fulfilling their intended design 
function. 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, it 
is concluded that the proposed changes do not create the possibility 
of a new or different kind of accident from any accident previously 
evaluated.
    C. The proposed amendment does not involve a significant 
reduction in a margin of safety.
    Operation at the 3455 MWt core power does not involve a 
significant reduction in a margin of safety. Extensive 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 regulatory acceptance criteria. 
The reduction in the EFPY for the Unit 2 heatup and cooldown curves 
does not reduce the margin of safety since the curves define the 
limits for ensuring the prevention of nonductile failure for the RCS 
and these curves remain unchanged.
    The pressure and temperature safety limits will be the same as 
those for the current operating cycle, thus ensuring that the fuel 
will be maintained within the same range of safety parameters that 
form the basis for the Final Safety Analysis Report (FSAR) accident 
evaluations.
    The power uprate represents a small increase in the energy 
production for the fuel cycle and is well within typical variations 
that occur as a result of increases in cycle length and capacity 
factor. The burnup of the fuel will increase proportionally with the 
increase in power, but will not challenge the current licensed 
burnup limit for Mark-BW fuel.
    The slight increase in core average linear heat rate will result 
in a slight loss of operating margin, but will not affect safety 
margins. The centerline fuel melt and transient cladding strain 
limits will not be affected by the power level uprate, but the 
margin to these limits will decrease slightly. The LOCA FQ [power 
peaking factor] limits will not be altered since the increase in 
core power is absorbed by reducing the power uncertainty used in 
determination of the limits.
    The power peaking limits that provide DNB [departure from 
nucleate boiling] protection are slightly lower resulting in a 
proportional loss in DNB margins. The mechanical evaluation of the 
fuel demonstrates that the power level uprate can be successfully 
accomplished in compliance with all design criteria.
    All FSAR Chapter 15 events have been evaluated and found to 
remain applicable for the power uprate. The radiological 
consequences analyses include an initial power assumption of 105 
percent of 3411 MWt and remain bounding for the 1.3 percent power 
uprate.
    The more restrictive limits for the power range high flux trip 
setpoint is based on calculations that ensure sufficient relief 
capacity to meet accident mitigation requirements. This change will 
appropriately limit reactor power levels, with inoperable MSSVs, 
such that the margin of safety is maintained at an equivalent level 
considering the proposed power uprate.
    As appropriate, all evaluations have been performed using 
methods that have either been reviewed and approved by the NRC or 
that are in compliance with all applicable regulatory review 
guidance and standards. All of the fuel and safety evaluations for 
the 1.3 percent power uprate were performed with the Framatome-ANP 
approved methodology listed in TS Section 6.9.1.14 of the SQN TSs. 
Therefore, it is concluded that 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: General Counsel, Tennessee Valley Authority, 
400 West Summit Hill Drive, ET 11A Knoxville, Tennessee 37902.
    NRC Section Chief: Richard P. Correia.

Tennessee Valley Authority, Docket No. 50-390, Watts Bar Nuclear Plant, 
Unit 1, Rhea County, Tennessee

    Date of amendment request October 31, 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 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

[[Page 64305]]

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: General Counsel, Tennessee Valley Authority, 
400 West Summit Hill Drive, ET 10H, Knoxville, Tennessee 37902.
    NRC Section Chief: Richard P. Correia.

Tennessee Valley Authority, Docket No. 50-390, Watts Bar Nuclear Plant, 
Unit 1, Rhea County, Tennessee

    Date of amendment request: November 13, 2001.
    Description of amendment request: The proposed amendment would 
revise the Watts Bar Nuclear Unit 1 (WBN) Technical Requirements Manual 
to add two new sections, TR 3.7.6, ``Shutdown Board Room (SDBR) Air 
Conditioning System (ACS),'' and TR 3.7.7, ``Elevation 772.0 480 Volt 
Board Room Air Conditioning (AC) Systems.'' Each section provides 
specific actions and associated completion times for various out-of-
service conditions associated with the safety-related air conditioning 
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:

    A. The proposed amendment does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    The proposed revision to the WBN Technical Requirements Manual 
(TRM) will provide formalized operational guidance for coping with 
partial or complete unavailability of shut down board room (SDBR) 
and 480V board room air conditioning (AC) equipment for limited 
periods of time. The change does not impact the frequency of an 
accident because failure of either the SDBR or the 480V board room 
AC systems is not an initiator of any accident scenario. The change 
does not modify any plant hardware including the air conditioning 
systems, and none of their automatic control features or redundant 
systems currently credited in failure analyses are being deleted, 
modified, or otherwise replaced by operator actions as a result of 
the proposed change.
    The proposed TRM revision changes current plant operating 
practice and WBN Final Safety Analysis Assumptions (FSAR) 
assumptions by allowing continued power operation with both trains 
of SDBR air conditioning concurrently inoperable and two 480V board 
room AC systems of the same unit to be concurrently inoperable for a 
limited duration, up to 12 hours. This condition is acceptable based 
on the low probability of the occurrence of postulated accidents 
resulting in core damage concurrent with multiple inoperable systems 
or trains of cooling equipment during this timeframe, and based on 
analyses which demonstrate that peak temperatures in each room 
served by these systems remain below mild environment temperature 
limits during this time period. Consequently, there is no 
significant adverse impact on the ability of required safety-related 
electrical equipment to continue to operate and perform their 
required functions, during both normal operation and during design 
basis events. Therefore, the proposed change does not involve a 
significant increase in the probability or consequences of an 
accident previously evaluated.
    B. The proposed amendment does not create the possibility of a 
new or different kind of accident from any accident previously 
evaluated.
    The proposed change does not modify any plant hardware including 
the subject air conditioning systems. The change provides specific 
operational guidance for coping with partial or complete 
unavailability of shut down board room and 480V board room air 
conditioning equipment. No new accident or event initiators are 
created by allowing multiple air conditioning systems to be 
unavailable for the limited time period of 12 hours. The supported 
electrical equipment remains capable of performing its intended 
function both during normal operations and post accident. Therefore, 
the proposed changes do not create the possibility of a new or 
different kind of accident from any accident previously evaluated.
    C. The proposed amendment does not involve a significant 
reduction in a margin of safety.
    The proposed TRM revision changes current FSAR assumptions by 
allowing continued power operation with both trains of SDBR air 
conditioning concurrently inoperable and allowing two 480V board 
room air conditioning systems of the same unit to be inoperable for 
a limited duration, up to 12 hours. This condition does not 
significantly reduce the margin of safety due to the low probability 
of the occurrence of a postulated accident resulting in core damage 
concurrent with multiple inoperable systems or trains of cooling 
equipment during the limited time period. In addition, transient 
temperature analyses demonstrate that peak temperatures in each room 
served by these systems remain below mild environment temperature 
limits for a period of 24 hours assuming a complete loss of air 
conditioning to all rooms served by the SDBR and 480V board room AC 
systems concurrently. The analysis is bounding for normal 
operational

[[Page 64306]]

conditions. Consequently, there is no significant adverse impact on 
the ability of required safety-related electrical equipment to 
continue to operate and perform their required functions during both 
normal operation and during design basis events. 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: General Counsel, Tennessee Valley Authority, 
400 West Summit Hill Drive, ET 10H, Knoxville, Tennessee 37902.
    NRC Section Chief: Richard P. Correia.

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

    Date of amendment request: October 2, 2001.
    Brief description of amendments: 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 Nuclear Station] 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 (TMI-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 Nuclear Regulatory Commission (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 2, 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, the 
emergency operating procedures, and site survey monitoring that 
support modification of emergency plan protective action 
recommendations.
    Therefore, the elimination of PASS requirements from the TSs 
(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: George L. Edgar, Esq., Morgan, Lewis and 
Bockius, 1800 M Street, NW., Washington, DC 20036.
    NRC Section Chief: Robert A. Gramm.

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

    Date of amendment request: October 25, 2001.

[[Page 64307]]

    Brief description of amendments: The proposed amendment would 
revise Technical Specification (TS) 4.2.1, ``Fuel Assemblies,'' for 
Comanche Peak Steam Electric Station (CPSES) Units 1 and 2 to allow the 
use of ZIRLOTM test assemblies and to further allow, `` * * 
* A limited number of lead test assemblies * * * be placed in non-
limiting core 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. Do the proposed changes involve a significant increase in the 
probability or consequences of an accident previously evaluated?
    Response: No.
    Changing the technical specifications within limits of the 
bounding accident analyses cannot change the probability of an 
accident previously evaluated, nor will it increase radiological 
consequences predicted by the analyses of record. Controlling the 
use of lead test assemblies according to limitations approved by the 
NRC [Nuclear Regulatory Commission] constrains fuel performance 
within limits bounded by existing design basis accident and 
transient analyses.
    Therefore, the proposed changes do not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    2. Do the proposed changes create the possibility of a new or 
different kind of accident from any accident previously evaluated?
    Response: No.
    Inclusion in the reactor core of lead test assemblies according 
to limitations set by the NRC for lead test assemblies and of a 
design approved by the NRC ensures that their effect on core 
performance remains within existing design limits. Use of fuel 
assemblies whose design has been previously approved by the NRC as 
lead test assemblies is consistent with current plant design bases, 
does not adversely affect any fission product barrier, and does not 
alter the safety function of safety significant systems, structures 
and components or their roles in accident prevention or mitigation. 
Currently licensed design basis accident and transient analyses of 
record remain valid.
    Therefore, the proposed change does not create the possibility 
of a new or different kind of accident from any previously 
evaluated.
    3. Do the proposed changes involve a significant reduction in a 
margin of safety?
    Response: No.
    The proposed change does not alter the manner in which Safety 
Limits, Limiting Safety System Setpoints, or Limiting Conditions for 
Operation are determined. This proposed clarification of TS 4.2.1 is 
bounded by existing limits on reactor operation. It leaves current 
limitations for use of lead test assemblies in place, conforms to 
plant design bases, is consistent with current safety analyses, and 
limits actual plant operation within analyzed and licensed 
boundaries.
    Therefore, the proposed change does not involve a reduction in a 
margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: George L. Edgar, Esq., Morgan, Lewis and 
Bockius, 1800 M Street, NW., Washington, DC 20036.
    NRC Section Chief: Robert A. Gramm.

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

    Date of amendment request: November 8, 2001.
    Brief description of amendments: The amendments would add the 
following to the Technical Specifications (TSs) for Comanche Peak Steam 
Electric Station (CPSES): (1) the phrase, ``* * * or if open, capable 
of being closed * * *'' to the TS Limiting Condition for Operation 
3.9.4 for the equipment hatch, during core alterations or movement of 
irradiated fuel assemblies inside containment; and (2) the requirement 
to verify the capability to install the equipment hatch in a new 
Surveillance Requirement (SR) 3.9.4.2. Nothing is proposed to be 
deleted from the TSs. Existing SR 3.9.4.2 would be renumbered SR 
3.9.4.3, but would not otherwise be changed. Item (1) will allow the 
equipment hatch to be open during the conditions stated above.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Do the proposed changes involve a significant increase in the 
probability or consequences of an accident previously evaluated?
    Response: No.
    The proposed changes will allow the equipment hatch to be open 
during CORE ALTERATIONS and movement of irradiated fuel assemblies 
inside containment. The status of the equipment hatch during 
refueling operations has no affect on the probability of the 
occurrence of any accident previously evaluated. The proposed 
revision does not alter any plant equipment or operating practices 
in such a manner that the probability of an accident is increased. 
Since the consequences of a fuel handling accident inside 
containment with an open equipment hatch are bounded by the current 
analysis described in the FSAR [Final Safety Analysis Report] and 
the probability of an accident is not affected by the status of the 
equipment hatch, the proposed change does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    2. Do the proposed changes create the possibility of a new or 
different kind of accident from any accident previously evaluated?
    Response: No.
    The proposed changes do not create any new failure modes for any 
system or component, nor do they adversely affect plant operation. 
No new equipment will be added and no new limiting single failures 
will be created. The plant will continue to be operated within the 
envelope of the existing safety analysis.
    Therefore, the proposed changes do not create a new or different 
kind of accident from any accident previously evaluated.
    3. Does the proposed change involve a significant reduction in a 
margin of safety?
    Response: No.
    The previously determined radiological dose consequences for a 
fuel handling accident inside containment with the personnel air 
lock doors open remain bounding for the proposed changes. These 
previously determined dose consequences were determined to be well 
within the limits of 10 CFR [Part] 100 and they meet the acceptance 
criteria of SRP [Standard Review Plan] section 15.7.4 and GDC 
[General Design Criterion] 19.
    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: George L. Edgar, Esq., Morgan, Lewis and 
Bockius, 1800 M Street, NW., Washington, DC 20036.
    NRC Section Chief: Robert A. Gramm.

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

    Date of application request: November 7, 2001.
    Description of amendment request: A change is proposed to Technical 
specification 3.0.3 to allow a longer period of time to perform a 
missed surveillance. The time is extended from the current limit of ``  
* * * up to 24 hours or up to the limit of the specified Frequency, 
whichever is less'' to ``  * * * up to 24 hours or up to the limit of 
the specified Frequency, whichever is greater.'' In addition, the 
following requirement would be added to the specification: ``A risk 
evaluation shall be performed for any Surveillance

[[Page 64308]]

delayed greater than 24 hours and the risk impact shall be managed.''
    The NRC staff issued a notice of opportunity for comment in the 
Federal Register on June 14, 2001 (66 FR 32400), on possible amendments 
concerning missed surveillances, 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 September 28, 2001 (66 FR 49714). The licensee affirmed the 
applicability of the following NSHC determination in its application 
dated November 7, 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 proposed change relaxes the time allowed to 
perform a missed surveillance. The time between surveillances is not 
an initiator of any accident previously evaluated. Consequently, the 
probability of an accident previously evaluated is not significantly 
increased. The equipment being tested is still required to be 
operable and capable of performing the accident mitigation functions 
assumed in the accident analysis. As a result, the consequences of 
any accident previously evaluated are not significantly affected. 
Any reduction in confidence that a standby system might fail to 
perform its safety function due to a missed surveillance is small 
and would not, in the absence of other unrelated failures, lead to 
an increase in consequences beyond those estimated by existing 
analyses. The addition of a requirement to assess and manage the 
risk introduced by the missed surveillance will further minimize 
possible concerns. Therefore, this change does not involve a 
significant increase in the probability or consequences of an 
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 proposed change does not involve a physical alteration of 
the plant (no new or different type of equipment will be installed) 
or a change in the methods governing normal plant operation. A 
missed surveillance will not, in and of itself, introduce new 
failure modes or effects and any increased chance that a standby 
system might fail to perform its safety function due to a missed 
surveillance would not, in the absence of other unrelated failures, 
lead to an accident beyond those previously evaluated. The addition 
of a requirement to assess and manage the risk introduced by the 
missed surveillance will further minimize possible concerns. Thus, 
this change does not create the possibility of a new or different 
kind of accident from any accident previously evaluated.

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

    The extended time allowed to perform a missed surveillance does 
not result in a significant reduction in the margin of safety. As 
supported by the historical data, the likely outcome of any 
surveillance is verification that the LCO [Limiting Condition for 
Operation] is met. Failure to perform a surveillance within the 
prescribed frequency does not cause equipment to become inoperable. 
The only effect of the additional time allowed to perform a missed 
surveillance on the margin of safety is the extension of the time 
until inoperable equipment is discovered to be inoperable by the 
missed surveillance. However, given the rare occurrence of 
inoperable equipment, and the rare occurrence of a missed 
surveillance, a missed surveillance on inoperable equipment would be 
very unlikely. This must be balanced against the real risk of 
manipulating the plant equipment or condition to perform the missed 
surveillance. In addition, parallel trains and alternate equipment 
are typically available to perform the safety function of the 
equipment not tested. Thus, there is confidence that the equipment 
can perform its assumed safety function.
    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 considerations.
    Attorney for licensee: John O'Neill, Esq., Shaw, Pittman, Potts & 
Trowbridge, 2300 N Street, N.W., Washington, DC 20037.
    NRC Section Chief: Stephen Dembek.

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

    Date of application request: November 7, 2001 (ULNRC-04557).
    Description of amendment request: The proposed amendment would 
revise Surveillance Requirements (SRs) 3.3.1.2 and 3.3.1.3 in the 
Technical Specifications (TSs) on reactor trip system (RTS) 
instrumentation. The proposed change to SR 3.3.1.2 would replace the 
reference to the nuclear instrumentation system (NIS) channel output by 
a reference to the power range channel output, and delete Note 1 to the 
SR. The change to SR 3.3.1.3 is editorial in nature.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. The proposed change does not involve a significant increase 
in the probability or consequences of an accident previously 
evaluated.
    Overall protection system performance will remain within the 
bounds of the previously performed accident analyses since there are 
no hardware changes. The RTS instrumentation will be unaffected. 
Protection systems will continue to function in a manner consistent 
with the plant design basis. All design, material, and construction 
standards that were applicable prior to the request are maintained.
    The probability and consequences of accidents previously 
evaluated in the FSAR [Final Safety Analysis Report] are not 
adversely affected because the change to the NIS power range channel 
daily surveillance assures the conservative response of the channel 
even at part-power levels.
    The proposed changes modify the NIS power range channel daily 
surveillance requirement to assure the NIS power range functions are 
tested in a manner consistent with the safety analysis and licensing 
basis.
    The proposed changes will not affect the probability of any 
event initiators. There will be no degradation in the performance 
of, or an increase in the number of challenges imposed on, safety-
related equipment assumed to function during an accident situation. 
There will be no change to normal plant operating parameters or 
accident mitigation performance.
    The proposed changes will not alter any assumptions or change 
any mitigation actions in the [accident] radiological consequence 
evaluations in the FSAR.
    Therefore, the proposed changes do not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    2. The proposed change does not create the possibility of a new 
or different kind of accident from any accident previously 
evaluated.
    There are no hardware changes nor are there any changes in the 
method by which any safety-related plant system performs its safety 
function. This amendment will not affect the normal method of plant 
operation or change any operating parameters. No performance 
requirements or response time limits will be affected; however, the 
proposed TS Bases changes impose explicit NIS power range high trip 
setpoint adjustment requirements prior to adjusting indicated power 
in a decreasing power direction. These requirements are consistent 
with assumptions made in the safety analysis and licensing basis.
    No new accident scenarios, transient precursors, failure 
mechanisms, or limiting single failures are introduced as a result 
of this amendment. There will be no adverse effect or challenges 
imposed on any safety-related system as a result of this amendment.

[[Page 64309]]

    This amendment does not alter the design or performance of the 
7300 Process Protection System, Nuclear Instrumentation System, or 
Solid State Protection System used in the plant protection systems.
    Therefore, the proposed changes do 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.
    The proposed changes require a revision to the criteria for 
implementation of NIS power range channel adjustments based on 
secondary power calorimetric calculations; however, the changes do 
not eliminate any RTS surveillances or alter the frequency of 
surveillances required by the Technical Specifications. The revision 
to the criteria for implementation of the daily surveillance will 
have a conservative effect on the performance of the NIS power range 
channels, particularly at part-power conditions. The nominal trip 
setpoints specified in the Technical Specification Bases and the 
safety analysis limits assumed in the transient and accident 
analyses are unchanged. None of the acceptance criteria for any 
accident analysis is changed.
    There will be no effect on the manner in which safety limits or 
limiting safety system settings are determined nor will there be any 
effect on those plant systems necessary to assure the accomplishment 
of protection functions. There will be no impact on the overpower 
limit, departure from nucleate boiling ratio (DNBR) limits, heat 
flux hot channel factor (FQ), nuclear enthalpy rise hot 
channel factor (FH), loss of coolant accident peak cladding 
temperature (LOCA PCT), peak local power density, or any other 
margin of safety. The radiological dose consequence acceptance 
criteria listed in the Standard Review Plan will continue to be met.
    The imposition of appropriate surveillance testing requirements 
will not reduce any margin of safety since the changes will assure 
that safety analysis assumptions on equipment operability are 
verified on a periodic frequency.
    Therefore, the proposed changes do not involve a significant 
reduction in a margin of safety.

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

Virginia Electric and Power Company, Docket No. 50-280, Surry Power 
Station, Unit No. 1, Surry County, Virginia

    Date of amendment request: October 15, 2001, as supplemented 
November 8, 2001.
    Description of amendment request: The proposed amendment would 
revise Technical Specifications Section 4.4. The proposed changes would 
permit a one-time 5-year extension of the 10-year performance-based 
Type A test interval established in NEI 94-01, ``Nuclear Energy 
Institute Industry Guideline for Implementing Performance-Based Option 
of 10 CFR Part 50, Appendix J,'' Revision 0, July 26, 1995.
    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?
    The proposed extension to Type A testing cannot increase the 
probability of an accident previously evaluated since extension of 
the containment Type A testing is not a physical plant modification 
that could alter the probability of accident occurrence nor, is an 
activity or modification by itself that could lead to equipment 
failure or accident initiation.
    The proposed extension to Type A testing does not result in a 
significant increase in the consequences of an accident as documented 
in NUREG-1493. The NUREG notes that very few potential containment 
leakage paths are not identified by Type B and C tests. It concludes 
that reducing the Type A (ILRT) testing frequency to once per twenty 
years leads to an imperceptible increase in risk.
    Surry provides a high degree of assurance through indirect testing 
and inspection that the containment will not degrade in a manner 
detectable only by Type A testing. The last two Type A tests identified 
containment leakage within acceptance criteria, indicating a very leak-
tight containment. Inspections required by the ASME Code are also 
performed in order to identify indications of containment degradation 
that could affect leak-tightness. Also, maintaining the containment 
subatmospheric during operations provides constant monitoring of the 
leaktightness of the containment structure. Separately, Type B and C 
testing, required by Technical Specifications, identifies any 
containment opening from design penetrations, such as valves, that 
would otherwise be detected by a Type A test. These factors establish 
that an extension to the Surry Type A test interval will not represent 
a significant increase in the consequences of an accident.
    2. Does the proposed license amendment create the possibility of a 
new or different kind of accident from any accident previously 
evaluated?
    The proposed revision to Technical Specifications adds a one-time 
extension to the current interval for Type A testing for Surry Unit 1. 
The current test interval of ten years, based on past performance, 
would be extended on a one-time basis to fifteen years from the last 
Type A test. The proposed extension to Type A testing does not create 
the possibility of a new or different type of accident since there are 
no physical changes being made to the plant and there are no changes to 
the operation of the plant that could introduce a new failure.
    3. Does the proposed amendment involve a significant reduction in a 
margin of safety?
    The proposed revision to Surry Technical Specifications adds a one-
time extension to the current interval for Type A testing. The current 
test interval of ten years, based on past performance, would be 
extended on a one-time basis to fifteen years from the last Type A test 
for Surry Unit 1. The proposed extension to Type A testing will not 
significantly reduce the margin of safety. The NUREG-1493 generic study 
of the effects of extending containment leakage testing found that a 
20-year interval in Type A leakage testing resulted in an imperceptible 
increase in risk to the public. NUREG-1493 found that, generically, the 
design containment leakage rate contributes about 0.1 percent of the 
overall risk and that decreasing the Type A testing frequency would 
have a minimal [effect] on this risk since 95% of the Type A detectable 
leakage paths would already be detected by Type B and C testing. In 
addition, the risk impact on the total integrated (fifteen year total) 
Surry Unit 1 plant risk above baseline, for those accident sequences 
influenced by Type A testing, is only 0.004%. Furthermore, for Surry, 
maintaining the containment subatmospheric during plant operations 
further reduces the risk of any containment leakage path going 
undetected.
    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: 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.

[[Page 64310]]

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: May 31, 2001 as supplemented October 17, 
2001.
    Description of amendment request: The proposed changes would revise 
the Technical Specifications and associated Bases to provide a separate 
allowed outage time for the backup air supply for the pressurizer 
power-operated relief valves (PORVs).
    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:
    Dominion has reviewed the requirements of 10 CFR 50.92 as they 
relate to the proposed change for Surry Units 1 and 2 and determined 
that a significant hazards consideration is not involved. The 
following is provided to support this conclusion.
    1. Does the change involve a significant increase in the 
probability or consequences of an accident previously evaluated.
    The proposed change does not introduce any new mechanisms for 
the initiation of transients or accidents or for the failure of 
equipment relied upon in the accident analyses to mitigate the 
consequences of accidents. The impact of the proposed change on the 
availability and reliability of the pressurizer PORVs is negligible. 
Therefore the accident analysis results and conclusions remain 
bounding.
    2. Does the change create the possibility of a new or different 
kind of accident from any accident previously evaluated.
    There are no modifications to the plant as a result of the 
changes. No new accident or event initiators are created by changing 
the required actions for various conditions of PORV inoperability. 
The proposed change will not introduce any new equipment failure 
modes that could initiate accidents or change the analysis results 
presented in the UFSAR [Updated Final Safety Analysis Report].
    3. Does the change involve a significant reduction in a margin 
of safety.
    The proposed change will not alter the limiting results of the 
safety analyses presented in Chapter 14 of the UFSAR. Provision of 
an allowed outage time for the pressurizer PORV backup air system 
and of more condition specific and appropriate actions for various 
types of PORV inoperability has an insignificant impact on the 
availability and reliability of the PORVs for performing their 
safety related functions.

    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: 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 e-mail to [email protected].

Calvert Cliffs Nuclear Power Plant, Inc., Docket Nos. 50-317 and 50-
318, Calvert Cliffs Nuclear Power Plant, Unit Nos. 1 and 2, Calvert 
County, Maryland

    Date of application for amendments: July 26, 2001.
    Brief description of amendments: The amendments modify Technical 
Specifications 5.5.14.b and 5.5.14.b.2, Technical Specification Bases 
Control Program, such that they are consistent with Title 10 of the 
Code of Federal Regulations (10 CFR 50.59).
    Date of issuance: November 21, 2001.
    Effective date: As of the date of issuance to be implemented within 
30 days.
    Amendment Nos.: 247 and 222.
    Renewed Facility Operating License Nos. DPR-53 and DPR-69: 
Amendments revised the Technical Specifications.
    Date of initial notice in Federal Register: September 5, 2001 (66 
FR 46475) The Commission's related evaluation of these amendments is 
contained in a Safety Evaluation dated November 21, 2001.
    No significant hazards consideration comments received: No.

Entergy Gulf States, Inc., and Entergy Operations, Inc., Docket No. 50-
458, River Bend Station, Unit 1, West Feliciana Parish, Louisiana

    Date of amendment request: April 19, 2001.
    Brief description of amendment: The amendment changes the River 
Bend Station Technical Specifications (TSs) to allow an increase in the 
number of spent fuel assemblies (SFAs) to be stored in the spent fuel 
pool from the current TS limit of 2680 SFAs to 3104 SFAs.
    Date of issuance: November 19, 2001.
    Effective date: As of the date of issuance and shall be implemented 
30 days from the date of issuance.
    Amendment No.: 123.
    Facility Operating License No. NPF-47: The amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: October 18, 2001 (66 FR 
52948) The Commission's related evaluation of the amendment is 
contained in a Safety Evaluation dated November 19, 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: July 23, 2001, as supplemented by letter 
dated October 25, 2001.
    Brief description of amendment: The change deletes Technical 
Specification

[[Page 64311]]

(TS) 3.9.12, ``Fuel Handling Building Ventilation System,'' and TS 
3.3.3.1 Surveillance Requirements for the Fuel Storage Pool area 
radiation monitors.
    Date of issuance: November 21, 2001.
    Effective date: As of the date of issuance and shall be implemented 
60 days from the date of issuance.
    Amendment No.: 176.
    Facility Operating License No. NPF-38: The amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: August 22, 2001 (66 FR 
44169). The October 25, 2001, supplement contained clarifying 
information that did not change the scope of the July 23, 2001, 
application nor the initial proposed no significant hazards 
consideration determination.
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated November 21, 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: August 22, 2001.
    Brief description of amendments: The amendments revise the 
Technical Specifications for St. Lucie Units 1 and 2 to allow small, 
controlled, safe insertions of positive reactivity while in shutdown 
modes.
    Date of Issuance: November 19, 2001.
    Effective Date: As of the date of issuance and shall be implemented 
within 60 days of issuance.
    Amendment Nos.: 179 and 122.
    Facility Operating License Nos. DPR-67 and NPF-16: Amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: September 19, 2001 (66 
FR 48287).
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated November 19, 2001.
    No significant hazards consideration comments received: No.

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

    Date of amendment request: June 21, 2001.
    Brief description of amendment request: The amendment revises Three 
Mile Island Nuclear Station, Unit 2 Technical Specifications 
Administrative Controls section to provide consistency with the changes 
to the revised subsection 50.59 of Title 10 of the Code of Federal 
Regulations, as published in the Federal Register on October 4, 1999 
(64 FR 53582).
    Date of issuance: November 28, 2001
    Effective date: As of the date of issuance and shall be implemented 
within 30 days.
    Amendment No.: 57.
    Facility Operating License No. DPR-73: Amendment revises the 
Technical Specifications.
    Date of initial notice in Federal Register: October 31, 2001 (66 FR 
55020).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated November 28, 2001.
    No significant hazards consideration comments received: No.

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

    Date of application for amendments: May 15, 2001.
    Brief description of amendments: The amendments change TS 3/
4.8.2.2, ``A. C. Distribution Shutdown,'' TS 3/4.8.2.4 ``D. C. 
Distribution--Shutdown,'' and TS 3/4.9.4, ``Containment Building 
Penetrations.'' The proposed amendments replaces the current required 
actions in TSs 3/4.8.2.2. and 3/4.8.2.4, to establish containment 
integrity within 8 hours if less than the specified minimum complement 
of A.C. or D.C. busses and equipment is operable in Modes 5 and 6 with 
new actions which require to immediately suspend operations involving 
core alterations, positive reactivity changes, and movement of 
irradiated fuel assemblies, to immediately initiate actions to restore 
the required busses and return equipment to operable status, and to 
immediately declare the associated required residual heat removal 
loop(s) inoperable. The proposed new actions are consistent with 
NUREG--1431, ``Standard Technical Specifications, Westinghouse 
Plants,'' Revision 1.
    In addition, the proposed amendments will change TS 3/4.9.4 to add 
options to use containment penetration closure methods that are 
equivalent to those that are currently required by the TSs during core 
alterations or movement of irradiated fuel in containment, and to allow 
unisolation of some penetrations under administrative control.
    Date of issuance: November 21, 2001.
    Effective date: As of the date of issuance and shall be implemented 
within 45 days.
    Amendment Nos.: 259 and 242.
    Facility Operating License Nos. DPR-58 and DPR-74: Amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: June 12, 2001 (66 FR 
31709).
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated November 21, 2001.
    No significant hazards consideration comments received: No.

Southern Nuclear Operating Company, Inc., Georgia Power Company, 
Oglethorpe Power Corporation, Municipal Electric Authority of Georgia, 
City of Dalton, Georgia, Docket Nos. 50-321 and 50-366, Edwin I. Hatch 
Nuclear Plant, Units 1 and 2, Appling County, Georgia

    Date of application for amendments: October 8, 2001.
    Brief description of amendments: The amendments revised the 
Technical Specifications to allow the main control room boundary to be 
opened intermittently under administrative controls and to allow 24 
hours to restore the main control room boundary to Operable status 
before requiring the plant to perform an orderly shutdown.
    Date of issuance: November 26, 2001.
    Effective date: As of the date of issuance and shall be implemented 
within 30 days from the date of issuance.
    Amendment Nos.: 225 and 168.
    Facility Operating License Nos. DPR-57 and NPF-5: Amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: October 26, 2001 (66 FR 
54301).
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated November 26, 2001.
    No significant hazards consideration comments received: No.

Tennessee Valley Authority, Docket No. 50-390, Watts Bar Nuclear Plant, 
Unit 1, Rhea County, Tennessee

    Date of application for amendment: September 7, 2001 (TS 01-09).
    Brief description of amendment: The amendment revised Technical 
Specifications (TS) Section 3.6.11, ``Ice Bed,'' Surveillance 
Requirement (SR) 3.6.11.2, SR 3.6.11.3, and the associated Bases, to 
lower the minimum average ice basket weight from 1236 pounds to 1110 
pounds.
    Date of issuance: November 29, 2001.
    Effective date: As of the date of its issuance and shall be 
implemented no later than Mode 4 during startup from Cycle 4 refueling 
outage.
    Amendment No.: 33.

[[Page 64312]]

    Facility Operating License No. NPF-90: Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: October 17, 2001 (66 FR 
52804).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated November 29, 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 3rd of December, 2001.

    For the Nuclear Regulatory Commission.
John A. Zwolinski,
Director, Division of Licensing Project Management, Office of Nuclear 
Reactor Regulation.
[FR Doc. 01-30455 Filed 12-11-01; 8:45 am]
BILLING CODE 7590-01-P