[Federal Register Volume 66, Number 95 (Wednesday, May 16, 2001)]
[Notices]
[Pages 27174-27184]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: 01-12192]


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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 April 23, 2001, through May 4, 2001. The 
last biweekly notice was published on May 2, 2001 (66 FR 22021).

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

    The Commission has made a proposed determination that the following 
amendment requests involve no significant hazards consideration. Under 
the Commission's regulations in 10 CFR 50.92, this means that operation 
of the facility in accordance with the proposed amendment would not (1) 
involve a significant increase in the probability or consequences of an 
accident previously evaluated; or (2) create the possibility of a new 
or different kind of accident from any accident previously evaluated; 
or (3) involve a significant reduction in a margin of safety. The basis 
for this proposed determination for each amendment request is shown 
below.
    The Commission is seeking public comments on this proposed 
determination. Any comments received within 30 days after the date of 
publication of this notice will be considered in making any final 
determination.
    Normally, the Commission will not issue the amendment until the 
expiration of the 30-day notice period. However, should circumstances 
change during the notice period such that failure to act in a timely 
way would result, for example, in derating or shutdown of the facility, 
the Commission may issue the license amendment before the expiration of 
the 30-day notice period, provided that its final determination is that 
the amendment involves no significant hazards consideration. The final 
determination will consider all public and State comments received 
before action is taken. Should the Commission take this action, it will 
publish in the Federal Register a notice of issuance and provide for 
opportunity for a hearing after issuance. The Commission expects that 
the need to take this action will occur very infrequently.
    Written comments may be submitted by mail to the Chief, Rules and 
Directives Branch, Division of Administrative Services, Office of 
Administration, U.S. Nuclear Regulatory Commission, Washington, DC 
20555-0001, and should cite the publication date and page number of 
this Federal Register notice. Written comments may also be delivered to 
Room 6D22, Two White Flint North, 11545 Rockville Pike, Rockville, 
Maryland from 7:30 a.m. to 4:15 p.m. Federal workdays. Copies of 
written comments received may be examined at the NRC Public Document 
Room, located at One White Flint North, 11555 Rockville Pike (first 
floor), Rockville, Maryland 20852. The filing of requests for a hearing 
and petitions for leave to intervene is discussed below.
    By June 15, 2001, the licensee may file a request for a hearing 
with respect to issuance of the amendment to the subject facility 
operating license and any person whose interest may be affected by this 
proceeding and who wishes to participate as a party in the proceeding 
must file a written request for a hearing and a petition for leave to 
intervene. Requests for a hearing and a petition for leave to intervene 
shall be filed in accordance with the Commission's ``Rules of Practice 
for Domestic Licensing Proceedings'' in 10 CFR part 2. Interested 
persons should consult a current copy of 10 CFR 2.714 which is 
available at the Commission's Public Document Room, located at One 
White Flint North, 11555 Rockville Pike (first floor), Rockville, 
Maryland 20852. Publicly available records will be accessible and 
electronically from the ADAMS Public Library component on the NRC Web 
site, http://www.nrc.gov (the Electronic Reading Room). If a request 
for a hearing or petition for leave to intervene is filed by the above 
date, the Commission or an Atomic Safety and Licensing Board, 
designated by the Commission or by the Chairman of the Atomic Safety 
and Licensing Board Panel, will rule on the request and/or petition; 
and the Secretary or the designated Atomic Safety and Licensing Board 
will issue a notice of a hearing or an appropriate order.
    As required by 10 CFR 2.714, a petition for leave to intervene 
shall set forth with particularity the interest of the petitioner in 
the proceeding, and how that interest may be affected by the results of 
the proceeding. The petition should specifically explain the reasons 
why intervention should be permitted with particular reference to the

[[Page 27175]]

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

Energy Northwest, Docket No. 50-397, Columbia Generating Station, 
Benton County, Washington

    Date of amendment request: April 16, 2001.
    Description of amendment request: Energy Northwest is requesting 
approval for a change to the facility as described by the Final Safety 
Analysis Report (FSAR). The change allows for an unisolable drain line 
between the reactor core isolation cooling (RCIC) and control rod 
drive/condensate (CRD/COND) pump rooms. This change will modify the 
requirements that the RCIC and CRD/COND pump rooms be water-resistant 
or watertight, and connected by an isolable drain line.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. The proposed change does not involve a significant increase 
in the probability or consequences of an accident previously 
evaluated.
    The proposed change to allow operation of the plant with an open 
drain line between the RCIC and CRD/COND pump rooms does not 
increase the chances of a flooding event occurring in the RCIC or 
CRD/COND pump rooms. Also, operating the plant with an open drain 
line between the RCIC and CRD/COND pump rooms does not increase the 
radiological consequences of any previously evaluated accidents. A 
conservative revision to the flooding safe shutdown analysis, which 
combines the effects on equipment of both rooms flooding 
simultaneously, shows that sufficient safe shutdown equipment 
remains available and safe plant shutdown can be accomplished. 
Remaining systems are the same as the equipment providing the safe 
shutdown path approved by the NRC for Appendix R post-fire safe 
shutdown scenarios. Furthermore, the effects of the postulated flood 
from a pipe crack plus other normal leakage spread over the large 
floor area of the combined RCIC and CRD/COND rooms results in a 
flood event that develops slowly. If credited, safety-related leak 
detection instrumentation is available to provide plant operators 
more time to terminate the flood and limit the amount of equipment 
potentially lost from the event. With consideration of operator 
action and mitigation, the flood could be terminated quickly with 
minimal components affected. Plant procedures provide direction for 
operators to take actions to mitigate floods.
    The proposed change to remove the requirement that pump room 
wall penetrations and doors located in the Reactor Building be 
``water-resistant'' or ``watertight'' does not contribute to the 
likelihood that a flooding event will occur, nor does it increase 
the radiological dose received in any previously evaluated 
accidents. Reactor Building pump room doors and penetrations will 
exhibit a minimal amount of leakage during a flooding event, and 
have seals that can leak yet still minimize flooding between rooms 
even with significant hydrostatic pressure generated from flooding 
water levels. The minimal water leakage past these seals is 
consistent with assumptions documented in the existing flooding 
analysis.
    Therefore, operation of Columbia Generating Station in 
accordance with the proposed amendment will not involve a 
significant increase in the probability or consequences of an 
accident previously evaluated.
    2. The proposed change does not create the possibility of a new 
or different kind of

[[Page 27176]]

accident from any accident previously evaluated.
    The proposed change to allow an unisolable drain line between 
the RCIC and CRD/COND pump rooms is accounted for in a revised and 
conservative flooding safe shutdown analysis. The flooding safe 
shutdown analysis documents the impact of flooding on equipment in 
the pump rooms, and on electrical circuits routed through, but not 
terminated in, the RCIC and CRD/COND pump rooms. From this analysis 
no link could be established between affected systems and mechanisms 
that could create a new or different kind of accident. The analysis 
also concluded that the effects of the unisolable drain line and 
subsequent flood would not cause a transient that would be imposed 
on the current analysis that assumes a flood with a single active 
failure. Therefore, the unisolable line between the RCIC and CRD/
COND rooms will not create the possibility of a new or different 
kind of accident. The proposed change to allow minimal water leakage 
past ECCS [Emergency Core Cooling System], RCIC and CRD/COND pump 
room doors and penetrations is consistent with and documented in 
existing flooding analysis assumptions. These rooms do not need to 
be water-resistant or watertight.
    Therefore, operation of Columbia Generating Station in 
accordance with the proposed amendment will not create the 
possibility of a new or different kind of accident from any accident 
previously evaluated.
    3. The proposed change does not involve a significant reduction 
in a margin of safety.
    The proposed change to allow an unisolable drain line between 
the RCIC and CRD/COND pump rooms does not result in a significant 
reduction in the margin of safety. The change is of very low risk 
significance, with an increase in core damage frequency of less than 
1E-10. Furthermore, a revised and conservative flooding safe 
shutdown analysis has concluded, as with previous flooding analysis 
for the ECCS, RCIC and CRD/COND pump rooms, that the ability to 
safely shutdown the plant has been preserved when considering a 
flooding scenario which impacts both the RCIC and CRD/COND pump 
rooms. In addition, safety-related leak detection instrumentation is 
available and could be credited to provide plant operators more time 
to terminate the flood and limit the amount of equipment potentially 
lost from the event. With consideration of operator action and 
mitigation, the flood could be terminated quickly with minimal 
components affected. Plant procedures provide direction for 
operators to take actions to mitigate flooding in ECCS, RCIC and 
CRD/COND pump rooms.
    The proposed change to allow minimal water leakage past ECCS, 
RCIC and CRD/COND pump room doors and penetrations does not result 
in a significant reduction in the margin of safety because it does 
not prevent the plant from achieving safe shutdown during a flooding 
event. Minimal water leakage is consistent with and documented in 
existing flooding analysis assumptions.
    Therefore, the operation of Columbia Generating Station in 
accordance with the proposed amendment will 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: Thomas C. Poindexter, Esq., Winston & 
Strawn, 1400 L Street, NW., Washington, DC 20005-3502.
    NRC Section Chief: Stephen Dembek.

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

    Date of amendment request: February 19, 2001.
    Description of amendment request: The proposed change to the 
Technical Specifications (TS) modifies TS 3.6.5, ``Vacuum Relief 
Valves,'' limiting condition for operation and extends the allowed 
outage time from 4 hours to 72 hours for returning an inoperable 
primary containment to annulus relief valve to OPERABLE status. In 
addition, Entergy proposes to delete Attachment 1 to the Waterford 
Steam Electric Station, Unit 3 Operating License and revise Condition 
2.C.1 to reflect the deletion.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Will operation of the facility in accordance with this 
proposed change involve a significant increase in the probability or 
consequences of an accident previously evaluated?
    The proposed changes do not create any new system interactions 
and have no impact on operation or function of any system or 
equipment in a way that could cause an accident. The primary 
containment to annulus vacuum relief valves are part of the 
containment vacuum relief system and are not initiators of any 
events nor affect any accident initiators of any events previously 
analyzed in Chapter 15 of the FSAR [Final Safety Analysis Report].
    The primary containment to annulus vacuum relief valves are 
designed to mitigate the consequences of an inadvertent containment 
spray system actuation during normal plant operation. The FSAR 
analysis determined that with one of the two containment vacuum 
lines failed, the resultant peak calculated external pressure load 
of 0.49 psi [pounds per square inch] on the containment was less 
than the design external pressure loading of 0.65 psi. These 
proposed changes do not affect any of the assumptions used in the 
analysis. Hence, the consequences of the design basis accident 
previously evaluated do not change.
    Therefore, this change does not involve a significant increase 
in the probability or consequences of any accident previously 
evaluated.
    2. Will operation of the facility in accordance with this 
proposed change create the possibility of a new or different kind of 
accident from any accident previously evaluated?
    The proposed changes do not alter the design, configuration, or 
the method of operation of the plant. There is no change being made 
to the parameters within which the plant is operated. The setpoints 
at which the protective or mitigative actions are initiated are 
unaffected by this change. As such, no new failure modes are being 
introduced that would involve any potential initiating events that 
would create any new or different kind of accident.
    Therefore, this change does not create the possibility of a new 
or different kind of accident from any previously evaluated.
    3. Will operation of the facility in accordance with this 
proposed change involve a significant reduction in a margin of 
safety?
    The proposed changes do not affect the bases used in or the 
results of the analysis to establish the margin of safety. The 
margin of safety is established through equipment design, operating 
parameters, and the setpoints at which automatic actions are 
initiated. None of these are impacted by the proposed change. The 
proposed change is acceptable because it assures at least one vacuum 
relief line will remain available in the event of a single failure. 
This further assures the ability to actuate upon demand for the 
purpose of mitigating the consequences of the design basis accident 
(inadvertent actuation of the containment spray system during normal 
operation). The remaining vacuum relief line provides sufficient 
vacuum relief capacity to prevent exceeding the design external 
pressure loading on containment of 0.65 psi. The resultant 
calculated peak external pressure loading on containment is 0.49 
psi.
    Therefore, this change does not involve a significant reduction 
in the margin of safety.

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

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

    Date of amendment request: April 2, 2001.
    Description of amendment request: Pursuant to 10 CFR 50.59, Entergy 
Operations, Inc. (the licensee) requests

[[Page 27177]]

review and approval of changes to the Waterford Steam Electric Station, 
Unit 3, design basis as described in the Updated Final Safety Analysis 
Report (UFSAR) for which it has been determined that an unreviewed 
safety question exists. The change concerns design requirements for the 
alignment of the Refueling Water Storage Pool (RWSP) boundary isolation 
valves to the RWSP Purification System.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    The proposed change will allow the manual valves (FS-423 and FS-
404) that isolate the RWSP from the RWSP Purification System to be 
maintained open. The RWSP Purification System is aligned to the RWSP 
to maintain the purity and clarity of the borated water contained in 
the pool. The RWSP is also one of two means of makeup to the Spent 
Fuel Pool (SFP), with the Condensate Storage Pool being the primary 
makeup source. These manual valves provide the boundary between the 
seismically qualified safety related RWSP and the non-seismic, non-
safety related RWSP Purification System.
    (1) The proposed activity does not involve a significant 
increase in the probability or consequences of any accident 
previously evaluated.
    The RWSP is not involved in any initiating event that could 
result in any accident. The RWSP has a safety function that assists 
in accident mitigation.
    The proposed change has been reviewed against Engineering 
Standards and Licensing requirements contained in the Updated Final 
Safety Analysis Report (UFSAR). This review has concluded that use 
of operator action to isolate the RWSP Purification System Boundary 
Isolation valves or secure the RWSP Purification pump, as necessary, 
will allow the RWSP to perform its safety function in any plant 
mode. The RWSP, however, is not required to perform a safety 
function concurrent with a seismic event. The highest estimated 
annual probability of a small Loss of Coolant Accident (LOCA) while 
the purification system is aligned to the SFP, and operator failure 
to isolate the purification system, causing a diversion of RWSP 
water that could affect the Emergency Core Cooling System (ECCS) 
pumps is about 2.5E-8 per year, which is considered a negligible 
risk.
    Therefore, the proposed activity does not involve a significant 
increase in the probability or consequences of any accident 
previously evaluated.
    (2) The proposed activity does not create the possibility of a 
new or different kind of accident from any accident previously 
evaluated.
    The RWSP Purification System was intended to be aligned 
periodically to the RWSP. The proposed change will allow the RWSP 
Purification System to be normally aligned to the RWSP through 
manually operated open valves. It has been shown that operator 
action can be credited to isolate the RWSP Purification System in a 
sufficient time to ensure the safety function of the RWSP is 
maintained. If Recirculation Actuation Signal (RAS) occurs the 
isolation valves will become inaccessible due to high dose rates in 
the general area. However if the RAS occurs before an operator can 
isolate the RWSP (i.e. 54 minutes), the RWSP Purification System 
would not have to be isolated because the RWSP would have fulfilled 
its required safety function.
    The proposed alignment to maintain the RWSP Purification System 
isolation valves open introduces a new system interaction during a 
LOCA. However, it has been demonstrated the the safety function of 
the RWSP is assured assuming the new system interaction.
    Therefore, the proposed activity does not create the possibility 
of a new or different kind of accident from any accident previously 
evaluated.
    (3) The proposed activity does not involve a significant 
reduction [in] a margin of safety.
    It has been evaluated that it will take no more than 54 minutes 
for operations personnel to isolate the RWSP. During this time, 
approximately 1.4% of the RWSP level could be depleted assuming 
maximum leakage. This volume will be incorporated into the 
analytical limit. The proposed analytical limit will continue to 
assure the safety limits evaluated in the Design Basis Accident 
(DBA) analyses are maintained.
    Therefore, this proposed activity does not involve a significant 
reduction in the margin of safety.

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

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

    Date of amendment request: April 2, 2001.
    Description of amendment request: The proposed amendment would 
revise the Technical Specifications (TSs) to relocate TSs 3/4.9.4, 
``Refueling Operations, Decay Time;'' 3/4.9.5, ``Refueling Operations, 
Communications;'' 3/4.9.6, ``Refueling Operations, Refueling 
Platform;'' and 
3/4.9.7, ``Refueling Operations, Crane Travel-Spent Fuel Storage 
Pool;'' to the Hope Creek Updated Final Safety Analysis Report (UFSAR). 
The proposed amendment would also modify the associated Bases pages and 
index pages.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. The proposed changes do not involve a significant increase in 
the probability or consequences of an accident previously evaluated.
    The requested amendments will not involve an increase in the 
probability or consequences of an accident previously evaluated. 
Relocation of the affected Technical Specification sections and 
their Bases to the Hope Creek UFSAR will have no affect on the 
probability that any accident will occur. Additionally, the 
consequences of an accident will not be impacted because the 
affected systems and components will continue to be utilized in the 
same manner as before. No impact on the plant response to accidents 
will be created.
    Based on the above, the proposed changes do not significantly 
increase the probability or consequences of any accident previously 
evaluated.
    2. The proposed change does not create the possibility of a new 
or different kind of accident from any accident previously 
evaluated.
    The proposed amendments will not create the possibility of a new 
or different kind of accident from any accident previously 
evaluated. No new accident causal mechanisms will be created as a 
result of the relocation of the affected Technical Specification 
requirements and their Bases to the Hope Creek UFSAR. Plant 
operation will not be affected by the proposed amendments and no new 
failure modes will be created. Therefore, the proposed changes do 
not create the possibility of a new or different kind of accident 
from any previously evaluated.
    3. The proposed change does not involve a significant reduction 
in a margin of safety.
    The proposed amendments will not involve a reduction in the 
margin of safety. Relocation of the affected Technical Specification 
requirements to the Hope Creek UFSAR is consistent with NUREG 1433, 
[``]Standard Technical Specifications, General Electric Plants, BWR/
4,[''] Revision 1, dated April 1995, and with the NRC's Final Policy 
Statement on Technical Specifications Improvements for Nuclear Power 
Reactors (58 FR 39132), dated July 22, 1993, which encourages 
utilities to propose amendments consistent with NUREG 1433. The 
margin of safety is unchanged; 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: Jeffrie J. Keenan, Esquire, Nuclear Business 
Unit--N21,

[[Page 27178]]

P.O. Box 236, Hancocks Bridge, NJ 08038.
    NRC Section Chief: James W. Clifford.

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

    Date of amendment request: April 17, 2001.
    Description of amendment request: The amendment would remove 
unnecessary details for certain secondary post-accident monitoring 
instrumentation from Technical Specification Table 3.2.6.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration which is presented below:

    1. Will the proposed changes involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    The post accident monitoring (PAM) instrumentation is not 
considered as an initiator or contributor to any previously 
evaluated accident. The proposed change will not impact the ability 
of the PAM instrumentation to perform its intended function, nor 
does it impact any Final Safety Analysis Report safety analysis. 
Therefore, the proposed change will not increase the probability of 
any accident previously evaluated.
    Additionally, while the PAM instrumentation provides information 
to the control room operator that may be used to mitigate an 
accident, this change does not affect the ability of the PAM 
instrumentation to perform this function. This change does not 
modify any parameters of previously analyzed events.
    Therefore, the proposed change will not increase the 
consequences of any accident previously evaluated.
    2. Will the proposed changes create the possibility of a new or 
different kind of accident from any accident previously evaluated?
    The proposed change does not involve any physical modification 
to the plant, change in Technical Specification setpoints, plant 
operation, or design basis of the plant. The PAM instrumentation 
provides information to the plant operator to assist in the 
mitigation of an accident, and the means for accomplishment of this 
function are unchanged. Under the proposed change, operability of 
the PAM instrumentation is not impacted. Therefore, this change does 
not create the possibility of a new or different kind of accident 
from any accident previously evaluated.
    3. Will the proposed changes involve a significant reduction in 
a margin of safety?
    The proposed change would delete instrument identification 
numbers and instrument ranges from Technical Specifications for 
certain PAM instrumentation. These details are not necessary to 
ensure the PAM instrumentation is maintained operable. The 
requirements of Technical Specification Limiting Condition for 
Operation and associated Surveillance Requirements are adequate to 
ensure the required instrumentation is maintained operable. The 
proposed change will not impact the ability of the PAM 
instrumentation to perform its intended function. Therefore, this 
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: Mr. David R. Lewis, Shaw, Pittman, Potts and 
Trowbridge, 2300 N Street, NW., Washington, DC 20037-1128.
    NRC Section Chief: James W. Clifford.

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

    Date of amendment request: March 22, 2001 (ET 01-0007).
    Description of amendment request: The amendment would delete the 
inequality () in front of the allowed temperature value and 
increase the allowed methyl iodide penetration values in item c of 
Technical Specification 5.5.11, ``Ventilation Filter Testing Program 
(VFTP),'' for the engineered safety feature (ESF) control room 
emergency ventilation system and auxiliary/fuel building emergency 
exhaust ventilation system.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. The proposed change does not involve a significant increase 
in the probability or consequences of an accident previously 
evaluated.
    The proposed changes revise the allowable methyl iodide 
penetration percent for the carbon in the Control Room Emergency 
Ventilation System and the Auxiliary/Fuel Building Emergency Exhaust 
System when tested in accordance with ASTM D3803-1989. The proposed 
change is based on the values that would be derived using a safety 
factor of 2 between credited and tested carbon efficiencies. This 
use of a safety factor of 2 is discussed in Generic Letter 99-02. 
Generic Letter 99-02 allows the reduction of the safety factor 
between the credited and tested carbon efficiencies from 5 (for 
systems with heaters) and 7 (for systems without heaters) to 2 (for 
systems with or without heaters) when tested in accordance with ASTM 
D3803-1989. Analyses of design-basis accidents assume a particular 
charcoal filter adsorption efficiency when calculating offsite and 
control room operator doses. A test of the charcoal filter samples 
determines whether the filter adsorber efficiency is greater than 
that assumed in the design-basis accident analysis. The laboratory 
test acceptance criteria contain a safety factor to ensure that the 
efficiency assumed in the accident analysis is still valid at the 
end of the operating cycle. Because ASTM D3803-1989 is a more 
accurate and demanding test, the use of a safety factor of 2 
provides an acceptable adsorption efficiency greater than that 
assumed in the safety analysis.
    Therefore, the proposed changes do not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    2. The proposed change does not create the possibility of a new 
or different kind of accident from any accident previously 
evaluated.
    The proposed changes revise the allowable methyl iodide 
penetration percent for the carbon in the Control Room Emergency 
Ventilation System and the Auxiliary/Fuel Building Emergency Exhaust 
System when tested in accordance with ASTM D3803-1989. The change in 
the allowable methyl iodide penetration percent is based on the 
values that would be derived using a safety factor of 2 as provided 
in Generic Letter 99-02. Generic Letter 99-02 allows the reduction 
of the safety factor between the credited and tested carbon 
efficiencies from 5 (for systems with heaters) and 7 (for systems 
without heaters) to 2 (for systems with or without heaters) when 
tested in accordance with ASTM D 3803-1989. No new or different 
accident scenarios, transient precursors, failure mechanisms, or 
limiting single failures will be introduced as a result of using a 
safety factor of 2 and deletion of the inequality sign associated 
with the temperature at which testing occurs.
    Therefore, the proposed changes do not create a new or different 
kind of accident from any accident previously evaluated.
    3. The proposed change does not involve a significant reduction 
in a margin of safety.
    The charcoal adsorber sample laboratory testing protocol 
accurately demonstrates the required performance of the adsorbers in 
the Control Room Emergency Ventilation System and the Auxiliary 
Building Emergency Exhaust System following a design basis accident 
or in the Fuel Building Emergency Exhaust System following a fuel 
handling accident. The change in safety factor and deletion of the 
inequality sign associated with the temperature at which testing 
occurs will not affect system performance or operation. This use 
[of] a safety factor of 2 is discussed in Generic Letter 99-02. 
Generic Letter 99-02 allows the reduction of the safety factor 
between the credited and tested carbon efficiencies from 5 (for 
systems with heaters) and 7 (for systems without heaters) to 2 (for 
systems with or without heaters) when tested in accordance with ASTM 
D3803-1989. Analyses of design-basis accidents assume a particular 
charcoal filter adsorption efficiency when calculating offsite

[[Page 27179]]

and control room operator doses. A test of the charcoal filter 
samples determines whether the filter adsorber efficiency is greater 
than that assumed in the design-basis accident analysis. The 
laboratory test acceptance criteria contain a safety factor to 
ensure that the efficiency assumed in the accident analysis is still 
valid at the end of the operating cycle. Because ASTM D3803-1989 is 
a more accurate and demanding test, the use of a safety factor of 2 
ensures the charcoal filter adsorption efficiency is greater than 
that assumed in the safety analysis when the penetration acceptance 
criterion is met. The offsite and control room dose analyses are not 
affected by this change and will remain within the limits of 10 CFR 
100 and 10 CFR 50, Appendix A.
    Therefore, the proposed changes do not involve a significant 
reduction in the margin of safety.

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

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 20852. 
Publicly available records will be accessible and electronically from 
the ADAMS Public Library component on the NRC Web site, http://www.nrc.gov (the Electronic Reading Room).

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

    Date of application for amendment: August 29, 2000.
    Brief description of amendment: The amendment revised the Technical 
Specifications to change the ``Administrative Controls'' section 
regarding certain position titles and the Shift Technical Advisor (STA) 
staffing requirement to allow one of the required on-shift Senior 
Reactor Operator (SRO) positions to be combined with the required STA 
position so as to serve in a dual SRO/STA position.
    Date of Issuance: April 27, 2001.
    Effective date: April 27, 2001 and shall be implemented within 30 
days of issuance.
    Amendment No.: 220
    Facilit Operating License No.DPR-16: Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: November 29, 2000 (65 
FR 71133).
    The Commission's related evaluation of this amendment is contained 
in a Safety Evaluation dated April 27, 2001.
    No significant hazards consideration comments received: No.

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

    Date of application for amendments: April 26, 2000, as supplemented 
November 6, 2000.
    Brief Description of amendments: The amendments change the 
Technical Specifications Ultimate Heat Sink maximum 24-hour average 
temperature from 89 degrees F to 90.5 degrees F and increase the lower 
temperature of the condition temperature range. In addition, they 
change a surveillance requirement to require verification that the 
temperature is less than or equal to 90.5 degrees F.
    Date of issuance: April 20, 2001.
    Effective date: April 20, 2001.
    Amendment Nos.: 213 and 240.
    Facility Operating License Nos. DPR-71 and DPR-62: Amendments 
change the Technical Specifications.
    Date of initial notice in Federal Register: May 17, 2000 (65 FR 
31356), superseded on March 21, 2001 (66 FR 15916). The November 6, 
2000, supplement contained additional information that expanded the 
scope of the initial application. Subsequently, the supplemented 
application was renoticed.
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated April 20, 2001.
    No significant hazards consideration comments received: No.

Consumers Energy Company, Docket No. 50-255, Palisades Plant, Van Buren 
County, Michigan

    Date of application for amendment: December 7, 2000.
    Brief description of amendment: The amendment changes the Technical 
Specifications regarding the Limiting Conditions for Operation for the 
containment cooling systems, the component cooling water system, and 
the service water system to be similar to changes to the ``Standard 
Technical Specifications, Combustion Engineering Plants,'' NUREG-1432, 
Revision 1, made by the Nuclear Energy Institute Technical 
Specifications Task Force Change Number 325, ``ECCS Conditions and 
Required Actions with  100% Equivalent ECCS Flow.''
    Date of issuance: May 3, 2001.
    Effective date: As of the date of issuance and shall be implemented 
within 90 days.
    Amendment No.: 199.
    Facility Operating License No. DPR-20. Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: January 24, 2001 (66 FR 
7677).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated May 3, 2001.
    No significant hazards consideration comments received: No.

Consumers Energy Company, Docket No. 50-255, Palisades Plant, Van Buren 
County, Michigan

    Date of application for amendment: December 7, 2000.
    Brief description of amendment: The amendment changes Technical

[[Page 27180]]

Specification (TS) 3.7.5 regarding the Limiting Conditions for 
Operation (LCOs) for the auxiliary feedwater system to be similar to 
changes to the ``Standard Technical Specifications, Combustion 
Engineering Plants,'' NUREG 1432, Revision 1, made by the Nuclear 
Energy Institute Technical Specifications Task Force Change Number 325, 
Revision 0.
    Date of issuance: May 3, 2001.
    Effective date: As of the date of issuance and shall be implemented 
within 90 days.
    Amendment No.: 200.
    Facility Operating License No. DPR-20. Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: January 24, 2001 (66 FR 
7674).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated May 3, 2001.
    No significant hazards consideration comments received: No.

Energy Northwest, Docket No. 50-397, Columbia Generating Station 
(formerly known as WNP-2), Benton County, Washington

    Date of application for amendment: September 5, 2000, as 
supplemented December 14, 2000.
    Brief description of amendment: The amendment revised Technical 
Specification (TS) Tables 3.3.5.1-1, 3.3.6.1-1 and 3.3.6.2-1. The 
changes add notes to the tables listing instrument channels that are 
common to, or support the operability of interrelated systems as 
governed by these technical specifications. Specifically:
    (1) Added note ``(e)'' to Table 3.3.5.1-1, ``Emergency Core Cooling 
System Instrumentation,'' Functions 1c, 1d, 2c and 2d in the column 
entitled ``Required Channels Per Function,'' indicating the 
applicability of the new footnote which reads as follows: ``(e) Also 
supports OPERABILITY of 230 kV offsite power circuit pursuant to LCO 
3.8.1 and LCO 3.8.2.''
    (2) Added note ``(e)'' to Table 3.3.6.1-1, ``Primary Containment 
Isolation Instrumentation,'' in Functions 2b and 2c, in the column 
entitled ``Required Channels Per Trip System,'' indicating the 
applicability of the new footnote which reads as follows: ``(e) Also 
required to initiate the associated LOCA Time Delay Function pursuant 
to LCO 3.3.5.1.''
    (3) Added note ``(c)'' to Table 3.3.6.2-1, ``Secondary Containment 
Isolation Instrumentation,'' in Functions 1 and 2, in the column 
entitled ``Required Channels Per Trip System,'' indicating the 
applicability of the new footnote which reads as follows: ``(c) Also 
required to initiate the associated LOCA Time Delay Function pursuant 
to LCO 3.3.5.1.''
    Date of issuance: April 30, 2001.
    Effective date: April 30, 2001, to be implemented within 30 days of 
the date of issuance.
    Amendment No.: 172.
    Facility Operating License No. NPF-21: The amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: November 15, 2000 (65 
FR 69059).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated April 30, 2001.
    No significant hazards consideration comments received: No.

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

    Date of application for amendment: September 7, 2000, as 
supplemented on April 2, 2001.
    Brief description of amendment: The amendment revises Technical 
Specification 3.7.3 to reflect planned modifications to the main 
feedwater system.
    Date of issuance: April 18, 2001.
    Effective date: April 18, 2001.
    Amendment No.: 207.
    Facility Operating License No. DPR-64: Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: November 15, 2000 (65 
FR 69062).
    The April 2, 2001, submittal contained clarifying information only, 
and did not change the initial no significant hazards consideration 
determination.
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated April 18, 2001.
    No significant hazards consideration comments received: No.

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

    Date of application for amendment: August 16, 2000.
    Brief description of amendment: The amendment (1) removes the 
``Offgas Treatment System Explosive Gas Mixing Instrumentation'' 
Technical Specification (TS) 3.7 from the Radiological Effluent 
Technical Specifications contained in Appendix B and adds a reference 
to the Offgas Treatment System Explosive Gas Monitoring Program to 
Administrative Section 6 of the TSs contained in Appendix A; (2) 
replaces the position title of Radiological and Environmental Services 
Manager contained in the Administrative Section 6 of Appendix A with 
Radiation Protection Manager; and (3) revises Plant Staff organization 
requirements contained in Administrative Section 6 to require either 
the Operations Manager or the Assistant Operations Manager to hold a 
Senior Reactor Operator license.
    Date of issuance: April 18, 2001.
    Effective date: As of the date of issuance to be implemented within 
30 days.
    Amendment No.: 270.
    Facility Operating License No. DPR-59: Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: September 20, 2000 (65 
FR 56956).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated April 18, 2001.
    No significant hazards consideration comments received: No.

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 application for amendment: November 10, 2000, as 
supplemented by letters dated February 15 and March 22, 2001.
    Brief description of amendment: The amendment changes the safety 
limit minimum critical power ratio (SLMCPR) for Cycle 12 operation with 
a mixed core of Siemens Power Corporation (now known as Framatome ANP 
Richland, Inc.) ATRIUM-10 reload fuel, and General Electric GE11 
reactor fuel. The amendment reflects a decrease of the two 
recirculation loop SLMCPR from 1.09 to 1.08, with the single 
recirculation loop SLMCPR remaining unchanged at 1.10. The amendment 
also revises Technical Specification 5.6.5 to update the list of 
references that are currently used to determine core operating limits.
    Date of issuance: April 26, 2001.
    Effective date: As of the date of issuance and shall be implemented 
within 30 days of the date of issuance.
    Amendment No: 146.
    Facility Operating License No. NPF-29: The amendment revises the 
Technical Specifications.
    Date of initial notice in Federal Register: December 27, 2000 (65 
FR 81917).
    The Commission's related evaluation of the amendment is contained 
in a

[[Page 27181]]

Safety Evaluation dated April 26, 2001. The February 15 and March 22, 
2001, supplements did not change the scope of the amendment or the 
original proposed no significant hazards consideration determination.
    No significant hazards consideration comments received: No.

Exelon Generation Company, LLC, Docket Nos. STN 50-454 and STN 50-455, 
Byron Station, Unit Nos. 1 and 2, Ogle County, Illinois

Docket Nos. STN 50-456 and STN 50-457, Braidwood Station, Unit Nos. 1 
and 2, Will County, Illinois

    Date of application for amendments: October 24, 2000, as 
supplemented on March 26, 2001.
    Brief description of amendments: The amendments revise the 
technical specifications to reference the generically approved 
Westinghouse Best-Estimate large break loss of coolant accident 
methodology for the plants.
    Date of issuance: April 18, 2001.
    Effective date: April 18, 2001.
    Amendment Nos.: 118, 118, 112 and 112
    Facility Operating License Nos. NPF-37, NPF-66, NPF-72 and NPF-77: 
The amendments revise the Technical Specifications.
    Date of initial notice in Federal Register: February 21, 2001 (66 
FR 11052).
    The March 26, 2001, letter provided clarifying information that did 
not change the scope of October 24, 2000, application or the initial no 
significant hazards consideration determination.
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated April 18, 2001.
    No significant hazards consideration comments received: No.

Exelon Generation Company, LLC, Docket Nos. STN 50-454 and STN 50-455, 
Byron Station, Unit Nos. 1 and 2, Ogle County, Illinois

Docket Nos. STN 50-456 and STN 50-457, Braidwood Station, Unit Nos. 1 
and 2, Will County, Illinois

    Date of application for amendments: July 5, 2000, as supplemented 
by letters dated November 27, 2000, December 21, 2000, January 31, 
2001, February 20, 2001, February 28, 2001, March 26, 2001, April 5, 
2001, and April 16, 2001.
    Brief description of amendments: The amendments revise the licenses 
and technical specifications to reflect approval of an increase in 
maximum thermal power from 3411 megawatts-thermal (MWt) to 3586.6 MWt.
    Date of issuance: May 4, 2001.
    Effective date: May 4, 2001.
    Amendment Nos.: 119, 119, 113 and 113.
    Facility Operating License Nos. NPF-37, NPF-66, NPF-72 and NPF-77: 
The amendments revised the License and the Technical Specifications.
    Date of initial notice in Federal Register: December 13, 2000.
    The supplements to the application provided clarifying information 
that did not change the scope of the July 5, 2000, application or the 
initial no significant hazards consideration determination.
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated May 4, 2001.
    No significant hazards consideration comments received: No.

Florida Power and Light Company, et al., Docket No. 50-389, St. Lucie 
Plant, Unit No. 2, St. Lucie County, Florida

    Date of application for amendment: November 17, 1999, as 
supplemented June 14, November 13, and December 4, 2000, and February 
21, 2001.
    Brief description of amendment: Increased the allowed outage time 
to restore an inoperable Emergency Diesel Generator to operable status 
from 72 hours to 14 days.
    Date of Issuance: April 26, 2001.
    Effective Date: Date of issuance, to be implemented within 60 days.
    Amendment No.: 115.
    Facility Operating License No. NPF-16: Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: December 15, 1999 (64 
FR 70089). The June 14, November 13, and December 4, 2000, and February 
21, 2001, supplements did not affect the original proposed no 
significant hazards determination, or expand the scope of the request 
as noticed in the Federal Register.
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated April 26, 2001.
    No significant hazards consideration comments received: No.

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

    Date of application for amendments: October 30, 2000, as 
supplemented February 28, 2001.
    Brief description of amendments: The amendments revised Technical 
Specification 5.3.2 for Turkey Point Units 3 and 4 to extend the 
residual heat removal (RHR) pump allowed outage time (AOT) from 72 
hours to 7 days to restore an inoperable RHR pump to operable status.
    Date of issuance: April 25, 2001.
    Effective date: Effective as of date of issuance, to be implemented 
within 60 days.
    Amendment Nos. 212 and 206.
    Facility Operating License Nos. DPR-31 and DPR-41: Amendments 
revised the Technical Specifications.
    Date of initial notice in  Federal Register: December 27, 2000 (65 
FR 81922). The February 28, 2001, supplemental letter provided 
clarifying information which did not change the proposed no significant 
hazards consideration determination or expand the scope of the request 
as noticed.
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated April 25, 2001.
    No significant hazards consideration comments received: No.

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

    Date of application for amendments: December 6, 2000.
    Brief description of amendments: These amendments revised Technical 
Specification Surveillance Requirement 4.6.1.3.c to extend the interval 
for testing the containment air lock interlock mechanisms from 6 months 
to 24 months. Additionally, the amendments corrected an unrelated 
administrative error in TS Table 3.3-2, ``Engineered Safety Features 
Actuation System Instrumentation.''
    Date of issuance: April 26, 2001.
    Effective date: April 26, 2001.
    Amendment Nos. 213 and 207.
    Facility Operating License Nos. DPR-31 and DPR-41: Amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: February 7, 2001 (66 FR 
9385).
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated April 26, 2001.
    No significant hazards consideration comments received: No.

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

    Date of application for amendment: October 16, 2000, as 
supplemented December 22, 2000.
    Brief description of amendment: The amendment revises the TSs to 
incorporate new pressure and temperature (P-T) limit curves. The 
reactor pressure vessel P-T limit curves are updated for inservice 
leakage and hydrostatic testing, non-nuclear heatup and cooldown, and 
criticality. The revised P-T limit curves are approved for an interim 
period not to exceed September 1, 2003.

[[Page 27182]]

    Date of issuance: April 30, 2001.
    Effective date: As of the date of issuance and shall be implemented 
within 30 days.
    Amendment No.: 238.
    Facility Operating License No. DPR-49: The amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: December 13, 2000 (65 
FR 77921).
    The December 22, 2000, letter was within the scope of the original 
Federal Register notice and did not change the staff's initial proposed 
no significant hazards consideration determination.
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated April 30, 2001.
    No significant hazards consideration comments received: No.

Pacific Gas and Electric Company, Docket No. 50-323, Diablo Canyon 
Nuclear Power Plant, Unit No. 2, San Luis Obispo County, California

    Date of application for amendment: June 2, 2000, as supplemented by 
letters dated December 15, 2000, and February 14, 2001.
    Brief description of amendments: The amendment revised Technical 
Specification (TS) Section 3.5.2, ``ECCS--Operating,'' Action A to 
allow a one-time increase in the allowed outage time for centrifugal 
charging pump (CCP) 2-1 during Unit 2's Cycle 10 from 72 hours to 7 
days. This change will allow for a potential on-line repair or a 
potential replacement of CCP 2-1. This pump is currently experiencing 
elevated vibration levels due to a structural resonance in the outboard 
bearing support structure and has been on an increased testing 
frequency since May 1996 due to high vibration.
    Date of issuance: April 20, 2001.
    Effective date: April 20, 2001, and shall be implemented within 30 
days from the date of issuance.
    Amendment No.: 146.
    Facility Operating License No. DPR-82: The amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: July 12, 2000 (65 FR 
43051).
    The December 15, 2000, and February 14, 2001, supplemental letters 
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 dated April 20, 2001.
    No significant hazards consideration comments received: No.

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

    Date of application for amendments: November 30, 2000.
    Brief description of amendments: The license amendments revised 
Technical Specifications (TS) Section TS 3.5.1, ``Accumulators'' to (1) 
reflect the values of the accumulator pressure and volume consistent 
with the analyses assumptions documented in the current DCPP Final 
Safety Analysis Report (FSAR) Update, and (2) align the DCPP TS with 
the standard TS for Westinghouse plants.
    Date of issuance: May 3, 2001.
    Effective date: May 3, 2001, and shall be implemented within 30 
days from the date of issuance.
    Amendment Nos.: Unit 1-147; Unit 2-147.
    Facility Operating License Nos. DPR-80 and DPR-82: The amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: December 27, 2000 (65 
FR 81928).
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated May 3, 2001.
    No significant hazards consideration comments received: No.

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

    Date of application for amendment: March 9, 2001.
    Brief description of amendment: The amendment revises the Technical 
Specifications only until the Fall 2001 refueling outage and allows 
Mode 2 (startup) operation with two required intermediate range monitor 
channels per trip system.
    Date of issuance: April 27, 2001.
    Effective date: As of the date of issuance and shall be implemented 
within 30 days from the date of issuance.
    Amendment No.: 166.
    Facility Operating License No. NPF-5: Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: March 20, 2001 (66 FR 
15768).
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated April 27, 2001.
    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: March 17, 2000.
    Brief description of amendments: The amendments revise Technical 
Specification (TS) 3/4.7.4, ``Essential Cooling Water System,'' to 
delete Surveillance Requirement 4.7.b.3, and to change Surveillance 
Requirements 4.7.4.b.1 and 4.7.4.b.2 to incorporate the wording from 
the Standard Technical Specifications for Westinghouse Plants (NUREG-
1431). Surveillance Requirement 4.7.4.b.3 requires verifying at least 
once per 18 months that each screen wash booster pump and the traveling 
screen start automatically on a Safety Injection test signal. The Bases 
for TS 3/4.7.4 were also changed.
    Date of issuance: April 30, 2001.
    Effective date: The amendments are effective as of the date of 
their issuance.
    Amendment Nos.: Unit 1--126; Unit 2--115.
    Facility Operating License Nos. NPF-76 and NPF-80: The amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: April 19, 2000 (65 FR 
21039).
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated April 30, 2001.
    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: November 18, 1999, as supplemented by 
letters dated November 29, 1999, and November 22, 2000.
    Brief description of amendments: Changes to Technical 
Specifications surveillance testing to satisfy the actions requested in 
Generic Letter 99-02. The November 29, 1999, and November 22, 2000, 
letters provided additional clarifying information that was within the 
scope of the original application and Federal Register notice and did 
not change the NRC staff's initial proposed no significant hazards 
consideration determination.
    Date of issuance: May 1, 2001.
    Effective date: As of the date of issuance and shall be implemented 
within 30 days from the date of issuance.
    Amendment Nos.: Unit 1--127; Unit 2--116.
    Facility Operating License Nos. NPF-76 and NPF-80: The amendments 
revised the Technical Specifications.

[[Page 27183]]

    Date of initial notice in Federal Register: December 29, 1999 (64 
FR 73099).
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated May 1, 2001.
    No significant hazards consideration comments received: No.

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

    Date of amendment request: May 17, 2000, as supplemented by letters 
dated August 31, 2000, and January 31, 2001.
    Brief description of amendments: The amendments revise the 
Allowable Values specified in Technical Specification (TS) Table 3.3.5-
1, ``Loss of Power (LOP) Diesel Generator (DG) Start Instrumentation'' 
to ensure that the 6.9 kiloVolt and 480 Volt undervoltage relays 
initiate the necessary actions when required. In addition, a change is 
made to Condition D of TS 3.3.5, ``Loss of Power (LOP) Diesel Generator 
(DG) Start Instrumentation,'' to eliminate the term ``undervoltage.'' 
This change is consistent with a change to TS Table 3.3.5-1.
    Date of issuance: April 20, 2001.
    Effective date: As of the date of issuance and shall be implemented 
within 30 days from the date of issuance.
    Amendment Nos.: 85 and 85.
    Facility Operating License Nos. NPF-87 and NPF-89: The amendments 
revised the Technical Specifications.
    Date of initial notice in  Federal Register: March 21, 2001 (66 FR 
15930).
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated April 20, 2001.
    No significant hazards consideration comments received: No.

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

    Date of application for amendment: November 30, 2000, as 
supplemented by letters dated March 8 and 12, 2001.
    Brief description of amendment: The changes revise the operability 
requirements for the refueling interlocks contained within TS 3.12.A as 
well as the surveillance requirements specified within 4.12.A. 
Clarifying changes are made to TS 3.12.D and 3.12.E to indicate that 
only the required interlocks need to be operable. In addition, TS 
3.12.F will be clarified to articulate that there must be a minimum of 
24 hours fission product decay prior to fuel handling. Some editorial 
changes were made in TS 3.12.B.
    Date of Issuance: April 20, 2001.
    Effective date: As of the date of issuance, and shall be 
implemented within 30 days.
    Amendment No.: 200.
    Facility Operating License No. DPR-28: Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: March 20, 2001 (66 FR 
15770).
    The Commission's related evaluation of this amendment is contained 
in a Safety Evaluation dated April 20, 2001.
    No significant hazards consideration comments received: No.

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

    Date of application for amendment: November 27, 2000.
    Brief description of amendment: The amendment eliminates the 
specifications associated with the 24 Vdc Emergency Core Cooling System 
(ECCS) instrumentation batteries and chargers. The 24 Vdc ECCS 
instrumentation loads will be transferred to the 125 Vdc main station 
batteries.
    Date of Issuance: April 20, 2001.
    Effective date: As of the date of issuance, and shall be 
implemented within 60 days.
    Amendment No.: 201.
    Facility Operating License No. DPR-28: Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: January 10, 2001 (66 FR 
2024).
    The Commission's related evaluation of this amendment is contained 
in a Safety Evaluation dated April 20, 2001.
    No significant hazards consideration comments received: No.

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

    Date of application for amendment: November 1, 2000.
    Brief description of amendment: The amendment revises the 
operability requirement for high pressure coolant injection (HPCI) and 
reactor core isolation cooling low steam line pressure isolation 
instrumentation to coincide with system operability requirements. The 
proposed change eliminates the need to open manual containment 
isolation valves under administrative control during reactor heatup, 
reduces the potential for operator error when closing these valves 
(potential for leaving valve mispositioned), and clarifies the steam 
line low pressure isolation function description. An administrative 
change to correct the HPCI High Steam Line d/p instrument component 
numbers was also made to ensure the accuracy of isolation 
instrumentation information.
    Date of Issuance: April 20, 2001.
    Effective date: As of the date of issuance, and shall be 
implemented within 60 days.
    Amendment No.: 202.
    Facility Operating License No. DPR-28: Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: December 13, 2000 (65 
FR 77928)
    The Commission's related evaluation of this amendment is contained 
in a Safety Evaluation dated April 20, 2001.
    No significant hazards consideration comments received: No.

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

    Date of application for amendment: December 19, 2000 as 
supplemented on February 13 and 23, 2001, and March 29, 2001.
    Brief description of amendment: The amendment revises the Technical 
Specifications (TSs) by changing the reactor vessel pressure/
temperature limit curves specified in TS 3.6.A.1, ``Reactor Coolant 
Systems--Pressure and Temperature Limitations,'' as graphically 
represented in Figures 3.6.1, Hydrostatic Pressure and Leak Tests, Core 
Not Critical, Figure 3.6.2, Normal Operation, Core Not Critical, and 
3.6.3, Normal Operation/Core Critical.
    Date of Issuance: May 4, 2001.
    Effective date: As of the date of issuance, and shall be 
implemented within 60 days.
    Amendment No.: 203.
    Facility Operating License No. DPR-28: Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: January 24, 2001 (66 FR 
7687).
    The February 13 and 23, and March 29, 2001, supplements provided 
clarifying information that did not expand the scope of the application 
as published in the Federal Register, or change the proposed no 
significant consideration determination.
    The Commission's related evaluation of this amendment is contained 
in a Safety Evaluation dated May 4, 2001.
    No significant hazards consideration comments received: No.

[[Page 27184]]

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

    Date of application for amendment: June 22, 2000, as supplemented 
September 19, 2000, and January 4, February 14, March 13, March 22, and 
April 11, 2001.
    Brief description of amendment: These amendments revise Technical 
Specification (TS) Figures 3.4-2 and 3.4-3, and the associated Bases. 
These amendments approve new pressure-temperature limits, low-
temperature overpressure protection (LTOP) system setpoints, and LTOP 
system effective temperature (Tenable) in the TS to a 
maximum of 32.3 effective full-power years (EFPY) for Unit 1 and 34.3 
EFPY for Unit 2. These changes were based, in part, on the use of the 
American Society of Mechanical Engineers Code Case N-641.
    Date of issuance: May 2, 2001.
    Effective date: As of the date of issuance, to be implemented 
within 30 days.
    Amendment Nos.: 226 and 207.
    Facility Operating License Nos. NPF-4 and NPF-7: Amendments change 
the TS.
    Date of initial notice in Federal Register: February 23, 2001 (66 
FR 11334). The January 4, 2001, submittal expanded the scope of the 
original June 22, 2000, application, which was noticed at 65 FR 48760. 
The February 14, March 13, March 22, and April 11, 2001, supplements 
contained clarifying information only, and did not change the February 
23, 2001, initial no significant hazards consideration determination or 
expand the scope of the Federal Register notice.
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated May 2, 2001.
    No significant hazards consideration comments received: No.

    Dated at Rockville, Maryland this 8th day of May 2001.

    For the Nuclear Regulatory Commission.
John A. Zwolinski,
 Director, Division of Licensing Project Management, Office of Nuclear 
Reactor Regulation.
[FR Doc. 01-12192 Filed 5-15-01; 8:45 am]
BILLING CODE 7590-01-P