[Federal Register Volume 66, Number 75 (Wednesday, April 18, 2001)]
[Notices]
[Pages 19998-20018]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: 01-9320]


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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 26 through April 6, 2001. The last 
biweekly notice was published on April 4, 2001 (66 FR 17962).

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 
Review and Directives Branch, Division of Freedom 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

[[Page 19999]]

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 May 18, 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 
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).

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

    Date of amendment request: March 1, 2001.
    Description of amendment request: The proposed amendment would 
revise the Technical Specification (TS) on page 4.5-3 to change the 
frequency of closure time testing of the main steam isolation valves 
(MSIVs). If approved, these tests would no longer occur during power 
operation. They would be conducted during each cold shutdown unless 
this test has been performed within the last 92 days.
    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 amendment does not:
    (1) Involve a significant increase in the probability or 
consequences of an accident previously evaluated.
    The proposed change revises Technical Specification 4.5.F.3 to 
require MSIV full-stroke testing during each cold shutdown rather 
than quarterly at power. Since this change only affects the 
frequency of testing

[[Page 20000]]

the isolation time of MSIVs, it does not impact the occurrence of 
accidents that the MSIVs are designed to mitigate. The 10% closure 
test that will be performed quarterly in order to test the MSIV 
closure scram instrumentation has some potential of causing an 
inadvertent closure of the MSIV. The current test of MSIV closure 
scram instrumentation is conducted with the quarterly full closure 
test and is performed at reduced power. The closure of an MSIV at 
the reduced power level does not result in a plant trip. Inadvertent 
closure of MSIVs is a transient of moderate frequency evaluated in 
the updated FSAR [final safety analysis report]. The small increase 
in potential for an MSIV full closure transient during the part-
stroke test is offset by the decrease in potential transients due to 
the plant power manipulation necessary to perform the full closure 
test.
    The proposed change affects the frequency of testing the MSIVs 
to ensure an acceptable level of reliability. Aligning the Oyster 
Creek test frequency for MSIVs with the ASME Code [American Society 
of Mechanical Engineers Boiler and Pressure Vessel Code] and 
industry practice assures adequate reliability for valve closure. 
Therefore, the MSIVs will be capable of closing to mitigate 
accidents.
    As a result of the discussion above, the change to the frequency 
of MSIV full closure testing does not involve a significant increase 
in the probability or consequences of an accident previously 
evaluated.
    (2) Create the possibility of a new or different kind of 
accident from any accident previously evaluated.
    There is no physical change in plant configuration associated 
with performing the MSIV full or partial closure tests. The MSIV 
closure scram is designed to anticipate the transient caused by 
valve closure with the plant in operation. Therefore, the proposed 
change does not create the possibility of a new or different kind of 
accident from any accident previously evaluated.
    (3) Involve a significant reduction in a margin of safety.
    The proposed change affects the method of assuring the 
reliability of the MSIVs. The change from a quarterly full-stroke 
closure test at power to full-stroke tests during cold shutdowns 
combined with quarterly part-stroke tests to ensure instrument 
function provides adequate means of assuring MSIV operability. The 
reliability of MSIVs to close within the required 3-10 seconds has 
been consistently demonstrated and it is expected that the valves 
will continue to pass this test when done on a cold shutdown basis. 
The quarterly 10% closure reactor protection system testing will 
assure that the valves will respond to a closure signal.
    Presently, the MSIVs are full-stroke closed quarterly at power 
in accordance with Technical Specification 4.5.F.3. The basis for 
the current quarterly full closure test at power and the proposed 
full closure test during cold shutdowns with part-stroking quarterly 
during instrument surveillance is consistent with the ASME Boiler 
and Pressure Vessel Code, Section XI. In addition, the proposed 
change is consistent with industry standard requirements contained 
in the Standard Technical Specifications, NUREG-1433, Revision 1. 
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: Kevin P. Gallen, Morgan, Lewis & Bockius, 
LLP, 1800 M Street, NW., Washington, DC 20036-5869.
    NRC Section Chief: Marsha Gamberoni.

Detroit Edison Company, Docket No. 50-16, Enrico Fermi Atomic Power 
Plant, Unit 1 (Fermi1), Monroe County, Michigan.

    Date of amendment request: November 6, 2000, (Reference NRC-00-
0084) and supplement by letter dated March 12, 2001, (Reference NRC-01-
0026).
    Description of amendment request: The proposed amendment will 
revise the Technical Specifications by: (1) deleting Specification A.1, 
the definition of Physical Barrier; (2) deleting Specification A.2, the 
definition of Protected Area; (3) deleting Specification A.4, the 
definition of Authorized Person; (4) Specification A.7.a, change the 
words ``Protected Area'' to ``facility''; (5) Specification B.1, delete 
the discussion on method for controlling facility access and add words 
noting that the method for controlling access to the facility will be 
included in the Fermi 1 Safety Analysis Report; (6) deleting 
Specification C.1, Reactor Building Access which specifies access 
limitations to this building; (7) Specification C.2, change the words 
``Protected Area'' to ``facility''; (8) deleting Specification E.1, 
Fuel and Repair Building which specifies access limitation to this 
building; (9) Specification H.4.a, change the words ``Protected Area'' 
to ``facility''; (10) Specification H.4.b, change the words ``Protected 
Area'' to ``facility''; (11) Specification I.6, deleting specific 
dosimetry requirements and replacing with a requirement that dosimetry 
will meet the requirements of 10 CFR Part 20; (12) Specification I.9.i, 
change the words ``Protected Area'' to ``facility''; and (13) Revise 
Figure B-1 to remove reference to the Protected Area Boundary and 
indication of structures that may be physically removed during the 
decommissioning process. In addition to these specific changes, the 
licensee will repaginate the Technical Specification. The above-listed 
are to support moving the licensee's program for controlling access to 
the Fermi 1 facility from the Fermi 1 license to the Fermi 1 Safety 
Analysis Report. This action would provide flexibility for the 
performance of decommissioning activities while maintaining controls 
commensurate with the small quantity of licensed material at Fermi 1 
and its status.
    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 using the standards in 10 CFR 50.92(c). The licensee's 
analysis is presented below:

    1. The proposed changes do not involve a significant increase in 
the probability or consequences of an accident.
    The changes to requirements regarding access to the facility and 
security do not affect the operation of any system. The requirements 
for having access control in the Fermi 1 Safety Analysis Report 
(F1SAR) will ensure people are aware when they are entering an area 
at Fermi 1 to which it is determined access should be controlled. 
The two analyzed accidents involved release of the activity in the 
liquid waste system and release of the activity in the residual 
sodium. These changes will not significantly increase the 
probability of these accidents since they do not affect operations 
of these systems. The possibility of an intruder entering the 
facility may slightly increase depending on the type of access 
controls implemented in the future which will be specified in the 
F1SAR. For example, if gates and doors are not closed and locked, 
but signs posted requiring permission for entry, the possibility of 
an unauthorized person entering the facility could increase. But, if 
such an intruder is intent on entering, he or she could do so under 
current access controls using common tools or equipment. The locks 
and barriers currently act as a reminder that the area is 
controlled. Future access control provisions will still provide that 
reminder.
    Allowing the Protected Area to be different from the Restricted 
Area required by 10 CFR 20 will not increase the probability of an 
accident. The requirement for limiting access to a restricted area 
or areas will still be required by 10 CFR 20. This requirement is 
for personnel protection and is unrelated to accident probability.
    Allowing the facility and the Protected Area required by 10 CFR 
Part 20 will not increase the probability of an accident. The 
requirement for limiting access to a restricted area or areas will 
still be required by 10 CFR Part 20. The requirement is for 
personnel protection and is unrelated to accident probability.
    The change deleting the specific requirements for dosimetry will 
not affect the probability of an accident since they apply to 
monitoring of personnel not control or operation of the facility. 
The requirements for monitoring are in 10 CFR Part 20 and will be 
referenced in the Technical Specifications.

[[Page 20001]]

    The radiological consequences of an accident will not be 
increased by the requested changes because the changes do not add 
radioactive material to the facility and the accidents analyses 
already assume release of all the activity in the primary sodium and 
liquid waste systems.
    2. The proposed changes do not create the possibility of a new 
or different accident from any previously evaluated.
    The requested changes do not change the method of operation of 
any system and so cannot create the possibility of a new or 
different accident. The analyzed accidents include release of all 
the radioactive material in the liquid waste system and release of 
all the radioactive material in the residual sodium remaining in the 
Protected Area which currently is the same as the licensed facility. 
Changing access, security, and boundary requirements cannot create a 
different type of accident.
    3. The proposed changes do not involve a significant reduction 
in a margin of safety.
    The proposed changes could slightly reduce the margin of safety, 
but only from the perspective of making it easier for an 
unauthorized individual to enter the facility or Protected Area. A 
determined unauthorized person could violate existing requirements 
or proposed requirements to gain entry. However, the Technical 
Specifications will still require access control requirements for 
portions of the facility. Access control requirements described in 
the F1SAR will ensure that personnel know the Fermi 1 Protected Area 
is a controlled access area and will prevent anyone from unknowingly 
entering the portions of the facility for which access control is 
determined appropriate per the F1SAR. There is a limited amount of 
radioactive material remaining at Fermi 1. The requirement for a 
restricted area or areas, as appropriate, will still apply per 10 
CFR Part 20 and is for individual protection not security. The 
proposed security measures are commensurate with the amount of 
radioactive material present in that neither 10 CFR Part 30 or 10 
CFR Part 70 established a security requirement for the amount of 
radioactive material at Fermi 1.
    Removing some buildings from Figure B-1 will not reduce the 
margin of safety. The figure will still fulfill the purpose of 
showing the facility boundary. The figure will still serve that 
purpose even if a building, structure, or barrier is modified or 
removed, since the provisions allowing changes requires that the 
boundary continues to encompass the area to which access is 
controlled. The accident analyses do not credit any building as a 
containment or as retaining any radioactive material during a 
possible accident. For the above reasons, the requested changes do 
not involve a significant reduction in margin of safety of Fermi 1.

    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, NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: John Flynn, Esquire, Detroit Edison Company, 
2000 Second Avenue, Detroit, Michigan 48226.
    NRC Branch Chief: Larry W. Camper.

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

    Date of amendment request: January 24, 2001.
    Description of amendment request: The amendment request proposes 
changes to the Technical Specifications (TSs) concerning certain 
operational conditions required when conducting core alterations or 
handling irradiated fuel in the primary containment. In addition, the 
licensee proposes to delete license condition 2.C.(17) and make certain 
editorial corrections.
    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 relaxation of TS OPERABILITY requirements for containment 
and control room ventilation systems during specific shutdown 
conditions do not affect the probability of any accident previously 
evaluated and do not alter current accident analyses consequences. 
During plant shutdown, these systems and structures are accident 
mitigating features for the postulated Fuel Handling Accident (FHA) 
and are not considered the initiator to any previously analyzed 
accident. They need not be required during CORE ALTERATIONS because 
the only accident postulated to result in significant fuel damage 
and radiation release during shutdown conditions is the FHA. The 
control room filtration, inlet radiation detection, and the air 
conditioning systems will continue to be required during the 
handling of any irradiated fuel assembly and during operations with 
the potential for draining the reactor vessel (OPDRVs). The 
containment will only be required during OPDRVs and when moving 
recently irradiated fuel assemblies. The current FHA analysis of 
record (approved by Amendment 110) assumes the containment is open 
after the irradiated fuel has undergone a sufficient decay period 
(i.e., has not been part of a critical reactor core within the 
previous 11 days). The analysis demonstrates that the offsite doses 
remain well within the Standard Review Plan Guidelines (less than 
25% of the 10 CFR [Part] 100 limits) and the control room doses 
remain less than the criteria of 10 CFR [Part] 50, Appendix A, 
General Design Criterion 19.
    The proposed changes regarding the removal of the SGT [Standby 
Gas Treatment] system from the ``Primary Coolant Sources Outside 
Containment'' leakage control program does not affect the 
reliability or filtration efficiency of the SGT system. Current TS 
surveillances test filtration efficiency and secondary containment 
in-leakage. There are no unfiltered pathways to the suction of the 
fans that require leakage testing.
    Therefore, these changes do 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 involve any design changes or any 
new modes of system operation. The proposed TS changes allow certain 
functions to be inoperable during CORE ALTERATIONS and during the 
handling of irradiated fuel that has undergone a sufficient 
radiation decay period. However, these out-of-service configurations 
are consistent with current design basis analyses. The removal of 
the SGT system from the ``Primary Coolant Sources Outside 
Containment'' leakage control program does not affect reliability or 
efficiency of the filtration system or otherwise affect the ability 
of the system to perform its safety function.
    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 reduce the margin of safety, as 
defined by SRP [Standard Review Plan] 15.7.4 Rev 1. The only 
accident postulated to occur during shutdown that results in 
significant fuel damage and subsequent radiation release is the FHA. 
The offsite and control room doses due to a FHA with an open 
containment have previously been evaluated with conservative 
assumptions and that analysis is not affected by the proposed 
changes. The analysis demonstrates that due to radioactive decay 
following reactor shutdown, the primary containment function is only 
required when handling recently irradiated fuel.
    The removal of the SGT system from the ``Primary Coolant Sources 
Outside Containment'' leakage control program does not affect 
reliability or efficiency of the filtration system or otherwise 
affect the ability of the system to perform its safety function.
    Therefore, this change does not involve a significant reduction 
in the margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: Mark Wetterhahn, Esq., Winston & Strawn,

[[Page 20002]]

1400 L Street, NW., Washington, DC 20005
    NRC Section Chief: Robert A. Gramm.

FirstEnergy Nuclear Operating Company, et al., Docket Nos. 50-334 and 
50-412, Beaver Valley Power Station, Unit Nos. 1 and 2 (BVPS-1 and 2), 
Shippingport, Pennsylvania

    Date of amendment request: December 27, 2000.
    Description of amendment request: The amendment request proposes 
various changes to the BVPS-1 and 2 technical specifications (TSs) and 
removal of a BVPS-1 license condition. The proposed BVPS-1 and 2 TS 
changes include (1) the revision of reactor trip system and engineered 
safety features actuation system instrumentation trip setpoints and 
allowable values; (2) the utilization of the Revised Thermal Design 
Procedure to generate additional departure from nucleate boiling (DNB) 
margin which facilitates revisions to the core safety limits, DNB 
parameters and Overtemperature and Overpower T trip setpoints; 
(3) the relocation of certain requirements from the TS to the core 
operating limit report; (4) the relocation of certain requirements from 
the TS to the Licensing Requirements Manual; and (5) miscellaneous 
changes that improve internal consistency of the BVPS TSs, simplify the 
presentation of requirements, provide clarifications, and improve 
consistency with the improved standard TSs. Changes to the TS Bases in 
support of the TS changes are also proposed. In addition, the deletion 
of BVPs-1 license condition regarding limitations on less than 3-loop 
operation is included in the amendment request.
    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. Revision to Setpoint and Allowable Values

    1. Does the change involve a significant increase in the 
probability or consequences of an accident previously evaluated?
    The RPS [reactor trip system] and ESFAS [engineered safety 
feature actuation system] trip functions are part of the accident 
mitigation response and are not themselves an initiator for any 
transient. Therefore, the probability of an accident previously 
evaluated is not significantly affected.
    This proposed amendment includes changes to RTS and ESFAS trip 
setpoints and allowable values that have been determined with the 
use of an approved methodology. The new values ensure that all 
automatic protective actions will be initiated at or before the 
condition assumed in the safety analysis. This change, which 
includes modification of the applicable Bases section(s), will allow 
the nominal trip setpoints to be adjusted within the calibration 
tolerance band allowed by the setpoint methodology. Plant operation 
with these revised values will not cause any design or analysis 
acceptance criteria to be exceeded. The structural and functional 
integrity of plant systems is unaffected. There will be no adverse 
effect on the ability of the channels to perform their safety 
functions as assumed in the safety analyses. Since there will be no 
adverse effect on the trip setpoints or the instrumentation 
associated with the trip setpoints, there will be no significant 
increase in the consequences of any accident previously evaluated.
    The proposed amendment does involve a hardware change. The 
hardware change involves the deletion of f(I) (BVPS Unit 
No. 1) and f2 (I) (BVPS Unit No. 2) for the Overpower 
T Trip Setpoint. This function is not modeled in the safety 
analysis nor included in the setpoint methodology calculation. 
Defeating this function, rather than leaving it in the equation with 
a setting of zero, eliminates the possibility that it will adversely 
contribute to the Overpower T Trip due to the limitations 
of the hardware and possible variations in the setpoint.
    Other changes in trip system function, content and format are 
proposed based on the current configuration of the trip system 
hardware at BVPS Unit No. 1. Similarly, since the ability of the 
instrumentation to perform its safety function is not adversely 
affected, there will be no significant increase in the consequences 
of any accident previously evaluated.
    The proposed editorial, administrative and format changes do not 
affect plant safety.
    Therefore, this change does not involve any 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 amendment includes changes to the format and 
magnitudes of nominal trip setpoints and allowable values that 
preserve all safety analysis assumptions related to accident 
mitigation. The protection system will continue to initiate the 
protective actions as assumed in the safety analysis. The proposed 
changes to Limiting Safety System Settings (LSSS) 2.2.1 and LCO 
3.3.2.1 will continue to ensure that the trip setpoints are 
maintained consistent with the setpoint methodology and the plant 
safety analysis. The proposed amendment does involve a hardware 
change. The hardware change involves the deletion of f(I) 
(BVPS Unit No. 1) and f2 (I) (BVPS Unit No. 2) for the 
Overpower T Trip Setpoint. This function is not modeled in 
the safety analysis nor included in the setpoint methodology 
calculation. Defeating this function, rather than leaving it in the 
equation with a setting of zero, eliminates the possibility that it 
will adversely contribute to the Overpower T Trip due to 
the limitations of the hardware and possible variations in the 
setpoint. Therefore, this hardware change does not create the 
possibility of a new or different kind of accident from any accident 
previously evaluated. Plant operation will not be changed.
    Other proposed changes are made so that the technical 
specifications more accurately reflect the plant-specific trip 
system hardware in BVPS Unit No. 1.
    Furthermore, the proposed changes do not alter the functioning 
of the RTS and ESFAS.
    Therefore, the proposed change does not create the possibility 
of a new or different kind of accident from any previously 
evaluated.
    3. Does the change involve a significant reduction in a margin 
of safety?
    The proposed RTS and ESFAS trip setpoints are calculated with an 
approved methodology. The proposed changes to LSSS 2.2.1 and LCO 
3.3.2.1 will continue to ensure that the trip setpoints are 
maintained consistent with the setpoint methodology and the plant 
safety analysis. Therefore, the response of the RTS and ESFAS to 
accident transients reported in the Updated Final Safety Analysis 
Report (UFSAR) is unaffected by this change. This proposed amendment 
does involve a hardware change. The hardware change involves the 
deletion of f(I) (BVPS Unit No. 1) and f2(I) (BVPS 
Unit No. 2) for the Overpower T Trip Setpoint. This 
function is not modeled in the safety analysis nor included in the 
setpoint methodology calculation. Defeating this function, rather 
than leaving it in the equation with a setting of zero, eliminates 
the possibility that it will adversely contribute to the Overpower 
T Trip due to the limitations of the hardware and possible 
variations in the setpoint. Therefore, accident analysis acceptance 
criteria are not affected. Other proposed changes are made so that 
the protection system technical specifications more accurately 
reflect the plant-specific trip system hardware in BVPS Unit No. 1.
    The proposed editorial, administrative, and format changes do 
not affect plant safety.
    Therefore, the proposed change does not involve a significant 
reduction in a margin of safety.

B. Revised Thermal Design Procedure (RTDP)--Overtemperature T 
and Overpower T

    1. Does the change involve a significant increase in the 
probability or consequences of an accident previously evaluated?
    The revised Figure 2.1-1 and the Overtemperature and Overpower 
T reactor trip functions do not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated because operation with these revised values 
will not cause any design or analysis acceptance criteria to be 
exceeded. The structural and functional integrity of all plant 
systems is unaffected. The Overtemperature and Overpower T 
reactor trip functions are part of the accident mitigation response 
and are not themselves an initiator for any transient. Therefore, 
the probability of occurrence previously evaluated is not 
significantly affected.
    The changes to Figure 2.1-1 and to the Overtemperature and 
Overpower T reactor trip functions do not affect the 
integrity of

[[Page 20003]]

the fission product barriers utilized for mitigation of radiological 
dose consequences as a result of an accident. Figure 2.1-1 provides 
restrictions to prevent overheating of the fuel and cladding, as 
well as possible cladding perforation, that would result in the 
release of fission products to the reactor coolant. It does not 
provide an automatic protective function but does provide the basis 
for the Overtemperature and Overpower T reactor trip 
functions. These trip functions ensure that automatic protective 
actions will be initiated at or before the condition assumed in the 
safety analyses. These changes produce no adverse effect on the 
ability of these functions to perform their safety functions assumed 
in the safety analyses. In addition, the off-site mass releases used 
as input to the dose calculations are unchanged from those 
previously assumed. Therefore, the off-site dose predictions remain 
within the acceptance criteria of 10 CFR 100 limits for each of the 
transients affected. Since it has been concluded that the transient 
analyses results are unaffected by the parameter modifications, it 
is concluded that the probability or consequences of an accident 
previously evaluated are not significantly increased.
    Therefore, the proposed change does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    2. Does the change create the possibility of a new or different 
kind of accident from any accident previously evaluated?
    The revised Figure 2.1-1 and Overtemperature and Overpower 
T reactor trip functions do not create the possibility of a 
new or different kind of accident from any accident previously 
evaluated because the setpoint adjustments do not affect accident 
initiation sequences. No new operating configuration is being 
imposed by the setpoint adjustments that would create a new failure 
scenario. In addition, no new failure modes or limiting single 
failures have been identified. Therefore, the types of accidents 
defined in the UFSAR continue to represent the credible spectrum of 
events to be analyzed which determine safe plant operation.
    Therefore, the proposed change does not create the possibility 
of a new or different kind of accident from any accident previously 
evaluated.
    3. Does the change involve a significant reduction in a margin 
of safety?
    The changes to Figure 2.1-1 and to the Overtemperature and 
Overpower T reactor trip functions do not involve a 
significant reduction in a margin of safety because the margin of 
safety associated with the Overtemperature and Overpower T 
reactor trip functions, as verified by the results of the accident 
analyses, are within acceptable limits. All transients impacted by 
implementation of the RTDP methodology have been analyzed and have 
met the applicable accident analyses acceptance criteria. The margin 
of safety required for each affected safety analysis is maintained. 
This conclusion is not changed by the Overtemperature and Overpower 
T setpoint modifications. The adequacy of the revised 
technical specification values to maintain the plant in a safe 
operating condition has been confirmed.
    Therefore, the proposed change does not involve a significant 
reduction in a margin of safety.

C. RTDP--Departure From Nucleate Boiling (DNB) Parameter Surveillance 
Requirements

    1. Does the change involve a significant increase in the 
probability or consequences of an accident previously evaluated?
    The intent of the change is to preserve the Safety Analyses 
Limits for DNB (TS 3/4.2.5). There is no increase in the probability 
or consequences of an accident previously evaluated because there is 
no change to any design or analysis acceptance criteria. The 
structural and functional integrity of any plant system is 
unaffected. The proposed license amendment revises the surveillance 
requirement acceptance criteria for the DNB parameters. The 
indicated DNB parameters preserve the assumptions used in the 
accident analysis and, therefore, there is no significant increase 
in probability or consequences previously evaluated.
    The changes to the DNB parameters do not affect the integrity of 
the fission product barriers utilized for mitigation of radiological 
dose consequences as a result of an accident. In addition, the off-
site mass releases used as input to the dose calculations are 
unchanged from those previously assumed. Therefore, the off-site 
dose predictions remain within the limits of the 10 CFR 100 for each 
of the transients affected. Since it has been determined that the 
transient results are unaffected by these parameter modifications, 
it is concluded that the consequences of an accident previously 
evaluated are not significantly increased.
    Therefore, the proposed change does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    2. Does the change create the possibility of a new or different 
kind of accident from any accident previously evaluated?
    The revised DNB parameter values do not create the possibility 
of a new or different kind of accident from any accident previously 
evaluated. The setpoint values do not affect the assumed accident 
initiation sequences. No new operating configuration is being 
imposed by changing these parameters that would create a new failure 
scenario. In addition, no new failure modes or single failures have 
been identified for any plant equipment. Therefore, the types of 
accidents defined in the UFSAR continue to represent the credible 
spectrum of events to be analyzed which determine safe plant 
operation.
    Therefore, the proposed change does not create the possibility 
of a new or different kind of accident from any accident previously 
evaluated.
    3. Does the change involve a significant reduction in a margin 
of safety?
    The DNB parameters are consistent with the UFSAR assumptions and 
maintain the required minimum DNBR [departure from nucleate boiling 
ratio] above the design limits throughout each analyzed transient. 
Thereby, the adequacy of the revised DNB parameter values to 
maintain the plant in a safe operating condition is confirmed.
    Therefore, the proposed change does not involve a significant 
reduction in a margin of safety.

D. Relocation to Colr [Core Operating Limits Report]

    1. Does the change involve a significant increase in the 
probability or consequences of an accident previously evaluated?
    The proposed amendment is a programmatic and administrative 
change that does not physically alter safety-related systems, nor 
does it affect the way in which safety-related systems perform their 
functions. Because the design of the facility and system operating 
parameters are not being changed, the proposed amendment does not 
involve a significant increase in the probability or consequences of 
any accident previously evaluated.
    The cycle-specific values relocated into the COLR will continue 
to be controlled by the BVPS programs and procedures. Each accident 
analysis addressed in the UFSAR will be examined with respect to 
changes in the cycle-dependent parameters, which are obtained from 
the use of NRC approved reload design methodologies, to ensure that 
the transient evaluation of new reloads are bounded by previously 
accepted analyses. This examination, which will be conducted per the 
requirements of 10 CFR 50.59, will ensure that future reloads will 
not involve a significant increase in the probability or 
consequences of an accident previously evaluated.
    Therefore, the proposed change does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    2. Does the change create the possibility of a new or different 
kind of accident from any accident previously evaluated?
    The proposed amendment is a programmatic and administrative 
change and does not result in any change in the manner in which the 
plant is operated or the way in which the Reactor Protection System 
provides plant protection. All of the accident transients analyzed 
in the UFSAR will continue to be protected by the same trip 
functions with the required trip setpoints. Removal of the cycle 
specific variables has no influence or impact on, nor does it 
contribute in any way to the probability or consequences of an 
accident. No safety-related equipment, safety function, or plant 
operation will be altered as a result of this proposed change. The 
cycle specific variables are calculated using the NRC approved 
methods, and submitted to the NRC to allow the staff to continue to 
review the values of these limits. The technical specifications will 
continue to require operation within the core operating limits, and 
appropriate actions will be required if these limits are exceeded.
    Therefore, the proposed change does not create the possibility 
of a new or different kind of accident from any accident previously 
evaluated.
    3. Does the change involve a significant reduction in a margin 
of safety?
    The proposed change to relocate core safety limits, trip 
setpoint parameter values,

[[Page 20004]]

and DNB parameter values to the Core Operating Limits Report 
represents an administrative change and no hardware changes are 
involved; therefore, no accident analysis acceptance criteria are 
affected.
    The margin of safety is not affected by the removal of cycle-
specific core operating limits from the technical specifications. 
The margin of safety presently provided by current technical 
specifications remains unchanged. Appropriate measures exist to 
control the values of these cycle specific limits. The proposed 
amendment continues to require operation within the core limits as 
obtained from NRC-approved methodologies, and the actions to be 
taken if a limit is exceeded will continue to require that the plant 
be placed in Hot Standby within one hour.
    The development of the limits for future reloads will continue 
to conform to those methods described in NRC approved documentation. 
In addition, each future reload will involve a 10 CFR 50.59 safety 
review to assure that operation of the unit within the cycle-
specific limits will not involve a significant reduction in the 
margin of safety.
    The proposed amendment is a programmatic and administrative 
change that provides assurance that plant operations continue to be 
conducted in a safe manner. The proposed amendment does not result 
in any change in the manner in which the plant is operated or the 
way in which the Reactor Protection System (RPS) provides plant 
protection. The proposed relocation does not alter the manner in 
which safety limits, limiting safety system setpoints or limiting 
conditions for operation are determined. Therefore, the response of 
the RPS to accident transients described in the UFSAR is unaffected 
by this change.
    As stated previously, this portion of the proposed amendment 
does not physically alter safety-related systems, nor does it affect 
the way in which safety-related systems perform their functions.
    The accident transients are unaffected and the safety analysis 
acceptance limits are unaffected. The design of the facility and 
system operating parameters are not being changed.
    Therefore, the proposed change does not involve a significant 
reduction in a margin of safety.

E. Relocation to Licensing Requirements Manual (LRM)

    1. Does the change involve a significant increase in the 
probability or consequences of an accident previously evaluated?
    The proposed change does not involve a significant increase in 
the probability of an accident previously evaluated because no 
changes are being made to any event initiator. Nor is any analyzed 
accident scenario being revised. The initiating conditions and 
assumptions for accidents described in the UFSAR remain as 
previously analyzed.
    The proposed change also does not involve a significant increase 
in the consequences of an accident previously evaluated. The change 
does not reduce the operability requirements for the affected 
instrumentation. The proposed relocation of TS requirements only 
affects the level of regulatory control involved in future changes 
to the requirements. The instrument setpoints will continue to be 
maintained in a similar manner as before. The conclusions and 
descriptions of the safety analyses described in the UFSAR remain 
unchanged.
    Therefore, the proposed change does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    2. Does the change create the possibility of a new or different 
kind of accident from any accident previously evaluated?
    The proposed change does not involve any physical changes to the 
plant or the modes of plant operation defined in the technical 
specifications. The proposed amendment does not involve the addition 
or modification of plant equipment nor does it alter the design or 
operation of any plant systems. No new accident scenarios, transient 
precursors, failure mechanisms, or limiting single failures are 
introduced as a result of these changes.
    There are no changes in this amendment which would cause the 
malfunction of safety-related equipment assumed to be operable in 
accident analyses. No new mode of failure has been created and no 
new equipment performance requirements are imposed. The proposed 
amendment has no effect on any previously evaluated accident.
    Therefore, the proposed change does not create the possibility 
of a new or different kind of accident from any accident previously 
evaluated.
    3. Does the change involve a significant reduction in a margin 
of safety?
    The margin of safety depends on the maintenance of specific 
operating parameters and systems within design requirements and 
safety analysis assumptions.
    The proposed change does not involve revisions to any safety 
limits or safety system setting that would adversely impact plant 
safety. The proposed change does not alter the functional 
capabilities assumed in a safety analysis for any system, structure, 
or component important to the mitigation and control of design bases 
accident conditions within the facility. Nor does this change revise 
any parameters or operating restrictions that are assumptions of a 
design basis accident. In addition, the proposed change does not 
affect the ability of safety systems to ensure that the facility can 
be placed and maintained in a shutdown condition for extended 
periods of time.
    The relocation of TS requirements does not reduce the 
effectiveness of the requirements being relocated. Rather, the 
relocation of the TS requirements results in a change in the 
regulatory control required for future changes made to the 
requirements. The relocated requirements will continue to be 
implemented by the appropriate plant procedures (e.g., operating and 
maintenance procedures) in the same manner as before. However, 
future changes to the relocated requirements will be controlled in 
accordance with 10 CFR 50.59 instead of 10 CFR 50.90. The provisions 
of 10 CFR 50.59 establish adequate controls over requirements 
removed from the TS and assure future changes to these requirements 
will be consistent with safe plant operation.
    Therefore, the proposed change does not involve a significant 
reduction in a margin of safety.

F. Miscellaneous Changes

    1. Does the change involve a significant increase in the 
probability or consequences of an accident previously evaluated?
    The administrative change, for BVPS Unit No. 1 only, pertaining 
to two loop operation and Reactor Coolant System isolation valve 
position, does not affect plant safety. The technical specification 
requirements in LCOs 3.4.1.1 and 3.4.1.4.1 will continue to prohibit 
two loop operation and ensure safe plant operation by properly 
controlling the operation and position of the reactor coolant loops 
and Reactor Coolant System isolation valves.
    The administrative change to delete line item 7.d, pertaining to 
Auxiliary Feedwater (AFW) Pump Auto-start on Emergency Bus 
Undervoltage, for BVPS Unit No. 1 only, from TS Tables 3.3-3, 3.3-4, 
and 4.3-2 will not affect plant safety because this function is not 
directly initiated by bus undervoltage. Rather, the automatic start 
of the motor-driven AFW pumps is accomplished by the combination of 
1) Emergency Bus feed breaker opening 2) valid start signal from 
ESFAS, and 3) Emergency Diesel Generator (EDG) sequencer actuation. 
Requirements for these items are included in the ESFAS related TS, 
Table 3.3-3 and 3.3-4 items 7.a, 7.c, 7.e, and EDG related TS 
4.8.1.1.2.b.3 (b). Therefore, since there is no change made to the 
plant hardware or its operation and requirements related to the AFW 
pump auto-start function are maintained elsewhere in the BVPS Unit 
No. 1 TS, deleting line item 7.d from BVPS Unit No. 1 TS Tables 3.3-
3, 3.3-4, and 4.3-2 will not significantly change the probability or 
consequences of any accident previously evaluated.
    The proposed change does not involve a significant increase in 
the probability of an accident previously evaluated because no 
changes are being made to any event initiator. Nor is any analyzed 
accident scenario being revised. The initiating conditions and 
assumptions for accidents described in the UFSAR remain as 
previously analyzed.
    The proposed change also does not involve a significant increase 
in the consequences of an accident previously evaluated. The change 
does not reduce the effectiveness or scope of the affected TS. The 
proposed changes are administrative in nature and do not affect any 
technical or equipment operability requirements. The conclusions and 
descriptions of the safety analyses described in the UFSAR remain 
unchanged.
    Therefore, the proposed change does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    2. Does the change create the possibility of a new or different 
kind of accident from any accident previously evaluated?
    The proposed change does not involve any physical changes to the 
plant or the modes of plant operation defined in the TS. The 
proposed amendment does not involve the

[[Page 20005]]

addition or modification of plant equipment nor does it alter the 
design or operation of any plant systems. No new accident scenarios, 
transient precursors, failure mechanisms, or limiting single 
failures are introduced as a result of these changes.
    There are no changes in this amendment which would cause the 
malfunction of safety-related equipment assumed to be operable in 
accident analyses. No new mode of failure has been created and no 
new equipment performance requirements are imposed. The proposed 
amendment has no effect on any previously evaluated accident.
    Therefore, the proposed change does not create the possibility 
of a new or different kind of accident from any accident previously 
evaluated.
    3. Does the change involve a significant reduction in a margin 
of safety?
    The margin of safety depends on the maintenance of specific 
operating parameters and systems within design requirements and 
safety analysis assumptions.
    The proposed change does not involve revisions to any safety 
limits or safety system setting that would adversely impact plant 
safety. The proposed change does not alter the functional 
capabilities assumed in a safety analysis for any system, structure, 
or component important to the mitigation and control of design bases 
accident conditions within the facility. Nor does this change revise 
any parameters or operating restrictions that are assumptions of a 
design basis accident. In addition, the proposed change does not 
affect the ability of safety systems to ensure that the facility can 
be placed and maintained in a shutdown condition for extended 
periods of time.
    The administrative change to delete line item 7.d, pertaining to 
AFW Pump Auto-start on Emergency Bus Undervoltage, BVPS Unit No. 1 
only, from TS Tables 3.3-3, 3.3-4, and 4.3-2 will not affect plant 
safety because this function is not directly initiated by bus 
undervoltage. Rather, the automatic start of the motor-driven AFW 
pumps is accomplished by the combination of (1) Emergency Bus feed 
breaker opening, (2) valid start signal from ESFAS, and (3) EDG 
sequencer actuation. Requirements for these items are included in 
the ESFAS related TS, Table 3.3-3 and 3.3-4 items 7.a, 7.c, 7.e, and 
EDG related TS 4.8.1.1.2.b.3 (b). Therefore, since there is no 
change made to the plant hardware or its operation and requirements 
related to the AFW pump auto-start function are maintained elsewhere 
in the BVPS Unit No.1 TS, deleting line item 7.d from BVPS Unit No. 
1 TS Tables 3.3-3, 3.3-4, and 4.3-2 will not involve a significant 
reduction in a margin of safety.
    The administrative change, for BVPS Unit No. 1 only, pertaining 
to two loop operation and Reactor Coolant System isolation valve 
position, does not affect plant safety. The technical specification 
requirements in LCOs 3.4.1.1 and 3.4.1.4.1 will continue to prohibit 
two-loop operation and ensure safe plant operation by properly 
controlling the operation and position of the reactor coolant loops 
and Reactor Coolant System isolation valve.
    The other proposed changes are also administrative in nature and 
only affect the format or presentation of information in the TS. The 
proposed changes have no affect on the conclusions or descriptions 
of the safety analyses described in the UFSAR.
    Therefore, the proposed change does not involve a significant 
reduction in a margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: Mary O'Reilly, FirstEnergy Nuclear Operating 
Company, FirstEnergy Corporation, 76 South Main Street, Akron, OH 
44308.
    NRC Section Chief: Marsha Gamberoni.

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

    Date of amendment request: March 12, 2001.
    Description of amendment request: The proposed amendments would 
revise the current 72-hour allowed outage time (AOT) specified in 
Technical Specification (TS) 3.8.1.1, Actions ``b'' and ``f,'' and 
associated TSs 3.4.3 and 3.5.2, to allow 14 days to restore an 
inoperable emergency diesel generator (EDG) to operable status. The 
proposed AOT is based on the licensee's integrated assessment of plant 
operations, deterministic design basis factors, and an evaluation of 
overall plant risk using probabilistic safety assessment techniques. 
Additionally, the proposed amendments would relocate TS Surveillance 
Requirement 4.8.1.1.2.g.1 to a licensee controlled maintenance program 
that will be incorporated by reference into the Updated Final Safety 
Analysis Report (UFSAR).
    The proposed amendments would also make administrative changes that 
consist of deleting footnotes on pages 
3/4 4-9, 3/4 5-4, 3/4 8-2, and 3/4 8-4, that are no longer applicable, 
and adding appropriate footnotes on pages 3/4 4-9 and 3/4 8-2 that are 
compatible with the revised EDG AOT.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    (1) Operation of the facility in accordance with the proposed 
amendment would not involve a significant increase in the 
probability or consequences of an accident previously evaluated.
    The proposed amendments for Turkey Point Unit 3 and Unit 4 will 
extend the AOT for a single inoperable EDG from 72 hours to 14 days. 
The EDGs are designed as backup AC power sources for essential 
safety systems in the event of a loss of offsite power. As such, the 
EDGs are not accident initiators, and an extended AOT to restore 
operability of an inoperable diesel generator would not 
significantly increase the probability of occurrence of accidents 
previously analyzed.
    The proposed Technical Specification revisions involve the AOT 
for a single inoperable EDG, and do not change the conditions, 
operating configuration, or minimum amount of operating equipment 
assumed in the plant safety analyses for accident mitigation. Plant 
defense-in-depth capabilities will be maintained with the proposed 
AOT, and the design basis for electric power systems will continue 
to conform with 10 CFR 50, Appendix A, General Design Criterion 17. 
In addition, a Probability Safety Assessment (PSA) was performed to 
quantitatively assess the risk-impact of the proposed amendment for 
each unit. The impact on the early radiological release probability 
for design basis events was also evaluated and it is concluded that 
the risk contribution from this proposed AOT is small and consistent 
with regulatory risk-assessment acceptance guidelines.
    The relocation of the TS Surveillance requirement 4.8.1.1.2.g.1 
from the Technical Specifications to a licensee controlled 
maintenance program referenced in the UFSAR is bounded by the risk 
assessment for the EDG AOT extension and therefore does not involve 
a significant increase in the probability or consequences of an 
accident previously evaluated.
    Therefore, facility operation in accordance with the proposed 
amendments would not involve a significant increase in the 
probability or consequences of an accident previously evaluated.
    (2) Operation of the facility in accordance with the proposed 
amendment would not create the possibility of a new or different 
kind of accident from any accident previously evaluated.
    The proposed amendments will not change the physical plant or 
the modes of operation defined in either facility license. The 
changes do not involve the addition of new equipment or the 
modification of existing equipment, nor do they alter the design of 
Turkey Point plant systems. Therefore, facility operation in 
accordance with the proposed amendments would not create the 
possibility of a new or different kind of accident from any accident 
previously evaluated.
    (3) Operation of the facility in accordance with the proposed 
amendment would not involve a significant reduction in a margin of 
safety.
    The proposed amendments are designed to improve EDG reliability 
by providing flexibility in the scheduling and performance of 
preventive and corrective maintenance activities. The proposed 
changes do not alter the basis for any Technical Specification that 
is related to the establishment of, or the

[[Page 20006]]

maintenance of, a nuclear safety margin, and design defense-in-depth 
capabilities are maintained. The relocation of the TS Surveillance 
requirement 4.8.1.1.2.g.1 from the Technical Specifications to a 
licensee controlled maintenance program referenced in the UFSAR is 
bounded by the risk assessment for the EDG AOT extension. An 
integrated assessment of the risk impact of extending the AOT for a 
single inoperable EDG has determined that the risk contribution is 
small and is within regulatory guidelines for an acceptable TS 
change. Therefore, facility operation in accordance with the 
proposed amendments would not involve a significant reduction in a 
margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    The proposed changes which consist of deleting four footnotes that 
are no longer applicable, and adding two footnotes that are compatible 
with the revised EDG AOT are administrative in nature. Therefore, the 
staff also proposes to determine that these proposed changes involve no 
significant hazards consideration.
    Attorney for licensee: M.S. Ross, Attorney, Florida Power & Light, 
P.O. Box 14000, Juno Beach, Florida 33408-0420.
    NRC Section Chief: Richard P. Correia.

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

    Date of amendment request: March 21, 2001.
    Description of amendment request: The licensee proposes to 
implement a repair roll (re-roll) process for the Crystal River Unit 3 
(CR-3) Once Through Steam Generator (OTSG) tubes applicable to the 
upper and lower tubesheets.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Involve a significant increase in the probability or 
consequences of an accident previously evaluated.
    The re-roll process is a method to create a new primary-to-
secondary pressure boundary joint in the upper tubesheet of Babcock 
& Wilcox (B&W) Once Through Steam Generators (OTSGs) manufactured 
with Inconel Alloy 600 tubes. The new pressure boundary is 
established by the re-roll to remove degradation of the existing 
roll joint from pressure boundary service. The re-roll process has 
been previously qualified as an acceptable repair methodology for 
use in the upper tubesheet of the Crystal River Unit 3 (CR-3) OTSGs 
by License Amendment No. 180. This proposed LAR incorporates 
Revision 4 of Topical Report BAW-2303P, ``OTSG Repair Roll 
Qualification Report.'' This proposed LAR also addresses several 
editorial changes which do not impact the current CR-3 accident 
analyses.
    The qualification of the OTSG tube re-roll methodology is based 
on establishing a mechanical joint length that will carry all 
structural loads imposed on the OTSG tubes while maintaining the 
required margins during normal and accident conditions. A series of 
tests and analyses were performed to establish the minimum 
acceptable length of the OTSG tube re-roll. Tests performed included 
leak, tensile, fatigue, ultimate load and eddy-current measurement 
uncertainty. The analyses evaluated plant operating and faulted load 
conditions. OTSG tube leakage remains bounded by the evaluation 
presented in the CR-3 Final Safety Analysis Report (FSAR) for a main 
steam line break (MSLB). The current CR-3 Improved Technical 
Specifications (ITS) include a description of the required 
inspection program for the OTSG tube re-rolls. The required ITS 
inspections following OTSG tube re-roll installation, and during 
future inservice inspections, ensure continuous monitoring of these 
tubes such that in service degradation of tubes repaired by the re-
roll process will be detected. Based on the qualification testing 
and analyses performed, as well as the industry experience with the 
use of the OTSG tube re-roll processes, there are no new safety 
issues associated with the use of the re-roll methodology. The 
probability of a steam generator tube rupture is not increased by 
the re-roll since it is a repair process not applied to defective 
OTSG tube areas. This repair process establishes a new pressure 
boundary roll joint which is free of degradation. Therefore, this 
change does not involve a significant increase in the probability or 
consequences of any accident previously evaluated.
    2. Create the possibility of a new or different kind of accident 
from previously evaluated accidents.
    The re-roll process creates no new failure modes or accident 
scenarios. The new pressure boundary joint created by the re-roll 
process has been demonstrated, by testing and analysis, to provide 
structural and leakage integrity equivalent to the original design 
and construction for all normal operating and accident conditions. 
Furthermore, testing and analysis demonstrate that the re-roll 
process creates no new adverse effects for the repaired tube and 
does not change the design or operating characteristics of the 
OTSGs. BAW-2303P, Revision 4, addresses limiting events for steam 
generator re-roll repairs. These events include Main Steam Line 
Break, Small Break Loss of Coolant Accident and other transients on 
the B&W Once Through Steam Generators. Therefore, this change does 
not create the possibility of a new or different kind of accident 
from any previously evaluated.
    3. Involve a significant reduction in a margin of safety.
    The re-roll process effectively removes the defective/degraded 
area of the tube from service by establishing a new pressure 
boundary. The re-roll interface created with the tubesheet satisfies 
the necessary structural, leakage and heat transfer requirements. 
Implementation of BAW-2303P, Revision 4, will result in assurance 
that parameters affecting the integrity of steam generator tubes 
continue to meet safety analyses and industry codes and standards. 
Therefore, the FSAR analyzed accident scenarios remain bounding, and 
the use of the re-roll process does not significantly reduce 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 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: R. Alexander Glenn, Associate General 
Counsel (MAC-BT15A), Florida Power Corporation, P.O. Box 14042, St. 
Petersburg, Florida 33733-4042.
    NRC Section Chief: Richard P. Correia.

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

    Date of amendment request: March 28, 2001.
    Description of amendment request: The proposed amendment would 
revise the Crystal River Unit 3 Improved Technical Specifications (ITS) 
3.7.18, ``Control Complex Cooling System'' to allow a one-time increase 
in the Completion Time for restoring an inoperable Control Complex 
Cooling System train from 7 days to 35 days.
    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 not involve a significant increase in the probability 
or consequences of an accident previously analyzed.
    The Control Complex Cooling System is not an initiator of any 
design basis accident. The Control Complex Cooling System is 
designed to provide sufficient cooling to ensure operability of 
safety-related equipment located in the control room and other 
portions of the control complex under normal and accident 
conditions.
    The proposed license amendment extends the Completion Time for 
restoring an inoperable Control Complex Cooling train from 7 days to 
35 days on a one-time basis for each train to allow on-line 
refurbishment

[[Page 20007]]

of the control complex chillers. The proposed amendment also 
specifies that the requirements of (LCO) 3.0.4 are not applicable to 
ITS 3.7.18 Condition A during the 35-day Completion Times. The 
design functions of the Control Complex Cooling System and the 
initial conditions for accidents that require the Control Complex 
Cooling System will not be affected by the change. The increased 
Completion Time requested by License Amendment Request (LAR) #259 
results in slight increases in core damage frequency and core damage 
probability; however, these increases are well below values that are 
considered risk significant. Therefore, the change will not 
significantly increase the probability or consequences of an 
accident previously evaluated.
    (2) Does not create the possibility of a new or different kind 
of accident from any accident previously analyzed.
    The proposed amendment extends the Completion Time for restoring 
an inoperable Control Complex Cooling System train on a one-time 
basis for each train to allow on-line performance of maintenance 
activities that will improve chiller reliability. The proposed 
amendment will not result in changes to the design, physical 
configuration or operation of the plant. Therefore, the proposed 
change will not create the possibility of a new or different kind of 
accident from any accident previously evaluated.
    (3) Does not involve a significant reduction in the margin of 
safety.
    The proposed license amendment increases the Completion Time for 
restoring an inoperable Control Complex Cooling System train from 7 
days to 35 days on a one-time basis for each train. The proposed 
amendment also specifies that the requirements of Limiting Condition 
for Operation (LCO) 3.0.4 are not applicable to ITS 3.7.18 Condition 
A during the one-time 35-day Completion Times. The proposed changes 
will maintain operational flexibility while allowing on-line 
refurbishment of the control complex chillers to improve their 
reliability and extend their useful lifetimes, thus increasing the 
long-term margin of safety of the system.
    The Control Complex Cooling System is designed to provide 
sufficient cooling to ensure operability of safety-related equipment 
located in the control room and other portions of the control 
complex under both normal and accident conditions. Either redundant 
train of the system is capable of performing this function; 
therefore, as long as one train is available, the margin of safety 
is maintained. Waiving the requirements of LCO 3.0.4 while the 
requested 35-day Completion Times are in effect will not impact the 
availability of the redundant system train, backup systems, or 
required support systems. In addition, since the heat removal 
requirements for the control room and other vital heat loads in the 
control complex are the same in Mode 1 as they are in Mode 3, 
allowing the plant to escalate Modes while chiller repairs are in 
progress will not impact the ability of the Control Complex Cooling 
System to fulfill its intended safety function. During the time that 
the required maintenance activities are being performed on each 
chiller, the availability of redundant system components will be 
maximized by administratively controlling preventive maintenance and 
surveillance activities performed on the Control Complex Cooling 
System and required support systems. Defense-in-depth measures will 
also be implemented to ensure the availability of temporary and 
permanently installed backup systems capable of providing cooling to 
the control room and the other vital equipment areas in the control 
complex. Although the increased Completion Time requested by LAR 
#259 results in a loss of redundancy and slight increases in core 
damage frequency and core damage probability, these increases are 
well below values that are considered risk significant. 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 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: R. Alexander Glenn, Associate General 
Counsel (MAC-BT15A), Florida Power Corporation, P.O. Box 14042, St. 
Petersburg, Florida 33733-4042.
    NRC Section Chief: Richard P. Correia.

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

    Date of amendment request: February 28, 2001.
    Description of amendment request: The proposed amendment changes 
the Seabrook Station Technical Specifications (TSs) 3/4.8.1.1 A.C. 
Sources--Operating. In addition, other changes are proposed either for 
clarity, which are reflective of the improved Standard Technical 
Specifications for Westinghouse Plants, NUREG-1431, Rev. 1 and Draft 
Rev. 2, or do not meet the four criteria of 10 CFR 50.36 for inclusion 
in TSs. Those requirements that do not meet the criteria for inclusion 
in the TSs will either be deleted or relocated to the Seabrook Station 
Technical Requirements manual.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration. The NRC staff has reviewed the licensee's analysis 
against the standards of 10 CFR 50.92(c). The NRC staff's analysis is 
presented below:

    1. The proposed changes do not involve a significant increase in 
the probability or consequences of an accident previously evaluated.
    The proposed changes do not involve a change in the operational 
limits, do not involve a change in physical design of the electrical 
power systems, do not change the function or operation of plant 
equipment or affect the response of that equipment if called upon to 
operate. The proposed allowed outage time extensions will not cause 
a significant increase in the probability or consequences of an 
accident previously evaluated. Therefore, the proposed changes do 
not involve a significant increase in the probability or 
consequences of accidents previously evaluated.
    2. The proposed changes do not create the possibility of a new 
or different kind of accident from any previously evaluated.
    No new or different kind of accident is created because the 
proposed changes do not involve a change in the operational limits, 
do not involve a change in physical design of the electrical power 
systems, do not change the function or operation of plant equipment 
or introduce any new failure mechanisms. The plant equipment will 
continue to respond per the design and analyses and there will not 
be a malfunction of a new or different type introduced by the 
proposed changes. Therefore, the proposed changes do not create the 
possibility of a new or different kind of accident from any 
previously evaluated.
    3. The proposed changes do not involve a significant reduction 
in the margin of safety.
    The margin of safety will remain the same because the proposed 
changes do not involve a change in the operational limits, do not 
involve a change in physical design of the electrical power systems, 
do not change the function or operation of plant equipment or affect 
the response of that equipment if it is called upon to operate. 
Therefore, the proposed changes do not involve a significant 
reduction in the margin of safety.

    Based on this review, it appears that the three standards of 
50.92(c) are satisfied. Therefore, the NRC staff proposes to determine 
that the amendment request involves no significant hazards 
consideration.
    Attorney for licensee: Lillian M. Cuoco, Esq., Senior Nuclear 
Counsel, Northeast Utilities Service Company, P.O. Box 270, Hartford, 
CT 06141-0270.
    NRC Section Chief: James W. Clifford.

Northeast Nuclear Energy Company, et al., Docket No. 50-336, Millstone 
Nuclear Power Station, Unit No. 2, New London County, Connecticut

    Date of amendment request: December 21, 2000.
    Description of amendment request: The proposed amendment would 
allow plant operation to continue if the temperature of the Ultimate 
Heat Sink (UHS) exceeds the Technical Specification limit of 75  deg.F 
provided the water temperature, averaged over the previous 24-hour 
period, is at or below 75  deg.F. The proposed operational

[[Page 20008]]

flexibility would only apply if the UHS temperature is between 75 
deg.F and 77  deg.F. The current action time requirements would still 
apply if the UHS temperature exceeds 77  deg.F, or if the 24-hour 
averaged value exceeds 75  deg.F. The current Technical Specification 
Limiting Condition for Operation (LCO) limit of 75  deg.F would not be 
changed. The Bases for the associated Technical Specification would 
also be modified.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration. The NRC staff has reviewed the licensee's analysis 
against the standards of 10 CFR 50.92(c). The NRC staff's analysis, 
which is based on the representations made by the licensee in the 
December 21, 2000, application, is presented below:

    1. Involve a significant increase in the probability or 
consequences of an accident previously evaluated.
    The proposed changes will allow plant operation to continue if 
the temperature of the UHS exceeds the Technical Specification limit 
of 75  deg.F provided that: (1) The water temperature, averaged over 
the previous 24 hour period, is at or below 75  deg.F, and (2) the 
UHS temperature is less than or equal to 77  deg.F. This increase in 
UHS temperature will not affect the normal operation of the plant to 
the extent which would make any accident more likely to occur. In 
addition, there exists adequate margin in the safety systems and 
heat exchangers to assure the safety functions are met at the higher 
temperature. An evaluation has confirmed that safe shutdown will be 
achieved and maintained for a loss of coolant accident (LOCA) with a 
loss of normal power (LNP) and a single active failure with an UHS 
water temperature as high as 77  deg.F.
    Thus, the proposed changes will have no adverse effect on plant 
operation, or the availability or operation of any accident 
mitigation equipment. The plant response to the design basis 
accidents will not change. In addition, the proposed changes can not 
cause an accident. Therefore, there will be no increase in the 
probability or consequences of an accident previously evaluated.
    2. Create the possibility of a new or different kind of accident 
from any accident previously evaluated.
    The proposed changes will allow plant operation to continue if 
the temperature of the UHS exceeds the Technical Specification limit 
of 75  deg.F provided that: (1) The water temperature, averaged over 
the previous 24 hour period, is at or below 75  deg.F, and (2) the 
UHS temperature is less than or equal to 77  deg.F. This will not 
alter the plant configuration (no new or different type of equipment 
will be installed) or require any new or unusual operator actions. 
The proposed changes will not alter the way any structure, system, 
or component functions and will not significantly alter the manner 
in which the plant is operated. There will be no adverse effect on 
plant operation or accident mitigation equipment. The proposed 
changes do not introduce any new failure modes. Also, the response 
of the plant and the operators following these accidents is 
unaffected by the changes. In addition, the UHS is not an accident 
initiator. Therefore, the proposed changes will not create the 
possibility of a new or different kind of accident from any 
previously analyzed.
    3. Involve a significant reduction in a margin of safety.
    The proposed changes will allow plant operation to continue if 
the temperature of the UHS exceeds the Technical Specification limit 
of 75  deg.F provided that: (1) The water temperature, averaged over 
the previous 24 hour period, is at or below 75  deg.F, and (2) the 
UHS temperature is less than or equal to 77  deg.F. The licensee 
performed an evaluation of the safety systems to ensure their safety 
functions can be met with a UHS water temperature of 77  deg.F. The 
evaluation determined that an increase in UHS temperature from 75 
deg.F to 77  deg.F would nominally cause a 2  deg.F temperature 
increase in service water system, reactor building closed cooling 
water system, and associated heat exchanger loads. This represents a 
slight reduction in the margins of safety in terms of these systems' 
abilities to remove accident heat loads, and in terms of the 
thermally induced pipe stresses within these systems during accident 
conditions. As part of its evaluation, however, the licensee 
verified that these safety systems will still be able to perform 
their design basis functions, and that pipe stresses will remain 
within allowable levels.
    Safe shutdown capability has been demonstrated for a UHS water 
temperature as high as 77  deg.F.
    The proposed changes will have no adverse effect on plant 
operation or equipment important to safety. The plant response to 
the design basis accidents will not change and the accident 
mitigation equipment will continue to function as assumed in the 
design basis accident analysis. Therefore, there will be no 
significant reduction in a margin of safety.

    Based on this review, it appears that the three standards of 10 CFR 
50.92(c) are satisfied. Therefore, the NRC staff proposes to determine 
that the amendment request involves no significant hazards 
consideration.
    Attorney for licensee: Lillian M. Cuoco, Esq., Senior Nuclear 
Counsel, Northeast Utilities Service Company, P.O. Box 270, Hartford, 
Connecticut.
    NRC Section Chief: James W. Clifford.

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

    Date of amendment request: January 5, 2001.
    Description of amendment request: The proposed change would amend 
the Salem Nuclear Generating Station, Unit Nos. 1 and 2 Technical 
Specifications (TSs) by adding a requirement to perform a Hydrogen 
Analyzer gas calibration at least once per 92 days, and changing the 
required frequency to perform a channel calibration of the Hydrogen 
Analyzer from once per 92 days to once per refueling.
    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 Hydrogen Analyzer provides detection and measurement of 
containment hydrogen concentration so that hydrogen concentration 
can be maintained below its flammable limit following a Loss of 
Coolant Accident. As such the Hydrogen Analyzer does not affect the 
probability of any previously evaluated accident.
    The proposed changes are consistent with the manufacturer's 
recommendations to ensure that the Hydrogen Analyzer will provide 
accurate indication of containment hydrogen concentration when 
required. Under the proposed change, a gas calibration consisting of 
all elements of the Hydrogen Analyzer channel calibration, with the 
exception of the calibration of the instrument's resistance 
temperature detector and pressure transducer, would be performed at 
least every 92 days. As a part of the gas calibration, a comparison 
of the indication of Hydrogen Analyzer resistance temperature 
detector and the pressure transducer against installed plant 
instrumentation measuring containment temperature and pressure would 
be performed. At least once per each refueling, a channel 
calibration of the Hydrogen Analyzers, including a calibration of 
the instrument's resistance temperature detector and pressure 
transducer using a secondary standard of a specified accuracy would 
be performed. Therefore, the proposed change would not affect the 
consequences of any previously evaluated accident.
    2. The proposed changes do not create the possibility of a new 
or different kind of accident from [any] accident previously 
evaluated.
    The proposed change affects only the specified calibration 
frequency of the Hydrogen Analyzers. The proposed surveillance 
frequency complies with the manufacturer's recommendations and will 
ensure that the Hydrogen Analyzers will provide accurate indication 
of containment hydrogen concentration when required. The change will 
not affect the design of any Salem Generating Station structure, 
system, or component, nor would it result in any new plant 
configuration. Therefore, the proposed changes do not create the 
possibility of a new or different kind of accident from [any] 
accident previously evaluated.
    3. The proposed changes do not involve a significant reduction 
in a margin of safety.
    The proposed change to the Hydrogen Analyzer calibration 
frequency will not affect the design or operating limits of any 
Salem

[[Page 20009]]

Generating Station structure, system, or component. The proposed 
surveillance frequency complies with the manufacturer's 
recommendations and will ensure that the Hydrogen Analyzers will 
provide accurate indication of containment hydrogen concentration 
when required. Therefore the proposed changes to the Technical 
Specifications 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, P.O. Box 236, Hancocks Bridge, NJ 08038.
    NRC Section Chief: James W. Clifford.

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

    Date of amendment request: December 28, 2000.
    Description of amendment request: This proposed amendment would 
revise Technical Specification (TS) 3/4.3.2, and Tables 3.3-3 and 3.3-4 
to incorporate consistent applicability and action for Engineered 
Safety Feature Actuation System (ESFAS) Instrumentation, Functional 
Unit 5.b. (Automatic Actuation Logic and Actuation Relay) Turbine Trip 
and Feedwater Isolation. This change will provide consistency between 
Tables 3.3-3, 3.3-4, and 4.3-2, and will be similar to the equivalent 
requirement in NUREG-1431, Revision 1.
    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 addition of an ACTION STATEMENT and the addition of an AOT 
[allowed outage time] (and its associated actions if not met) for a 
TS action statement are neither an accident initiator or precursor. 
The ESFAS actuates in response to an accident and has a mitigating 
function. Increasing the TS requirements for specific TS instrument 
loops provides additional assurance that the channels will be 
capable of performing their design function in the event of a DBA 
[design-basis accident]. The ability of the operations staff to 
respond to an evaluated accident or plant transient will not be 
hampered. This change provides conservative requirements to assure 
that the design basis of the plant is maintained.
    Addition of conservative changes to the Engineered Safety 
Feature Actuation System Instrumentation do not contribute to the 
initiation of any accident evaluated in the FSAR [Final Safety 
Analysis Report]. Supporting factors are as follows:
     The changes provide consistency between Tables 3.3-2, 
3.3-3, and 4.3-2, resulting in a one-for-one correlation between the 
functional units in those tables. These changes are conservative and 
consistent with the Standard Technical Specifications, NUREG-1431, 
Rev. 1. There are no deletions from the Technical Specifications 
made by these changes, nor relaxation in any applicability, action, 
or surveillance requirements.
     Overall plant performance and operation is not altered 
by the proposed changes. There are to be no plant hardware changes 
as a result of this proposed change and only minimal procedural 
changes.
    Therefore, since the Engineered Safety Feature Actuation System 
Instrumentation are treated more conservatively, the probability of 
occurrence or consequences of an accident evaluated in the VCSNS 
FSAR will be no greater than the original design basis of the plant.
    Therefore, the change does not involve a significant increase in 
the probability or consequences of an accident previously evaluated.
    2. Does the change create the possibility of a new or different 
kind of accident from any accident previously evaluated?
    The proposed changes provide consistency between Tables 3.3-2, 
3.3-3, and 4.3-2, resulting in a one-for-one correlation between the 
functional units in those tables. Additionally, the addition of an 
ACTION STATEMENT and an AOT with conservative requirements are 
intended to assure that the plant is in a safe configuration and can 
meet accident analyses assumptions. These changes are conservative 
and consistent with the Improved Technical Specifications, NUREG-
1431, Rev. 1. No new accident initiator mechanisms are introduced 
since:
     No physical changes to the Engineered Safety Feature 
Actuation System Instrumentation are made.
     No deletions from the Technical Specifications are 
made.
     No relaxation in any applicability, action, or 
surveillance requirements are made.
    Since the safety and design requirements continue to be met and 
the integrity of the reactor coolant system pressure boundary is not 
challenged, no new accident scenarios have been created. Therefore, 
the types of accidents defined in the FSAR continue to represent the 
credible spectrum of events to be analyzed which determine safe 
plant operation.
    3. Does this change involve a significant reduction in margin of 
safety?
    The proposed change requires that an instrument channel for an 
Engineered Safety Feature remain operable or be restored to 
operability within a reasonable time period, otherwise a controlled 
shutdown is required. This conforms to the safety analysis where the 
plant and its systems, structures and components must be capable of 
performing the safety function while a DBA is occurring, in the 
presence of a worst case single failure.
    This is not a reduction in a margin of safety, since it restores 
the margin that was designed into the plant.
    The proposed changes provide consistency between Tables 3.3-2, 
3.3-3, and 4.3-2, resulting in a one-for-one correlation between the 
functional units in those tables. These changes are conservative and 
consistent with the Standard Technical Specifications, NUREG-0452, 
Rev. 5.
    The proposed changes impose more restrictive operating 
limitations, and their use provides increased assurance that the 
Engineered Safety Feature Actuation System Instrumentation remains 
operable. Since the changes are conservative additions, it is 
concluded that the changes do not involve a significant reduction in 
the margin of safety. This is not a reduction in a margin of safety, 
since it restores the margin that was designed into the plant.

    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 G. Eppink, South Carolina Electric & 
Gas Company, Post Office Box 764, Columbia, South Carolina 29218.
    NRC Section Chief: Richard L. Emch, Jr.

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

    Date of amendment requests: March 21, 2001.
    Description of amendment requests: The amendment application 
proposes to revise the Facility Operating License No. NPF-10, and 
Facility Operating License No. NPF-15 for San Onofre Nuclear Generating 
Station, Units 2 and 3, respectively. The licensee proposed to simplify 
the Facility Operating Licenses by deleting those license conditions 
which have been completed and are no longer required to be identified 
in the licenses.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    (1) Involve a significant increase in the probability or 
consequences of any accident previously evaluated?
    Response: No.
    This proposed change is administrative since it only deletes 
completed San Onofre Units 2 and 3 license conditions, providing

[[Page 20010]]

appropriate references and discussion of the actions taken which 
document their completion. There is no physical plant change or 
change to plant operation which could increase the probability or 
consequences of any accident previously evaluated.
    Therefore, the probability or consequences of any accident 
previously evaluated is not increased.
    (2) Create the possibility of a new or different kind of 
accident from any previously evaluated?
    Response: No.
    This proposed change is administrative because it only deletes 
completed Onofre Units 2 and 3 license conditions and there is no 
physical plant change or change to plant operation which could 
introduce any mechanism which could create a new or different kind 
of accident.
    Therefore, the possibility of a new or different kind of 
accident from any previously evaluated is not created.
    (3) Involve a significant reduction in a margin of safety?
    Response: No.
    This change is administrative because it only deletes completed 
San Onofre Units 2 and 3 license conditions and there is no physical 
plant change or change to plant operation, therefore there is no 
impact in a margin of safety.
    Therefore, a significant reduction in a margin of safety is not 
involved.

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

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

    Date of amendment requests: March 21, 2001.
    Description of amendment requests: The amendment application 
proposes to revise the San Onofre Nuclear Generating Station, Units 2 
and 3, Technical Specification (TS) to clarify the methodology used to 
test the Control Room Emergency Air Cleanup System and Post-Accident 
Cleanup Filter System High Efficiency Particulate Air (HEPA) filters. 
Specifically, in TS 5.5.2.12, ``Ventilation Filter Testing Program 
(VFTP),'' the reference to the American Society for Mechanical 
Engineers (ASME) Code ASME N510-1989 will be revised to the American 
National Standards Institute (ANSI) N510-1975. Also, in TS 5.5.2.12.d, 
references to Regulatory Guide (RG) 1.52, Revision 2, and ASME N510-
1989 will be deleted.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    (1) Will operation of the facility in accordance with this 
proposed change involve a significant increase in the probability or 
consequences of an accident previously evaluated?
    Response: No.
    The proposed change is to change the reference to ASME Code in 
subsections 5.5.2.12.a and 5.5.2.12.b from ASME N510-1989 to ASME 
N510-1975. Technical Specification (TS) 3.7.11, ``Control Room 
Emergency Air Cleanup System (CREACUS)'' Surveillance Requirement 
(SR) 3.7.11.2 and TS 3.7.14, ``Fuel Handling Building Post-Accident 
Cleanup Filter System (PACU),'' SR 3.7.14.2 require CREACUS and PACU 
filter testing ``in accordance with the Ventilation Filter Testing 
Program (VFTP).''
    San Onofre Nuclear Generating Station (SONGS) TS 5.5.2.12.a, 
``Ventilation Filter Testing Program,'' states that the in-place 
HEPA filter testing is performed in accordance with Regulatory Guide 
(RG) 1.52, Revision 2 and ASME N510-1989. However, the CREACUS in-
place HEPA filter testing uses a method (``Alternate Shroud Test'') 
which is no longer specified in ASME N510-1989. But this method is 
specified in ANSI N510-1975 and was used when the plant was 
licensed. In addition, the PACU in-place HEPA filter testing 
methodology which is employed at SONGS has a downstream point 
location which differs from the location suggested in ASME N510-
1989. ANSI N510-1975, while providing a suggestion where downstream 
sample could be located, nevertheless does not provide a specific 
location. The test acceptance criteria are the same for methods 
cited in ANSI N510-1975 and ASME N510-1989. The method which is 
employed at SONGS provides more conservative results because the 
test is performed on individual HEPA filters, which ensures that 
each of the HEPA filters in the tested bank meets the acceptance 
criteria.
    The locations of the PACU HEPA downstream sample points are 
different from the location suggested in ASME N510-1989, though they 
meet the requirements delineated in ANSI N510-1975. ANSI N510-1975 
requires that a single representative downstream sample point be 
established, if possible, at the location where adequate mixing may 
be achieved, or at a point downstream of a fan, or multiple 
downstream sampling points may be used (such as in the Alternate 
Shroud Technique used in the CREACUS system) if a single downstream 
sample point is not feasible.
    Since the HEPA filters are tested to the same acceptance 
criteria, and the testing methodology is permitted by ANSI N510-
1975, to which the plant is licensed, it is concluded that the 
proposed change will not involve a significant increase in the 
probability or consequences of an accident previously evaluated.
    Section 5.5.12.d will be modified by the proposed change by 
deleting the references to RG 1.52, Revision 2 and ASME N510-1989. 
There are no requirements for pressure drop test across combined 
HEPA filters, the prefilters, and the charcoal absorbers in RG 1.52, 
Revision 2, and ASME N510-1989. The proposed version of section 
5.5.2.12.d reads:
    ``Testing to demonstrate the pressure drop across the combined 
HEPA filters, the prefilters, and the charcoal absorbers, when 
tested at the appropriate system flowrate.''
    The proposed change clarifies the statement of section 
5.5.2.12.d. Pressure drop testing across combined HEPA filters, the 
prefilters, and the charcoal adsorbers is industry-wide practice 
which is based on good engineering practice and operating 
experience.
    Therefore, the probability or consequences of an accident 
previously evaluated will not be increased by operating the facility 
in accordance with this proposed change.
    (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?
    Response: No.
    The proposed change does not change the design or configuration 
of the plant. The proposed change is to change the reference to ASME 
Code in subsection 5.5.2.12.a and 5.5.2.12.b from ASME N510-1989 to 
ANSI N510-1975 to more clearly reflect the standard used. Also, 
subsection 5.5.2.12.d will be changed by deleting the references to 
RG 1.52, Revision 2, and ASME N510-1989 regarding pressure drop test 
across HEPA filters. RG 1.52, Revision 2, and ASME N510-1989 do not 
require pressure drop test across HEPA filters.
    Therefore, this proposed change will not create the possibility 
of a new or different kind of accident from any accident that has 
been previously evaluated.
    (3) Will operation of the facility in accordance with this 
proposed change involve a significant reduction in a margin of 
safety?
    Response: No.
    The proposed change is to change the reference to the ASME Code 
in subsections 5.5.2.12.a and 5.5.2.12.b from ASME N510-1989 to ANSI 
N510-1975. The CREACUS units' HEPA filters are currently tested to 
ANSI N510-1975. Although the test methodology is slightly different 
than that in ASME N510-1989, the acceptance criteria are the same 
and the current methodology is conservative. Thus the current 
testing satisfies the acceptance criteria of ASME N510-1989, even 
though the test method is different.
    The current methodology for HEPA filter testing will not change 
as a result of the proposed change. Also, deletion of reference to 
RG 1.52, Revision 2, and ASME N510-1989 from subsection 5.5.2.12.d 
clarifies this section because these standards do not require HEPA 
filters pressure drop test. Consequently, there is no change to the 
design or operation of the plant as a result of this change.

[[Page 20011]]

    Therefore, operation of the facility in accordance with this 
proposed change will not involve a significant reduction in a margin 
of safety.

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

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

    Date of application for amendments: March 12, 2001 (TS 99-18).
    Brief description of amendments: The proposed amendment would 
revise the Sequoyah Nuclear Plant (SQN) Technical Specifications (TSs). 
The revision would delete TS 4.7.7.a and add proposed TS 3/4.7.13 
regarding the control room air conditioning system to make the SQN TSs 
more consistent with the Westinghouse Standard TSs.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), Tennessee Valley 
Authority (TVA), the licensee, has provided its analysis of the issue 
of no significant hazards consideration, which is presented below:

    A. The proposed amendment does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    TVA has identified Surveillance Requirement (SR) 4.7.7.a, which 
determines operability of the main control room emergency 
ventilation system (CREVS) relative to temperature, to be inadequate 
and nonconservative. TVA proposes to deleted this SR coincident with 
the addition of a new TS 3/4.7.13. The proposed TS addition for the 
main control room air-conditioning system (CRACS) provides a more 
adequate SR for determination of operability with associated actions 
to take for inoperability; resolves an inadequate TS in accordance 
with the guidance in NRC [U.S. Nuclear Regulatory Commission] 
Administrative Letter 98-10; establishes clarity between CRACS and 
CREVS; and provides greater consistency with NUREG-1431 and TSTF-51 
[TS Task Force issue Traveler No. 51], Revision 2. These proposed 
revisions are conservative and are not the result of a change to 
plant equipment, system design, testing methods, or operating 
practices. Since the proposed revisions will increase conservatism 
and the systems will continue to meet their required safety function 
without plant modification or operating practices, the change does 
not involve a significant increase in the probability or 
consequences of an accident previously evaluated.
    B. The proposed amendment does not create the possibility of a 
new or different kind of accident from any accident previously 
evaluated.
    The proposed revisions to the SQN TSs will not alter plant 
equipment or operating practices. The change will not result in the 
installation of any new equipment or systems. The intent of deleting 
the SR and adding a specification is to address a nonconservative 
TS, provide clarification of plant systems, and improve consistency 
with NUREG-1431. Since the systems' functions are associated with 
accident mitigation and will continue to perform without change and 
were not previously considered to contribute to accident generation, 
the proposed changes will not create the possibility of a new or 
different kind of accident.
    C. The proposed amendment does not involve a significant 
reduction in a margin of safety.
    Both the main control room (MCR) emergency ventilation and air-
conditioning systems provide for the safe, uninterrupted occupancy 
of the MCR during an accident and the subsequent recovery period. 
The proposed TS revisions will not change the methods of operating 
the plant or setpoints associated with safety-related equipment in 
the implementation of this request. Therefore, the proposed 
revisions do not involve a reduction in the margin of safety.

    The NRC has reviewed the licensee's analysis and, based on this 
review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment 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 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, Docket No. 50-289, Three Mile Island 
Nuclear Station, Unit 1, Dauphin County, Pennsylvania

    Date of application for amendment: January 15, 2001.
    Brief description of amendment: The amendment revised TS 5.4.2(f) 
to remove the fuel assembly U-235 loading criterion for fuel assemblies 
stored in the spent fuel storage pool.
    Date of issuance: March 26, 2001.
    Effective date: As of the date of issuance and shall be implemented 
within 30 days of issuance including issuance of approval of changes to 
the Updated Final Safety Analysis Report as described in the Nuclear 
Regulatory Commission staff's safety evaluation.
    Amendment No.: 231.
    Facility Operating License No. DPR-50. Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: February 21, 2001 (66 
FR 11051).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated March 26, 2001.
    No significant hazards consideration comments received: No.

[[Page 20012]]

AmerGen Energy Company, LLC, Docket No. 50-289, Three Mile Island 
Nuclear Station, Unit 1, Dauphin County, Pennsylvania

    Date of application for amendment: October 20, 2000, as 
supplemented March 14, 2001.
    The March 14, 2001, letter provided additional clarifying 
information, which did not change the initial proposed no significant 
hazards consideration determination or expand the amendment beyond the 
scope of the original notice.
    Brief description of amendment: The amendment revised the frequency 
for maintenance inspections of the emergency diesel generators from 
annually to once every 2 years and stated that the inspections shall be 
conducted in accordance with procedures developed in conjunction with 
applicable Fairbanks Morse Owners Group and manufacturer's 
recommendations.
    Date of issuance: March 29, 2001.
    Effective date: As of the date of issuance and shall be implemented 
within 30 days.
    Amendment No.: 232.
    Facility Operating License No. DPR-50. Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: January 10, 2001 (66 FR 
2012).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated March 29, 2001.
    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: December 1, 2000.
    Brief description of amendments: The amendments revise the value of 
the minimum departure from nucleate boiling ratio (DNBR) from 
`` 1.30'' in the current Technical Specifications (TSs) to 
`` 1.3 (through operating cycle 10)'' and `` 1.34 
(operating cycle 11 and later)'' in the safety limits TS 2.1.1.1 and in 
function 15, DNBR--Low, in Table 3.3.1-1, ``Reactor Protective System 
Instrumentation.'' The amendments are structured such that the 
`` 1.34'' would become effective for each unit in operating 
cycle 11 and later. Operating cycle 11 begins in spring 2002 for Unit 
2, in fall 2002 for Unit 1, and in spring 2003 for Unit 3. From now to 
operating cycle 11, the `` 1.30'' will remain the minimum 
DNBR requirement for the three units.
    Date of issuance: March 28, 2001
    Effective date: March 28, 2001, and shall be implemented within 60 
days of the date of issuance.
    Amendment Nos.: Unit 1-133, Unit 2-133, Unit 3-133
    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 24, 2001 (66 FR 
7670)
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated March 28, 2001.
    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: December 5, 2000.
    Brief description of amendments: The amendments revise the action 
statement for Specification 3.7.5, ``Auxiliary Feedwater (AFW) 
System,'' of the Technical Specifications (TSs). The amendments 
incorporate NRC-approved TS Task Force (TSTF) Traveler Number TSTF-340, 
Revision 3, to allow a 7-day completion time for the turbine-driven AFW 
pump if inoperability occurs in reactor Mode 3 following a refueling 
outage, and if Mode 2 had not been entered.
    Date of issuance: March 29, 2001.
    Effective date: March 29, 2001, and shall be implemented within 45 
days of the date of issuance.
    Amendment Nos.: Unit 1-134, Unit 2-134, Unit 3-134.
    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 24, 2001.
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated March 29, 2001.
    No significant hazards consideration comments received: No.

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

    Date of application for amendments: January 25, 2000, as 
supplemented on October 31, 2000, and December 18, 2000.
    Brief description of amendments: The amendments revise the Calvert 
Cliffs Technical Specifications to eliminate response time testing for 
those pressure sensors which were discussed and approved in the 
Combustion Engineering Owners Group Topical Report NPSD-1167.
    Date of issuance: March 22, 2001.
    Effective date: As of the date of issuance to be implemented within 
30 days.
    Amendment Nos.: 244 and 218.
    Renewed Facility Operating License Nos. DPR-53 and DPR-69: 
Amendments revised the Technical Specifications.
    Date of initial notice in Federal Register: February 9, 2000 (65 FR 
6403).
    The October 31 and December 18, 2000, letters provided clarifying 
information that did not change the initial proposed no significant 
hazards consideration determination. The Commission's related 
evaluation of these amendments is contained in a Safety Evaluation 
dated March 22, 2001.
    No significant hazards consideration comments received: No.

Carolina Power & Light Company, et al., Docket Nos. 50-32 and 50-324, 
Brunswick Steam Electric Plant, Units 1 and 2, Brunswick County, North 
Carolina

    Date of application for amendments: December 1, 2000.
    Brief Description of amendments: The amendments change the 
Technical Specifications for the submittal date of the ``Radioactive 
Effluent Release Report'' to ``prior to May 1'' of each year.
    Date of issuance: March 21, 2001.
    Effective date: March 21, 2001.
    Amendment Nos.: 212 and 239.
    Facility OperatingLicense Nos. DPR-71 and DPR-62: Amendments change 
the Technical Specifications.
    Date of initial notice in Federal Register: February 7, 2001 (66 FR 
9381).
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated March 21, 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, as revised by 
letter dated January 12, 2001.
    Brief description of amendment: The amendment changes the Technical 
Specifications (TSs) to allow Type B and C containment leak rate 
testing to be performed in accordance with 10 CFR Part 50, Appendix J, 
Option B. The amendment also increases the interval in TS Surveillance 
Requirement 3.6.2.2 for containment air lock door interlock testing 
from 18 months to 24 months.

[[Page 20013]]

    Date of issuance: March 30, 2001.
    Effective date: As of the date of issuance, to be implemented 
within 60 days.
    Amendment No.: 194.
    Facility Operating License No. DPR-20. Amendment changed the 
Technical Specifications.
    Date of initial notice in Federal Register: January 24, 2001 (66 FR 
7676).
    The licensee's letter dated January 12, 2001, 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 March 30, 2001.
    No significant hazards consideration comments received: No.

Duke Energy Corporation, et al., Docket No. 50-413, Catawba Nuclear 
Station, Unit 1, York County, South Carolina

    Date of application for amendment: February 20, 2001.
    Brief description of amendment: The amendment revised the Technical 
Specifications Table 3.3.2-1 for Catawba Nuclear Station Unit 1. It 
modified the required actions for the Engineered Safety Feature 
Actuation System Table 3.3.2-1, function 6.f (auxiliary feedwater 
(AFW), auxiliary feedwater pump train A and train B suction transfer on 
suction pressure--low) on a one-time basis. The proposed one-time 
change will require that if more than one channel of low suction 
pressure instrumentation becomes inoperable, in lieu of requiring unit 
shutdown within 7 hours, the licensee will immediately enter the 
applicable condition(s) or required action(s) for the associated AFW 
train made inoperable by the inoperable channels. This modification 
will support the timely replacement of a broken pressure switch in the 
Train B of AFW suction transfer on low suction pressure function.
    Date of issuance: April 6, 2001.
    Effective date: As of the date of issuance and shall be implemented 
within 30 days from the date of issuance.
    Amendment No.: 190.
    Facility Operating License No. NPF-35: Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: February 27, 2001 (66 
FR 12568). The Commission's related evaluation of the amendment is 
contained in a Safety Evaluation dated April 6, 2001.
    No significant hazards consideration comments received: No.

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

    Date of application for amendment: November 2, 2000.
    Brief description of amendment: The amendment made the following 
changes: (1) added a new Technical Specification (TS) 3.3.1.3, 
``Oscillation Power Range Monitoring (OPRM) Instrumentation,'' (2) 
revised TS 3.4.1, ``Recirculation Loops Operating,'' to remove 
monitoring specifications that are no longer needed upon activation of 
the automatic OPRM instrumentation, and (3) revised TS 5.6.5 to include 
in the Core Operating Limits Report the applicable operating limits for 
the OPRM and also reference the topical report that describes the 
analytical methods used to determine the setpoint values for the OPRM.
    Date of issuance: April 5, 2001.
    Effective date: April 5, 2001, and shall be implemented prior to 
restart from Refuel Outage 15.
    Amendment No.: 171.
    Facility Operating License No. NPF-21: The amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: December 13, 2000 (65 
FR 77916).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated April 5, 2001.
    No significant hazards consideration comments received: No.

Energy Nuclear Generation Company, Docket No. 50-293, Pilgrim Nuclear 
Power Station, Plymouth County, Massachusetts

    Date of application for amendment: September 1, 2000.
    Brief description of amendment: The amendment approves a change to 
the Pilgrim Technical Specification Table 4.6-3. The change modifies 
the reactor pressure vessel surveillance capsule withdrawal schedule by 
substituting ``21 (approx)'' under the column ``Effective Full Power 
Years (EFPY)'' for the current ``18 (approx).''
    Date of issuance: April 2, 2001.
    Effective date: As of the date of issuance, and shall be 
implemented within 30 days.
    Amendment No.: 188.
    Facility Operating License No. DPR-35: Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: November 1, 2000 (65 FR 
65342).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated April 2, 2001.
    No significant hazards consideration comments received: No.

Entergy Operations, Inc., Docket No. 50-313, Arkansas Nuclear One, Unit 
No. 1, Pope County, Arkansas

    Date of amendment request: September 28, 2000, as supplemented by 
letters dated October 26, 2000, and February 19, 2001.
    Brief description of amendment: The amendment changes the Arkansas 
Nuclear One, Unit 1 technical specifications to allow a revised reroll 
repair process for the steam generators.
    Date of issuance: March 28, 2001.
    Effective date: As of the date of issuance and shall be implemented 
within 30 days from the date of issuance.
    Amendment No.: 212.
    Facility Operating License No. DPR-51: Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: December 13, 2000 (65 
FR 77519).
    The supplemental letters dated October 26, 2000, and February 19, 
2001, provided additional information that did not change the initial 
proposed no significant hazards consideration determination or expand 
the scope of the application.
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated March 28, 2001.
    No significant hazards consideration comments received: No.

Entergy Operations, Inc., Docket No. 50-313, Arkansas Nuclear One, Unit 
No. 1, Pope County, Arkansas

    Date of amendment request: August 29, 2000, as supplemented by 
letter dated March 2, 2001.
    Brief description of amendment: The amendment revises the TS to 
allow steam generator tubes to remain in service with indications of 
outer diameter intergranular attack (ODIGA) in the upper tubesheet 
region of the steam generators.
    Date of issuance: March 28, 2001.
    Effective date: As of the date of issuance and shall be implemented 
within 30 days from the date of issuance.
    Amendment No.: 213.
    Facility Operating License No. DPR-51: Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: December 13, 2000 (65 
FR 77917).
    The supplemental letter dated March 2, 2001, provided clarifying 
information that did not change the initial proposed no significant 
hazards consideration

[[Page 20014]]

determination or expand the scope of the application.
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated March 28, 2001.
    No significant hazards consideration comments received: No.

Exelon Generation Company, LLC, Docket No. 50-353, Limerick Generating 
Station, Unit 2, Montgomery County, Pennsylvania

    Date of application for amendment: November 20, 2000.
    Brief description of amendment: Revised Technical Specification 
Figure 3.4.6.1-1, which affects heatup, cooldown and inservice test 
Pressure-Temperature limitations. The revisions are applicable until 
the end of Operating Cycle 7.
    Date of issuance: March 23, 2001.
    Effective date: As of the date of issuance, to be implemented 
within 30 days.
    Amendment No.: 111.
    Facility Operating License No. NPF-85. This amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: February 21, 2001 (66 
FR 11058).
    The December 20, 2000, letter provided additional information that 
did not change the initial proposed no significant hazards 
consideration determination or expand the application beyond the scope 
of the original Federal Register notice.
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated March 23, 2001.
    No significant hazards consideration comments received: No.

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

    Date of application for amendments: January 18, 2001, as 
supplemented February 20, and March 26, 2001.
    Brief description of amendments: The amendments changed the 
Technical Specification (TS) Table 1.2, ``Operational Conditions,'' and 
TS 3/4.9.1, ``Reactor Mode Switch,'' to allow movement of a single 
control rod with the reactor in hot shutdown or cold shutdown for post-
maintenance and surveillance testing of the control rod and the control 
rod drive. The amendments also changed TS Table 3.3.1-1 to require the 
nuclear instrumentation system intermediate range monitors (IRMs) to be 
operable when moving a control rod in hot shutdown or cold shutdown. TS 
Table 4.3.1.1-1 is changed to add surveillance requirements for the 
IRMs in hot or cold shutdown.
    Date of issuance: April 5, 2001.
    Effective date: As of the date of issuance, to be implemented 
within 30 days.
    Amendments Nos.: 149 and 112.
    Facility Operating License Nos. NPF-39 and NPF-85: The amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: February 21, 2001 (66 
FR 11060). The February 20, and March 26, 2001, letters provided 
clarifying information that did not change the initial proposed no 
significant hazards consideration determination or expand the 
application beyond the scope of the original Federal Register notice.
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated April 5, 2001.
    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: January 18, 2001.
    Brief description of amendments: The amendments changed Technical 
Specification Surveillance Requirement 4.9.2.d to allow the shorting 
links to remain in place if adequate shutdown margin has been 
demonstrated.
    Date of issuance: April 5, 2001.
    Effective date: As of the date of issuance, to be implemented 
within 30 days.
    Amendments Nos.: 150 and 113.
    Facility Operating License Nos. NPF-39 and NPF-85: The amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: February 21, 2001 (66 
FR 11059).
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated April 5, 2001.
    No significant hazards consideration comments received: No.

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

    Date of application for amendments: December 27, 1999.
    Brief description of amendments: The amendments revise Technical 
Specifications (TSs) to increase allowable out-of-service times (AOTs) 
and surveillance test intervals (STIs) for selected actuation 
instrumentation. The amendments implement AOT/STI changes based on 
Topical Reports by General Electric Company and the Boiling Water 
Reactor Owners' Group which have previously been reviewed and approved 
by the Nuclear Regulatory Commission (NRC).
    Date of issuance: March 28, 2001.
    Effective date: Immediately, to be implemented within 120 days.
    Amendment Nos.: 198 and 194.
    Facility Operating License Nos. DPR-29 and DPR-30: The amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: August 9, 2000 (65 FR 
48746).
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated March 28, 2001.
    No significant hazards consideration comments received: No.

FirstEnergy Nuclear Operating Company, et al., Docket Nos. 50-334 and 
50-412, Beaver Valley Power Station (BVPS), Unit Nos. 1 and 2, 
Shippingport, Pennsylvania

    Date of application for amendments: May 12, 2000, as supplemented 
by letters dated June 19, November 2, and December 1, 2000, and January 
29, 2001.
    Brief description of amendments: This amendment authorized changes 
to the BVPS-1 and 2, Updated Final Safety Analysis Reports (UFSARs) 
with regard to selected design-basis accident dose consequence 
calculations. For BVPS-1, changes involve the following DBAs: loss of 
offsite alternating-current (AC) power, fuel-handling accident, 
accidental release of waste gas, steam generator tube rupture, rod 
cluster control assembly ejection, single reactor coolant pump locked 
rotor, and loss of reactor coolant for small ruptured pipes/loss-of-
coolant accidents. For BVPS-2, changes involve the following DBAs: 
steam system piping failures (or main steam line break), loss of AC 
power, reactor coolant pump shaft seizure, rod cluster control assembly 
ejection, failure of small lines carrying primary coolant outside 
containment, steam generator tube rupture, loss-of-coolant accidents, 
and waste gas system failure.
    Date of issuance: March 22, 2001.
    Effective date: This license amendment is effective as of the date 
of its issuance and shall be implemented by the next update to the 
UFSAR as required by 10 CFR 50.71(e). Implementation of the amendment 
requires the incorporation in the UFSAR of the changes to the 
description of the facility as described in the licensee's application 
dated May 12, 2000, as supplemented June 19, November 2, and December 
1, 2000, and January 29, 2001.
    Amendment Nos.: 237 and 119

[[Page 20015]]

    Facility Operating License Nos. DPR-66 and NPF-73: Amendments 
authorize revisions to the BVPS-1 and 2 UFSARs.
    Date of initial notice in Federal Register: September 6, 2000 (65 
FR 54086).
    The June 19, 2000, letter revised the licensee's no significant 
hazards evaluation and was used, with the original submittal, as a 
basis for the staff's proposed no significant hazards consideration 
determination which was published on September 6, 2000. The November 2, 
and December 1, 2000, and January 29, 2001, letters provided clarifying 
information that did not change the initial proposed no significant 
hazards consideration determination or expand the amendment beyond the 
scope of the original notice.
    The Commission's related evaluation of the UFSAR changes is 
contained in a Safety Evaluation dated March 22, 2001.
    No significant hazards consideration comments received: No.

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

    Date of application for amendment: November 27, 2000, as 
supplemented March 12, 2001.
    Brief description of amendment: The amendments delete Technical 
Specifications Section 6.8.4.3, ``Post-Accident Sampling,'' thereby 
eliminating the requirements to have and maintain the post-accident 
sampling system.
    Date of Issuance: March 27, 2001.
    Effective Date: March 27, 2001.
    Amendment No.: 174 and 114.
    Facility Operating License Nos. DPR-67 and NPF-16: Amendment 
revised the Technical Specifications.
    Date of initial notice in Federal Register: December 27, 2000 (65 
FR 81921). The March 12, 2001, supplement 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 March 27, 2001.
    No significant hazards consideration comments received: No.

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

    Date of application for amendment: August 31, 2000.
    Brief description of amendment: This amendment revised Technical 
Specification (TS) and associated Bases pages Table 3.3.18-1, ``Remote 
Shutdown System Instrumentation.'' The list of instruments that would 
be used by operators to place and maintain the plant in a safe shutdown 
condition from outside the control room have been modified consistent 
with recent plant modifications and changes to the approach to achieve 
and maintain a safe shutdown condition.
    Date of issuance: March 22, 2001.
    Effective date: Date of issuance, to be implemented prior to Fall 
2001 Restart.
    Amendment No.: 196.
    Facility Operating License No. DPR-72: Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: October 4, 2000 (65 FR 
59223).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated March 22, 2001.
    No significant hazards consideration comments received: No.

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

    Date of application for amendments: November 15, 2000, as 
supplemented March 7, 2001.
    Brief description of amendments: The proposed amendment would 
revise Technical Specification (TS) 3.2.6, ``Allowable Power Level--
APL,'' and TS 1.38, ``Allowable Power Level (APL),'' definitions of APL 
to make them consistent throughout the TSs.
    Date of issuance: March 29, 2001.
    Effective date: As of the date of issuance and shall be implemented 
within 30 days.
    Amendment Nos.: 251, 233.
    Facility Operating License Nos. DPR-58 and DPR-74: Amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: December 27, 2000 (65 
FR 81924). The supplement contained clarifying information and did not 
change the initial no significant hazards consideration determination 
or expand the scope of the original Federal Register notice.
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated March 29, 2001.
    No significant hazards consideration comments received: No.

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

    Date of application for amendment: January 2, 2001, as supplemented 
March 5, 2001.
    Brief description of amendment: The amendment revises Technical 
Specifications (TS) 3/4.6.2.2.a for the Unit 1 spray additive tank to 
require a contained volume between 4000 and 4600 gallons of between 30 
and 34 percent by weight sodium hydroxide (NaOH) solution. In addition, 
the amendment makes four types of format changes to the TS pages for 
Unit 1.
    Date of issuance: March 29, 2001.
    Effective date: As of the date of issuance and shall be implemented 
within 30 days.
    Amendment No.: 252.
    Facility Operating License No. DPR-58: Amendment revises the 
Technical Specifications.
    Date of initial notice in Federal Register: January 24, 2001 (66 FR 
7681).
    The March 5, 2001, supplemental letter did not change the scope of 
the proposed action and did not change the NRC's proposed no 
significant hazards consideration determination.
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated March 29, 2001.
    No significant hazards consideration comments received: No.

Niagara Mohawk Power Corporation, Docket Nos. 50-387 and 50-388, 
Susquehanna Steam Electric Station, Units 1 and 2, Luzerne County, 
Pennsylvania

    Date of application for amendment: January 13, 2000 (submitted by 
PP&L, Inc., the licensee before July 1, 2000), as supplemented 
September 6, 2000 (submitted by PPL Susquehanna, LLC, the licensee on 
and after July 1, 2000).
    Brief description of amendment: The amendments made administrative 
changes to the Technical Specifications correcting the wording of the 
legends in Figure 3.4.10.1, ``Reactor Vessel Pressure vs. Minimum 
Vessel Temperature,'' for both units, and correcting administrative 
errors in Section 5.6.5.b, regarding the Core Operating Limits Report, 
for Unit 2.
    Date of issuance: March 30, 2001.
    Effective date: As of the date of issuance to be implemented within 
30 days.
    Amendment Nos.: 192 and 167.
    Facility Operating License Nos. NPF-14 and NPF-22: Amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: April 5, 2000 (65 FR 
17918).
    The staff's related evaluation of the amendment is contained in a 
Safety Evaluation dated March 30, 2001.
    No significant hazards consideration comments received: No.

[[Page 20016]]

Northeast Nuclear Energy Company, et al., Docket No. 50-245, Millstone 
Nuclear Power Station, Unit No. 1, New London County, Connecticut

    Date of amendment request: August 31, 2000, as supplemented October 
12 and November 8, 2000, and February 16, 2001.
    Brief description of amendment: This amendment conforms the license 
to reflect the transfer of Operating License No. DPR-21 for the 
Millstone Nuclear Power Station, Unit No. 1 to the extent held by the 
Selling Owners to Dominion Nuclear Connecticut, Inc., as previously 
approved by an Order dated March 9, 2001.
    Date of issuance: March 31, 2001.
    Effective date: As of the date of issuance and shall be implemented 
within 60 days from the date of issuance.
    Amendment No.: 109.
    Facility Operating License No. DPR-21: The amendment revised the 
Operating License and Technical Specifications.
    Date of initial notice in Federal Register: October 24, 2000 (65 FR 
63630).
    The October 12 and November 8, 2000, and February 16, 2001 
supplements provided clarifying information and did not expand the 
scope of the application as originally published.
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated March 9, 2001.
    No significant hazards consideration comments received: No.

Northeast Nuclear Energy Company, et al., Docket No. 50-336, Millstone 
Nuclear Power Station, Unit No. 2, New London County, Connecticut

    Date of application for amendment: July 31, 2000, as supplemented 
January 4, 2001.
    Brief description of amendment: This amendment authorizes changes 
to the Millstone Nuclear Power Station, Unit No. 2 Final Safety 
Analysis Report (FSAR) to allow the use of the Siemens Power 
Corporation U.S. Nuclear Regulatory Commission-approved methodology for 
determining the fuel centerline melt linear heat rate limit (FCMLHRL) 
on a cycle-by-cycle basis. Northeast Nuclear Energy Company evaluated 
this method of calculating FCMLHRL utilizing the criteria of 10 CFR 
50.59 and determined that this change required NRC approval before 
implementation. Technical Specification Bases 2.1.1, ``Reactor Core,'' 
has also been revised accordingly.
    Date of issuance: March 29, 2001.
    Effective date: As of the date of issuance and shall be implemented 
no later than the date of submission of the next update of the FSAR.
    Amendment No.: 255.
    Facility Operating License No. DPR-65: Amendment authorized changes 
to the FSAR and revised the Technical Specifications.
    Date of initial notice in Federal Register: January 24, 2001 (66 FR 
7684).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated March 29, 2001.
    No significant hazards consideration comments received: No.

Northeast Nuclear Energy Company, et al., Docket No. 50-336 and 50-423, 
Millstone Nuclear Power Station, Unit Nos. 2 and 3, New London County, 
Connecticut

    Date of application for amendment: August 31, 2000, as supplemented 
October 12 and November 8, 2000, and February 16, 2001.
    Brief description of amendment: These amendments conform the 
licenses to reflect the transfer of Operating Licenses Nos. DPR-65 and 
NPF-49 for the Millstone Nuclear Power Station, Unit Nos. 2 and 3, to 
the extent held by the Selling Owners to Dominion Nuclear Connecticut, 
Inc., as previously approved by an Order dated March 9, 2001.
    Date of issuance: March 31, 2001.
    Effective date: As of the date of issuance and shall be implemented 
within 60 days from the date of issuance.
    Amendment Nos.: 256 and 196.
    Facility Operating License Nos. DPR-65 and NPF-49: These amendments 
revise the operating licenses and Millstone, Unit 2, Technical 
Specifications.
    Date of initial notice in Federal Register: October 24, 2000 (65 FR 
63630).
    The October 12 and November 8, 2000, and February 16, 2001, 
supplements provided clarifying information and did not expand the 
scope of the application as originally published.
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated March 9, 2001.
    No significant hazards consideration comments received: No.

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

    Date of application for amendment: February 5, 2001.
    Brief description of amendment: The amendment revised the Kewaunee 
Nuclear Power Plant (KNPP) Technical Specifications (TSs) 3.1.d.2 to 
reduce the maximum allowable leakage of primary system reactor coolant 
to the secondary system from 500 gallons per day (gpd) through any one 
steam generator to 150 gpd through any one steam generator. In 
addition, the amendment removes reference to the voltage based repair 
criteria.
    Date of issuance: March 27, 2001.
    Effective date: As of the date of issuance and shall be implemented 
within 30 days.
    Amendment No.: 153.
    Facility Operating License No. DPR-43: Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: February 21, 2001 (66 
FR 11062).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated March 27, 2001.
    No significant hazards consideration comments received: No.

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

    Date of application for amendment: December 13, 2000, as 
supplemented April 3, 2001.
    Brief description of amendment: The amendment changes License 
Condition 2.C.4 to conform to NRC Generic Letter (GL) 86-10, 
``Implementation of Fire Protection Requirements.'' The amendment also 
relocates the Fire Protection Program (FPP) elements from the Technical 
Specifications to the licensee-controlled FPP, in accordance with GL 
86-10 and GL 88-12, ``Removal of Fire Protection Requirements from 
Technical Specifications.''
    Date of issuance: April 5, 2001.
    Effective date: As of the date of issuance and shall be implemented 
within 45 days.
    Amendment No.: 119.
    Facility Operating License No. DPR-22. Amendment revised the 
Operating License and Technical Specifications.
    Date of initial notice in Federal Register: January 24, 2001 (66 FR 
7684).
    The April 3, 2001, 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 5, 2001.
    No significant hazards consideration comments received: No.

[[Page 20017]]

Omaha Public Power District, Docket No. 50-285, Fort Calhoun Station, 
Unit No. 1, Washington County, Nebraska

    Date of amendment request: October 18, 2000.
    Brief description of amendment: The amendment revised Section 
2.1.2, Figures 2-1A and 2-1B, and the associated Bases of the Fort 
Calhoun Station Technical Specifications to extend the existing 
pressure-temperature (P-T) curves from 20 effective full power years 
(EFPY) to 24.25 EFPY. Additionally, the amendment deletes Figure 2-3, 
``Predicted Radiation Induced NDTT Shift'' and updates the fluence 
analysis for projecting RTNDT at 24.25 EFPY.
    Date of issuance: March 27, 2001.
    Effective date: March 27, 2001, and shall be implemented within 30 
days from the date of issuance.
    Amendment No.: 197.
    Facility Operating License No. DPR-40. Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: December 27, 2000 (65 
FR 81925).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated March 27, 2001.
    No significant hazards consideration comments received: No.

Omaha Public Power District, Docket No. 50-285, Fort Calhoun Station, 
Unit No. 1, Washington County, Nebraska

    Date of amendment request: April 14, 2000, as supplemented by 
letters dated June 2, July 28, and December 1, 2000, and January 31, 
2001.
    Brief description of amendment: The amendment changed the 
surveillance requirements for laboratory testing of the charcoal 
adsorbers for the control room, the spent fuel pool storage area and 
the safety injection pump rooms. In addition, the amendment deletes the 
laboratory testing requirements for the containment charcoal adsorbers. 
The changes comply with the guidance of Generic Letter 99-02, 
``Laboratory Testing of Nuclear-Grade Activated Charcoal.''
    Date of issuance: April 4, 2001.
    Effective date: April 4, 2001, and shall be implemented within 60 
days from the date of issuance.
    Amendment No.: 198.
    Facility Operating License No. DPR-40. Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: March 5, 2001 (66 FR 
13355).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated April 4, 2001.
    No significant hazards consideration comments received: No.

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

    Date of application for amendments: May 15, 2000, as supplemented 
on August 25, 2000.
    Brief description of amendments: The amendments revise Technical 
Specifications (TSs) requirements to test the remaining diesel 
generators (DGs) when one of the two independent offsite power sources 
is inoperable, as described in Section 3/4.8.1, Action a; and when a DG 
is inoperable for other than preventive maintenance reasons, as 
described in Section 3/4.8.1, Action b, of the TSs. The amendments also 
expand the DG loading band from existing 2500-2600 KW to 2330-2600 KW 
for the monthly, 6-month, and the 2-hour loaded prerequisite for the 
hot restart tests, and correct an administrative oversight.
    Date of issuance: April 2, 2001.
    Effective date: As of the date of issuance, and shall be 
implemented within 60 days.
    Amendment Nos.: 242 and 223.
    Facility Operating License Nos. DPR-70 and DPR-75: The amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: July 12, 2000 (65 FR 
43052).
    The August 25, 2000, supplement did not change the conclusions made 
in the Commission's initial proposed no significant hazards 
consideration determination.
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated April 2, 2001.
    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 11, 2001.
    Brief description of amendments: The amendments delete Technical 
Specifications (TS) Section 5.5.2.2, ``Post Accident Sampling 
Program,'' for San Onofre Nuclear Generating Station, Units 2 and 3, 
and thereby eliminate the requirements to have and maintain the post-
accident sampling systems (PASS). The amendments also revise TS 
5.5.2.8, ``Primary Coolant Sources Outside Containment Program,'' to 
reflect the elimination of PASS. Additionally, the amendments delete 
PASS-related License Conditions 2.c(19)i for Unit 2 and 2.C.(17)d for 
Unit 3.
    Date of issuance: March 26, 2001.
    Effective date: March 26, 2001, to be implemented within 60 days of 
issuance.
    Amendment Nos.: Unit 2--178; Unit 3-169.
    Facility Operating License Nos. NPF-10 and NPF-15: The amendments 
revised the Facility Operating Licenses and Technical Specifications.
    Date of initial notice in Federal Register: February 21, 2001 (66 
FR 11063).
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated March 26, 2001.
    No significant hazards consideration comments received: No.

Tennessee Valley Authority, Docket No. 50-260, Browns Ferry Nuclear 
Plant, Unit 2, Limestone County, Alabama

    Date of application for amendment: February 5, 2001.
    Brief description of amendment: The amendment authorizes a one 
cycle delay in removal of the second capsule.
    Date of issuance: April 2, 2001.
    Effective date: April 2, 2001.
    Amendment No.: 271.
    Facility Operating License No. DPR-52: The amendment revises the 
Reactor Vessel Material Surveillance Program.
    Date of initial notice in Federal Register: February 28, 2001.
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated April 2, 2001.
    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 22, 2001 (TS 00-01).
    Brief description of amendments: These amendments revised the 
Technical Specifications (TSs) by revising the surveillance test 
requirement to assess flow blockage in the ice condenser containment.
    Date of issuance: March 22, 2001.
    Effective date: March 22, 2001.
    Amendment Nos.: 267 and 258.
    Facility Operating License Nos. DPR-77 and DPR-79: Amendments 
revised the TSs.
    Date of initial notice in Federal Register: February 7, 2001 (66 FR 
9388).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated March 22, 2001.
    No significant hazards consideration comments received: No.

[[Page 20018]]

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

    Date of application for amendment: September 26, 2000.
    Brief description of amendment: This amendment revises Technical 
Specification (TS) requirements regarding secondary containment 
systems, including the Standby Gas Treatment System. The affected TS 
sections are 1.0, Definitions; 3/4.7.B, Standby Gas Treatment System; 
and 3/4.7.C, Secondary Containment System. In addition, a new TS 
section, 3/4.7.E, Reactor Building Automatic Ventilation System 
Isolation Valves is proposed. Some of the proposed changes are 
administrative in nature and do not affect the technical aspects of the 
requirements. Associated changes to the TS Bases are also being made to 
conform to the changed TS. The proposed changes provide certain 
additional flexibility in operations when equipment is made or found to 
be inoperable, while also ensuring appropriate actions are taken to 
place the plant in a safe condition under such conditions.
    Date of Issuance: March 23, 2001.
    Effective date: As of the date of issuance, and shall be 
implemented within 60 days.
    Amendment No.: 197.
    Facility Operating License No. DPR-28: Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: October 18, 2000 (65 FR 
62394).
    The Commission's related evaluation of this amendment is contained 
in a Safety Evaluation dated March 23, 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: October 25, 2000.
    Brief description of amendment: The amendment revises the 125-volt 
DC station battery system Technical Specifications Section 3.10.A.2.b 
to reflect the availability of a second, fully qualified battery 
charger, for each main station battery system.
    Date of Issuance: March 27, 2001.
    Effective date: As of the date of issuance, and shall be 
implemented within 60 days.
    Amendment No.: 198.
    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 March 27, 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 7, 2000.
    Brief description of amendment: The amendment revises Technical 
Specification 3.5.A.1 by adding a note regarding operability of the Low 
Pressure Coolant Injection system (LPCI) under certain restrictive 
conditions. The subject change would provide a clarification of system 
operability that would result in additional flexibility in operations 
during hot shutdown conditions.
    Date of Issuance: March 30, 2001.
    Effective date: As of the date of issuance, and shall be 
implemented within 60 days.
    Amendment No.: 199.
    Facility Operating License No. DPR-28: Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register:
    January 24, 2001 (66 FR 7686).
    The Commission's related evaluation of this amendment is contained 
in a Safety Evaluation dated March 30, 2001.
    No significant hazards consideration comments received: No.

    Dated at Rockville, Maryland this 10th day of April 2001.
    For the Nuclear Regulatory Commission.
John A. Zwolinski,
Director, Division of Licensing Project Management, Office of Nuclear 
Reactor Regulation.
[FR Doc. 01-9320 Filed 4-17-01; 8:45 am]
BILLING CODE 7590-01-P