[Federal Register Volume 67, Number 5 (Tuesday, January 8, 2002)]
[Notices]
[Pages 924-939]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: 02-301]


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NUCLEAR REGULATORY COMMISSION


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

I. Background

    Pursuant to Public Law 97-415, the U.S. Nuclear Regulatory 
Commission (the Commission or NRC staff) is publishing this regular 
biweekly notice. Public Law 97-415 revised section 189 of the Atomic 
Energy Act of 1954, as amended (the Act), to require the Commission to 
publish notice of any amendments issued, or proposed to be issued, 
under a new provision of section 189 of the Act. This provision grants 
the Commission the authority to issue and make immediately effective 
any amendment to an operating license upon a determination by the 
Commission that such amendment involves no significant hazards 
consideration, notwithstanding the pendency before the Commission of a 
request for a hearing from any person.
    This biweekly notice includes all notices of amendments issued, or 
proposed to be issued from December 7, 2001, through December 27, 2001. 
The last biweekly notice was published on December 26, 2001 (66 FR 
64461).

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 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 February 7, 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

[[Page 925]]

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. Publicly 
available records will be accessible electronically from the Agencywide 
Document Access and Management System's (ADAMS) Public Electronic 
Reading Room on the Internet at the NRC Web site, http://www.nrc.gov/reading-rm/adams.html. If a request for a hearing or petition for leave 
to intervene is filed by the above date, the Commission or an Atomic 
Safety and Licensing Board, designated by the Commission or by the 
Chairman of the Atomic Safety and Licensing Board Panel, will rule on 
the request and/or petition; and the Secretary or the designated Atomic 
Safety and Licensing Board will issue a notice of a hearing or an 
appropriate order.
    As required by 10 CFR 2.714, a petition for leave to intervene 
shall set forth with particularity the interest of the petitioner in 
the proceeding, and how that interest may be affected by the results of 
the proceeding. The petition should specifically explain the reasons 
why intervention should be permitted with particular reference to the 
following factors: (1) The nature of the petitioner's right under the 
Act to be made a party to the proceeding; (2) the nature and extent of 
the petitioner's property, financial, or other interest in the 
proceeding; and (3) the possible effect of any order which may be 
entered in the proceeding on the petitioner's interest. The petition 
should also identify the specific aspect(s) of the subject matter of 
the proceeding as to which petitioner wishes to intervene. Any person 
who has filed a petition for leave to intervene or who has been 
admitted as a party may amend the petition without requesting leave of 
the Board up to 15 days prior to the first prehearing conference 
scheduled in the proceeding, but such an amended petition must satisfy 
the specificity requirements described above.
    Not later than 15 days prior to the first prehearing conference 
scheduled in the proceeding, a petitioner shall file a supplement to 
the petition to intervene which must include a list of the contentions 
which are sought to be litigated in the matter. Each contention must 
consist of a specific statement of the issue of law or fact to be 
raised or controverted. In addition, the petitioner shall provide a 
brief explanation of the bases of the contention and a concise 
statement of the alleged facts or expert opinion which support the 
contention and on which the petitioner intends to rely in proving the 
contention at the hearing. The petitioner must also provide references 
to those specific sources and documents of which the petitioner is 
aware and on which the petitioner intends to rely to establish those 
facts or expert opinion. Petitioner must provide sufficient information 
to show that a genuine dispute exists with the applicant on a material 
issue of law or fact. Contentions shall be limited to matters within 
the scope of the amendment under consideration. The contention must be 
one which, if proven, would entitle the petitioner to relief. A 
petitioner who fails to file such a supplement which satisfies these 
requirements with respect to at least one contention will not be 
permitted to participate as a party.
    Those permitted to intervene become parties to the proceeding, 
subject to any limitations in the order granting leave to intervene, 
and have the opportunity to participate fully in the conduct of the 
hearing, including the opportunity to present evidence and cross-
examine witnesses.
    If a hearing is requested, the Commission will make a final 
determination on the issue of no significant hazards consideration. The 
final determination will serve to decide when the hearing is held.
    If the final determination is that the amendment request involves 
no significant hazards consideration, the Commission may issue the 
amendment and make it immediately effective, notwithstanding the 
request for a hearing. Any hearing held would take place after issuance 
of the amendment.
    If the final determination is that the amendment request involves a 
significant hazards consideration, any hearing held would take place 
before the issuance of any amendment.
    A request for a hearing or a petition for leave to intervene must 
be filed with the Secretary of the Commission, U.S. Nuclear Regulatory 
Commission, Washington, DC 20555-0001, Attention: Rulemaking and 
Adjudications Branch, or may be delivered to the Commission's Public 
Document Room, located at One White Flint North, 11555 Rockville Pike 
(first floor) Rockville, Maryland, 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 
Access and Management System's (ADAMS) Public Electronic Reading Room 
on the Internet at the NRC Web site, http://www.nrc.gov/reading-rm./adams.html. Persons who 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 amendments request: November 19, 2001.
    Description of amendments request: The proposed amendments would 
change the Loss of Feedwater Flow analysis in the Updated Final Safety 
Analysis Report (UFSAR). The current analysis contained several non-
conservative assumptions, which would be corrected by the reanalysis. 
These corrections include: incorporating a single-failure of the 
Auxiliary Feedwater System, including steam generator blowdown, 
accounting for the change in density of water once the feedwater flow 
has stopped, and assuming sludge deposition in the steam generators. 
Also assumed in the analysis is the installation of a modification to 
isolate steam generator blowdown on an auxiliary feedwater actuation 
signal and an operator action to adjust the auxiliary feedwater flow 
after the event initiation.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration. The NRC staff has reviewed the licensee's analysis 
against the standards of 10 CFR 50.92(c). The NRC staff's review is 
presented below:


[[Page 926]]


    1. Does the amendment involve a significant increase in the 
probability or consequences of an accident previously evaluated?
    No. The proposed amendments would modify several assumptions in 
the UFSAR Loss of Feedwater Flow analysis to more accurately reflect 
the plant response to the event. The significant changes to the 
accident include the addition of an operator action to increase 
auxiliary feedwater (AFW) flow and the implementation of an 
automatic steam generator blowdown isolation following the event 
initiation. These changes do not affect any accident initiators or 
precursors because they only alter the operation of the plant 
following the accident initiation. Thus, the proposed amendments do 
not increase the probability of occurrence of any previously 
analyzed accidents. In addition, besides the aforementioned changes, 
the proposed amendments would not alter any design parameter, 
condition, equipment configuration, or manner in which Calvert 
Cliffs Units 1 and 2 are operated. Furthermore, the proposed 
modifications do not alter or prevent the ability of existing 
structures, systems, or components to perform their intended safety 
or accident-mitigating functions depicted in the UFSAR. The proposed 
modifications to the analysis account for a single-failure and 
update the assumptions to more recent standards. With these changes, 
the plant continues to meet the current acceptance criteria. 
Therefore, the proposed amendments do not involve a significant 
increase in the consequences of an accident previously evaluated.
    2. Does the change create the possibility of a new or different 
kind of accident from any accident previously evaluated?
    No. The proposed amendments alter the design function of the 
steam generator blowdown valves to isolate upon receipt of an 
auxiliary feedwater actuation system signal. The isolation of the 
steam generator blowdown will not create the possibility of a new or 
a different kind of accident because blowdown isolation already 
occurs on a high radiation signal or a containment spray actuation 
signal. The operator action to increase AFW flow only alters the 
operation of the AFW in the conservative direction. The other 
changes to the accident analysis do not alter any design parameter, 
condition, equipment configuration, or manner in which the units are 
operated. Furthermore, none of the changes alter or prevent the 
ability of structures, systems, or components to perform their 
intended safety or accident mitigating functions. Accordingly, the 
proposed amendment does not create any new or different kind of 
accident from any accident previously evaluated.
    3. Does the change involve a significant reduction in a margin 
of safety?
    No. The proposed amendments do not change any design parameter, 
safety limit, or acceptance criteria. However, the amendments do 
change an analysis methodology. This change, performed with the 
objective of imposing conservative assumptions on the accident 
analysis, keeps the accident within the acceptance criteria for 
anticipated operational occurrences. Therefore, operation in 
accordance with the proposed amendment will not involve a 
significant reduction in a margin of safety.

    Based on the NRC staff's review, it appears that the three 
standards of 10 CFR 50.92(c) are satisfied. Therefore, the NRC staff 
proposes to determine that the proposed amendment involves no 
significant hazards consideration.
    Attorney for licensee: Jay E. Silberg, Esquire, Shaw, Pittman, 
Potts and Trowbridge, 2300 N Street, NW., Washington, DC 20037.
    NRC Section Chief: L. Raghavan, Acting.

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

    Date of amendment request: November 19, 2001.
    Description of amendment request: The proposed amendment would 
provide a one-time extension, from 10 to 14 days, of the allowed outage 
time (AOT) for one train of the Control Room Emergency Ventilation 
System (CREVS) to be inoperable due to the emergency power supply being 
inoperable.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration. The NRC staff has reviewed the licensee's analysis 
against the standards of 10 CFR 50.92(c). The NRC staff's review is 
presented below:

    1. Does the amendment involve a significant increase in the 
probability or consequences of an accident previously evaluated?
    No. The proposed amendment would allow a one-time extension of 
the AOT of one train of the CREVS. Since the CREVS is an accident 
mitigation system, the AOT extension would not affect any accident 
initiators or precursors. Therefore, the AOT extension would not 
increase the probability of an accident previously evaluated. 
Similarly, since the consequences of a design-basis accident 
coincident with a failure of the redundant CREVS train during a 14-
day outage are the same as those during the already approved 10-day 
outage, the proposed change does not increase the consequences of an 
accident previously evaluated.
    2. Does the change create the possibility of a new or different 
kind of accident from any accident previously evaluated?
    No. The proposed amendment does not affect accident initiators 
or precursors because the CREVS is not being modified nor will any 
unusual operator actions be required. The CREVS is not an accident 
initiator, but is an accident mitigator. Therefore, the proposed 
change does not create the possibility of a new or different kind of 
accident from any accident previously evaluated.
    3. Does the change involve a significant reduction in a margin 
of safety?
    No. The proposed amendment would allow a one-time extension of 
the AOT of one train of the CREVS from the currently allowed 10 days 
to 14 days. This action decreases the margin of safety. However, 
based upon the licensee's management of plant risk, the change 
increases the Core Damage Frequency (CDF) and Large Early Release 
Frequency (LERF) to less than the Regulatory Guide 1.174 criteria of 
1E-6 per reactor year for CDF and 1E-07 per reactor year for LERF. 
Therefore, the proposed amendment does not involve a significant 
reduction in a margin of safety.

    Based on the NRC staff's review, it appears that the three 
standards of 10 CFR 50.92(c) are satisfied. Therefore, the NRC staff 
proposes to determine that the proposed request involves no significant 
hazards consideration.
    Attorney for licensee: Jay E. Silberg, Esquire, Shaw, Pittman, 
Potts and Trowbridge, 2300 N Street, NW., Washington, DC 20037.
    NRC Section Chief: L. Raghavan, Acting.

Carolina Power & Light Company, et al., Docket Nos. 50-325 and 50-324, 
Brunswick Steam Electric Plant (BSEP), Units 1 and 2, Brunswick County, 
North Carolina

    Date of amendments request: November 26, 2001.
    Description of amendments request: The proposed amendments will add 
Technical Specification 5.5.12.f, ``Programs and Manuals, Primary 
Containment Leakage Rate Testing Program.'' This addition will provide 
a one-time exception to the frequency of the performance-based leakage 
rate testing program.
    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 license amendments do not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    The proposed change to Technical Specification 5.5.12 provides a 
one-time extension to the testing frequency for containment 
integrated leakage rate (i.e., Type A) testing. The existing 10-year 
test interval is based on past test performance. The proposed 
Technical Specification change will extend the Type A testing 
frequency to 15 years, one month from the last Type A test for Unit 
1 and to 15 years for Unit 2. The proposed Technical Specification 
change does not involve a physical change to the plant or a change 
in the manner in which the plant is operated or controlled. The 
primary containment is designed to provide an

[[Page 927]]

essentially leak tight barrier against the uncontrolled release of 
radioactivity to the environment for postulated accidents. As such, 
the primary containment does not involve the prevention or 
identification of any precursors of an accident. Therefore, the 
proposed Technical Specification change does not involve a 
significant increase in the probability of an accident previously 
evaluated.
    The proposed change involves only a one-time change to the 
interval between Type A containment leakage tests. Type B and C 
containment leakage testing will continue to be performed at the 
frequency currently required by the BSEP Technical Specifications. 
As documented in NUREG-1493, ``Performance-Based Containment 
Leakage-Test Program,'' industry experience has shown that Type B 
and C containment leakage tests have identified a very large 
percentage of containment leakage paths and that the percentage of 
containment leakage paths that are detected only by Type A testing 
is very small. In fact, an analysis of 144 integrated leak rate 
tests results, including 23 failures, found that no failures were 
due to containment liner breach. NUREG-1493 also concluded, in part, 
that reducing the frequency of Type A containment leakage rate 
testing to once per 20 years was found to lead to an imperceptible 
increase in risk. The BSEP, Unit 1 and 2 test history and risk-based 
evaluation of the proposed extension to the Type A test frequency 
supports this conclusion. The design and construction requirements 
of the primary containment, combined with the containment 
inspections performed in accordance with the American Society of 
Mechanical Engineers (ASME) Code, Section XI and the Maintenance 
Rule (i.e., 10 CFR 50.65) provide a high degree of assurance that 
the primary containment will not degrade in a manner that is 
detectable only by Type A testing. Therefore, the proposed Technical 
Specification change does not involve a significant increase in the 
consequences of an accident previously evaluated.
    2. The proposed license amendments will not create the 
possibility of a new or different kind of accident from any accident 
previously evaluated.
    The proposed change to the Technical Specification 5.5.12 
involves a one-time extension to the testing interval for Type A 
containment leakage rate testing. The primary containment and the 
testing requirements invoked to periodically demonstrate the 
integrity of the primary containment exist to ensure the ability to 
mitigate the consequences of an accident. The primary containment 
and its associated testing requirements do not involve the 
prevention or identification of any precursors of an accident. The 
proposed change to the Type A leakage rate testing frequency does 
not involve any physical changes being made to the facility. In 
addition, the proposed changes to the Type A leakage rate testing 
frequency [do] not change the operation of the plants such that a 
new failure mode involving the possibility of a new or different 
kind of accident from any accident previously evaluated is created. 
Therefore, the proposed change does not create the possibility of a 
new or different kind of accident from any accident previously 
evaluated.
    3. The proposed license amendments do not involve a significant 
reduction in a margin of safety.
    The proposed extension to the Type A testing frequency 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 extension for 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 a very 
small amount to the individual risk and that the decrease in Type A 
testing frequency would have a minimal [effect] on this risk since 
most potential leakage paths are detected by Type B and C testing. 
The proposed change involves only an extension of the frequency for 
Type A containment leakage testing; the overall primary containment 
leakage rate limit specified by Technical Specifications is being 
maintained. Type B and C containment leakage testing will continue 
to be performed at the frequency currently required by the BSEP 
Technical Specifications. The regular containment inspections being 
performed in accordance with the ASME, Section XI, and the 
Maintenance Rule (i.e., 10 CFR 50.65) provide a high degree of 
assurance that the containment will not degrade in a manner that is 
only detectable by Type A testing. In addition, the on-line 
containment monitoring capability that is inherit to [a] boiling 
water reactor using an inert containment atmosphere allows for the 
detection of gross containment leakage that may develop during power 
operation. The combination of these factors ensures that the margin 
of safety is maintained. Therefore, the proposed license amendments 
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: 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.

FirstEnergy Nuclear Operating Company, Docket No. 50-440, Perry Nuclear 
Power Plant, Unit 1, Lake County, Ohio

    Date of amendment request: December 5, 2001.
    Description of amendment request: The proposed amendment 
incorporates Technical Specification Task Force (TSTF) Standard 
Technical Specification Traveler, TSTF-364, Revision 0, ``Revision to 
Technical Specification Bases Control Program to Incorporate Changes to 
10 CFR 50.59.'' The proposed change deletes reference to the term 
``unreviewed safety question,'' and replaces it with the phrase 
``requires NRC approval pursuant to 10 CFR 50.59.''
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration which is presented below:

    1. The proposed change does not involve a significant increase 
in the probability or consequences of an accident previously 
evaluated.
    The proposed change is consistent with the changes described in 
TSTF change TSTF-364, ``Revision to Technical Specification Bases 
Control Program to Incorporate Changes to 10 CFR 50.59.'' 
Specifically, the proposed change deletes the reference to the term 
``unreviewed safety question'' as defined in 10 CFR 50.59 (pre-1999 
revision) and replaces it with the phrase ``requires NRC approval 
pursuant to 10 CFR 50.59.'' The deletion of the definition of 
``unreviewed safety question'' was approved by the NRC in the 
current revision of the 10 CFR 50.59 regulation (October 1999). 
Changes to the Technical Specification Bases will still be evaluated 
in accordance with the requirements of 10 CFR 50.59. Therefore, the 
proposed change does not involve a significant increase in the 
probability or consequences of an accident previously evaluated.
    2. The proposed change does not create the possibility of a new 
or different kind of accident from any accident previously 
evaluated.
    The proposed change is to an administrative program. The change 
does not involve any physical modifications to the facility nor add 
new equipment. The methods of plant operation have not been altered. 
Therefore, the proposed change does not create the possibility of a 
new or different kind of accident from any accident previously 
evaluated.
    3. The proposed change does not involve a significant reduction 
in a margin of safety.
    The proposed change is administrative in nature, based upon the 
current version of 10 CFR 50.59 regulation. Changes to the Technical 
Specification Bases will still be evaluated by 10 CFR 50.59. The 
proposed change has no direct impact upon any plant safety analyses. 
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: Mary E. O'Reilly, Attorney,

[[Page 928]]

FirstEnergy Corporation, 76 South Main Street, Akron, OH 44308

    NRC Section Chief: Anthony J. Mendiola.

Nine Mile Point Nuclear Station, LLC, Docket No. 50-220, Nine Mile 
Point Nuclear Station, Unit No. 1, Oswego County, New York

    Date of amendment request: October 19, 2001.
    Description of amendment request: The proposed amendment would 
revise the Technical Specifications (TSs) by the following: (1) 
Implement programmatic controls for radiological effluent technical 
specifications in the Administrative Controls section of the TSs, (2) 
relocate existing procedural details to licensee-controlled documents 
or new programs to accommodate the incorporation of Generic Letter 89-
01 and relevant portions of the Improved Standard Technical 
Specifications (NUREG-1433), and (3) update the references to current 
regulatory requirements such as those set forth in 10 CFR 20.1-20.262 
and 10 CFR 20.1001-20.2402.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee provided 
its analysis of the issue of no significant hazards consideration. The 
NRC staff reviewed the licensee's analysis and has performed its own, 
which is presented below:

    1. Does the amendment involve a significant increase in the 
probability or consequences of an accident previously evaluated?
    No. The proposed amendment does not affect accident initiators 
or precursors because it does not alter any design parameter, 
condition, equipment configuration, or manner in which the unit is 
operated. Furthermore, it does not alter or prevent the ability of 
existing structures, systems, or components to perform their 
intended safety or accident-mitigating functions depicted in the 
Updated Final Safety Analysis Report. The proposed amendment is 
administrative and it only alters the format and location of 
programmatic controls and procedural details. These changes will not 
prevent the unit to continue to comply with applicable regulatory 
requirements. As a result, the proposed amendment will not alter the 
conditions or assumptions used in previous accident analyses. 
Therefore, operation in accordance with the proposed amendment will 
not involve a significant increase in the probability or 
consequences of an accident previously evaluated.
    2. Does the change create the possibility of a new or different 
kind of accident from any accident previously evaluated?
    No. The proposed amendment does not affect accident initiators 
or precursors because it does not alter any design parameter, 
condition, equipment configuration, or manner in which the unit is 
operated. Furthermore, it does not alter or prevent the ability of 
structures, systems, or components to perform their intended safety 
or accident mitigating functions. Accordingly, the proposed 
amendment does not create any new or different kind of accident from 
any accident previously evaluated.
    3. Does the change involve a significant reduction in a margin 
of safety?
    No. The proposed amendment does not change any design parameter, 
analysis methodology, safety limits or acceptance criteria. It is 
administrative as outlined above, with the objective to assure 
continued compliance with applicable regulatory requirements 
governing the radiation protection plan, radioactive effluents, 
radioactive sources, and radiological environmental monitoring. 
Therefore, operation in accordance with the proposed amendment will 
not involve a significant reduction in a margin of safety.

    Based on the NRC staff's review, it appears that the three 
standards of 10 CFR 50.92(c) are satisfied. Therefore, the NRC staff 
proposes to determine that the proposed amendment involves no 
significant hazards consideration.
    Attorney for licensee: Mark J. Wetterhahn, Esquire, Winston & 
Strawn, 1400 L Street, NW., Washington DC 20005-3502.
    NRC Section Chief: L. Raghavan, Acting.

Nine Mile Point Nuclear Station, LLC, Docket No. 50-220, Nine Mile 
Point Nuclear Station, Unit No. 1 (NMP-1), Oswego County, New York

    Date of amendment request: October 26, 2001.
    Description of amendment request: Section 6.0 of the Technical 
Specifications (TSs) delineates the required administrative controls, 
including plant management responsibilities, station organization, 
staff qualifications and training, review and audit activities, 
procedures, reporting requirements, record retention, radiation area 
control, and various plant programs. The licensee proposed an amendment 
to revise Section 6.0 of the TSs to make it consistent with its 
counterpart, Section 5.0, of the Nine Mile Point Nuclear Station, Unit 
No. 2 (NMP-2) TSs. The NMP-2 TSs was fully converted to the format and 
style of the Improved Standard Technical Specifications (NUREG-1433 and 
NUREG-1434) by Amendment No. 91, dated February 15, 2000. While NMP-1 
and NMP-2 are of different reactor designs, the administrative controls 
are, by necessity, either identical or very similar. Consistency of 
administrative controls between the two units is essential to avoid 
confusion and to improve efficiency.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration. The NRC staff has reviewed the licensee's analysis, and 
performed its own, which is presented below:

    1. Does the amendment involve a significant increase in the 
probability or consequences of an accident previously evaluated?
    No. The proposed amendment is concerned only with administrative 
controls, and does not affect accident initiators or precursors 
because it does not alter any design parameter, condition, equipment 
configuration, or manner in which NMP-1 is operated. Furthermore, it 
does not alter or prevent the ability of existing structures, 
systems, or components to perform their intended safety or accident-
mitigating functions depicted in the Updated Final Safety Analysis 
Report. The proposed amendment only affects the administrative 
controls in accordance with NUREG-1433 and NUREG-1434. These changes 
will not prevent NMP-1 to continue to comply with applicable 
regulatory requirements. As a result, the proposed amendment will 
not alter the conditions or assumptions used in previous accident 
analyses. Therefore, operation in accordance with the proposed 
amendment will not involve a significant increase in the probability 
or consequences of an accident previously evaluated.
    2. Does the change create the possibility of a new or different 
kind of accident from any accident previously evaluated?
    No. The proposed amendment does not affect accident initiators 
or precursors because it does not alter any design parameter, 
condition, equipment configuration, or manner in which the unit is 
operated. Furthermore, it does not alter or prevent the ability of 
structures, systems, or components to perform their intended safety 
or accident mitigating functions. Accordingly, the proposed 
amendment does not create any new or different kind of accident from 
any accident previously evaluated.
    3. Does the change involve a significant reduction in a margin 
of safety?
    No. The proposed amendment does not change any design parameter, 
analysis methodology, safety limits or acceptance criteria. It is 
administrative as outlined above, with the objective to assure 
continued compliance with applicable regulatory requirements 
governing the various topics of administrative controls. Therefore, 
operation in accordance with the proposed amendment will not involve 
a significant reduction in a margin of safety.

    Based on the NRC staff's review, it appears that the three 
standards of 10 CFR 50.92(c) are satisfied. Therefore, the NRC staff 
proposes to determine that the

[[Page 929]]

proposed amendment involves no significant hazards consideration.
    Attorney for licensee: Mark J. Wetterhahn, Esquire, Winston & 
Strawn, 1400 L Street, NW., Washington, DC 20005-3502.
    NRC Section Chief: L. Raghavan (Acting).

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

    Date of amendment requests: October 17, 2001.
    Description of amendment requests: The proposed license amendments 
would modify Technical Specification 3.9.4, ``Containment 
Penetrations,'' to allow the equipment hatch, both personnel air lock 
doors and both emergency air lock doors to remain open, and penetration 
flow path(s) providing direct access from the containment atmosphere to 
the outside atmosphere to be unisolated under administrative control, 
during core alterations and movement of irradiated fuel assemblies. In 
addition, the amendments revise Technical Specification 1.1, 
``Definitions,'' for Dose Equivalent I-131, to allow the use of the 
thyroid dose conversion factors, listed in the International Commission 
on Radiological Protection Publication 30, ``Limits for Intakes of 
Radionuclides by Workers.''
    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 change would allow the containment equipment hatch, 
Personnel Air Lock (PAL) doors, Emergency Air Lock (EAL) doors, and 
penetrations to remain open during fuel movement and core 
alterations. These penetrations are normally closed during this time 
period in order to prevent the escape of radioactive material in the 
event of a Fuel Handling Accident (FHA) inside containment. These 
penetrations are not initiators of any accident and the probability 
of a FHA is unaffected by the position of these penetrations.
    The new FHA analysis with an open containment demonstrates the 
maximum offsite doses are well within (less than 25%) the limits 
specified in 10 CFR 100. These offsite dose values are also well 
within the acceptable limits provided in NUREG-0800, Section 15.7.4. 
This FHA analysis results in a maximum offsite dose of 60.62 Rem to 
the thyroid and 0.4281 Rem to the whole body. The calculated control 
room dose is also well below the acceptance criteria specified in 
General Design Criteria (GDC) 19. The analysis results in thyroid 
and whole body doses to the control room operator of 11.56 Rem and 
0.0072 Rem, respectively. Although the offsite and control room dose 
values are increased by the proposed changes, the resulting values 
are still well within acceptable limits and do not significantly 
increase the consequences of a FHA.
    Therefore, the proposed changes do not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    2. The proposed change does not create the possibility of a new 
or different kind of accident from any accident previously 
evaluated.
    The proposed change does not involve the addition or 
modification of any plant equipment. However, the proposed change 
does alter the containment closure configuration and method of 
operation of the plant during certain operational activities. The 
proposed change involves a change to the technical specification 
(TS) that would allow the equipment hatch door, the PAL doors, the 
EAL doors, and containment penetrations to be open during core 
alterations and fuel movement inside containment. This change only 
affects the containment barrier configuration of the plant during 
certain operational activities. Even allowing these doors and 
penetrations to be open, all of the resulting radiological 
consequences remain within acceptable limits and this configuration 
does not create the possibility of a new or different accident than 
previously evaluated.
    Therefore, the proposed change does not create the possibility 
of a new or different kind of accident from any previously 
evaluated.
    3. The proposed change does not involve a significant reduction 
in a margin of safety.
    This proposed change creates the potential for increased dose in 
the control room and at the site boundary due to a FHA. However, the 
new analysis demonstrates that the resultant doses are well within 
the 10 CFR 100 limits and well below the GDC [General Design 
Criterion] 19 limits. In the case of the offsite dose values, they 
remain less than 25% of the 10 CFR 100 limits, which is considered 
acceptable in NUREG-0800 [Standard Review Plan], Section 15.7.4. 
Based on this, even though the dose values have increased from the 
previously calculated values, the margin of safety is not 
significantly reduced.
    In the new analysis, the offsite and control room doses due to a 
FHA with an open containment have been evaluated using conservative 
assumptions, such as all airborne activity caused by the FHA in the 
containment is released instantaneously to the outside atmosphere, 
which ensures the calculation bounds the expected dose. The new 
analysis also assumes closure of the containment within two hours. 
As a result, requiring immediate initiation of the closure of the 
containment and completion of closure within approximately 30 
minutes following a containment evacuation requirement from the FHA 
will reduce the potential offsite doses in the event of a FHA, and 
provides additional margin to the calculated offsite doses.
    Therefore, the proposed change does not involve a significant 
reduction in a margin of safety.

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

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

    Date of amendment requests: November 13, 2001.
    Description of amendment requests: The proposed license amendments 
would modify Technical Specifications 5.5.9, ``Steam Generator (SG) 
Tube Surveillance Program,'' and 5.6.10, ``SG Tube Inspection Report,'' 
of the Diablo Canyon Power Plant, Unit Nos. 1 and 2 Technical 
Specifications, to add new surveillance and reporting requirements 
associated with SG tube inspection and repair. The new requirements 
establish alternate repair criteria for axial primary water stress 
corrosion cracking at dented tube support plate intersections.
    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.
    Examination of crack morphology for primary water stress 
corrosion cracking (PWSCC) at dented intersections has been found to 
show one or two microcracks well aligned with only a few uncorroded 
ligaments and little or no other inside diameter axial cracking at 
the intersection. This relatively simple morphology is conducive to 
obtaining good accuracy in nondestructive examination (NDE) sizing 
of these indications. Accordingly, alternate repair criteria (ARC) 
are established based on crack length and average and maximum depth 
within the thickness of the tube support plate (TSP).
    The application of the ARC requires a Monte Carlo condition 
monitoring assessment to determine the as-found condition of the 
tubing. The condition

[[Page 930]]

monitoring analysis described in WCAP-15573, Revision 1, is 
consistent with NRC Generic Letter 95-05 [``Voltage-Based Repair 
Criteria for Westinghouse Steam Generator Tubes Affected by Outside 
Diameter Stress Corrosion Cracking''] requirements.
    The application of the ARC requires a Monte Carlo operational 
assessment to determine the need for tube repair. The repair bases 
are obtained by projecting the crack profile to the end of the next 
operating cycle and determining the burst pressure and leakage for 
the projected profile using Monte Carlo analysis techniques 
described in WCAP-15573, Revision 1. The burst pressure and leakage 
are compared to the requirements in WCAP-15573, Revision 1. Separate 
analyses are required for the total crack length and the length 
outside the TSP due to differences in requirements. If the projected 
end of cycle (EOC) requirements are satisfied, the tube will be left 
in service.
    A steam generator (SG) tube rupture event is one of a number of 
design basis accidents that are analyzed as part of a plant's 
licensing basis. A single or multiple tube rupture event would not 
be expected in a SG in which the ARC has been applied. The ARC 
requires repair of any indication having a maximum crack depth 
greater than or equal to 40 percent outside the TSP, thus limiting 
the potential length of a deep crack outside the TSP at EOC 
conditions and providing margin against burst and leakage for free 
span indications.
    For other design basis accidents such as a MSLB [main steam line 
break], MFLB [main feed line break], control rod ejection, and 
locked reactor coolant pump motor, the tubes are assumed to retain 
their structural integrity.
    Therefore, the proposed change does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    2. The proposed change does not create the possibility of a new 
or different kind of accident from any accident previously 
evaluated.
    Implementation of the proposed SG tube ARC does not introduce 
any significant changes to the plant design basis. A single or 
multiple tube rupture event would not be expected in a SG in which 
the ARC has been applied. Both condition monitoring and operational 
assessments are completed as part of the implementation of ARC to 
determine that structural and leakage margin exists prior to 
returning SGs to service following inspections. If the condition 
monitoring requirements are not satisfied for burst or leakage, the 
causal factors for EOC indications exceeding the expected values 
will be evaluated. The methodology and application of this ARC will 
continue to ensure that tube integrity is maintained during all 
plant conditions consistent with the requirements of Regulatory 
Guide (RG) 1.121 [``Bases for Plugging Degraded PWR Steam Generator 
Tubes''] and Revision 1 of RG 1.83 [Inservice Inspection of 
Pressurized Water Reactor Steam Generator Tubes].
    In the analysis of a SG tube rupture event, a bounding primary-
to-secondary leakage rate equal to the operational leakage limits in 
the Technical Specifications (TS), plus the leak rate associated 
with the double-ended rupture of a single tube, is assumed. For 
other design basis accidents, the tubes are assumed to retain their 
structural integrity and exhibit primary-to-secondary leakage within 
the limits assumed in the current licensing basis accident analyses. 
MSLB leakage rates from the proposed PWSCC ARC are combined with 
leakage rates from other approved ARC (i.e., voltage-based ARC and 
W* ARC). The combined leakage rates will not exceed the limits 
assumed in the current licensing basis accident analyses.
    The 40 percent maximum depth repair limit for free span 
indications provides a very low likelihood of free span leakage 
under design basis or severe accident conditions. Leakage from 
indications inside the TSP is limited by the constraint of the TSP 
even under severe accident conditions, and leakage behavior in a 
severe accident would be similar to that found acceptable by the NRC 
under approved ARC for axial outside diameter stress corrosion 
cracking (ODSCC) at TSP intersections. Therefore, even under severe 
accident conditions, it is concluded that application of the 
proposed ARC for PWSCC at dented TSP locations results in a 
negligible difference in risk of a tube rupture or large leakage 
event, when compared to current 40 percent repair limits or 
previously approved ARC.
    Diablo Canyon Power Plant (DCPP) continues to implement a 
maximum operating condition leak rate limit of 150 gallons per day 
per SG to preclude the potential for excessive leakage during all 
plant conditions.
    The possibility of a new or different kind of accident from any 
previously evaluated is not created because SG tube integrity is 
maintained by inservice inspection, condition monitoring, 
operational assessment, tube repair, and primary-to-secondary 
leakage monitoring.
    3. The proposed change does not involve a significant reduction 
in a margin of safety.
    Tube repair limits provide reasonable assurance that tubes 
accepted for continued service without repair will exhibit adequate 
tube structural and leakage integrity during subsequent plant 
operation. The implementation of the proposed ARC is demonstrated to 
maintain SG tube integrity consistent with the criteria of draft NRC 
RG 1.121. The guidelines of RG 1.121 describe a method acceptable to 
the NRC staff for meeting General Design Criteria (GDC) 2, 4, 14, 
15, 31, and 32 by ensuring the probability or the consequences of SG 
tube rupture remain within acceptable limits. This is accomplished 
by determining the limiting conditions of degradation of SG tubing, 
for which tubes with unacceptable cracking should be removed from 
service.
    Upon implementation of the proposed ARC, even under the worst-
case conditions, the occurrence of PWSCC at the tube support plate 
elevations is not expected to lead to a SG tube rupture event during 
normal or faulted plant conditions. The ARC involves a computational 
assessment to be completed for each indication left in service 
ensuring that performance criteria for tube integrity and leak 
tightness are met until the next scheduled outage.
    As discussed below, certain tubes are excluded from application 
of ARC. Existing tube integrity requirements apply to these tubes, 
and the margin of safety is not reduced. In addressing the combined 
loading effects of a loss-of-coolant (LOCA) and safe shutdown 
earthquake (SSE) on the SGs (as required by GDC 2), the potential 
exists for yielding of the TSP in the vicinity of the wedge groups, 
accompanied by deformation of tubes and a subsequent postulated in-
leakage. Tube deformation could lead to opening of pre-existing 
tight through wall cracks, resulting in secondary to primary in-
leakage following the event, which could have an adverse affect on 
the Final Safety Analysis Report (FSAR) results. Based on a DCPP 
analysis of LOCA and SSE, SG tubes located in wedge region exclusion 
zones are susceptible to deformation, and are excluded from 
application of ARC.
    A DCPP tube stress analysis for MFLB/MSLB plus SSE loading 
determined that high bending stresses occur in certain SG tubes at 
the seventh TSP, because the stresses exceed the maximum imposed 
bending stress for existing test data (equal to approximately the 
lower tolerance limit yield stress). These tubes are located in rows 
11 to 15 and 36 to 46, and are excluded from application of ARC.
    Tube intersections that contain TSP ligament cracking are also 
excluded from application of ARC.
    Based on the above, it is concluded that the proposed license 
amendment request does not result in a significant reduction in 
margin with respect to the plant safety analyses as defined in the 
FSAR or 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: Christopher J. Warner, Esq., Pacific Gas and 
Electric Company, P.O. Box 7442, San Francisco, California 94120.
    NRC Section Chief: Stephen Dembek.

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

    Date of amendment requests: November 16, 2001.
    Description of amendment requests: The proposed license amendments 
would modify Technical Specification 5.5.16, ``Containment Leakage Rate 
Testing Program,'' to allow a one-time extension of the ten-year 
interval for the performance-based leakage rate testing program for 
Type A tests as prescribed by Nuclear Energy Institute (NEI) Report NEI 
94-01, Revision 0, ``Industry

[[Page 931]]

Guideline for Implementing Performance-Based Option of 10 CFR part 50, 
Appendix J,'' and applied by 10 CFR part 50, Appendix J, Option B.
    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 extension to the Type A testing interval from 1-in-
10 years to 1-in-15 years will not increase the probability of an 
accident previously evaluated. The determination of containment 
integrity is not an accident initiator. The containment Type A 
testing interval extension does not involve a plant modification and 
the testing interval extension is not of a type that could lead to 
equipment failure or accident initiation.
    The proposed extension to the Type A testing interval does not 
involve a significant increase in the consequences of an accident. 
Research documented in NUREG-1493 has determined that Type B and C 
tests can identify the vast majority (approximately 97 percent) of 
all potential leakage paths. Experience at Diablo Canyon Power Plant 
(DCPP) demonstrates that excessive containment leakage paths are 
detected by Type B and C local leakage rate tests. Type B and C 
testing will identify any containment opening, such as a valve, that 
would otherwise be detected by the Type A tests.
    NUREG-1493 concluded that increasing the Type A test interval to 
1-in-20 years leads to an imperceptible increase in risk. A DCPP 
plant specific probabilistic risk assessment of the change in the 
Type A testing interval from 1-in-10 years to 1-in-15 years 
determined the total integrated risk of the associated specific 
accident sequences increases by 0.03 percent. This risk impact when 
compared to other severe accident induced risks is negligible. The 
increase in the Regulatory Guide (RG) 1.174 [An Approach for Using 
Probabilistic Risk Assessment in Risk-Informed Decisions on Plant-
Specific Changes to the Licensing Basis] large early release 
fraction (LERF) figure-of-merit criteria resulting from a change in 
the Type A test interval from 1-in-10 years to 1-in-15 years is risk 
insignificant.
    Testing and inspection provide a high degree of assurance that 
the containment will not degrade in a manner detectable only by Type 
A testing. The structural capability of the containment has been 
shown by the Type A testing results that have established that DCPP 
has had acceptable containment leakage rates with considerable 
margin. Inspections required by 10 CFR 50.65 and American Society of 
Mechanical Engineers code are performed in order to identify 
indications of containment degradation that could affect leak 
tightness. The results of containment concrete examination have 
concluded the containment concrete has had no loss of structural 
capacity and no areas of the concrete shell have experienced 
accelerated degradation or aging. The results of containment liner 
inspections have not identified any significant degradations that 
could adversely impact the containment structural integrity or leak 
tightness, such as through-holes in the containment liner. Due to 
the large containment leakage rate margin available, and no 
identified mechanism that would cause significant degradation of 
containment, a 5 year extension of the ILRT interval would not be 
expected to result in containment leakage above the acceptable 
limit.
    Therefore, the proposed change does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    2. The proposed change does not create the possibility of a new 
or different kind of accident from any accident previously 
evaluated.
    The proposed extension of the Type A testing interval will not 
create the possibility of a new or different type of accident from 
any previously evaluated. There are no physical changes being made 
to the plant and there are no changes in operation of the plant that 
could introduce a new failure mode, creating an accident.
    The containment structure is passive. Under normal operating 
conditions, there is no significant environmental or operational 
stress present that would contribute to its degradation. Passive 
failures resulting in significant containment structural leakage are 
therefore extremely unlikely to develop between Type A tests.
    Therefore, the proposed change does not create the possibility 
of a new or different kind of accident from any accident previously 
evaluated.
    3. The proposed change does not involve a significant reduction 
in a margin of safety.
    The proposed extension of the Type A testing interval 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 results in an 
imperceptible increase in risk to the public. NUREG-1493 found that, 
generically, the design containment leakage rate contributes about 
0.1 percent to the individual risk and that the increase in the Type 
A testing interval would have a minimal effect on this risk because 
97 percent of the potential leakage paths are detected by Type B and 
C testing.
    A DCPP plant specific probabilistic risk assessment of the 
change in the Type A testing interval from 1-in-10 years to 1-in-15 
years determined the total integrated risk of the associated 
specific accident sequences increases by 0.03 percent. This risk 
impact when compared to other severe accident induced risks is 
negligible. The increase in RG 1.174 LERF figure-of-merit criteria 
resulting from a change in the Type A test interval from 1-in-10 
years to 1-in-15 years is risk insignificant.
    Deferral of Type A testing for DCPP does not increase the level 
of risk to the public due to loss of capability to detect and 
measure containment leakage or loss of containment structural 
capability. Other containment testing methods and inspections will 
assure all limiting conditions of operation will continue to be met. 
The margin of safety inherent in existing accident analyses is 
maintained.
    Therefore, the proposed change does not involve a significant 
reduction in a margin of safety.

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

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

    Date of amendment requests: November 16, 2001.
    Description of amendment requests: The proposed license amendments 
would modify Technical Specification (TS) 1.1, ``Definitions, Dose 
Equivalent I-131,'' to allow the use of the thyroid dose conversion 
factors listed in the International Commission on Radiological 
Protection (ICRP) Publication 30, ``Limits for Intakes of Radionuclides 
by Workers,'' 1979, in the steam generator tube rupture and main steam 
line break radiological consequences analyses.
    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 revision of Technical Specification (TS) 1.1, Definitions, 
``Dose Equivalent I-131,'' to allow use of the iodine thyroid dose 
conversion factors from the International Commission on Radiological 
Protection (ICRP) Publication 30, 1979, and the revised steam 
generator tube rupture (SGTR) and main steam line break (MSLB) 
radiological consequences analyses are used to determine post-
accident dose. They are not related to any accident initiator. 
Therefore, this change cannot increase the probability of an 
accident.
    The revised SGTR thermal and hydraulic analysis input 
assumptions are consistent with actual plant limits and parameters.
    The revised MSLB offsite and control room radiological 
consequences analysis dose results are within 10 CFR Part 100 limits 
and

[[Page 932]]

the NUREG-0800 Standard Review Plan (SRP) section 15.1.5 and section 
6.4 guideline values.
    The revised SGTR control room radiological consequences analysis 
dose results are within the SRP section 6.4 guideline values.
    The revised SGTR offsite radiological consequences analysis dose 
results are within the 10 CFR part 100 dose limits. The SGTR offsite 
dose results also meet the SRP section 15.6.3 and section 6.4 
guideline values, with the exception of the 2 hour Exclusion Area 
Boundary (EAB) thyroid dose. The calculated 2 hour EAB thyroid dose 
of 30.5 Rem is 1.5 percent above the SRP 15.6.3 guideline value of 
30 Rem. The 2 hour EAB thyroid dose has been compared against the 
conservative thyroid dose SRP 15.6.3 guideline value of 30 Rem. The 
2 hour EAB dose thyroid dose would be equivalent to a Regulatory 
Guide (RG) 1.183 methodology Total Effective Dose Equivalent (TEDE) 
of approximately 1.25 Rem, which is well below the RG 1.183 TEDE 
limit of 2.5 Rem for the accident-initiated iodine spike case. 
Therefore, the 2 hour EAB thyroid dose of 30.5 Rem is not considered 
to be a significant increase in dose.
    Therefore, the proposed change does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    2. The proposed change does not create the possibility of a new 
or different kind of accident from any accident previously 
evaluated.
    The use of the iodine thyroid dose conversion factors from ICRP 
Publication 30 and the revised SGTR and main steam line break MSLB 
radiological consequences analyses do not involve any physical plant 
changes. The change does not involve changes in operation of the 
plant that could introduce a new failure mode for creating an 
accident or affect the mitigation of an accident.
    Therefore, the proposed change does not create the possibility 
of a new or different accident from any previously evaluated.
    3. The proposed change does not involve a significant reduction 
in a margin of safety.
    The use of ICRP Publication 30 thyroid dose conversion factors 
to calculate the radiological consequences for a SGTR and MSLB 
accident is endorsed by RG 1.183, ``Alternative Radiological Source 
Terms for Evaluating Design Basis Accidents at Nuclear Power 
Reactors,'' US Nuclear Regulatory Commission, July 2000. Therefore, 
the revision of TS 1.1, Definitions, ``Dose Equivalent I-131,'' to 
allow use of the iodine thyroid dose conversion factors from ICRP 
Publication 30 does not result in a significant reduction in the 
margin provided by TS 1.1. The revised SGTR thermal and hydraulic 
analysis input assumptions are consistent with actual plant limits 
and parameters.
    The revised MSLB offsite and control room radiological 
consequences analysis dose results are within 10 CFR part 100 limits 
and the NUREG-0800 SRP section 15.1.5 and section 6.4 guideline 
values.
    The revised SGTR control room radiological consequences analysis 
dose results are within the SRP section 6.4 guideline values.
    The revised SGTR radiological consequences analysis dose results 
are within the 10 CFR part 100 dose limits. The SGTR dose results 
also meet the SRP section 15.6.3 and section 6.4 guideline values, 
with the exception of the 2 hour EAB thyroid dose. The calculated 2 
hour EAB thyroid dose of 30.5 Rem is 1.5 percent above the SRP 
15.6.3 guideline value of 30 Rem. The 2 hour EAB dose thyroid dose 
would be equivalent to a RG 1.183 methodology TEDE of approximately 
1.25 Rem, which is well below the RG 1.183 TEDE limit of 2.5 Rem for 
the accident-initiated iodine spike case. Therefore, the 2 hour EAB 
thyroid dose of 30.5 Rem is not a significant reduction in the 
margin of safety.
    Therefore, the proposed change does not involve a significant 
reduction in a margin of safety.

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

Southern California Edison Company, et al., Docket Nos. 50-361 and 50-
362, San Onofre Nuclear Generating Station, Units 2 and 3, San Diego 
County, California

    Date of amendment requests: December 11, 2001.
    Description of amendment requests: The proposed amendments would 
revise Technical Specification (TS) 3.8.3, ``Diesel Fuel Oil, Lube Oil, 
and Starting Air,'' on the emergency diesel generators. The revisions 
would change (1) Conditions A and C of the Actions for Limiting 
Condition for Operation 3.8.3, and (2) Surveillance Requirement 
3.8.3.1. The proposed amendments would change the minimum required 
diesel fuel oil storage to support (1) using California Diesel fuel 
rather than the existing Environmental Protection Agency (EPA) Clear 
diesel fuel, (2) revising the diesel generator load profile in reactor 
Modes 1 through 4, and (3) changing the units of the required diesel 
fuel oil storage. A ``greater than or equal to'' would also be changed 
to a ``greater than'' sign.
    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?
    Response: No.
    This proposed change revises the minimum amount of stored diesel 
fuel. The change is required to (1) support the use of California 
Diesel fuel rather than the existing EPA Clear diesel fuel, and (2) 
reflect a change in the diesel generator load profile in Modes 1 
through 4.
    In addition, this proposed change revises the units for the 
minimum diesel fuel storage requirements from tank level to a 
minimum required volume of fuel in gallons. A ``greater than or 
equal to'' sign is revised to a ``greater than'' sign for 
consistency. These are administrative changes only.
    Technical Specification (TS) 3.8.3, ``Diesel Fuel Oil, Lube Oil, 
and Starting Air,'' requires that each diesel generator have 
sufficient fuel to operate for a period of 7 days, while the Diesel 
Generator (DG) is supplying maximum post Loss of Coolant Accident 
(LOCA) load demand. This requirement is currently expressed as a 
minimum tank level limit. In Modes 1 through 4, the existing tank 
level limit is 89%, which ensures that a 7-day supply of fuel is 
available. TS 3.8.3, Condition A, states that during Modes 1 through 
4, if one or more Diesel Generators (DG) has a fuel level in the 
storage tank less than 89% and greater than or equal to 76%, then 
fuel oil level must be restored to within limits within 48 hours. 
The 76% level requirement is based on maintaining a 6-day supply of 
diesel fuel in Modes 1 through 4. If the tank level is at or below 
76% (6-day supply), the associated DG must be declared inoperable 
immediately.
    Similarity, for Modes 5 and 6, the existing tank level limit is 
72%, which ensures that a 7-day supply of fuel is available. TS 
3.8.3, Condition C, states that during Modes 5 and 6, if one 
required DG has a fuel level in the storage tank less than 72% and 
greater than 63%, then [the] fuel oil level must be restored to 
within limits within 48 hours. The 63% level requirement is based on 
maintaining a 6-day supply of diesel fuel in Modes 5 and 6. If the 
tank level is at or below 63% (6-day supply), the associated diesel 
generator must be declared inoperable immediately.
    As described in the Bases to TS 3.8.3, these tank level 
requirements are based on fuel volume requirements. In Modes 1 
through 4, 89% and 76% level limits are based on a 7-day (49,724 
gallons) and 6-day (42,960 gallons) fuel supply, respectively [plus 
an allowance for instrument Total Loop Uncertainty (TLU)]. In Modes 
5 and 6, the 72% and 63% tank level limits are based on a 7-day 
(40,472 gallons) and 6-day (34,960 gallons) fuel supply, 
respectively (plus an allowance for instrument TLU).
    Because the Lower Heating Value (LHV) per gallon of California 
Diesel fuel is less than that of EPA Clear diesel fuel, it was 
necessary to recalculate the amount of fuel required to supply 
necessary loads for the required time periods. For Modes 1 through 
4, the resulting minimum volumes of California Diesel fuel are 
45,662 gallons and 39,468 gallons for the 7-day and 6-day fuel 
supply, respectively. For Modes 5 and 6, the required volumes of 
California Diesel fuel are

[[Page 933]]

41,691 gallons and 35,735 gallons for a 7-day supply and a 6-day 
supply, respectively.
    It should be noted that the minimum volumes for Modes 1 through 
4 are decreased due to a change in the calculated [diesel generator] 
load profile. SONGS no longer requires the third-of-a-kind High 
Pressure Safety Injection (HPSI) pump to be started following a Loss 
of Coolant Accident (LOCA) or Main Steam Line Break (MSLB). 
Operation of the third-of-a-kind HPSI pump is no longer assumed as 
part of the DG load profile in Modes 1 through 4. This resulted in a 
net decrease in the amount of required stored diesel fuel in Modes 1 
through 4, even when the use of the California Diesel fuel [with the 
lower heating value] is taken into account.
    The diesel generators and the associated support systems such as 
the fuel oil storage and transfer systems are designed to mitigate 
accidents and are not accident initiators. Revising the minimum 
volumes of stored [diesel] fuel in the storage tanks will not result 
in a significant increase in the probability of any accident 
previously evaluated. [The revisions are to maintain the current 
requirements for a 7-day and 6-day supply of stored diesel fuel].
    Following implementation of this proposed change, there will be 
no change in the ability of the diesel generators to supply maximum 
post-LOCA load demand for 7 days. The proposed minimum volumes of 
fuel, 45,662 gallons and 39,468 gallons, ensure that a 7-day and 6-
day supply of fuel, respectively, are available in Modes 1 through 
4. The proposed minimum volumes of fuel, 41,691 gallons and 35,735 
gallons, ensure that a 7-day and a 6-day supply, respectively, of 
fuel is available in Modes 5 and 6. This is identical to the current 
requirements. Therefore this change will not result in a significant 
increase in the consequences of any accident previously evaluated.
    Therefore, the proposed change does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    2. Does the proposed change create the possibility of a new or 
different kind of accident from any accident previously evaluated?
    Response: No.
    Following this change, the diesel generators will still be able 
to supply maximum post-LOCA load demand. The current 7-day and 6-day 
fuel supply requirements will be maintained following this change. 
[The diesel generators fuel oil storage and transfer systems are not 
accident initiators].
    The changes in units from tank level percentage to fuel volume 
in gallons is an administrative change only. The change from a 
``greater than or equal to'' sign to a ``greater than'' sign in TS 
3.8.3, Condition A, is for consistency with other parts of TS 3.8.3 
and is also an administrative change.
    Therefore, this proposed change will not create the possibility 
of a new or different kind of accident from any accident that has 
been previously evaluated.
    3. Does the proposed change involve a significant reduction in a 
margin of safety?
    Response: No.
    The Bases to TS 3.8.3 states that ``Each diesel generator (DG) 
is provided with a storage tank having a fuel oil capacity 
sufficient to operate that diesel for a period of 7 days, while the 
DG is supplying maximum post loss of coolant accident load demand.'' 
When the fuel oil tank level is less than required to support [7 
days] of operation, the required action depends on whether or not a 
6-day supply of fuel is available. [The proposed tank level limits 
for Modes 1 through 4 will maintain the 7-day and 6-day fuel supply 
requirements following changeout to California Diesel fuel and the 
change in the DG load profile for Modes 1 through 4].
    The proposed tank level limits for Modes 5 and 6 will maintain 
these 7-day and 6-day fuel supply requirements following changeout 
to California Diesel fuel.
    The change in units from tank level percentage to fuel volume in 
gallons is an administrative change only. The change from a 
``greater than or equal to'' sign to a ``greater than'' sign in TS 
3.8.3, Condition A, is for consistency with other parts of the TS 
3.8.3 and is also an administrative change.
    Therefore, the proposed change does not involve a significant 
reduction in a margin of safety.

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

Tennessee Valley Authority, Docket No. 50-296, Browns Ferry Nuclear 
Plant, Unit 3, Limestone County, Alabama

    Date of amendment request: November 1, 2001.
    Description of amendment request: The proposed amendment would 
revise Technical Specification (TS) 2.1.1.2, ``Reactor Core Safety 
Limits,'' by modifying the safety limit minimum critical power ratio 
(SLMCPR).
    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:

    As required by 10 CFR 50.91(a), TVA 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 amendment establishes revised SLMCPR values for two 
recirculation loop operation and for single recirculation loop 
operation. The probability of an evaluated accident is derived from 
the probabilities of the individual precursors to that accident. The 
proposed SLMCPRs preserve the existing margin to transition boiling 
and the probability of fuel damage is not increased. Since the 
change does not require any physical plant modifications or 
physically affect any plant components, no individual precursors of 
an accident are affected and the probability of an evaluated 
accident is not increased by revising the SLMCPR values.
    The consequences of an evaluated accident are determined by the 
operability of plant systems designed to mitigate those 
consequences. The revised SLMCPRs have been performed using NRC-
approved methods and procedures. The basis of the MCPR Safety Limit 
is to ensure no mechanistic fuel damage is calculated to occur if 
the limit is not violated. These calculations do not change the 
method of operating the plant and have no effect on the consequences 
of an evaluated accident. Therefore, the proposed TS change does not 
involve an 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 license amendment involves a revision of the SLMCPR 
for two recirculation loop operation and for single loop operation 
based on the results of an analysis of the Cycle 11 core. Creation 
of the possibility of a new or different kind of accident would 
require the creation of one or more new precursors of that accident. 
New accident precursors may be created by modifications of the plant 
configuration, including changes in the allowable methods of 
operating the facility. This proposed license amendment does not 
involve any modifications of the plant configuration or changes in 
the allowable methods of operation. Therefore, the proposed TS 
change does 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 margin of safety as defined in the TS bases will remain the 
same. The new SLMCPRs are calculated using NRC-approved methods and 
procedures which are in accordance with the current fuel design and 
licensing criteria. The SLMCPRs remain high enough to ensure that 
greater than 99.9% of all fuel rods in the core are expected to 
avoid transition boiling if the limit is not violated, thereby 
preserving the fuel cladding integrity. Therefore, the proposed TS 
changes do not involve a 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.

[[Page 934]]

    Attorney for licensee: General Counsel, Tennessee Valley Authority, 
400 West Summit Hill Drive, ET l0H, 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, Hamilton County, Tennessee

    Date of application for amendments: September 12, 2001 (TS 01-04).
    Brief description of amendments: The proposed amendments would 
change the Sequoyah (SQN) Unit 1 and 2 Technical Specification (TS) 3/4 
6.5.1 and associated Bases to reflect an increase in the ice condenser 
basket weight from 1071 pounds to 1145 pounds and the total ice 
condenser ice weight from 2,082,024 pounds to 2,225,880 pounds. This 
change is being made in response to a reanalysis by Westinghouse 
Electric Company that identified a modeling input error used in the 
original analysis.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), Tennessee Valley 
Authority (TVA), the licensee, has provided its analysis of the issue 
of no significant hazards consideration, which is presented below:

    A. The proposed amendment does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    The analyzed accidents of consideration in regards to changes 
affecting the ice condenser are a loss-of-coolant accident (LOCA) 
and a main steam line break (MSLB) inside containment. The ice 
condenser is a passive system and is not postulated as being the 
initiator of any LOCA or main steam line break (MSLB) and is 
designed to remain functional following a design basis earthquake.
    In addition, the ice condenser does not interconnect or interact 
with any systems that have an interface with the reactor coolant or 
main steam systems.
    For SQN, the LOCA is the more severe accident in terms of 
containment pressure and ice bed meltout and is therefore the more 
limiting accident. SQN's LOCA Containment Integrity Analysis 
calculates the post-LOCA peak containment pressure to be 11.44 
pounds per square inch gauge (psig), which is below SQN's 
containment design pressure of 12.0 psig. The analysis contains an 
assumed ice mass that is an input value to the calculation to ensure 
that sufficient heat removal capability is available from SQN's ice 
condenser to limit the accident peak pressure inside containment. 
The analyzed peak accident pressure must remain below the 
containment design pressure.
    TVA's proposed TS revision reflects the ice mass assumed in the 
SQN ['s] Containment Integrity Analysis. Accordingly, TVA's proposed 
change ensures that ice mass values retain the existing margin 
between the calculated peak containment accident pressure and SQN's 
containment design pressure.
    Since the proposed changes to the TS and TS bases are solely to 
revise ice weight values to reflect current margins within SQN's 
analysis, and are not the result of or require any physical changes 
to the ice condenser, there is no change in the probability of an 
accident previously evaluated in the Safety Analysis Report.
    Based on the above discussions, the proposed changes do not 
involve an increase in the probability or consequences of an 
accident previously evaluated.
    B. The proposed amendment does not create the possibility of a 
new or different kind of accident from any accident previously 
evaluated.
    Because the TS and TS bases changes do not involve any physical 
changes to the ice condenser, or chemical changes to the ice 
contained therein, or make any changes in the operational aspects of 
the ice condenser as required by the TS, there are no new or 
different kind of accidents created from those already identified 
and evaluated.
    C. The proposed amendment does not involve a significant 
reduction in a margin of safety.
    The ice condenser TSs ensure that during a LOCA or MSLB the ice 
condenser will initially pass sufficient air and steam mass to 
preclude over-pressurizing lower containment, that it will absorb 
sufficient heat energy initially and over a prescribed time period 
to assist in precluding containment vessel failure, and that it will 
not alter the bulk containment sump pH and boron concentration 
assumed in the accident analysis.
    TVA's proposed change does not physically alter the ice 
condenser, but rather accounts for changes to input assumptions for 
SQN's containment pressure analysis to correct a computer model 
input error. The correction to the model provides a more accurate 
accounting of the pressure response inside containment following a 
LOCA. The error correction requires an increase the ice mass assumed 
in the analysis to ensure that SQN's post-LOCA peak containment 
pressures remain unchanged. The margin that exists between the 
accident peak pressure and the containment design pressure is 
unaffected. Accordingly, TVA' s proposed change does not reduce the 
margin of safety.

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

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

    Date of amendment request: November 20, 2001.
    Description of amendment request: The proposed amendment would 
change the Technical Specifications Table 3.2.6 by revising the Allowed 
Outage Times (AOTs) and associated action requirements for certain 
post-accident monitoring (PAM) 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. The NRC staff has reviewed the licensee's analysis 
against the standards of 10 CFR 50.92(c). The NRC staff's review is 
presented below:

    1. Will the proposed changes involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    The PAM instrumentation is not considered as an initiator or 
contributor to any previously evaluated accident. The proposed 
change will not affect any Final Safety Analysis Report safety 
analysis. Because there are no credible failures of the PAM 
instrumentation that could initiate any accidents previously 
evaluated, changing the AOTs and related actions for PAM 
instrumentation will not increase the probability of any accident 
previously evaluated. The operability or inoperability of this 
instrumentation will not cause an accident because this 
instrumentation was not intended to and does not serve a function 
for preventing accidents. Therefore, the proposed change does not 
involve a significant increase in the probability of an accident 
previously evaluated.
    The availability and use of PAM instrumentation ensures that the 
prescribed operator (manual) actions for mitigating the consequences 
of an accident will be implemented when necessary, and that the 
operator has sufficient information to verify required automatic 
actions have occurred as intended. The availability and use of PAM 
instrumentation provide assurance that the consequences of accidents 
will not be greater than previously evaluated. Changes to allowed 
outage times and shutdown completion times do not affect the 
consequences of accidents. The proposed change does not modify any 
parameters or assumptions contained in previously analyzed design-
basis events. The continued availability and use of this 
instrumentation ensures that the prescribed manual operator actions 
for mitigating the consequences of an accident will be implemented 
when necessary, and that the operator has sufficient information to 
verify required automatic actions have occurred as intended. The 
requirements of the revised TS are adequate to ensure the required 
instrumentation is maintained operable such

[[Page 935]]

that PAM instrumentation will be available to perform its intended 
safety function. Therefore, the proposed change does not involve a 
significant increase in the consequences of an accident previously 
evaluated.
    2. Will the proposed changes create the possibility of a new or 
different kind of accident from any accident previously evaluated?
    The proposed change does not involve any physical modification 
to the plant, change in TS setpoints, plant design-basis, or the 
manner in which the plant is operated. Because the PAM 
instrumentation serves a passive function and does not provide any 
automatic action, there are no credible failures of the PAM 
instrumentation that could initiate a new or different kind of 
accident from any accident previously evaluated.
    The change to AOTs and related action requirements for PAM 
instrumentation will not result in a failure mode not previously 
analyzed. This instrumentation is not considered an accident 
precursor because its existence or availability does not have any 
adverse impact in the pre-accident state of the reactor.
    Therefore, this change does not create the possibility of a new 
or different kind of accident from any accident previously 
evaluated.
    3. Will the proposed changes involve a significant reduction in 
a margin of safety?
    PAM instrumentation is assumed to be used by operators for 
monitoring only after an accident occurs and performs no automatic 
functions. The continued availability and use of this 
instrumentation ensures that the prescribed manual operator actions 
for mitigating the consequences of an accident will be implemented 
when necessary, and that the operator has sufficient information to 
verify required automatic actions have occurred as intended. The 
requirements of the revised TS are adequate to ensure the required 
instrumentation is maintained operable such that PAM instrumentation 
will be available to perform its intended safety function.
    Therefore, this change does not involve a significant reduction 
in a margin of safety.

    Based on this review, it appears that the three standards of 10 CFR 
50.92(c) are satisfied. Therefore, the NRC staff proposes to determine 
that the amendment request involves no significant hazards 
consideration.
    Attorney for licensee: Mr. David R. Lewis, Shaw, Pittman, Potts and 
Trowbridge, 2300 N Street, NW., Washington, DC 20037-1128.
    NRC Section Chief: James W. Clifford.

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

    Date of amendment request: December 11, 2001.
    Description of amendment request: A change is proposed to 
Surveillance Requirement (SR) 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 SR 3.0.3: ``A risk 
evaluation shall be performed for any Surveillance 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 December 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: Jay Silberg, Esq., Shaw, Pittman, Potts and 
Trowbridge, 2300 N Street, NW., Washington, DC 20037.
    NRC Section Chief: Stephen Dembek.

[[Page 936]]

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 electronically from the Agencywide 
Document Access and Management System's (ADAMS) Public Electronic 
Reading Room on the Internet at the NRC Web site, http://www.nrc.gov/reading-rm/adams.html. Persons who do not have access to ADAMS or who 
encounter problems in accessing the documents located in ADAMS, should 
contact the NRC Public Document Room Reference staff by telephone at 1-
800-397-4209, 301-415-4737, or by e-mail to [email protected].

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

    Date of application for amendment: December 29, 2000, supplemented 
on October 11, 2001.
    Brief description of amendment: The amendment revises the offsite 
power source identified in Technical Specification 3.7.A.3 to remove 
one listed source and add a different source. Also, the bases have been 
revised to reflect the availability of the offsite sources and to 
revise minor administrative changes.
    Date of Issuance: December 27, 2001.
    Effective date: December 27, 2001, and shall be implemented within 
30 days of issuance.
    Amendment No.: 222.
    Facility Operating License No. DPR-16: Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: April 4, 2001 (66 FR 
17965).
    The October 11, 2001, letter provided clarifying information within 
the scope of the original application and did not change the staff's 
initial proposed no significant hazards consideration determination.
    The Commission's related evaluation of this amendment is contained 
in a Safety Evaluation dated December 27, 2001.
    No significant hazards consideration comments received: No.

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

    Date of application for amendment: April 11, 2001, as supplemented 
June 14, 2001.
    Brief description of amendment: The amendment revises the list of 
documents that describe the analytical methods used to determine the 
core operating limits specified in Technical Specification 6.9.1.8b. 
The revision consists of updating the list of documents to include the 
latest NRC-approved methodologies, along with deleting the revision 
numbers and dates of all documents in the list.
    Date of issuance: December 19, 2001.
    Effective date: As of the date of issuance and shall be implemented 
prior to restart from refueling outage 14 which is currently scheduled 
in early February of 2002.
    Amendment No.: 260.
    Facility Operating License No. DPR-65: Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: July 25, 2001 (66 FR 
38760).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated December 19, 2001.
    No significant hazards consideration comments received: No.

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

    Date of application for amendment: June 28, 2001.
    Brief description of amendment: The amendment deletes the post-
maintenance testing Surveillance Requirement 4.6.3.1 of containment 
isolation valves.
    Date of issuance: December 21, 2001.
    Effective date: As of the date of issuance and shall be implemented 
within 30 days from the date of issuance.
    Amendment No.: 200.
    Facility Operating License No. NPF-49: Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: October 17, 2001 (66 FR 
52799).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated December 21, 2001.
    No significant hazards consideration comments received: No.

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

    Date of application for amendment: June 12, 2001, as supplemented 
by letters dated October 15 and November 16, 2001.
    Brief description of amendment: The amendment revised certain 
emergency diesel generator (EDG) Technical Specifications (TSs) to 
remove the requirement for an accelerated test frequency, remove the 
requirement to subject the EDGs to an inspection in accordance with the 
manufacturer's recommendations, allow that certain EDG tests may be 
done in modes other than shutdown, and remove the EDG special reporting 
requirements.
    Date of issuance: December 17, 2001.
    Effective date: As of the date of issuance to be implemented within 
60 days from the date of issuance.
    Amendment No.: 237.
    Facility Operating License No. NPF-6: Amendment revised the TSs.
    Date of initial notice in Federal Register: July 11, 2001 (66 FR 
36340). The October 15 and November 16, 2001, supplemental letters 
provided clarifying information and revised TSs that were within the 
scope of the original Federal

[[Page 937]]

Register notice and did not change the staff's initial no significant 
hazards consideration determination.
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated December 17, 2001.
    No significant hazards consideration comments received: No.

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

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

    Date of application for amendments: February 22, as supplemented by 
letters dated May 4, and September 13, 2001.
    Brief description of amendments: The amendments revise reactor 
vessel water level--low scram and isolation setpoints in order to 
minimize unnecessary reactor scrams that might result from events 
involving a temporary reduction in feedwater flow. The revision to 
these setpoints was originally requested as part of the power uprate 
licensing amendment. However, since the setpoint reduction will provide 
a similar benefit when operating at the current thermal power, Exelon 
has requested to implement the requested changes prior to power uprate 
approval as part of efforts to improve summer reliability at Dresden 
and Quad Cities Nuclear Power Stations.
    Date of issuance: December 18, 2001.
    Effective date: As of the date of issuance and shall be implemented 
within 30 days.
    Amendment Nos.: Dresden Units 2 and 3--190/184; Quad Cities Units 1 
and 2--200/196.
    Facility Operating License Nos. DPR-19, DPR-25, DPR-29 and DPR-30. 
The amendments revised the Technical Specifications.
    Date of initial notice in Federal Register: June 6, 2001 (66 FR 
30490).
    The supplemental letters contained clarifying information and did 
not change the initial no significant hazards consideration 
determination and did not expand the scope of the original Federal 
Register notice.
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated December 18, 2001.
    No significant hazards consideration comments received: No.

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

    Date of application for amendments: September 29, 2000, as 
supplemented by letters dated March 1, July 13, August 9, August 13, 
August 27, and October 17, 2001.
    Brief description of amendments: The amendments change the 
technical specifications to reflect a change in fuel vendors from 
Siemens Power Corporation to General Electric, and a transition to GE14 
fuel. As part of the transition, changes are made to (1) increase the 
number of required automatic depressurization system valves from four 
to five, and (2) remove an allowance to continue operating for 72 hours 
if certain combinations of emergency core cooling systems are 
inoperable.
    Date of issuance: December 20, 2001.
    Effective date: As of the date of issuance and shall be implemented 
as follows: for Unit 1, prior to reaching Startup (i.e., Mode 2) 
following refueling outage 17, scheduled for completion in November 
2002; for Unit 2, prior to reaching Startup (i.e., Mode 2) following 
refueling outage 16, scheduled for completion in February 2002.
    Amendment Nos.: 201 and 197.
    Facility Operating License Nos. DPR-29 and DPR-30: The amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: December 27, 2000, (65 
FR 81912) and August 22, 2001 (66 FR 44172).
    The submittals dated July 13, August 9, August 13, August 27, and 
October 17, 2001, did not change the scope of the amendment or the 
proposed no significant hazards findings dated December 27, 2000, (65 
FR 81912) and August 22, 2001 (66 FR 44172).
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated December 20, 2001.
    No significant hazards consideration comments received: No.

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

    Date of application for amendments: October 17, 2001.
    Brief description of amendments: Revised St. Lucie Unit 1 and 2 
Technical Specifications (TS) actions regarding inoperable redundant 
components when an Emergency Diesel Generator becomes inoperable, such 
that required actions will be based on the TS for the inoperable 
redundant components.
    Date of Issuance: December 17, 2001.
    Effective Date: As of the date of issuance and shall be implemented 
within 60 days of issuance.
    Amendment Nos.: 180 and 123.
    Facility Operating License Nos. DPR-67 and NPF-16: Amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: November 14, 2001 (66 
FR 57121).
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated December 17, 2001.
    No significant hazards consideration comments received: No.

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

    Date of application for amendments: October 23, 2000.
    Brief description of amendments: The amendments removed references 
to the containment hydrogen monitors in Technical Specification (TS) 
Tables 3.3-5, ``Accident Monitoring Instrumentation,'' and 4.3-4, 
``Accident Monitoring Instrumentation Surveillance Requirements.'' In 
addition, the amendments deleted TS 3/4.6.5, ``Combustible Gas 
Control--Hydrogen Monitors,'' and TS 3/4.6.6, ``Post Accident 
Containment Vent System.''
    Date of issuance: December 20, 2001.
    Effective date: As of the date of issuance and shall be implemented 
within 120 days of issuance.
    Amendment Nos: 217 and 211.
    Facility Operating License Nos. DPR-31 and DPR-41: Amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: January 10, 2001 (66 FR 
2014).
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated December 20, 2001.
    No significant hazards consideration comments received: No.

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

    Date of application for amendment: October 26, 2001.
    Brief description of amendment: Amendment changes the Operating 
License to extend certain Technical Specification surveillance 
requirement intervals on a one-time basis.
    Date of issuance: December 19, 2001.

[[Page 938]]

    Effective date: As of the date of issuance and shall be implemented 
within 60 days.
    Amendment No.: 206.
    Facility Operating License No. DPR-20. Amendment revised the 
Operating License.
    Date of initial notice in Federal Register: November 13, 2001 (66 
FR 56865).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated December 19, 2001.
    No significant hazards consideration comments received: No.

PPL Susquehanna, LLC, Docket No. 50-387, Susquehanna Steam Electric 
Station, Unit 1, Luzerne County, Pennsylvania

    Date of application for amendment: September 19, 2001, as 
supplemented October 26, 2001.
    Brief description of amendment: The amendment authorized a one-
cycle delay in removal of the second reactor pressure vessel material 
surveillance capsule.
    Date of issuance: December 20, 2001.
    Effective date: As of date of issuance.
    Amendment No.: 197.
    Facility Operating License No. NPF-14: The amendment authorized a 
change to the Reactor Vessel Material Surveillance Program.
    Date of initial notice in Federal Register: November 14, 2001 (66 
FR 57122).
    The October 26, 2001, letter provided clarifying information that 
did not change the initial proposed no significant hazards 
consideration determination.
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated December 20, 2001.
    No significant hazards consideration comments received: No.

PSEG Nuclear LLC, Docket Nos. 50-272, Salem Nuclear Generating Station, 
Unit No. 1, Salem County, New Jersey

    Date of application for amendments: December 10, 2001, as 
supplemented on December 21, and December 24, 2001.
    Brief description of amendments: The amendment allows a one-time 
change to Technical Specifications (TSs) Limiting Condition for 
Operation 3/4.7.4, ``Service Water System,'' to increase the Allowed 
Outage Time (AOT) from 72 hours to 10 days. The increase in the TS 3/
4.7.4 AOT is necessary in order to allow repairs to a portion of the 12 
Service Water System piping while remaining at power.
    Date of issuance: December 27, 2001.
    Effective date: As of the date of issuance.
    Amendment No.: 248.
    Facility Operating License Nos. DPR-70: The amendment revised the 
Technical Specifications.
    Public comments requested as to proposed no significant hazards 
consideration: No.
    The Commission's related evaluation of the amendment, finding of 
emergency circumstances, and final determination of no significant 
hazards consideration determination are contained in a Safety 
Evaluation dated December 27, 2001.
    No significant hazards consideration comments received: No.

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

    Date of application for amendments: January 5, 2001.
    Brief description of amendments: The amendments revise Technical 
Specification 3/4.6.4, ``Containment Systems, Combustible Gas Control, 
Hydrogen Analyzers,'' to reduce the channel calibration frequency of 
the Hydrogen Analyzers from quarterly to a frequency of once per 
refueling outage. The change also adds an additional surveillance 
requirement to perform a quarterly gas calibration.
    Date of issuance: December 17, 2001.
    Effective date: As of the date of issuance, and shall be 
implemented within 60 days.
    Amendment Nos.: 247 and 228.
    Facility Operating License Nos. DPR-70 and DPR-75: The amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: April 18, 2001 (66 FR 
20008).
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated December 17, 2001.
    No significant hazards consideration comments received: No.

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

    Date of application for amendment: October 1, 2001.
    Brief description of amendment: This amendment deletes Technical 
Specification 6.8.4.d requiring a program for post-accident sampling, 
and thereby eliminates the requirements to have and maintain the Post-
Accident Sampling System at Virgil C. Summer Nuclear Station, Unit No. 
1.
    Date of issuance: December 20, 2001.
    Effective date: December 20, 2001.
    Amendment No.: 152.
    Facility Operating License No. NPF-12: Amendment revises the 
Technical Specifications.
    Date of initial notice in Federal Register: October 31, 2001 (66 FR 
55023).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated December 20, 2001.
    No significant hazards consideration comments received: No.

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

    Date of amendments request: December 8, 2000.
    Brief description of amendments: The amendments delete or modify 
existing license conditions from the Unit 1 and Unit 2 Operating 
License, which have been completed or are otherwise no longer in 
effect. These activities have now been completed and the license 
conditions are either obsolete or are no longer needed.
    Date of issuance: December 7, 2001.
    Effective date: As of the date of issuance and shall be implemented 
within 30 days from the date of issuance.
    Amendment Nos.: 152 and 144.
    Facility Operating License Nos. NPF-2 and NPF-8: Amendments revise 
the Operating License.
    Date of initial notice in Federal Register: March 7, 2001 (66 FR 
13808).
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated December 7, 2001.
    No significant hazards consideration comments received: No.

Virginia Electric and Power Company, Docket Nos. 50-338 and 50-339, 
North Anna Power Station, Units 1 and 2, Louisa County, Virginia

    Date of application for amendment: September 10, 2001.
    Brief description of amendment: These amendments eliminate the 
Technical Specification requirements to have and maintain a post-
accident sampling system (PASS) at North Anna Power Station, Units 1 
and 2. In addition, for North Anna Power Station, Unit 2, the amendment 
deletes a license condition associated with the implementation of PASS.
    Date of issuance: December 19, 2001.
    Effective date: As of the date of issuance and shall be implemented 
within 120 days from the date of issuance.
    Amendment Nos.: 229 and 210.

[[Page 939]]

    Facility Operating License Nos. NPF-4 and NPF-7: Amendments change 
the Technical Specifications for both units and the license for Unit 2 
only.
    Date of initial notice in Federal Register: October 31, 2001 (66 FR 
55025).
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated December 19, 2001.
    No significant hazards consideration comments received: No.

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

    Date of application for amendments: September 10, 2001.
    Brief Description of amendments: These amendments eliminate the 
Technical Specification requirements to have and maintain a post-
accident sampling system at Surry Power Station, Unit Nos. 1 and 2.
    Date of issuance: December 18, 2001.
    Effective date: December 18, 2001.
    Amendment Nos.: 229 and 229.
    Facility Operating License Nos. DPR-32 and DPR-37: Amendments 
change the Technical Specifications.
    Date of initial notice in Federal Register: October 31, 2001 (66 FR 
55026).
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated December 18, 2001.
    No significant hazards consideration comments received: No.

    Dated at Rockville, Maryland, this 31st day of December 2001.

    For the Nuclear Regulatory Commission.
Stuart A. Richards,
Acting Director, Division of Licensing Project Management, Office of 
Nuclear Reactor Regulation.
[FR Doc. 02-301 Filed 1-7-02; 8:45 am]
BILLING CODE 7590-01-P