[Federal Register Volume 67, Number 63 (Tuesday, April 2, 2002)]
[Notices]
[Pages 15619-15636]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: 02-7799]


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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 March 8, 2002 through March 21, 2002. The 
last biweekly notice was published on March 19, 2002 (67 FR 12597).

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

[[Page 15620]]

margin of safety. The basis for this proposed determination for each 
amendment request is shown below.
    The Commission is seeking public comments on this proposed 
determination. Any comments received within 30 days after the date of 
publication of this notice will be considered in making any final 
determination.
    Normally, the Commission will not issue the amendment until the 
expiration of the 30-day notice period. However, should circumstances 
change during the notice period such that failure to act in a timely 
way would result, for example, in derating or shutdown of the facility, 
the Commission may issue the license amendment before the expiration of 
the 30-day notice period, provided that its final determination is that 
the amendment involves no significant hazards consideration. The final 
determination will consider all public and State comments received 
before action is taken. Should the Commission take this action, it will 
publish in the Federal Register a notice of issuance and provide for 
opportunity for a hearing after issuance. The Commission expects that 
the need to take this action will occur very infrequently.
    Written comments may be submitted by mail to the Chief, Rules and 
Directives Branch, Division of Administrative Services, Office of 
Administration, U.S. Nuclear Regulatory Commission, Washington, DC 
20555-0001, and should cite the publication date and page number of 
this Federal Register notice. Written comments may also be delivered to 
Room 6D22, Two White Flint North, 11545 Rockville Pike, Rockville, 
Maryland, from 7:30 a.m. to 4:15 p.m. Federal workdays. Copies of 
written comments received may be examined at the NRC's Public Document 
Room (PDR), 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 May 2, 2002, the licensee may file a request for a hearing with 
respect to issuance of the amendment to the subject facility operating 
license and any person whose interest may be affected by this 
proceeding and who wishes to participate as a party in the proceeding 
must file a written request for a hearing and a petition for leave to 
intervene. Requests for a hearing and a petition for leave to intervene 
shall be filed in accordance with the Commission's ``Rules of Practice 
for Domestic Licensing Proceedings'' in 10 CFR part 2. Interested 
persons should consult a current copy of 10 CFR 2.714, which is 
available at the NRC's PDR, located at One White Flint North, 11555 
Rockville Pike (first floor), Rockville, Maryland. Publicly available 
records will be accessible from the Agencywide Documents Access and 
Management Systems (ADAMS) Public Electronic Reading Room on the 
internet at the NRC web site, http://www.nrc.gov/reading-rm/doc-collections/cfr/. 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 Staff, or may be delivered to the Commission's PDR, 
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

[[Page 15621]]

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 PDR, located at One White Flint North, 11555 Rockville 
Pike (first floor), Rockville, Maryland. Publicly available records 
will be accessible from the Agencywide Documents Access and Management 
Systems (ADAMS) Public Electronic Reading Room on the internet at the 
NRC Web site, http://www.nrc.gov/reading-rm.html. If you do not have 
access to ADAMS or if there are problems in accessing the documents 
located in ADAMS, contact the NRC PDR Reference staff at 1-800-397-
4209, 304-415-4737 or by e-mail to [email protected].

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

    Date of amendment request: November 16, 2001.
    Description of amendment request: The proposed amendment would 
revise Technical Specification (TS) Section 3.6.2.2, ``Suppression Pool 
Water Level,'' and TS 3.6.2.4, ``Suppression Pool Makeup (SMPU) 
System'' to revise the allowable operating range for the Suppression 
Pool water level and the modes of applicability for the upper 
containment pools. The amendment would permit draining of the reactor 
cavity pool portion of the upper containment pool with unit in Mode 3, 
``Hot Shutdown,'' and reactor pressure less than 235 pounds per square 
inch gauge (psig). Draining of the upper containment pool is required 
as part of the refueling preparations and is currently not permissible 
in Mode 1, ``Power Operations,'' Mode 2, ``Startup,'' or Mode 3 by TS 
Section 3.6.2.4.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration which is presented below:

    Does the change involve a significant increase in the 
probability or consequences of an accident previously evaluated?
    The proposed changes revise the required water levels in the 
upper containment pools and suppression pool during Mode 3. The 
probability of an accident previously evaluated is unrelated to the 
water levels in the pools since they are mitigative systems. The 
operation or failure of a mitigative system does not contribute to 
the occurrence of an accident. No active or passive failure 
mechanisms that could lead to an accident are affected by these 
proposed changes.
    The consequences of a previously evaluated accident are not 
significantly increased. The changes have no impact on the ability 
of any of the Emergency Core Cooling Systems (ECCS) to function 
adequately, since adequate net positive suction head (NPSH) is 
provided with reduced water volumes. The post-accident containment 
temperature is not significantly affected by the proposed reduction 
in total heat sink volume. The increase in suppression pool water 
level to compensate for the reduction in upper containment pool 
volume will provide reasonable assurance that the minimum post-
accident vent coverage is adequate to assure the pressure 
suppression function of the suppression pool is accomplished. The 
suppression pool water will be raised only after the reactor 
pressure has been reduced sufficiently to assure that the 
hydrodynamic loads from a loss of coolant accident will not exceed 
the design values. The reduced reactor pressure will also ensure 
that the loads due to main steam safety relief valve actuation with 
an elevated pool level are within the design loads. The change in 
exposure rate expected due to draining the upper containment pool in 
Mode 3 is small (i.e., by approximately two orders of magnitude) 
compared to the measured exposure rates in the reactor cavity during 
refueling preparations. Therefore, these changes do not have an 
adverse impact on the ability to maintain refueling exposure rates 
as low as reasonably achievable.
    Therefore, the proposed changes do not significantly increase 
the consequences of an accident previously evaluated.
    Does the change create the possibility of a new or different 
kind of an accident from any accident previously evaluated?
    The proposed changes to the water level requirements for the 
upper containment pool and the suppression pool do not involve the 
use or installation of new equipment. Installed equipment is not 
operated in a new or different manner. No new or different system 
interactions are created, and no new processes are introduced. The 
increased suppression pool water level does not increase the 
probability of flooding in the drywell. No new failures have been 
created by the change in the water level requirements.
    Therefore, the proposed changes do not create the possibility of 
a new or different kind of accident from any accident previously 
evaluated.
    Does the change involve a significant reduction in a margin of 
safety?
    The proposed changes to the upper containment pool and 
suppression pool water levels do not introduce any new setpoints at 
which protective or mitigative actions are initiated. No current 
setpoints are altered by this change. The design and functioning of 
the containment pressure suppression system is unchanged. The 
proposed total water volume is sufficient to provide high confidence 
that the pressure suppression and containment systems will be 
capable of mitigating large and small break accidents. All analyzed 
transient results remain well within the design values for the 
structures and equipment. Therefore, the proposed changes do not 
involve a significant reduction in a margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: Robert Helfrich, Mid-West Regional Operating 
Group, Exelon Generation Company, LLC, 4300 Winfield Road, Warrenville, 
IL 60555.
    NRC Section Chief: Anthony J. Mendiola.

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

    Date of amendment requests: March 1, 2002.
    Description of amendment requests: 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 Nuclear Regulatory Commission (NRC) staff issued a notice of 
opportunity for comment in the Federal Register on June 14, 2001 (66 FR 
32400), on possible amendments concerning missed surveillances, 
including a model safety evaluation and model no significant hazards 
consideration (NSHC) determination, using the consolidated line item 
improvement process. The NRC staff subsequently issued a notice of 
availability of the models for referencing in license amendment 
applications in the Federal Register on September 28, 2001 (66 FR 
49714).
    The licensee affirmed the applicability of the following NSHC 
determination in its request for amendments dated March 1, 2002.
    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:


[[Page 15622]]



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 Accident Previously 
Evaluated.

    The proposed change does not involve a physical alteration of 
the plant (no new or different type of equipment will be installed) 
or 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 request for amendments 
involves no significant hazards consideration.
    Attorney for licensee: Nancy C. Loftin, Esq., Corporate Secretary 
and Counsel, Arizona Public Service Company, P.O. Box 53999, Mail 
Station 9068, Phoenix, Arizona 85072-3999.
    NRC Section Chief: Stephen Dembek.

Entergy Operations, Inc., System Energy Resources, Inc., South 
Mississippi Electric Power Association, and Entergy Mississippi, Inc., 
Docket No. 50-416, Grand Gulf Nuclear Station, Unit 1, Claiborne 
County, Mississippi

    Date of amendment request: January 31, 2002.
    Description of amendment request: Entergy Operations, Inc. is 
proposing that the Grand Gulf Nuclear Station, Unit 1, Operating 
License be amended to reflect a 1.7 percent increase in the licensed 
100 percent reactor core thermal power level (an increase in reactor 
power level from 3,833 megawatts thermal to 3,898 megawatts thermal). 
These changes result from increased accuracy of the feedwater flow and 
temperature measurements to be achieved by utilizing high accuracy 
ultrasonic flow measurement instrumentation. The basis for this change 
is consistent with the revision, issued in June 2000, to appendix K to 
part 50 of title 10 of the Code of Federal Regulations, allowing 
operating reactor licensees to use an uncertainty factor of less than 2 
percent of rated reactor thermal power in analyses of postulated design 
basis loss-of-coolant accidents.
    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.
    The comprehensive analytical efforts performed to support the 
proposed change included a review of the Nuclear Steam Supply System 
(NSSS) systems and components that could be affected by this change. 
All systems and components will function as designed, and the 
applicable performance requirements have been evaluated and found to 
be acceptable.
    The comprehensive analytical efforts performed to support the 
proposed uprate conditions included a review and evaluation of all 
components and systems that could be affected by this change. 
Evaluation of accident analyses confirmed the effects of the 
proposed uprate are bounded by the current dose analyses. All 
systems will function as designed, and all performance requirements 
for these systems have been evaluated and found acceptable. Because 
the integrity of the plant will not be affected by operation at the 
uprated condition, it is concluded that all structures, systems, and 
components required to mitigate a transient remain capable of 
fulfilling their intended functions. The reduced uncertainty in the 
flow input to the power calorimetric measurement allows the current 
safety analyses to be used, with small changes to the core operating 
limits, to support operation at a core power of 3,898 megawatts 
thermal (MWt). As such, all Final Safety Analysis Report (FSAR) 
Chapter 15 accident analyses continue to demonstrate compliance with 
the relevant event acceptance criteria. Those analyses performed to 
assess the effects of mass and energy releases remain valid. The 
source terms used to assess radiological consequences have been 
reviewed and determined to either bound operation at the 1.7 percent 
uprated condition, or new analyses were performed to verify all 
acceptance criteria continue to be met.
    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.
    No new accident scenarios, failure mechanisms, or limiting 
single failures are introduced as a result of the proposed changes. 
All systems, structures, and components previously required for the 
mitigation of a transient remain capable of fulfilling their 
intended design functions. The proposed changes have no adverse 
effects on any safety-related system or component and do not 
challenge the performance or integrity of any safety related system.
    Therefore, the proposed change does not create the possibility 
of a new or different kind of accident from any previously 
evaluated.
    3. Does the proposed change involve a significant reduction in a 
margin of safety?
    Response: No.
    Operation at the uprated power condition does not involve a 
significant reduction in a margin of safety. Analyses of the primary 
fission product barriers have concluded that all relevant design 
criteria remain satisfied,

[[Page 15623]]

both from the standpoint of the integrity of the primary fission 
product barrier and from the standpoint of compliance with the 
required acceptance criteria.
    Therefore, 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: Nicholas S. Reynolds, Esquire, Winston and 
Strawn, 1400 L Street, NW., 12th Floor, Washington, DC 20005-3502.
    NRC Section Chief: Robert A. Gramm.

Entergy Operations, Inc., System Energy Resources, Inc., South 
Mississippi Electric Power Association, and Entergy Mississippi, Inc., 
Docket No. 50-416, Grand Gulf Nuclear Station, Unit 1, Claiborne 
County, Mississippi

    Date of amendment request: January 31, 2002.
    Description of amendment request: Entergy Operations, Inc. requests 
an amendment for the Grand Gulf Nuclear Station, Unit 1, Technical 
Specifications to extend the allowed out-of-service time from 72 hours 
to 14 days for a Division 1 or Division 2 Emergency Diesel Generator 
(DG) during reactor operational modes 1, 2, or 3. The proposed changes 
are intended to provide flexibility in performance of corrective and 
preventive maintenance on the DGs during power operation.
    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.
    The proposed Technical Specification (TS) changes do not affect 
the design, operational characteristics, function, or reliability of 
the DGs. The DGs are not the initiators of previously evaluated 
accidents. The DGs are designed to mitigate the consequences of 
previously evaluated accidents including a loss of offsite power. 
Extending the allowed outage time (AOT) for a single DG would not 
significantly affect the previously evaluated accidents since the 
remaining DGs supporting the redundant ESF systems would continue to 
perform the accident mitigating functions as designed.
    The duration of a TS AOT is determined considering that there is 
a minimal possibility that an accident will occur while a component 
is removed from service. A risk-informed assessment was performed 
which concluded that the increase in plant risk is small and 
consistent with the USNRC [U.S. Nuclear Regulatory Commission] 
``Safety Goals for the Operations of Nuclear Power Plants; Policy 
Statement,'' Federal Register, Vol. 51, p. 30028 (51 FR 30028), 
August 4, 1986, as further described by NRC [Nuclear Regulatory 
Commission] Regulatory Guide 1.177.
    The current TS requirements establish controls to ensure that 
redundant systems relying on the remaining DGs are Operable. In 
addition to these requirements, administrative controls will be 
established to provide assurance that the AOT extension is not 
applied during adverse weather conditions that could potentially 
affect offsite power availability. Administrative controls are also 
implemented to avoid or minimize risk-significant plant 
configurations during the time when a DG is removed from service.
    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.
    The proposed TS changes do not involve a change in the design, 
configuration, or method of operation of the plant that could create 
the possibility of a new or different kind of accident. The proposed 
change extends the AOT currently allowed by the TS.
    Therefore, the proposed change does not create the possibility 
of a new or different kind of accident from any previously 
evaluated.
    3. Does the proposed change involve a significant reduction in a 
margin of safety?
    Response: No.
    The Engineered Safety Feature (ESF) systems required to mitigate 
the consequences of postulated accidents consist of three 
independent divisions. The ESF systems of any two of the three 
divisions provide for the minimum safety functions necessary to shut 
down the unit and maintain it in a safe shutdown condition. Each of 
the three independent ESF divisions can be powered from one of the 
offsite power sources or its associated on-site DG. This design 
provides adequate defense-in-depth to ensure that the ESF equipment 
needed to mitigate the consequences of an accident will have diverse 
power sources available to accomplish the required safety functions. 
Thus, with one DG out of service, there are sufficient means to 
accomplish the safety functions and prevent the release of 
radioactive material in the event of an accident.
    The proposed AOT change does not affect any of the assumptions 
or inputs to the safety analyses of the FSAR and does not erode the 
decrease in severe accident risk achieved with the issuance of the 
Station Blackout (SBO) Rule, 10 CFR 50.63 ``Loss of All Alternating 
Current Power.''
    The proposed extended AOT deviates from the recommended 72 hour 
AOT of Regulatory Guide (RG) 1.93. However, an extension of the 72 
hour AOT to 14 days has been demonstrated to be acceptable based on 
deterministic and risk-informed analyses. The proposed changes are 
not in conflict with any other approved codes or standards 
applicable to the onsite AC [Alternating Current] power sources.
    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: Nicholas S. Reynolds, Esquire, Winston and 
Strawn, 1400 L Street, NW., 12th Floor, Washington, DC 20005-3502.
    NRC Section Chief: Robert A. Gramm.

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

    Date of amendment request: February 8, 2002.
    Description of amendment request: The proposed amendment would 
replace referenced control requirements for access to high radiation 
areas with the actual requirements of 10 CFR part 20. The referenced 
document in Technical Specifications Section 6.11 would no longer 
exist.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Does the change involve a significant increase in the 
probability or consequences of an accident previously evaluated?
    The proposed changes replace referenced control requirements 
affecting access to high radiation areas with the actual 
requirements. This proposed change does not involve any changes to 
system or equipment configuration. The reliability of systems and 
components relied upon to prevent or mitigate the consequences of 
accidents previously evaluated is not affected by the proposed 
changes. Therefore, these changes 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?
    The proposed changes are administrative in nature and do not 
involve a change to the plant design or operation. No new or 
different types of equipment will be installed as a result of this 
change. The proposed change is administrative in nature and replaces 
referenced control requirements for

[[Page 15624]]

access to high radiation areas with the actual requirements. No new 
accident modes or equipment failure modes are created by these 
changes. Therefore, these proposed changes do not create the 
possibility of a new or different kind of accident from any accident 
previously evaluated.
    3. Does the change involve a significant reduction in a margin 
of safety?
    The proposed change does not impact or have a direct effect on 
any safety analysis assumptions. The proposed change is 
administrative in nature and replaces referenced control 
requirements for access to high radiation areas with the actual 
requirements.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: Ernest L. Blake, Jr., Esquire, Shaw, 
Pittman, Potts & Trowbridge, 2300 N Street, NW., Washington, DC 20037.
    NRC Section Chief: Robert A. Gramm.

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

    Date of amendment requests: January 14, 2002.
    Description of amendment requests: The proposed amendments would 
add an allowable plus or minus () 1 percent (%) as-left 
setpoint tolerance for the pressurizer code safety valves to Unit 1 and 
Unit 2 technical specification (TS) 3.4.2 and TS 3.4.3. In addition, 
the proposed amendments would revise Unit 2 TS 3.4.2 and TS 3.4.3 to 
increase the allowable as-found setpoint tolerance for the Unit 2 
pressurizer code safety valves from  1 % to  
3%.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Does the proposed change involve a significant increase in 
the probability of occurrence or consequences of an accident 
previously evaluated?
    Response: No.
    Probability of Occurrence of an Accident Previously Evaluated--
    The proposed changes to pressurizer code safety valve as-found 
and as-left setpoint tolerance do not affect any accident initiators 
or precursors. There are no new failure modes for the pressurizer 
code safety valves created by this change in setpoint tolerance. No 
adverse interactions with the RCS are created by this change in 
setpoint tolerance. The lowest possible setpoint of any of the 
pressurizer code safety valves (including the  3% 
tolerance) is higher than the highest RCS pressures anticipated 
during shutdown, startup, normal operating, and anticipated 
operational occurrence conditions. The lowest possible pressurizer 
code safety valve setpoint is also higher than the setpoint of the 
PORVs. Therefore, there would not be an adverse interaction between 
the pressurizer code safety valves and the PORVs. Thus, the 
probability of occurrence of an accident previously evaluated is not 
significantly increased.
    The format changes for the Unit 2 TS 3.4.3 page do not impact 
any accident initiators or precursors. Thus, the probability of 
occurrence of an accident previously evaluated is not significantly 
increased.
    Consequences of an Accident Previously Evaluated--
    The proposed change to add an allowable as-left setpoint 
tolerance for the Unit 1 and 2 pressurizer code safety valves does 
not adversely affect any of the accident and safety analyses. In 
addition, the proposed increase in the Unit 2 as-found pressurizer 
code safety valve setpoint tolerance does not adversely affect any 
of the accident and safety analyses. Both the as-left setpoint of 
 1% and the as-found setpoint of  3% of the 
nominal lift pressure of 2485 psig provides reasonable assurance 
that the pressurizer code safety valves are capable of performing 
their design function as assumed in the accident and safety 
analyses. Even at the highest allowable lift pressure, the 
pressurizer code safety valves, in conjunction with the RPS, remain 
capable of limiting the RCS pressure within the Safety Limit of 110% 
of design pressure (or 2735 psig). Thus, there will be no increase 
in offsite doses and the consequences of an accident previously 
analyzed are not increased.
    The format changes for the Unit 2 TS 3.4.3 page do not impact 
the pressurizer code safety valve's function. Thus, there will be no 
increase in offsite doses, and the consequences of an accident 
previously analyzed are not increased.
    2. Does the proposed change create the possibility of a new or 
different kind of accident from any accident previously evaluated?
    Response: No.
    The proposed changes to pressurizer code safety valve as-found 
and as-left setpoint tolerance do not create any new or different 
accident initiators or precursors. There are no new failure modes 
for the pressurizer code safety valves created by this change in 
setpoint tolerance. No adverse interactions with the RCS are created 
by this change in setpoint tolerance. The lowest possible setpoint 
of any of the pressurizer code safety valves (including the 
 3% tolerance) is higher than the highest RCS pressures 
anticipated during shutdown, startup, normal operating, and 
anticipated operational occurrence conditions. The lowest possible 
pressurizer code safety valve setpoint is also higher than the 
setpoint of the PORVs. Therefore, there would not be an adverse 
interaction between the pressurizer code safety valves and the 
PORVs. Thus, the proposed changes do not create the possibility of a 
new or different kind of accident from any previously evaluated.
    The format changes for the Unit 2 TS 3.4.3 page do not create 
any new or different accident initiators or precursors. Thus, the 
possibility of a new or different kind of accident from any 
previously evaluated is not created.
    3. Does the proposed change involve a significant reduction in a 
margin of safety?
    Response: No.
    The proposed changes do not impact pressurizer code safety valve 
capability to perform the design function required by the accident 
and safety analyses, nor do the proposed changes impact the 
operational characteristics of the pressurizer code safety valves. 
The pressurizer code safety valves, in conjunction with the RPS, 
ensure that the RCS Safety Limit of 110% of design pressure (or 2735 
psig) is not exceeded for any analyzed event. Therefore, the 
proposed changes do not involve a significant reduction in margin of 
safety.
    The format changes for the Unit 2 TS 3.4.3 page do not involve a 
significant reduction in a margin of safety.

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

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

    Date of amendment request: January 21, 2002.
    Description of amendment request: The proposed revised Technical 
Specification (TS) Requirement will modify TS Surveillance Requirement 
(SR) 3.7.3.1 to improve consistency with Cooper Nuclear Station (CNS) 
License Amendment No. 185, approved on March 13, 2001, and eliminate 
unnecessary restrictions regarding how the Reactor Equipment Cooling 
(REC) System surge tank level is monitored.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Does the change involve a significant increase in the 
probability or consequences of an accident previously evaluated?
    This change eliminates the specific details regarding performing 
the SR 3.7.3.1 verification of Reactor Equipment Cooling (REC) surge 
tank level. This change will not result in a significant increase in 
the probability of an accident previously

[[Page 15625]]

evaluated because the method of verifications of REC surge tank 
level has no effect on the initiators of any analyzed events.
    The method of performing the surveillance on REC surge tank 
level does not affect the performance of the minimum equipment 
credited in the mitigation of any analyzed event. As a result, no 
analysis assumptions or mitigative functions are impacted. 
Therefore, this change will not result in 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.
    The proposed change does not involve a physical alteration of 
the plant. No new equipment is being introduced, and installed 
equipment is not being operated in a new or different manner. There 
is no change being made to the parameters within which the plant is 
operated. There are no setpoints, at which protective or mitigative 
actions are initiated, affected by this change. This change will not 
alter the manner in which equipment operation is initiated, nor will 
the function demands on credited equipment be changed. No alteration 
in the procedures which ensure the plant remains within analyzed 
limits is being proposed, and no change is being made to the 
procedures relied upon to an off-normal event. As such, no new 
failure modes are being introduced. The change does not alter 
assumptions made in the safety analysis and licensing basis. 
Therefore, the change does not create the possibility of a new or 
different kind of accident from any accident previously evaluated.
    3. Does this change involve a significant reduction in a margin 
of safety?
    The margin of safety is established through equipment design, 
operating parameters, and the setpoints at which automatic actions 
are initiated. Credited equipment remains available to actuate upon 
demand for the purpose of mitigating an analyzed event. The proposed 
change is acceptable because the operability of the REC System is 
unaffected, there is no detrimental impact on any equipment design 
parameter, and the plant will still be required to operate within 
assumed conditions. The normal procedural controls on methods of 
surveillance performance provide adequate assurance that the REC 
System will be capable of performing its intended safety function. 
Detailing the performance method within the TSs does not impact the 
margin of safety (which is more closely related to tank volume than 
the method of verifying volume). Therefore, the change does not 
involve a significant reduction in the margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: Mr. John R. McPhail, Nebraska Public Power 
District, Post Office Box 499, Columbus, NE 68602-0499.
    NRC Section Chief: Robert A. Gramm.

Nuclear Management Company, LLC, Docket No. 50-331, Duane Arnold Energy 
Center, Linn County, Iowa

    Date of amendment request: February 8, 2002.
    Description of amendment request: The proposed amendment would 
revise the Technical Specifications (TS) to change TS Section 5.0, 
Administrative Controls, to adopt TSTF-258 Revision 4. Revisions to the 
TS are proposed to Section 5.2.2, Unit Staff, to delete details of 
staffing requirements and delete requirements for the Shift Technical 
Advisor (STA) as a separate position while retaining the function. 
Section 5.5.4, Radioactive Effluent Controls Program, would be revised 
to be consistent with the intent of 10 CFR part 20. Section 5.6.4, 
Monthly Operating Reports, would be revised by deleting periodic 
reporting requirements for main steam safety/relief valve challenges to 
be consistent with Generic Letter 97-02. Section 5.7, High Radiation 
Area, would be revised in accordance with 10 CFR 20.1601(c).
    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 amendment will not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    This request for amendment to Duane Arnold Energy Center's TS 
provides for adoption of the NRC-approved generic change TSTF item 
TSTF-258, Revision 4. The Amendment request includes revisions to TS 
Section 5.0, ``Administrative Controls,'' to delete details of 
staffing requirements, delete requirements for the STA as a separate 
position while retaining the function, revise the Radioactive 
Effluent Controls Program to be consistent with the intent of 10 CFR 
20, delete periodic reporting requirements of challenges to main 
steam safety/relief valves, and revise radiological control 
requirements for radiation areas to be consistent with those 
specified in 10 CFR 20.1601(c).
    The proposed TS changes are administrative in nature and do not 
impact the operation, physical configuration, or function of plant 
equipment or systems. The changes do not impact the initiators or 
assumptions of analyzed events, nor do they impact mitigation of 
accidents or transient events. Therefore, these proposed changes do 
not involve a significant increase in the probability or 
consequences of an accident previously evaluated.
    2. The proposed amendment will not create the possibility of a 
new or different kind of accident from any accident previously 
evaluated.
    The proposed TS changes are administrative in nature and do not 
alter plant configuration, require that new equipment be installed, 
alter assumptions made about accidents previously evaluated or 
impact the operation or function of plant equipment or systems. The 
proposed changes do not introduce any new modes of plant operation 
or make any changes to system setpoints. The proposed changes do not 
create the possibility of a new or different kind of accident due to 
credible new failure mechanisms, malfunctions, or accident 
initiators not considered in the design and licensing bases. 
Therefore, the possibility of a new or different kind of accident 
from any accident previously evaluated has not been created.
    3. The proposed amendment will not involve a significant 
reduction in a margin of safety.
    The proposed TS changes are administrative in nature and do not 
involve physical changes to plant structures, systems, or components 
(SSCs), or the manner in which these SSCs are operated, maintained, 
modified, tested, or inspected. The proposed changes do not involve 
a change to any safety limits, limiting safety system settings, 
limiting conditions for operation, or design parameters for any SSC. 
The proposed changes do not impact any safety analysis assumptions 
and do not involve a change in initial conditions, system response 
times, or other parameters affecting any accident analysis. 
Regarding the deletion of the requirement for the STA as a separate 
position, the function will be retained, so there will be no 
reduction in the margin of safety. As a result, there is no 
significant reduction in a margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: Mr. Alvin Gutterman, Morgan Lewis, 1111 
Pennsylvania Avenue NW., Washington, DC 20004.
    NRC Section Chief: William D. Reckley, Acting Section Chief.

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

    Date of amendment request: February 12, 2002.
    Description of amendment request: The proposed amendment would 
revise Surveillance Requirement (SR) 4.0.E to extend the delay period 
before entering a limiting condition for operation following a missed 
surveillance. The delay period would be extended from the current limit 
of ``* * * up to 24 hours or up to the limit of the time interval, 
whichever is less'' to ``* * *

[[Page 15626]]

up to 24 hours or up to the limit of the time interval, whichever is 
greater.'' In addition, the following requirement would be added to SR 
4.0.E: ``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 February 12, 2002.
    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.

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

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

    Date of amendment request: November 15, 2001, as supplemented by 
letter dated January 31, 2002.
    Description of amendment request: The proposed amendment request 
modifies License Condition 2.C(10) associated with loading and 
contingency unloading of spent fuel casks in the fuel building due to 
changes in the dry storage system design.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. The requested license amendment does not involve a 
significant increase in the probability or consequences of an 
accident previously evaluated.
    Accidents previously evaluated are those addressed in the Trojan 
Nuclear Plant (TNP) Defueled Safety Analysis Report (DSAR), the TNP 
Decommissioning Plan and License Termination Plan (``Decommissioning 
Plan''), and LCA [license change application] 237, Revision 3, and 
LCA 246, Revision 0. [Since their approval via Amendments 199 and 
200 to the TNP License on April 23, 1999, Revision 3 of LCA 237 and 
Revision 0 of LCA 246, have undergone revision per 10 CFR 50.59, as 
allowed by TNP License Condition 2.C(10). The current revisions are 
LCA 237, Revision 4, and LCA 246, Revision 1.] The basis for the 
conclusion that the probability or consequences of an accident 
previously evaluated in the DSAR or Decommissioning Plan is not 
significantly increased is not materially changed from the 
significant hazards consideration determination provided in the 
current LCA 237, Revision 4, and LCA 246, Revision 1. Loading and 
contingency unloading of the MPC [multi-purpose canister] as 
described in the proposed Revision 5 of LCA 237 and Revision 2 of 
LCA 246 consist of activities that are functionally the same as 
those for loading and contingency unloading a PWR [pressurized water 
reactor] Basket under the previous Trojan Storage System design. 
With the original Transfer Cask, PWR Basket, and its shield and 
structural lids and associated welds replaced under the new design 
by the Holtec Transfer Cask, MPC, and its MPC redundant closures 
(i.e., lid, vent and drain port cover plates, closure ring, and 
associated welds), respectively, these and associated Trojan Storage 
System design changes do not significantly impact the activities 
that will be conducted during ISFSI [independent spent fuel storage 
installation] loading/unloading. Furthermore, the safety evaluations 
in the proposed Revision 5 of LCA 237 and Revision 2 of LCA 246 show 
that the Trojan ISFSI design changes do not significantly impact the 
potential for or consequences of off-normal events or accidents 
during ISFSI loading and contingency unloading. Thus, the 
probability or consequences of an accident previously evaluated in 
the DSAR or Decommissioning Plan is not significantly increased.
    The postulated events previously evaluated in Revision 3 of LCA 
237 and Revision 0 of LCA 246 include drops, tipovers, mishandling, 
operational errors, and support system malfunctions that could 
potentially

[[Page 15627]]

occur during loading and contingency unloading operations.
    As discussed in proposed Revision 5 to LCA 237 and Revision 2 to 
LCA 246, the Trojan Storage System design changes do not 
significantly affect the conclusions with respect to the potential 
for or consequences of a Transfer Cask and/or MPC drop, tipover, or 
mishandling event. The design safety factors, load testing 
requirements, and administrative controls (i.e., procedures, 
training, maintenance, and inspections) for the fuel handling 
equipment are materially unaffected by the Trojan Storage System 
design changes, such that there is no significant increase in 
probability of a Transfer Cask and/or MPC drop, tipover, or 
mishandling event. As described in the safety evaluation in proposed 
Revision 5 to LCA 237 and Revision 2 to LCA 246, the calculated 
consequence of a Transfer Cask drop, tipover, or mishandling event 
prior to the MPC lid being welded to the MPC is approximately 0.003 
rem whole body dose at the site boundary, which is the same as was 
calculated for these events in LCA 237, Revision 3. This calculated 
consequence, which is well below the EPA PAG [Environmental 
Protection Agency protective action guide] of 1 rem whole body dose 
for the early phase of an event, has accumulated additional 
conservatism since the submittal and NRC approval of LCA 237, 
Revision 3, applicable to loading the PWR Basket. The additional 
conservatism is the result of the calculation assumption that five 
years have elapsed for cooling of the fuel, combined with the fact 
that approximately five additional years have passed since this 
event was originally analyzed for LCA 237, Revision 3, during which 
additional cooling of the TNP spent nuclear fuel has occurred. Thus, 
there is no significant increase in consequences of a Transfer Cask 
drop, tipover, or mishandling event.
    The Trojan Storage System design changes also do not 
significantly increase the probability or consequences of 
operational errors and/or support system malfunctions that could 
potentially occur during loading/unloading operations. As discussed 
in the safety evaluation in proposed Revision 5 to LCA 237 and 
Revision 2 to LCA 246, the changes to pressures associated with the 
ISFSI confinement boundary do not impact the conclusion that the 
postulated inadvertent over-pressurization of the MPC during 
draining and/or drying operations is not considered credible, since 
multiple equipment failures and a procedural error are still 
required in order for the event to occur. With the revised design 
decay heat load as summarized above, the longer time period required 
for boiling to occur in the MPC further reduces the potential for a 
postulated over-pressurization event.
    As shown in proposed Revision 5 of LCA 237 and Revision 2 of LCA 
246, the higher operating pressures during loading operations (e.g., 
pressure testing and MPC blowdown and backfill pressures) and the 
redesign of several of the systems involved in MPC closure 
operations (e.g., vacuum drying, blowdown system, and helium 
recirculation cooling), do not significantly impact the probability 
or consequences of equipment failures. The maximum normal design 
pressure ratings of the MPC, vacuum drying system, helium 
recirculation system, and helium backfill system, including their 
associated pressurized lines and system components, are such that 
the operating pressure increase does not significantly increase the 
probability of a passive failure of a pressurized line on the MPC. 
However, because of the increased operating and test pressures 
associated with the Holtec-designed MPC as compared to the PWR 
Basket, the consequence of a bounding scenario involving the passive 
failure of a pressurized line is increased. However, this increase 
is not considered to be significant since, as detailed in Section 
5.2.5.2.2 of proposed Revision 5 to LCA 237 and Revision 2 to LCA 
246, the dose consequence remains well below the EPA PAG of 1 rem 
whole body for the early phase of an event.
    Based on the above, the impacts of the Trojan Storage System 
design changes on cask loading/unloading operations would not 
significantly increase the probability or consequences of any 
accident previously evaluated.
    2. The requested license amendment does not create the 
possibility of a new or different kind of accident from any accident 
previously evaluated.
    The aforementioned design changes for the Trojan Storage System 
do not create the possibility of a new or different kind of accident 
from any accident previously evaluated, including those evaluated in 
Revision 3 of LCA 237 and Revision 0 of LCA 246 approved by the NRC 
on April 23, 1999. With the original Transfer Cask, PWR Basket, and 
its shield and structural lids and associated welds replaced under 
the new design by the Holtec Transfer Cask, MPC, and its MPC 
redundant closures (i.e., lid, vent and drain port cover plates, 
closure ring, and associated welds), respectively, these and 
associated Trojan Storage System design changes do not significantly 
impact the functional activities that will be conducted during ISFSI 
loading/unloading. Thus, the loading procedure and system design 
changes do not introduce any new types of accidents not previously 
analyzed in Revision 3 of LCA 237 and Revision 0 of LCA 246.
    3. The requested license amendment does not involve a 
significant reduction in the margin of safety.
    The basis for the conclusion that a significant reduction in the 
margin of safety is not involved is not materially changed from the 
significant hazards consideration determination provided in the 
current LCA 237, Revision 4, and LCA 246, Revision 1. Specifically, 
the TNP Permanently Defueled Technical Specifications (PDTS) contain 
four limiting conditions of operation that address: (1) Spent Fuel 
Pool water level, (2) Spent Fuel Pool boron concentration, (3) Spent 
Fuel Pool temperature, and (4) Spent Fuel Pool load restrictions. 
These Technical Specifications will remain in effect as long as 
spent fuel is stored in the Spent Fuel Pool, which is in accordance 
with their applicability statements. As discussed below, the Trojan 
Storage System design changes and their impact on ISFSI loading/
unloading activities will not affect the PDTS or their bases.
    Loading and contingency unloading of the MPC as described in the 
proposed Revision 5 of LCA 237 and Revision 2 of LCA 246 consist of 
activities that are functionally the same as those for loading and 
contingency unloading a PWR Basket under the previous Trojan Storage 
System design. The Cask Loading Pit, where spent fuel will be loaded 
into the MPC, is immediately adjacent to the Spent Fuel Pool. The 
gate between the Cask Loading Pit and Spent Fuel Pool will be opened 
to allow spent fuel assemblies to be moved from the spent fuel 
storage racks in the Spent Fuel Pool to the MPC in the Cask Loading 
Pit. Opening the gate will allow free exchange of the water between 
the Cask Loading Pit and the Spent Fuel Pool. The water in the Cask 
Loading Pit must be at essentially the same level, boron 
concentration, and temperature as the Spent Fuel Pool prior to the 
first opening of the gate to ensure that the limiting conditions of 
operation are continuously satisfied for the Spent Fuel Pool. 
Therefore, the Cask Loading Pit will be filled, to about the same 
level as the Spent Fuel Pool, with water that is about the same 
boron concentration and temperature as the Spent Fuel Pool. With 
these precautions, the limiting conditions of operation pertaining 
to Spent Fuel Pool level, boron concentration, and temperature will 
be continuously maintained for the Spent Fuel Pool and the margin of 
safety will be unaffected. Except for small changes to accommodate 
lid lift rigging, the level in the Cask Loading Pit will not be 
reduced until the MPC lid has been placed on the loaded MPC. This 
configuration is consistent with the objective of keeping the 
radiological exposure to personnel as low as reasonably achievable 
(ALARA). The contingency unloading sequence is essentially the 
reverse of the loading sequence. Thus, the loading and contingency 
unloading processes for the MPC with the Trojan Storage System 
design changes incorporated do not involve a significant reduction 
in the margin of safety.
    As with the previous design, the Trojan Storage System design 
changes will be implemented such that when lifting and moving heavy 
loads, loads that will be carried over fuel in the Spent Fuel Pool 
racks and the heights at which they may be carried will be limited 
in such a way as to preclude impact energies, in the unlikely event 
of a drop, from exceeding 240,000 in-lbs in accordance with Limiting 
Condition for Operation (LCO) 3.1.4, ``Spent Fuel Pool Load 
Restrictions.'' With this precaution, the LCO pertaining to load 
restrictions over the Spent Fuel Pool will be satisfied for fuel 
stored in the Spent Fuel Pool racks and the margin of safety will be 
unaffected. The safe load path for heavy loads being lifted and 
moved outside the Spent Fuel Pool will be located sufficiently far 
from the Spent Fuel Pool as to not have an adverse effect on the 
Spent Fuel Pool in the unlikely event of a load drop. In addition, 
the Trojan Storage System design changes do not affect the 
implementation of mechanical stops and electrical interlocks on the 
Fuel Building overhead crane that provide additional assurance that 
heavy loads are not carried

[[Page 15628]]

over the fuel in the Spent Fuel Pool racks. Thus, the Trojan Storage 
System design changes and their impact on ISFSI loading and 
contingency unloading activities do not involve a significant 
reduction in the margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: Douglas R. Nichols, Esq., Portland General 
Electric Company, 121 S.W. Salmon Street, Portland, Oregon 97204.
    NRC Section Chief: Robert A. Gramm.

Southern Nuclear Operating Company, Inc., et al., Docket Nos. 50-424 
and 50-425, Vogtle Electric Generating Plant, Units 1 and 2, Burke 
County, Georgia

    Date of amendment request: October 30, 2001, as supplemented by 
letter dated February 11, 2002.
    Description of amendment request: The proposed amendments would 
revise Technical Specifications Table 3.3.1-1, ``Reactor Trip System 
Instrumentation'' and the associated Bases B 3.3.1. A limit or 
``clamp'' on the Over Temperature Delta Temperature (OTDT) reactor trip 
function is proposed to address design issues related to fuel rod 
design under transient conditions. In addition, editorial revisions to 
Bases B 3.3.1 are included.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Does the proposed change involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    The proposed clamp on the OTDT reactor trip function is not 
credited in the safety analyses. Implementation of the limit or 
``clamp'' on the OTDT reactor trip function, along with the 
corresponding changes to the AFD [axial flux difference] modifier 
f1 (AFD) and RAOC [relaxed axial offset control] band, 
will ensure the prevention of stress failure of the fuel rod 
cladding for Condition I and II reactor coolant system cooldown 
events. This demonstrates continued compliance with 10 CFR 50, 
Appendix A, Criterion 10, i.e., that the specified acceptable fuel 
design limits are not exceeded.
    There is no change in the radiological consequences of any 
accident since the fuel clad, the reactor coolant system pressure 
boundary, and the containment are not changed, nor will the 
integrity of these physical barriers be challenged. In addition, the 
proposed modification will not change, degrade, or prevent any 
reactor trip system actuations.
    Therefore, the proposed change does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    2. Does the proposed change create the possibility of a new or 
different kind of accident from any accident previously evaluated?
    The proposed clamp on the OTDT reactor trip function is not 
credited in the safety analyses. Implementation of the limit or 
``clamp'' on the OTDT reactor trip function, along with the 
corresponding changes to the AFD modifier f1 (AFD) and 
RAOC band, will ensure the prevention of stress failure of the fuel 
rod cladding for Condition I and II reactor coolant system cooldown 
events.
    The design basis of the OTDT reactor trip setpoint is to ensure 
DNB [departure from nucleate boiling] protection and to preclude 
vessel exit boiling. The installation of the OTDT clamp would 
continue to ensure this same protection and that the OTDT design 
basis would remain unaffected. The introduction of the OTDT clamp 
would not create any new transients nor would it invalidate the OTDT 
design basis. In addition, there are no transients analyzed in the 
VEGP [Vogtle Electric Generating Plant] FSAR [final safety analysis 
report] that result in a reduction in the reactor coolant 
temperature which rely on OTDT as the primary reactor trip function, 
as cooldown events tend to be non-limiting with respect to the 
criterion of DNB.
    Therefore, the proposed change does not create the possibility 
of a new or different kind of accident from any accident previously 
evaluated.
    3. Does the proposed change involve a significant reduction in 
the margin of safety?
    The proposed clamp on the OTDT reactor trip function is not 
credited in the safety analyses. Implementation of the limit or 
``clamp'' on the OTDT reactor trip function, along with the 
corresponding changes to the AFD modifier f1 (AFD) and 
RAOC band, will ensure the prevention of stress failure of the fuel 
rod cladding for Condition I and II RCS [reactor coolant system] 
cooldown events. This demonstrates continued compliance with 10 CFR 
50, Appendix A, Criterion 10, i.e., that the specified acceptable 
fuel design limits are not exceeded.
    The design basis of the OTDT reactor trip setpoint is to ensure 
DNB [departure from nucleate boiling] protection and to preclude 
vessel exit boiling. The installation of the OTDT clamp would 
continue to ensure this same protection and that the OTDT design 
basis would remain unaffected.
    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 request involves no significant hazards consideration.
    Attorney for licensee: Mr. Arthur H. Domby, Troutman Sanders, 
NationsBank Plaza, Suite 5200, 600 Peachtree Street, NE., Atlanta, 
Georgia 30308-2216.
    NRC Section Chief: Richard J. Laufer, Acting.

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

    Date of amendment request: February 14, 2002.
    Description of amendment request: The proposed amendment revises 
Technical Specifications (TS) 3.4.2.2, ``Reactor Coolant System.'' to 
relax the lift setting tolerance of the pressurizer safety valves from 
2 percent to 3 percent. The current TS 
requirements that the as left lift setting be within 1 
percent will remain intact.
    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.
    The proposed TS change takes credit for the assumptions made in 
the reanalysis of the turbine trip and rod withdrawal from power 
events already evaluated in the UFSAR [Updated Final Safety Analysis 
Report]. 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.
    The proposed TS change takes credit for the assumptions made in 
the reanalysis of the turbine trip and rod withdrawal from power 
events already evaluated in the UFSAR. Therefore, the change does 
not create the possibility of a new or different kind of accident 
from any accident previously evaluated.
    3. Does the proposed change involve a significant reduction in a 
margin of safety?
    Response: No.
    Margin of safety is associated with confidence in the ability of 
the fission product barriers (i.e., fuel and fuel cladding, reactor 
coolant pressure boundary, and containment structure) to limit the 
level of radiation dose to the public. The proposed TS change takes 
credit for the assumptions made in the reanalysis of the turbine 
trip and rod withdrawal from power events already evaluated in the 
UFSAR. Those analyses demonstrated that (1) the fuel design limits 
were maintained by the reactor protection system since the DNBR 
[departure from

[[Page 15629]]

nucleate boiling ratio] was maintained above the limit value, and 
(2) the plant design is such that a turbine trip presents no hazard 
to the integrity of the RCS [reactor coolant system] or the main 
steam system pressure boundary. 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 standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
request for amendments involves no significant hazards consideration.
    Attorney for licensee: Morgan Lewis, 1111 Pennsylvania Avenue, NW., 
Washington, DC 20004.
    NRC Section Chief: Robert A. Gramm.

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

    Date of amendment request: February 14, 2002.
    Description of amendment request: The proposed amendment revises 
Technical Specifications to eliminate shutdown actions associated with 
radiation monitoring instrumentation. The proposed changes will enhance 
plant reliability by reducing exposure to unnecessary shutdowns and 
increase operational flexibility, and relax certain other restrictions.
    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.
    The radiation monitors affected by the proposed amendment are 
not potential accident initiators. Adequate measures are available 
to compensate for radiation monitors that are out of service. The 
proposed amendment does not affect how the affected radiation 
monitors function or their role in the response of an operator to an 
accident or transient. The core damage frequency in the STP [South 
Texas Project] PRA [probabilistic risk assessment] is not impacted 
by the proposed changes. Therefore, STPNOC [South Texas Project 
Nuclear Operating Company] concludes that there is no 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 previously evaluated?
    Response: No.
    The radiation monitors affected by the proposed amendment are 
not credited for the prevention of any accident not evaluated in the 
safety analysis. The proposed amendment involves no changes in the 
way the plant is operated or controlled. It involves no change in 
the design configuration of the plant. No new operating environments 
are created. Therefore, STPNOC concludes the proposed change does 
not create the possibility of a new or different kind of accident 
from any accident previously evaluated.
    3. Does the proposed change involve a significant reduction in a 
margin of safety?
    Response: No.
    The proposed change has no significant effect on functions that 
are supported by the affected radiation monitors. There will be no 
significant effect on the availability and reliability of the 
affected radiation monitors. Adequate measures are available to 
compensate for radiation monitors that are out of service. 
Therefore, STPNOC concludes 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 standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
request for amendments involves no significant hazards consideration.
    Attorney for licensee: Morgan Lewis, 1111 Pennsylvania Avenue, NW., 
Washington, DC 20004.
    NRC Section Chief: Robert A. Gramm.

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

    Date of amendment request: February 14, 2002.
    Description of amendment request: The proposed amendment revises 
Technical Specifications governing radiation monitoring instrumentation 
and reactor coolant system leakage detection to eliminate the 
associated shutdown action requirements and relax certain other 
restrictions.
    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.
    The radiation monitors and leakage detection instrumentation 
affected by the proposed amendment are not potential accident 
initiators. Adequate measures are available to compensate for 
instrumentation that is out of service. The proposed amendment does 
not affect how the affected instrumentation normally functions or 
its role in the response of an operator to an accident or transient. 
The core damage frequency in the STP [South Texas Project] PRA 
[probabilistic risk assessment] is not impacted by the proposed 
changes. Therefore, STPNOC [South Texas Project Nuclear Operating 
Company] concludes that there is no 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.
    The instrumentation affected by the proposed amendment is not 
credited for the prevention of any accident not evaluated in the 
safety analysis. The proposed amendment involves no changes in the 
way the plant is operated or controlled. It involves no change in 
the design configuration of the plant. No new operating environments 
are created. Therefore, STPNOC concludes the proposed change does 
not create the possibility of a new or different kind of accident 
from any accident previously evaluated.
    3. Does the proposed change involve a significant reduction in a 
margin of safety?
    Response: No.
    The proposed change has no significant effect on functions that 
are supported by the affected instrumentation. There will be no 
significant effect on the availability and reliability of the 
affected instrumentation. Adequate measures are available to 
compensate for instrumentation that is out of service. Therefore, 
STPNOC concludes 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 standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
request for amendments involves no significant hazards consideration.
    Attorney for licensee: Morgan Lewis, 1111 Pennsylvania Avenue, NW., 
Washington, DC 20004.
    NRC Section Chief: Robert A. Gramm.

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

    Date of amendment request: January 14, 2002.
    Description of amendment request: The proposed amendment would 
revise Technical Specification (TS) 3.4.16, applicable Bases ``Reactor 
Coolant System Specific Activity,'' and Surveillance Requirement (SR) 
3.4.16.2, from 1.0 microcuries per gram (uCi/gm) iodine-131 to 0.265 
uCi/gm iodine-131. TS 3.4.16, Figure 3.4.16-1, ``Reactor Coolant Dose 
Equivalent Iodine-131 Specific Activity Limit Versus Percent of Rated 
Thermal Power,'' is being deleted and the maximum value of 21 uCi/gm 
iodine-131 is being added to TS Required Action 3.14.16.A and 3.4.16.C. 
In addition, TS Section 3.3.7, ``CREVS [Control Room Emergency 
Ventilation System] Actuation Instrumentation,'' Table 3.3.7-1 changes 
the allowable

[[Page 15630]]

value to the Control Room Radiation and Control Room Air Intakes for SR 
3.3.7.1, 3.3.7.2, and 3.3.7.4 from less than or equal to () 
5.77E-04 uCi/cubic centimeter (cc) (20,199 counts per minute (cpm)) to 
9.45E-05 uCI/cc (3,307 cpm).
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    A. The proposed amendment does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    The proposed Technical Specification[s] change[s] to reduce the 
steady state and 48[-]hour reactor coolant system (RCS) allowable 
iodine concentrations, and to revise the surveillance requirement 
value for the Main Control Room [MCR] air intake radiation monitors 
[do] not change any operator actions nor [do they] change plant 
systems or structures. Therefore, the proposed change[s] to WBN Unit 
1 Technical Specification[s] [do] not result in a significant 
increase in the probability of an accident.
    The calculated radiological consequences at the Exclusion Area 
Boundary (EAB) and Low Population Zone (LPZ) are larger than 
currently discussed in the Final Safety Analysis Report (FSAR) 
accidents for the main steam line break (MSLB) and steam generator 
tube rupture (SGTR) (with the exception of thyroid and beta doses 
being slightly lower for STGR) accidents. The radiological 
consequences for the SGTR and MSLB accidents increased due to 
utilizing more conservative methodologies and more conservative 
assumptions in the calculation. However, the calculated radiological 
consequences remain within the limits identified in 10 CFR 100, 
``Reactor Site Criteria,'' and General Design Criteria (GDC)-19, 
``Control Room,'' and are consistent with NUREG-0800, ``Standard 
Review Plan,'' acceptance criteria.
    The surveillance requirement radiation limit for the Main 
Control Room air intake radiation monitors will be reduced to 
compensate for the change in source terms which resulted from the 
use of the methodology changes in the SGTR accident. This change 
ensures the monitors perform their safety function of control room 
isolation during accident conditions and does not increase the 
probability or consequences of an accident previously evaluated.
    In summary, the control room dose, the LPZ dose, and the EAB 
dose for the SGTR and MSLB remain bounded by the acceptance criteria 
of NUREG-0800 and continue to satisfy an appropriate fraction of the 
10 CFR 100 dose limits and the GDC-19 dose limits. The surveillance 
requirement changes for the Main Control Room radiation monitors 
ensure the monitors perform their intended design function. 
Therefore, the proposed change does not result in a significant 
increase in the [probability or] consequences of an accident 
previously analyzed.
    B. The proposed amendment does not create the possibility of a 
new or different kind of accident from any accident previously 
evaluated.
    The proposed TS change does not alter the configuration of the 
plant. The changes do not directly affect plant operation. The 
change will not result in the installation of any new equipment or 
system or the modification of any existing equipment or systems. No 
new operation procedures, conditions or modes will be created by 
this proposed change. Therefore, the possibility of a new or 
different kind of accident from any accident previously evaluated is 
not created.
    C. The proposed amendment does not involve a significant 
reduction in a margin of safety.
    The methods for calculating the radiological consequences are 
revised for the MSLB and SGTR analysis to utilize the thyroid dose 
conversion factors in International Commission on Radiation 
Protection Publication 30 (ICRP-30) to calculate the dose and 
ARCON96 methodology to calculate atmospheric dispersion 
coefficients.
    The calculated radiological consequences at the EAB and LPZ are 
slightly larger than those noted in the FSAR accidents for the MSLB 
and SGTR (thyroid and beta doses slightly lower for SGTR) accidents. 
The radiological dose consequences for the SGTR and MSLB accidents 
increased due to utilizing more conservative methodologies and more 
conservative assumptions in the calculation. The calculated dose 
consequences of the evaluated accidents remain less than the dose 
limits identified in 10 CFR 100 and GDC-19, and are consistent with 
NUREG-0800 acceptance criteria. The surveillance requirement for the 
MCR radiation monitors is being reduced for consistency with lower 
source terms and to ensure the monitors perform their intended 
design function of isolating the Main Control Room subsequent to an 
accident. Therefore, it is concluded that the proposed change to 
lower the RCS Specific Activity and subsequent changes to the Main 
Control Room radiation monitors are required to ensure the Main 
Control Room dose and the offsite dose are below the acceptable 
limits. Therefore these changes do not result in a significant 
reduction in the margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: General Counsel, Tennessee Valley Authority, 
400 West Summit Hill Drive, ET 10H, Knoxville, Tennessee 37902.
    NRC Section Chief: Richard P. Correia.

Notice of Issuance of Amendments to Facility Operating Licenses

    During the period since publication of the last biweekly notice, 
the Commission has issued the following amendments. The Commission has 
determined for each of these amendments that the application complies 
with the standards and requirements of the Atomic Energy Act of 1954, 
as amended (the Act), and the Commission's rules and regulations. The 
Commission has made appropriate findings as required by the Act and the 
Commission's rules and regulations in 10 CFR Chapter I, which are set 
forth in the license amendment.
    Notice of Consideration of Issuance of Amendment to Facility 
Operating License, Proposed No Significant Hazards Consideration 
Determination, and Opportunity for a Hearing in connection with these 
actions was published in the Federal Register as indicated.
    Unless otherwise indicated, the Commission has determined that 
these amendments satisfy the criteria for categorical exclusion in 
accordance with 10 CFR 51.22. Therefore, pursuant to 10 CFR 51.22(b), 
no environmental impact statement or environmental assessment need be 
prepared for these amendments. If the Commission has prepared an 
environmental assessment under the special circumstances provision in 
10 CFR 51.12(b) and has made a determination based on that assessment, 
it is so indicated.
    For further details with respect to the action see (1) the 
applications for amendment, (2) the amendment, and (3) the Commission's 
related letter, Safety Evaluation and/or Environmental Assessment as 
indicated. All of these items are available for public inspection at 
the Commission's Public Document Room, located at One White Flint 
North, 11555 Rockville Pike (first floor), Rockville, Maryland. 
Publicly available records will be accessible from the Agencywide 
Documents Access and Management Systems (ADAMS) Public Electronic 
Reading Room on the internet at the NRC web site, http://www.nrc.gov/reading-rm.html. If you do not have access to ADAMS or if there are 
problems in accessing the documents located in ADAMS, contact the NRC 
Public Document Room (PDR) Reference staff at 1-800-397-4209, 301-415-
4737 or by e-mail to [email protected].

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

    Date of application for amendment: August 13, 2001.
    Brief description of amendment: The amendment defers withdrawal of 
the first set of reactor vessel surveillance specimens until 10.4 
effective full

[[Page 15631]]

power years, expected to be one additional operating cycle.
    Date of issuance: March 8, 2002.
    Effective date: As of the date of issuance and shall be implemented 
within 30 days.
    Amendment No.: 143.
    Facility Operating License No. NPF-62: The amendment changes the 
updated safety analysis report.
    Date of initial notice in Federal Register: October 17, 2001 (66 FR 
52796). The Commission's related evaluation of the amendment is 
contained in a Safety Evaluation dated March 8, 2002.
    No significant hazards consideration comments received: No.

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

    Date of application for amendment: June 21, 2001, as supplemented 
by letter dated January 18, 2002.
    Brief description of amendment: The amendment modifies the 
technical specification requirement that the main steamline safety 
relief valves (SRVs) open when they are manually actuated by instead 
requiring that the SRV valve actuators stroke on a manual actuation.
    Date of issuance: March 19, 2002.
    Effective date: As of the date of issuance and shall be implemented 
within 30 days.
    Amendment No.: 144.
    Facility Operating License No. NPF-62: The amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: October 3, 2001 (66 FR 
50465). The supplemental letter 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 amendment is contained 
in a Safety Evaluation dated March 19, 2002.
    No significant hazards consideration comments received: No.

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

    Date of application for amendments: November 9, 2001.
    Brief description of amendments: The amendments would revise 
Technical Specification 5.6.5b to add NRC-approved Topical Report 
CENPD-404-P-A, ``Implementation of ZIRLO TM Cladding 
Material in CE Nuclear Power Fuel Assembly Designs,'' into the list of 
analytical methods used to determine core operating limits and thus, 
enable use of ZIRLO clad fuel in Palo Verde Nuclear Generating Station 
units.
    Date of Issuance: March 12, 2002.
    Effective date: March 12, 2002, and shall be implemented within 60 
days of the date of issuance.
    Amendment Nos.: Unit 1-140, Unit 2-140, Unit 3-140.
    Facility Operating License Nos. NPF-41, NPF-51, and NPF-74: The 
amendments revised the Technical Specifications.
    Date of initial notice in Federal Register: January 22, 2002 (67 FR 
2919). The Commission's related evaluation of the amendment is 
contained in a Safety Evaluation dated March 12, 2002.
    No significant hazards consideration comments received: No.

Carolina Power & Light Company, Docket Nos. 50-325, Brunswick Steam 
Electric Plant, Unit 1, Brunswick County, North Carolina

    Date of amendment request: November 26, 2001, as supplemented 
January 31, 2002, February 5, 2002, and February 11, 2002.
    Description of amendment request: The amendment revises the 
Improved Technical Specification 5.5.12 to allow a one-time interval 
increase for the Type A Integrated Leakage Rate Test for no more than 3 
years, 2 months.
    Date of issuance: March 6, 2002.
    Effective date: March 6, 2002.
    Amendment Nos: 216.
    Facility Operating License No. DPR-71: The amendment changes the 
Technical Specifications.
    Date of initial notice in Federal Register: January 8, 2002 (67 FR 
926). The January 31, 2002, and February 5, 2002, supplements contained 
clarifying information only, and did not change the initial no 
significant hazards consideration determination or expand the scope of 
the initial Federal Register notice. The February 11, 2002, supplement 
revised the original request, but the initial no significant hazards 
consideration determination bounded the revised request.
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated March 6, 2002.
    No significant hazards consideration comments received: No.

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

    Date of amendment request: June 26, 2001, as supplemented January 
14, and February 1, 2002.
    Description of amendment request: The amendments revise the 
Technical Specifications to support installation of the General 
Electric Nuclear Measurement Analysis and Control Digital Power Range 
Neutron Monitoring System.
    Date of issuance: March 8, 2002.
    Effective date: March 8, 2002.
    Amendment Nos: 217 and 243.
    Facility Operating License Nos. DPR-71 and DPR-62: Amendments 
change the Technical Specifications.
    Date of initial notice in Federal Register: July 25, 2001 (66 FR 
38759). The January 14, and February 1, 2002, supplements contained 
clarifying information only and did not change the initial no 
significant hazards consideration determination or expand the scope of 
the initial application.
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated March 8, 2002.
    No significant hazards consideration comments received: No.

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

    Date of amendment request: August 1, 2001, as supplemented February 
4, 2002.
    Description of amendment request: The amendment revises the 
Technical Specifications to incorporate NRC-approved Technical 
Specification Task Force Traveler Item 51, ``Revise containment 
requirements during handling irradiated fuel and core alterations,'' 
Revision 2.
    Date of issuance: March 14, 2002.
    Effective date: March 14, 2002.
    Amendment Nos: 218 and 244.
    Facility Operating License Nos. DPR-71 and DPR-62: Amendments 
change the Technical Specifications.
    Date of initial notice in Federal Register: September 5, 2001 (66 
FR 46477). The February 4, 2002, supplement contained clarifying 
information only, and did not change the initial no significant hazards 
consideration determination or expand the scope of the initial Federal 
Register notice.
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated March 14, 2002.
    No significant hazards consideration comments received: No.

[[Page 15632]]

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

    Date of amendment request: November 7, 2001.
    Description of amendment request: The amendments revise Technical 
Specification (TS) 3.1.4, ``Control Rod Scram Times,'' to delineate 
more specific requirements for testing control rod scram times 
following refueling outages. TS 5.1 is revised to reference Title 10 of 
the Code of Federal Regulations (10 CFR) Section 50.59. The amendment 
incorporates the Nuclear Regulatory Commission-approved Technical 
Specification Task Force (TSTF) Item 222, Revision 1, ``Control Rod 
Scram Testing,'' and TSTF Item 364, Revision 0, ``Revision to TS Bases 
Control Program to Incorporate Changes to 10 CFR 50.59.''
    Date of issuance: March 19, 2002.
    Effective date: As of the date of issuance and shall be implemented 
within 60 days from the date of issuance.
    Amendment Nos: 219/245.
    Facility Operating License Nos. DPR-71 and DPR-62: Amendments 
change the Technical Specifications.
    Date of initial notice in Federal Register: November 28, 2001 (66 
FR 59502). The Commission's related evaluation of the amendments is 
contained in a Safety Evaluation dated March 19, 2002.
    No significant hazards consideration comments received: No.

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

    Date of application for amendments: August 6, 2001.
    Brief description of amendments: The amendments revised Technical 
Specification (TS) 3.3.2 for Engineered Safety Feature Actuation System 
Instrumentation, and TS 3.3.6 for Containment Purge and Exhaust 
Isolation Instrumentation. The amendments excluded the Containment 
Purge Ventilation System and the Hydrogen Purge System containment 
isolation valves from the instrumentation testing requirements in TS 
3.3.2 and TS 3.3.6. The amendments also made appropriate changes in the 
Bases for TS 3.3.6 and TS 3.6.3.
    Date of issuance: March 20, 2002.
    Effective date: As of the date of issuance and shall be implemented 
within 30 days from the date of issuance.
    Amendment Nos.: 196/189.
    Facility Operating License Nos. NPF-35 and NPF-52: Amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: December 12, 2001 (66 
FR 64291). The Commission's related evaluation of the amendments is 
contained in a Safety Evaluation dated March 20, 2002.
    No significant hazards consideration comments received: No.

Duke Energy Corporation, Docket Nos. 50-269, 50-270, and 50-287, Oconee 
Nuclear Station, Units 1, 2, and 3, Oconee County, South Carolina

    Date of application of amendments: August 14, 2001.
    Brief description of amendments: The proposed amendments would 
revise TS Surveillance Requirement 3.3.5.2 by changing the Engineered 
Safeguards Protective System Analog Instrument channel functional test 
frequency from 31 days to 92 days.
    Date of Issuance: March 18, 2002.
    Effective date: As of the date of issuance and shall be implemented 
within 30 days from the date of issuance.
    Amendment Nos.: 321/321/322.
    Renewed Facility Operating License Nos. DPR-38, DPR-47, and DPR-55: 
Amendments revised the Technical Specifications.
    Date of initial notice in Federal Register: September 5, 2001 (66 
FR 46478). The Commission's related evaluation of the amendments is 
contained in a Safety Evaluation dated March 18, 2002.
    No significant hazards consideration comments received: No.

Exelon Generation Company, Docket No. 50-352, Limerick Generating 
Station, Unit 1, Montgomery County, Pennsylvania

    Date of application for amendment: December 21, 2001, as 
supplemented February 15, 2002.
    Brief description of amendment: This amendment revises the minimum 
critical power ratio safety limits for operating cycle 10.
    Date of issuance: March 12, 2002.
    Effective date: As of the date of issuance, and shall be 
implemented within 30 days.
    Amendment No.: 156.
    Facility Operating License No. NPF-39: This amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: January 22, 2002 (67 FR 
2924). The February 15, 2002, 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 March 12, 2002.
    No significant hazards consideration comments received: No.

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

    Date of application for amendments: May 23, 2001.
    Brief description of amendments: These amendments deleted Technical 
Specification 3.4.2, Limiting Condition for Operation, Action Statement 
b, concerning operator actions with stuck open safety/relief valves.
    Date of issuance: As of date of issuance and shall be implemented 
within 30 days.
    Effective date: March 20, 2002.
    Amendment Nos.: 157 and 119.
    Facility Operating License Nos. NPF-39 and NPF-85. The amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: August 22, 2001 (66 FR 
44171). The Commission's related evaluation of the amendments is 
contained in a Safety Evaluation dated March 20, 2002.
    No significant hazards consideration comments received: No.

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

    Date of application for amendments: June 26, 2001, as supplemented 
by letter dated November 15, 2001.
    Brief description of amendments: The amendments revised Technical 
Specification 3/4.3.3, Emergency Core Cooling System, Actions 36 and 37 
of Table 3.3.3-1, and associated TS Bases. The change to Action 36 
clarifies equipment affected by inoperable components. The change to 
Action 37 takes advantage of the inherent overlap of the degraded 
voltage relays' characteristics such that inoperable relays that define 
a channel can be taken out of service without placing its associated 
source breaker in the trip position.
    Date of issuance: March 20, 2002.
    Effective date: As of date of issuance and shall be implemented 
within 30 days.
    Amendment Nos.: 158 and 120.
    Facility Operating License Nos. NPF-39 and NPF-85: The amendments 
revised the Technical Specifications.

[[Page 15633]]

    Date of initial notice in Federal Register: August 22, 2001 (66 FR 
44171). The November 15, 2001, letter provided clarifying information 
that did not change the initial proposed no significant hazards 
consideration determination.
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated March 20, 2002.
    No significant hazards consideration comments received: No.

National Aeronautics and Space Administration, Docket Nos. 50-30 and 
50-185, the Plum Brook Test Reactor and the Plum Brook Mockup Reactor, 
Sandusky, Ohio

    Date of application for amendments: December 20, 1999, as 
supplemented by letters dated March 26, November 19, and December 20, 
2001, and January 24, 2002.
    Brief description of amendments: The amendment allows 
decommissioning of the PBRF in accordance with NASA's application as 
supplemented. Pursuant to 10 CFR 50.82(b)(5), the approved 
decommissioning plan will be a supplement to the Safety Analysis Report 
or equivalent.
    Date of issuance: March 20, 2002.
    Effective date: March 20, 2002.
    Amendment Nos.: Amendment No. 11 to Plum Brook Test Reactor and 
Amendment No. 7 to the Plum Brook Mockup Reactor.
    Facility Operating License Nos. TR-3 and R-93: These amendments 
consist of changes to the Facility Licenses.
    Date of initial notice in Federal Register: January 22, 2002 (67 FR 
2924). The January 24, 2002, supplemental letter provided additional 
clarifying information, did not expand the scope of the application as 
originally noticed, and did not change the staff's original proposed no 
significant hazards consideration determination. The Commission's 
related evaluation of the amendments is contained in a Safety 
Evaluation enclosed with the amendments dated March 20, 2002.
    No significant hazards consideration comments received: No.

Nine Mile Point Nuclear Station, LLC, Docket No. 50-410, Nine Mile 
Point Nuclear Station, Unit 2, Oswego County, New York

    Date of application for amendment: November 20, 2001, as 
supplemented January 28 and February 21, 2002.
    Brief description of amendment: The amendment revised the Technical 
Specifications, Section 2.1.1.2, to reflect the results of cycle-
specific calculations performed for the upcoming Operating Cycle 9, and 
Section 5.6.5.b, to delete two redundant references.
    Date of issuance: March 13, 2002.
    Effective date: As of the date of issuance, to be implemented prior 
to startup from Refueling Outage 8.
    Amendment No.: 105.
    Facility Operating License No. NPF-69: Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: December 26, 2001 (66 
FR 66468). The licensee's January 28 and February 21, 2002, 
supplemental letters provided clarifying information that was within 
the scope of the amendment request and did not change the initial 
proposed no significant hazards consideration determination.
    The staff's related evaluation of the amendment is contained in a 
Safety Evaluation dated March 13, 2002.
    No significant hazards consideration comments received: No.

Nine Mile Point Nuclear Station, LLC, Docket No. 50-410, Nine Mile 
Point Nuclear Station, Unit 2, Oswego County, New York

    Date of application for amendment: December 26, 2001.
    Brief description of amendment: The amendment revises Table 
3.6.1.3-1, ``Secondary Containment Bypass Leakage Paths Leakage Rate 
Limits,'' to reflect the NRC staff's approval of the licensee's 
proposed modification of two primary containment isolation valves on 
feedwater piping from air-operated to become simple check valves.
    Date of issuance: March 8, 2002.
    Effective date: As of the date of issuance to be implemented prior 
to startup from Refueling Outage 8.
    Amendment No.: 104.
    Facility Operating License No. NPF-69: Amendment revises the 
Technical Specifications.
    Date of initial notice in Federal Register: February 5, 2002 (67 FR 
5329).
    The staff's related evaluation of the amendment is contained in a 
Safety Evaluation dated March 8, 2002.
    No significant hazards consideration comments received: No.

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

    Date of amendment request: February 28, 2001, as supplemented July 
31, 2001, and December 21, 2001.
    Description of amendment request: The amendment changes Seabrook 
Station Technical Specification 3/4.8.1.1 A.C. Sources--Operating. The 
changes are related to allowed outage time for restoration or 
verification of the operability of offsite power sources and to 
emergency diesel generator surveillance requirements.
    Date of issuance: March 7, 2002.
    Effective date: As of its date of issuance, and shall be 
implemented within 60 days.
    Amendment No.: 80.
    Facility Operating License No. NPF-86: Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: April 18, 2001 (66 FR 
20007). The July 31, 2001, and December 21, 2001, letters were within 
the scope of and did not affect the staff's finding of no significant 
hazards considerations.
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated March 7, 2002.
    No significant hazards consideration comments received: No.

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

    Date of application for amendments: July 30, 2001, as supplemented 
September 7, October 16, and December 5, 2001, and January 18, 2002.
    Brief description of amendments: The amendments revised Technical 
Specification 5.5.12, ``Primary Containment Leakage Rate Testing 
Program,'' to allow a one-time deferral of the Type A containment 
integrated leakage rate test (ILRT) at the Susquehanna Steam Electric 
Station (SSES), Units 1 and 2. The Unit 1 test may be deferred to no 
later than May 3, 2007, and the Unit 2 test may be deferred to no later 
than October 30, 2007, resulting in an extended interval of 15 years 
for performance of the next ILRT at each unit. Additionally, the 
amendments allow a one-time deferral of the drywell-to-suppression 
chamber bypass leakage test, Surveillance Requirement (SR) 3.6.1.1.2, 
so that it will continue to be conducted along with the ILRT.
    Date of issuance: March 8, 2002.
    Effective date: As of date of issuance and shall be implemented 
within 60 days.
    Amendment Nos.: 202, 176.
    Facility Operating License Nos. NPF-14 and NPF-22. The amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: February 5, 2002 (67 FR 
5330). The Commission's related evaluation of the amendments is 
contained in a Safety Evaluation dated March 8, 2002.
    No significant hazards consideration comments received: No.

[[Page 15634]]

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: August 7, 2001.
    Brief description of amendment: This amendment adds a response time 
requirement to the Technical Specifications for the Source Range 
Neutron Flux Reactor Trip function.
    Date of issuance: March 8, 2002.
    Effective date: March 8, 2002.
    Amendment No.: 157.
    Facility Operating License No. NPF-12: Amendment revises the 
Technical Specifications.
    Date of initial notice in Federal Register: February 5, 2002 (67 FR 
5332). The Commission's related evaluation of the amendment is 
contained in a Safety Evaluation dated March 8, 2002.
    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: June 19, 2001.
    Brief description of amendment: This amendment approves inclusion 
of two upgraded 7300 Process Protection System instrument cards (NLP--
Loop Power Supply and Isolator card, and NSA--Summing Amplifier card) 
into the response time testing elimination population. The associated 
Bases for Technical Specification 3/4.3.1 is being revised to reflect 
this change.
    Date of issuance: March 12, 2002.
    Effective date: March 12, 2002.
    Amendment No.: 158.
    Facility Operating License No. NPF-12: Amendment revises the 
Technical Specifications.
    Date of initial notice in Federal Register: July 25, 2001 (66 FR 
38766).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated March 12, 2002.
    No significant hazards consideration comments received: No.

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 application for amendments: January 9, 2002.
    Brief description of amendments: The amendments revise the 
Technical Specification 5.4, ``Technical Specifications (TS) Bases 
Control'' to delete the term ``unreviewed safety question.''
    Date of issuance: March 19, 2002.
    Effective date: March 19, 2002, to be implemented within 60 days of 
issuance.
    Amendment Nos.: Unit 2-184; Unit 3-175.
    Facility Operating License Nos. NPF-10 and NPF-15: The amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: February 5, 2002 (67 FR 
5333). The Commission's related evaluation of the amendments is 
contained in a Safety Evaluation dated March 19, 2002.
    No significant hazards consideration comments received: No.

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

    Date of application for amendments: December 14, 2001.
    Brief description of amendments: The amendments revise Surveillance 
Requirement (SR) 3.0.3 to extend the delay period, before entering a 
Limiting Condition for Operation, following a missed surveillance. The 
delay period 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 is 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.''
    Date of issuance: March 8, 2002.
    Effective date: As of the date of issuance and shall be implemented 
by August 1, 2002.
    Amendment Nos.: 228/170.
    Facility Operating License Nos. DPR-57 and NPF-5: Amendments revise 
the Technical Specifications and associated Bases.
    Date of initial notice in Federal Register: February 5, 2002 (67 FR 
5333). The Commission's related evaluation of the amendments is 
contained in a Safety Evaluation dated March 8, 2002.
    No significant hazards consideration comments received: No.

Southern Nuclear Operating Company, Inc., et al., Docket Nos. 50-424 
and 50-425, Vogtle Electric Generating Plant, Units 1 and 2, Burke 
County, Georgia

    Date of application for amendments: April 27, 2001.
    Brief description of amendments: The amendments revised the 
Technical Specifications 3.3.6, ``Containment Ventilation Isolation 
Instrumentation,'' to extend the surveillance test interval for Potter 
and Brumfield type motor-driven slave relays in the containment 
ventilation isolation system from 92 days to 18 months. The associated 
Bases for SR 3.3.6.5 will be revised to reflect this change.
    Date of issuance: February 21, 2002.
    Effective date: As of the date of issuance and shall be implemented 
within 30 days from the date of issuance.
    Amendment Nos.: 124/102.
    Facility Operating License Nos. NPF-68 and NPF-81: Amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: June 21, 2001 (66 FR 
31714). The Commission's related evaluation of the amendments is 
contained in a Safety Evaluation dated February 21, 2002.
    No significant hazards consideration comments received: No.

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

    Date of amendment request: May 30, 2001.
    Brief description of amendments: The proposed amendment permits 
relaxation of the allowed outage times and bypass test times for 
limiting conditions for operation outlined in Technical Specifications 
3.3.1, ``Reactor Trip System Instrumentation,'' and 3.3.2, ``Engineered 
Safety Features Actuation System Instrumentation.''
    Date of issuance: March 19, 2002.
    Effective date: The amendments are effective as of the date of 
issuance, and shall be implemented within 30 days of the day of 
issuance.
    Amendment Nos.: Unit 1-136; Unit 2-125.
    Facility Operating License Nos. NPF-76 and NPF-80: The amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: August 22, 2001 (66 FR 
44177). The Commission's related evaluation of the amendments is 
contained in a Safety Evaluation dated March 19, 2002.
    No significant hazards consideration comments received: No.

[[Page 15635]]

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

    Date of amendment request: August 2, 2001.
    Brief description of amendments: The amendments consist of revision 
to Technical Specifications 3/4.6.1.6 regarding containment structural 
integrity.
    Date of issuance: March 19, 2002.
    Effective date: As of the date of issuance, and shall be 
implemented within 30 days from the date of issuance.
    Amendment Nos.: Unit 1-137; Unit 2-126.
    Facility Operating License Nos. NPF-76 and NPF-80: The amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: January 22, 2002 (67 FR 
2929). The Commission's related evaluation of the amendments is 
contained in a Safety Evaluation dated March 19, 2002.
    No significant hazards consideration comments received: No.

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: March 12, 2001.
    Brief description of amendments: The amendments delete Sequoyah 
Technical Specification (TS) Surveillance Requirement 4.7.7.a from TS 
3/4.7.7, ``Control Room Emergency Ventilation Systems,'' and adds a new 
Section 3/4.7.13, ``Control Room Air-Conditioning System (CRACS),'' to 
the TS. This TS addition will also provide the necessary requirements, 
consistent with NUREG-1431, to address the condition when main control 
room chillers and air handling units are inoperable.
    Date of issuance: February 27, 2002.
    Effective date: February 27, 2002.
    Amendment Nos.: 273 and 262.
    Facility Operating License Nos. DPR-77 and DPR-79: Amendments 
revised the TSs.
    Date of initial notice in Federal Register: April 18, 2001 (66 FR 
20011). The Commission's related evaluation of the amendment is 
contained in a Safety Evaluation dated February 27, 2002.
    No significant hazards consideration comments received: No.

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: January 15, 2002 (TS 01-13).
    Brief description of amendments: The amendments revised Technical 
Specifications (TSs) Section 4.0.5.c to provide an exception to the 
recommendations of Regulatory Position c.4.b NRC Regulatory Guide 1.14, 
Revision 1, ``Reactor Coolant Pump Flywheel Integrity,'' dated August 
1975. The exception allows either (a) a qualified in-place ultrasonic 
volumetric examination over the volume from the inner bore of the 
flywheel to the circle of one-half the outer radius or (b) a surface 
examination (magnetic particle testing and/or liquid penetrant testing) 
of exposed surfaces of the removed flywheel to be conducted at 
approximately 10-year intervals.
    Date of issuance: March 8, 2002.
    Effective date: Date of issuance, to be implemented within 45 days 
of issuance.
    Amendment Nos.: 274/263.
    Facility Operating License Nos. DPR-77 and DPR-79: Amendments 
revised the TSs.
    Date of initial notice in Federal Register: February 5, 2002 (67 FR 
5339). The Commission's related evaluation of the amendment is 
contained in a Safety Evaluation dated March 8, 2002.
    No significant hazards consideration comments received: No.

Virginia Electric and Power Company, Docket No. 50-339, North Anna 
Power Station, Unit 2, Louisa County, Virginia

    Date of application for amendment: January 9, 2001.
    Brief description of amendment: This amendment revises the Facility 
Operating License (FOL) to remove expired license conditions, make 
editorial changes in the FOL, relocate license conditions, and remove 
license conditions associated with completed modifications.
    Date of issuance: March 19, 2002.
    Effective date: As of the date of issuance and shall be implemented 
within 60 days from the date of issuance.
    Amendment Nos.: 211.
    Facility Operating License No. NPF-7: Amendment changes the FOL.
    Date of initial notice in Federal Register: February 21, 2001 (66 
FR 11065). The Commission's related evaluation of the amendments is 
contained in a Safety Evaluation dated March 19, 2002.
    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: April 11, 2000, as supplemented 
August 28, and November 20, 2000, April 11, July 31, November 19, and 
December 20, 2001, and February 8, 2002.
    Brief Description of amendments: These amendments revise the 
Technical Specifications requirements to be consistent with an 
alternative source term in accordance with the requirements of 10 CFR 
50.67, ``Accident Source Term.''
    Date of issuance: March 8, 2002.
    Effective date: March 8, 2002.
    Amendment Nos.: 230 and 230.
    Facility Operating License Nos. DPR-32 and DPR-37: Amendments 
change the Technical Specifications.
    Date of initial notice in Federal Register: June 27, 2001 (66 FR 
34289). The supplements contained clarifying information only, and did 
not change the initial no significant hazards consideration 
determination or expand the scope of the initial application.
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated March 8, 2002.
    No significant hazards consideration comments received: No.

Yankee Atomic Electric Co., Docket No. 50-29, Yankee Nuclear Power 
Station (YNPS) Franklin County, Massachusetts

    Date of application for amendment: September 28, 2000, as 
supplemented by letters dated October 12, 2000, April 18, May 29 and 
June 28, 2001, and March 4, 2002.
    Brief description of amendment: The amendment revises License 
Condition 2.C.(3) to reference the revisions of the Physical Security 
Plan, Guard Training and Qualification Plan, and Safeguards Contingency 
Plan which provide for movement of the spent nuclear fuel from the 
spent fuel pool to the Independent Spent Fuel Storage Installation.
    Date of issuance: March 13, 2002.
    Effective date: March 13, 2002.
    Amendment No.: 156.
    Facility Operating License No. DPR-3: The amendment revised the 
License.
    Date of initial notice in Federal Register: March 26, 2001 (66 FR 
16501). The April 18, May 29, and June 28, 2001, and March 4, 2002, 
supplemental letters provided additional clarifying information that 
did not expand the scope of the application as originally noticed and 
did not change the staff's original proposed no significant hazards 
consideration determination.

[[Page 15636]]

    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated March 13, 2002.
    No significant hazards consideration comments received: No.

    Dated at Rockville, Maryland, this 25th day of March, 2002.

    For the Nuclear Regulatory Commission.
John A. Zwolinski,
Director, Division of Licensing Project Management, Office of Nuclear 
Reactor Regulation.
[FR Doc. 02-7799 Filed 4-1-02; 8:45 am]
BILLING CODE 7590-01-P