[Federal Register Volume 64, Number 221 (Wednesday, November 17, 1999)]
[Notices]
[Pages 62704-62722]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: 99-29846]


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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 October 23, 1999, through November 5, 1999. 
The last biweekly notice was published on November 3, 1999 (64 FR 
59796).

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

    The Commission has made a proposed determination that the following 
amendment requests involve no significant hazards consideration. Under 
the Commission's regulations in 10 CFR 50.92, this means that operation 
of the facility in accordance with the proposed amendment would not (1) 
involve a significant increase in the probability or consequences of an 
accident previously evaluated; or (2) create the possibility of a new 
or different kind of accident from any accident previously evaluated; 
or (3) involve a significant reduction in a margin of safety. The basis 
for this proposed determination for each amendment request is shown 
below.
    The Commission is seeking public comments on this proposed 
determination. Any comments received within 30 days after the date of 
publication of this notice will be considered in making any final 
determination.
    Normally, the Commission will not issue the amendment until the 
expiration of the 30-day notice period. However, should circumstances 
change during the notice period such that failure to act in a timely 
way would result, for example, in derating or shutdown of the facility, 
the Commission may issue the license amendment before the expiration of 
the 30-day notice period, provided that its final determination is that 
the amendment involves no significant hazards consideration. The final 
determination will consider all public and State comments received 
before action is taken. Should the Commission take this action, it will 
publish in the Federal Register a notice of issuance and provide for 
opportunity for a hearing after issuance. The Commission expects that 
the need to take this action will occur very infrequently.
    Written comments may be submitted by mail to the Chief, Rules and 
Directives Branch, Division of Administration Services, Office of 
Administration, U.S. Nuclear Regulatory Commission, Washington, DC 
20555-0001, and should cite the publication date and page number of 
this Federal Register notice. Written comments may also be delivered to 
Room 6D22, Two White Flint North, 11545 Rockville Pike, Rockville, 
Maryland from 7:30 a.m. to 4:15 p.m. Federal workdays. Copies of 
written comments received may be examined at the NRC Public Document 
Room, the Gelman Building, 2120 L Street, NW., Washington, DC. The 
filing of requests for a hearing and petitions for leave to intervene 
is discussed below.
    By December 17, 1999, 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, the Gelman 
Building, 2120 L Street, NW., Washington, DC, 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 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

[[Page 62705]]

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: Rulemakings and 
Adjudications Staff, or may be delivered to the Commission's Public 
Document Room, the Gelman Building, 2120 L Street, NW., Washington DC, 
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, the Gelman Building, 2120 L 
Street, NW., Washington, DC, and electronically from the ADAMS Public 
Library component on the NRC Web site, http://www.nrc.gov (the 
Electronic Reading Room).

Carolina Power & Light Company, et al., Docket No. 50-400, Shearon 
Harris Nuclear Power Plant, Unit 1, Wake and Chatham Counties, North 
Carolina

    Date of amendment request: October 21, 1999.
    Description of amendment request: The proposed amendment would 
revise the Technical Specifications (TS) for the Harris Nuclear Plant 
(HNP) to implement selected improvements described in NRC Generic 
Letter (GL) 93-05, ``Line-Item Technical Specifications To Reduce 
Surveillance Requirements For Testing During Power Operation,'' dated 
September 27, 1993. Specifically, HNP proposes to modify the following 
TS to be consistent with GL 93-05: (1) TS 4.1.3.1.2--Change the 
frequency of the control rod movement test to quarterly; (2) TS 
4.6.4.1--Change the frequency of the Hydrogen Monitor analog channel 
operational test to quarterly; (3) TS 4.3.3.1 (Table 4.3-3)--Change the 
Radiation Digital Channel Operational Test to quarterly; (4) TS 
4.4.6.2.2.b.--Change the time for remaining in cold shutdown without 
leak testing the Reactor Coolant System Pressure Isolation Valves to 7 
days; (5) TS 4.4.3.2--Change the testing of the capacity of pressurizer 
heaters to once per 18 months; (6) TS 4.6.4.2.a.--Change the Hydrogen 
Recombiner functional test to once per 18 months; and (7) TS 
4.7.1.2.1.a--Change frequency of testing Auxiliary Feedwater Pumps to 
quarterly.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. The proposed amendment does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    There are no systems being modified as a result of this change. 
Additionally, the way in which equipment is tested is not affected 
by this change. Reducing surveillance intervals for TS components 
(such as control rod testing) may reduce the probability of an 
accident (rod drop accident) by reducing actions that could cause an 
accident to occur (rod movement).
    Therefore, the proposed change does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    2. The proposed amendment does not create the possibility of a 
new or different kind of accident from any accident previously 
evaluated.
    No system, structure, or component is being modified as a result 
of this change. Additionally, there are no changes to the way 
equipment is operated as a result of this change. Operating 
parameters are not being modified as a result of this change.
    Therefore, the proposed change does not create the possibility 
of a new or different kind of accident from any accident previously 
evaluated.
    3. The proposed amendment does not involve a significant 
reduction in the margin of safety.
    These proposed changes are in accordance with NRC Generic Letter 
93-05, dated September 27, 1993 and NUREG-1366, dated December 1992. 
These changes pertain to testing requirements for TS equipment which 
help ensure operability requirements are met. This change does not 
modify the required safety function or operating parameters for 
equipment described in HNP TS.
    Therefore, the proposed change does not involve a significant 
reduction in the margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: William D. Johnson, Vice President and 
Corporate Secretary, Carolina Power & Light Company, Post Office Box 
1551, Raleigh, North Carolina 27602.
    NRC Section Chief: Kahtan Jabbour, Acting.

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

    Date of amendment request: October 15, 1999.

[[Page 62706]]

    Description of amendment request: The amendments would revise 
Section 5.5.7, ``Reactor Coolant Pump Flywheel Inspection Program,'' of 
the Technical Specifications. Section 5.5.7 currently specifies that 
inspections be done according to Regulatory Position c.4.b of 
Regulatory Guide 1.14, Revision 1, such that an in-place ultrasonic 
volumetric examination of the areas of higher stress concentration at 
the bore and keyway be performed at approximately 3-year intervals. The 
licensee proposed to revise this to require a qualified in-place 
ultrasonic examination over the volume from the inner bore of the 
flywheel to the circle of one half the outer radius, or a surface 
examination (magnetic particle and/or penetration testing) of exposed 
surfaces defined by the volume of the disassembled flywheel. The 
licensee stated that the technical basis has been set forth in 
Westinghouse Topical Report WCAP-14535A, and cited similar amendments 
already granted to other nuclear plants.
    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:

First Standard

    Would implementation of the changes proposed in this LAR involve 
a significant increase in the probability or consequences of an 
accident previously evaluated?
    No. There are no accident probabilities or consequences impacted 
by this LAR [license amendment request]. As discussed in Attachment 
3 [the licensee's description of the proposed amendment], following 
a reduction in the scope and frequency of the examinations currently 
required by the applicable Technical Specifications and Regulatory 
Guide 1.14, Revision I, an adequate inservice inspection program 
will continue to be maintained for the reactor coolant pump 
flywheels. Since the integrity of the flywheels will continue to be 
ensured, these components will continue to be available to fulfill 
their existing design function during pump coastdown flow 
transients. Additionally, there is no more risk that the flywheels 
will become a source of missile generation. Consequently, there is 
no significant increase in the probability or consequences of an 
accident previously evaluated.

Second Standard

    Would implementation of the changes proposed in this LAR create 
the possibility of a new or different kind of accident from any 
previously evaluated?
    No. The proposed changes contained in this LAR only reduce the 
existing inspection requirements for the reactor coolant pump 
flywheels. This LAR proposes no changes to the plants' design, 
equipment, or method of operation at either McGuire or Catawba 
Nuclear Station. Furthermore, the reduction in the inspection 
requirements for the flywheels has been generically approved by the 
NRC and is justified by WCAP-14535A. Therefore, since implementation 
of this LAR results in no actual impact upon either of the Duke 
nuclear plants, and since the integrity of the flywheels will 
continue to be ensured at an acceptable level, no new or different 
kinds of accidents are being created.

Third Standard

    Would implementation of the changes proposed in this LAR involve 
a significant reduction in a margin of safety?
    No. Margin of safety is related to the confidence in the ability 
of the fission product barriers to perform their design functions 
during and following an accident situation. These barriers include 
the fuel cladding, the reactor coolant system, and the containment 
system. These barriers are unaffected by the changes proposed in 
this LAR. As discussed in WCAP-14535A, a reduction in the frequency 
for performing the inservice inspections currently done in 
accordance with Regulatory Guide 1.14, Revision I, will not preclude 
the ability to accurately demonstrate the integrity of the reactor 
coolant pump flywheels. This LAR creates no additional threat to the 
integrity of the fission product barriers from the standpoint of 
missile generation or otherwise. Therefore, implementation of the 
changes proposed in this LAR does not impact the assumption of the 
integrity of the flywheels, the fission product barriers, or any 
other accident analyses assumptions. Consequently, no margin of 
safety will be significantly impacted by this LAR.

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

Duke Energy Corporation, Docket Nos. 50-369 and 50-370, McGuire Nuclear 
Station, Units 1 and 2, Mecklenburg County, North Carolina

    Date of amendment request: October 15, 1999.
    Description of amendment request: The amendments would revise 
Section 5.5.7, ``Reactor Coolant Pump Flywheel Inspection Program,'' of 
the Technical Specifications. Section 5.5.7 currently specifies that 
inspections be done according to Regulatory Position c.4.b of 
Regulatory Guide 1.14, Revision 1, such that an in-place ultrasonic 
volumetric examination of the areas of higher stress concentration at 
the bore and keyway be performed at approximately 3-year intervals. The 
licensee proposed to revise this to require a qualified in-place 
ultrasonic examination over the volume from the inner bore of the 
flywheel to the circle of one half the outer radius, or a surface 
examination (magnetic particle and/or penetration testing) of exposed 
surfaces defined by the volume of the disassembled flywheel. The 
licensee stated that the technical basis has been set forth in 
Westinghouse Topical Report WCAP-14535A, and cited similar amendments 
already granted to other nuclear plants.
    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:

First Standard

    Would implementation of the changes proposed in this LAR involve 
a significant increase in the probability or consequences of an 
accident previously evaluated?
    No. There are no accident probabilities or consequences impacted 
by this LAR [license amendment request]. As discussed in Attachment 
3 [the licensee's description of the proposed amendment], following 
a reduction in the scope and frequency of the examinations currently 
required by the applicable Technical Specifications and Regulatory 
Guide 1.14, Revision I, an adequate inservice inspection program 
will continue to be maintained for the reactor coolant pump 
flywheels. Since the integrity of the flywheels will continue to be 
ensured, these components will continue to be available to fulfill 
their existing design function during pump coastdown flow 
transients. Additionally, there is no more risk that the flywheels 
will become a source of missile generation. Consequently, there is 
no significant increase in the probability or consequences of an 
accident previously evaluated.

Second Standard

    Would implementation of the changes proposed in this LAR create 
the possibility of a new or different kind of accident from any 
previously evaluated?
    No. The proposed changes contained in this LAR only reduce the 
existing inspection requirements for the reactor coolant pump 
flywheels. This LAR proposes no changes to the plants' design, 
equipment, or method of operation at either McGuire or Catawba 
Nuclear Station. Furthermore, the reduction in the inspection 
requirements for the flywheels has been generically approved by the 
NRC and is justified by WCAP-14535A. Therefore, since implementation 
of this LAR results in no actual impact upon either of the Duke 
nuclear plants, and since the integrity of the flywheels will 
continue to be ensured at an acceptable level, no new or different 
kinds of accidents are being created.

[[Page 62707]]

Third Standard

    Would implementation of the changes proposed in this LAR involve 
a significant reduction in a margin of safety?
    No. Margin of safety is related to the confidence in the ability 
of the fission product barriers to perform their design functions 
during and following an accident situation. These barriers include 
the fuel cladding, the reactor coolant system, and the containment 
system. These barriers are unaffected by the changes proposed in 
this LAR. As discussed in WCAP-14535A, a reduction in the frequency 
for performing the inservice inspections currently done in 
accordance with Regulatory Guide 1.14, Revision I, will not preclude 
the ability to accurately demonstrate the integrity of the reactor 
coolant pump flywheels. This LAR creates no additional threat to the 
integrity of the fission product barriers from the standpoint of 
missile generation or otherwise. Therefore, implementation of the 
changes proposed in this LAR does not impact the assumption of the 
integrity of the flywheels, the fission product barriers, or any 
other accident analyses assumptions. Consequently, no margin of 
safety will be significantly impacted by this LAR.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: Mr. Albert Carr, Duke Energy Corporation, 
422 South Church Street, Charlotte, North Carolina.
    NRC Section Chief: Richard L. Emch, Jr.

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

    Date of amendment request: September 29, 1999.
    Description of amendment request: The proposed amendments would 
revise the Containment Inservice Inspection (ISI) Program Technical 
Specifications (TS) 5.5.2, ``Containment Leakage Testing Program,'' and 
TS 5.5.7, ``Pre-Stressed Concrete Containment Tendon Surveillance 
Program.'' The proposed amendments would permit the American Society of 
Mechanical Engineers (ASME) Boiler and Pressure Vessel Code, Section 
XI, Subsection IWL visual examinations to be performed in lieu of 
concrete and post-tensioning system general visual examinations 
required by 10 CFR 50, Appendix J and Regulatory Guide 1.163 between 
Type A tests. In addition, the amendment would permit general visual 
examinations of the concrete and post-tensioning system that can be 
performed with a unit in operation to be performed prior to the 
beginning of a refueling outage during which a Type A test is 
scheduled.
    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. Involve a significant increase in the probability or 
consequences of an accident previously evaluated?
    No. Implementation of this amendment would not involve a 
significant increase in the probability or consequences of an 
accident previously evaluated. Approval of this amendment will have 
no significant effect on accident probabilities or consequences. The 
containment is not an accident initiating system or structure; 
therefore, there will be no impact on any accident probabilities by 
the approval of this amendment. The containment serves an important 
function to mitigate consequences of postulated accidents previously 
evaluated and the examination frequencies proposed in this amendment 
will not result in a reduction in the capacity of the containment to 
meet its intended function. The requested flexibility in scheduling 
containment visual examinations has no significant impact on the 
validity of the examinations or of containment structural integrity.
    Additionally, the change to Technical Specification 5.5.7 and 
the planned revision to Selected Licensee Commitment 16.6.2 
described in this amendment application reflect the adoption of an 
ASME Section XI, Subsection IWE and IWL Inservice Inspection Program 
as required by 10 CFR 50 Section 55a(g)(4). Implementation of this 
program will not result in a reduction in the capacity of the 
containment to meet its intended function.
    Therefore, the probability or consequences of an accident 
previously evaluated will not be increased by approval of the 
requested changes.
    B. Create the possibility of a new or different kind of accident 
from the accident previously evaluated?
    No. Implementation of this amendment would not create the 
possibility of a new or different kind of accident from any accident 
previously evaluated. No new accident causal mechanisms are created 
as a result of NRC approval of this amendment request. No changes 
are being made to the plant that would introduce any new accident 
causal mechanisms. This amendment request does not impact any plant 
systems that are accident initiators, since the containment 
functions primarily as an accident mitigator.
    C. Involve a significant reduction in a margin of safety?
    No. Implementation of this amendment would not involve a 
significant reduction in a margin of safety. Margin of safety is 
related to the confidence in the ability of the fission product 
barriers to perform their design functions during and following an 
accident situation, including the performance of the containment. 
This component is already capable of performing as intended, and its 
function is verified by visual examination, post-tensioning system 
examinations, and leakage rate testing.
    The examination requirements of ASME XI, Subsection IWL, are 
essentially identical to those contained in Regulatory Guide 1.35, 
Rev. 3, and are more rigorous than those required by 10 CFR 50, 
Appendix J and Regulatory Guide 1.163. Previous visual examinations 
of containment concrete and post-tensioning system surfaces have not 
revealed any indications of abnormal degradation of the containment. 
The five-year frequency for IWL examinations is adequate in lieu of 
the general visual examination frequency specified in Regulatory 
Guide 1.163 for containment concrete and post-tensioning system 
examinations.
    The ability of the containment to perform its design function 
will not be impaired by the implementation of this amendment at 
Oconee Nuclear Station. Consequently, no safety margins will be 
impacted.

    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: Anne W. Cottingham, Winston and Strawn, 1200 
17th Street, NW., Washington, DC.
    NRC Section Chief: Richard L. Emch, Jr.

Duquesne Light Company, et al., Docket No. 50-412, Beaver Valley Power 
Station, Unit 2, Shippingport, Pennsylvania

    Date of amendment request: June 17, 1999.
    Description of amendment request: The proposed amendment would 
revise Technical Specification (TS) Section 3.4.9.1 and associated 
figures to extend the applicability of the heatup and cooldown curve 
pressure and temperature limits from 10 effective full power years 
(EFPY) to 15 EFPY. The proposed changes include new heatup and cooldown 
curves developed in accordance with the methodology provided in 
Regulatory Guide 1.99, Revision 2, and Code Case N-640. The 
applicability of TS Section 3.4.9.3, Overpressure Protection Systems, 
is also updated to 15 EFPY, and the maximum allowable power operated 
relief valve (PORV) setpoints for the over pressure protection system 
are revised. Revisions to the TS Bases are also made.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:


[[Page 62708]]


    1. Does the change involve a significant increase in the 
probability or consequences of an accident previously evaluated?
    The proposed heatup and cooldown curves have been revised by 
changing the applicability from 10 effective full power years (EFPY) 
to 15 EFPY. The curves have been developed in accordance with the 
methodology provided in Regulatory Guide 1.99, Revision 2 and Code 
Case N-640. The proposed heatup and cooldown curves define limits 
that still ensure the prevention of nonductile failure for the 
reactor vessel. The design basis events that were protected against 
have not changed; therefore, the probability of an accident is not 
increased.
    The overpressure protection system (OPPS) has been revised such 
that the applicability has changed from 10 EFPY to 15 EFPY. This 
system protects the Reactor Coolant System (RCS) at low temperatures 
so that the integrity of the Reactor Coolant Pressure Boundary 
(RCPB) is not compromised by violating the pressure/temperature (P/
T) limits. These changes were determined in accordance with the 
methodologies set forth in the regulations to provide an adequate 
margin of safety to ensure the reactor vessel will withstand the 
effects of normal cyclic loads due to temperature and pressure 
changes as well as the loads associated with postulated faulted 
events. The lower limit on pressure during the design basis OPPS 
mass injection and heat addition transients is established based on 
operational consideration for the RCP number one seal limit which 
requires a nominal differential pressure across the seal faces for 
proper film-riding performance. As part of the OPPS setpoint 
evaluation, margin to the RCP number one seal limit is evaluated.
    This limit corresponds to a differential pressure across the 
seal of 200 psid, which corresponds to the gage pressures. The 
pressure undershoot below the PORV setpoint during a design basis 
mass injection or heat addition event can exceed 100 psi. Therefore, 
with the PORV setpoints developed for the 15 EFPY heatup and 
cooldown curves, there is the potential for RCS pressure to violate 
the RCP number one seal limit at the lowest RCS temperatures.
    Undershoot below the PORV setpoint can be significantly higher 
if both PORVs actuate during an OPPS event, and it is anticipated 
that the pump seal limit would be exceeded. However, staggering the 
setpoints minimizes the likelihood that both PORVs will actuate 
simultaneously during credible OPPS events. Similarly, WCAP 14040-
NP-A indicates that when there is insufficient range between the 
upper and lower pressure limits to select PORV setpoints that 
provide protection against violating both limits, then the setpoint 
selection that provides protection against the upper limit violation 
takes precedence. WCAP-4040-NP, Revision 1 was approved by the NRC 
by letter dated October 16, 1995, which was incorporated in Revision 
2 of the approved WCAP issued in January 1996.
    Modification of the heatup and cooldown curves and OPPS 
setpoints does not alter any assumptions previously made in the 
radiological consequence evaluations nor affect mitigation of the 
radiological consequences of an accident described in the Updated 
Final Safety Analysis Report (UFSAR). Therefore, the proposed 
changes will not significantly increase 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 heatup and cooldown curves applicable for the first 
15 EFPY were generated using approved methodology and Code Case N-
640. Generating these curves with Code Case N-640 reduced the excess 
conservatism that exists in the current curves and results in an 
increase in the safety of the plant, as the likelihood of RCP seal 
failures and/or fuel problems will decrease. The change does not 
cause the initiation of any accident nor create any new single 
failure.
    The modification of the OPPS setpoints ensures that the RCPB 
integrity is protected at low temperatures. The new setpoints were 
selected using conservative assumptions to ensure that sufficient 
margin is available to prevent violation of the P/T limits due to 
anticipated mass and heat input transients. The modification of the 
setpoints does not change, degrade, or prevent the safe response of 
the RCS to accident scenarios, as described in UFSAR Chapter 15. The 
proposed change does not cause the initiation of any accident nor 
create any new credible single failure.
    Therefore, the proposed license amendment 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 new P/T curves define the limits for ensuring prevention of 
nonductile failure for the reactor vessel, and does not 
significantly reduce the margin of safety for the plant. The 
methodology provided in Code Case N-640 removed some of the excess 
conservatism from the current Appendix G analysis. However, this 
improved overall plant safety by expanding the operating window 
relative to the RCP seal requirements. The probability of damaging 
the RCP seals is reduced. Therefore, the margin of safety is not 
significantly reduced.
    The OPPS setpoints will continue to ensure the RCS pressure 
boundary will be protected from pressure transients. They were 
generated using the proposed heatup and cooldown curves as input. 
The OPPS setpoints include additional margin by including instrument 
uncertainties not included in the current setpoints. Therefore, the 
margin of safety is not significantly reduced.

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

Duquesne Light Company, et al., Docket Nos. 50-334 and 50-412, Beaver 
Valley Power Station, Unit Nos. 1 and 2, Shippingport, Pennsylvania

    Date of amendment request: July 15, 1999.
    Description of amendment request: The license amendment request 
(LAR) proposes to revise the Technical Specifications frequency for the 
Quench and Recirculation Spray Systems nozzle air flow test from 5 
years to 10 years. This LAR also includes a revision to correct the 
terminology used in an action requirement as well as miscellaneous 
editorial and format changes.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Does the change involve a significant increase in the 
probability or consequences of an accident previously evaluated?
    The proposed extension of the testing frequency of the Quench 
Spray and Recirculation Spray Systems' nozzles to ten years does not 
change the way these systems are operated or their operability 
requirements. The proposed change to the surveillance frequency of 
safety equipment has no impact on the probability of an accident 
occurrence nor can it create a new or different type of accident. 
NUREG-1366, ``Improvements to Technical Specifications Surveillance 
Requirements,'' dated December 1992, and Generic Letter 93-05, 
``Line Item Technical Specifications Improvements to Reduce 
Surveillance Requirements for Testing During Power Operation,'' 
dated September 27, 1993, concluded that the corrosion of stainless 
steel piping is negligible during the extended surveillance interval 
for nozzle testing. The results of the above NRC study were 
evaluated by Duquesne Light Company and found to be applicable to 
Beaver Valley Power Station (BVPS) Unit 1 and 2. Since the Quench 
Spray and Recirculation Spray Systems are maintained dry, there is 
no additional mechanism that could cause blockage of the spray 
nozzles. Thus, the nozzles in these spray systems are expected to 
remain operable during the ten year surveillance interval to 
mitigate the consequence of an accident previously evaluated. No 
obstructed or clogged spray systems' nozzles have been observed 
during the five year frequency surveillance tests at either BVPS 
Unit 1 or Unit 2 to date. Testing of the spray systems' nozzles at 
the proposed reduced frequency will not increase the probability of 
occurrence of a postulated accident or the consequences of an 
accident previously evaluated.
    This license amendment also revises the Action criteria in the 
BVPS Unit 1 and 2 Axial Flux Difference [AFD] technical

[[Page 62709]]

specification to correct the terminology referring to the Core 
Operating Limits Report (COLR) limits. The proposed change 
incorporates the terminology (acceptable operation limits) used in 
the corresponding Action condition of the ISTS [Improved Standard 
Technical Specifications]. The proposed change does not alter the 
AFD limits specified in the COLR and the AFD specification continues 
to assure plant operation within those limits. With AFD within the 
acceptable operation limits specified in the COLR, the resulting 
axial power distribution remains within the initial conditions 
assumed in the safety analyses. Therefore, these changes will not 
increase the probability of occurrence of a postulated accident or 
the consequences of an accident previously evaluated.
    2. Does the change create the possibility of a new or different 
kind of accident from any accident previously evaluated?
    The proposed reduced frequency testing of the Quench Spray and 
Recirculation Spray Systems' nozzles does not change the way the 
spray systems are operated. The reduced frequency of testing the 
spray nozzles does not change the plant operation or system 
readiness. The reduced frequency testing of the Quench Spray and 
Recirculation Spray Systems' nozzles does not generate any new 
accident precursors. Therefore, the possibility of a new or 
different kind of accident previously evaluated is not created by 
the proposed changes in surveillance frequency of the spray systems' 
nozzles.
    This license amendment also revises the Action criteria in the 
BVPS Unit 1 and 2 Axial Flux Difference technical specification to 
correct the terminology referring to the Core Operating Limits 
Report (COLR) limits. This addresses an incorrect use of terminology 
and the revision does not involve a technical intent change. 
Therefore, the possibility of a new or different kind of accident 
previously evaluated is not created by the proposed terminology 
correction.
    3. Does the change involve a significant reduction in a margin 
of safety?
    The proposed amendment does not involve revisions to any safety 
limits or safety system setting that would adversely impact plant 
safety. The proposed amendment does not affect the ability of 
systems, structures or components important to the mitigation and 
control of design bases accident conditions within the facility. In 
addition, the proposed amendment does not affect the ability of 
safety systems to ensure that the facility can be maintained in a 
shutdown or refueling condition for extended periods of time.
    Reduced testing of the Quench Spray and Recirculation Spray 
Systems' nozzles does not change the way these spray systems are 
operated or these spray systems' operability requirements. Generic 
Letter 93-05 and NUREG-1366 concluded that the corrosion of 
stainless steel piping is negligible during the extended 
surveillance interval for nozzle testing. The results of the above 
NRC study were evaluated by Duquesne Light Company and found to be 
applicable to BVPS Unit 1 and 2. Since the Quench Spray and 
Recirculation Spray Systems are maintained dry, there is no 
additional mechanism that could cause blockage of these spray 
systems' nozzles. Thus, the proposed reduced testing frequency is 
adequate to ensure spray nozzle operability. The surveillance 
requirements do not affect the margin of safety in that the 
operability requirements of the Quench Spray and Recirculation Spray 
Systems remain unaltered. The existing safety analyses remain 
bounding. Therefore, the margin of safety is not adversely affected.
    This license amendment also revises the Action criteria in the 
BVPS Unit 1 and 2 Axial Flux Difference technical specification to 
correct the terminology referring to the Core Operating Limits 
Report (COLR) limits. This addresses an incorrect use of terminology 
and the revision does not involve a technical intent change. The 
operating criteria on Axial Flux Difference are not altered from 
their intended requirements. Therefore, the margin of safety is not 
adversely affected by the proposed terminology correction.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: Jay E. Silberg, Esquire, Shaw, Pittman, 
Potts & Trowbridge, 2300 N Street, NW., Washington, DC 20037.
    NRC Section Chief: Sheri R. Peterson
    Duquesne Light Company, et al., Docket Nos. 50-334 and 50-412, 
Beaver Valley Power Station, Unit Nos. 1 and 2, Shippingport, 
Pennsylvania
    Date of amendment request: July 20, 1999
    Description of amendment request: The licensee amendment request 
proposes to relocate the following Technical Specifications items to 
the Licensing Requirements Manual:
In-core Detectors (Unit 1 and 2),
Chlorine Detection System (Unit 1 and 2),
Turbine Over-speed Protection (Unit 2 only),
Crane Travel Spent Fuel Storage Pool Building (Unit 1 and 2).

    In addition to the relocation, certain editorial and format changes 
are proposed. Also, it is proposed that certain information on the 
Remote Shutdown Panel Monitoring Instrumentation be moved to the 
Updated Final Safety Analysis Report (USFAR).

    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?
    Consistent with the guidance provided in Generic Letter (GL) 95-
10 and the content of the Improved Standard Technical Specifications 
(ISTS) contained in NUREG-1431, Rev. 1, this license amendment 
request (LAR) proposes the relocation of the following TS to the 
Licensing Requirements Manual (LRM):

3/4.3.3.2 Incore Detectors (Unit 1 and 2)
3/4.3.3.7 Chlorine Detection System (Unit 1 and 2)
3/4.3.4 Turbine Overspeed Protection (Unit 2 only)

    In order to completely relocate the chlorine detection system 
requirements from the Technical Specifications (TS), portions of the 
Unit 1 Specifications 3/4.7.7, Control Room Habitability Systems and 
3/4.9.15, Control Room Emergency Habitability Systems, as well as 
the Unit 2 Specification, 3/4.7.7, Control Room Emergency Air 
Cleanup and Pressurization System are proposed to be revised to 
reflect the removal of the chlorine detection system from the TS. 
The applicable surveillance requirements, and modes of applicability 
from these specifications are proposed to be relocated to the LRM 
along with the associated chlorine detection system TS. In addition, 
new actions have been added to the chlorine detection system 
specifications to integrate the new requirements.
    In addition to the TS identified for relocation by the NRC in GL 
95-10, this LAR proposes the relocation of another TS that does not 
meet the criteria of 10 CFR 50.36 and is not included in the ISTS. 
The additional TS proposed to be relocated to the LRM is 3/4.9.7 
Crane Travel Spent Fuel Storage Pool Building (Unit 1 and 2).
    This LAR also proposes that the TS Bases section associated with 
each of the TS listed above be relocated to the LRM as well. The 
appropriate TS pages (i.e., LCO, Bases, Table of Contents, etc.) are 
revised to reflect the removal of these Specifications and Bases 
from the TS.
    The TS and bases discussed above and proposed for relocation 
will be moved into the BVPS LRM. The Unit 1 and Unit 2 LRM are 
appendices of the associated unit UFSAR. As part of the UFSAR any 
changes made to the LRM must be in accordance with the provisions of 
10 CFR 50.59.
    In addition to the relocation of the above listed TS, this LAR 
includes the removal of the ``Measurement Range'' information from 
the Unit 1 and 2 TS Table 3.3-9, Remote Shutdown Panel Monitoring 
Instrumentation. This design information is being moved from the TS 
to an applicable Updated Final Safety Analysis Report (UFSAR) 
section. The removal of this detail from the TS is consistent with 
the level of detail in the corresponding ISTS Specification. As part 
of the UFSAR any changes made to the measurement range information 
must be in accordance with the provisions of 10 CFR 50.59.
    LAR 1A-251/2A-121 includes two Bases enhancements. Additional 
information is being added to the reactor trip system 
instrumentation Bases to discuss diverse and anticipatory protection 
features not credited in the accident analyses. The reactor trip 
system instrumentation Bases is also revised

[[Page 62710]]

to more clearly describe the source and intermediate range neutron 
flux protection features required during shutdown modes.
    The proposed changes include the addition of license numbers to 
some of the TS pages contained in this LAR. In addition, this LAR 
contains changes that update the format of the affected TS pages and 
make editorial corrections. These changes are administrative in 
nature and do not impact the technical content of the affected TS 
pages.
    The proposed changes regarding the relocation of information 
from the TS in this LAR follow the guidance provided in Generic 
Letter 95-10, the NRC ``Final Policy Statement on Technical 
Specifications Improvements for Nuclear Power Reactors'' (58 FR 
39132) dated July 22, 1993, and are consistent with the content of 
the ISTS. In addition, the proposed location for this information 
(UFSAR and LRM) ensures that future changes to the relocated 
requirements will be in accordance with the provisions of 10 CFR 
50.59 and that NRC review and approval will be requested should a 
change to this information involve an unreviewed safety question.
    The proposed amendment does not involve a significant increase 
in the probability of an accident previously evaluated because no 
changes are being made to any accident initiator. No analyzed 
accident scenario is being changed. The initiating conditions and 
assumptions for accidents described in the UFSAR remain as 
previously analyzed. The failure of any of the systems or components 
affected by this LAR, except for turbine overspeed protection, is 
not an accident initiating event. Due to the low likelihood of 
equipment damage or failure resulting from turbine missiles 
generated by a turbine overspeed event, assumptions related to the 
turbine overspeed protection system are not part of an initial 
condition of a design basis accident or transient.
    The proposed amendment also does not involve a significant 
increase in the consequences of an accident previously evaluated. 
The amendment does not reduce the current requirements for the 
systems and components proposed for relocation. The amendment only 
requests that the requirements be retained in a more appropriate 
document. The systems and components proposed for relocation in this 
amendment perform no active role in mitigating a design basis 
accident described in the UFSAR. The systems or components proposed 
for relocation are not part of the initial conditions assumed in a 
safety analysis for a design basis accident described in the UFSAR. 
In addition, the affected systems and components do not function to 
actuate any protective equipment, nor are they part of the primary 
success path assumed in the safety analyses to mitigate any design 
basis accident described in the UFSAR.
    The bases enhancements included in this LAR are administrative 
in nature and serve only to provide additional descriptive 
information. These changes do not impact plant safety.
    Therefore, the proposed amendment 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 does not involve any physical changes to 
the plant or the modes of plant operation defined in Appendix A of 
the operating license. 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.
    Moving specifications to the LRM or design information to the 
UFSAR will not change the physical plant or the modes of plant 
operation. Whether these specifications are located in the TS or the 
LRM has no effect on any previously evaluated accident. The 
relocation of TS information does not involve a change in the 
configuration of equipment nor does it alter the design or operation 
of plant systems.
    Expanding the Bases for both units to discuss additional 
information regarding the protective functions not credited in the 
safety analysis or the neutron flux trip functions required in 
shutdown modes provides additional information to enhance the 
awareness of the protective instrumentation functions. The proposed 
bases changes do not result in any adjustments or physical 
alteration to the affected protective instrumentation functions. The 
Reactor Protection System will continue to function as currently 
designed and assumed in the accident analyses.
    Therefore, operation of the facility in accordance with the 
proposed amendment will not create the possibility of a new or 
different kind of accident from any accident previously evaluated.
    3. 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 amendment does not involve revisions to any safety 
limits or safety system setting that would adversely impact plant 
safety. The proposed amendment does not affect the ability of 
systems, structures or components important to the mitigation and 
control of design bases accident conditions within the facility. In 
addition, the proposed amendment does not affect the ability of 
safety systems to ensure that the facility can be maintained in a 
shutdown or refueling condition for extended periods of time, and 
sufficient instrumentation and control capability is available for 
monitoring and maintaining the unit status.
    The relocation of TS requirements and information to the LRM or 
UFSAR does not reduce the requirements for the affected systems and 
components to be maintained operable and function within design 
requirements. The relocation of TS requirements and information to 
the LRM and UFSAR will allow changes to this information to be made 
in accordance with the provisions of 10 CFR 50.59 and continues to 
ensure that NRC review and approval will be requested should a 
change to this information involve an unreviewed safety question.
    Expanding the Bases for both units to discuss additional 
information regarding the protective functions not credited in the 
safety analysis or the neutron flux trip functions required in 
shutdown modes provides additional information to enhance the 
awareness of the protective instrumentation functions. The addition 
of descriptive text to the TS bases does not affect the TS 
requirements for the affected equipment to be maintained operable 
and function within the applicable design requirements. The Reactor 
Protection System will continue to function as currently designed 
and assumed in the accident analyses.
    Therefore, operation of the facility in accordance with the 
proposed amendment 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 request involves no significant hazards consideration.
    Attorney for licensee: Jay E. Silberg, Esquire, Shaw, Pittman, 
Potts & Trowbridge, 2300 N Street, NW., Washington, DC 20037.
    NRC Section Chief: Sheri R. Peterson.

Duquesne Light Company, et al., Docket Nos. 50-334 and 50-412, Beaver 
Valley Power Station, Unit Nos. 1 and 2, Shippingport, Pennsylvania

    Date of amendment request: September 20, 1999.
    Description of amendment request: The proposed amendments would 
revise the standard to which the control room ventilation charcoal and 
Supplementary Leak Collection and Release System (SLCRS) charcoal must 
be laboratory tested as specified in: Beaver Valley Power Station, Unit 
No. 1 (BVPS-1), Technical Specification (TS) 4.7.7.1.1.c.2 for the 
Control Room Emergency Habitability Systems; BVPS-1 TS 4.7.8.1.b.3 for 
the SLCRS; Beaver Valley Power Station, Unit No. 2 (BVPS-2), TS 
4.7.7.1.d for the Control Room Emergency Air Cleanup and Pressurization 
System; and BVPS-2 TS 4.7.8.1.b.3 for the SLCRS. NRC Generic Letter 99-
02, ``Laboratory Testing of Nuclear-Grade Activated Charcoal,'' dated 
June 3, 1999, requested licensees to revise their TS criteria 
associated with laboratory testing of ventilation charcoal to a valid 
test protocol, which included American Society for Testing Materials 
(ASTM) D3803-1989. This license amendment request revises the charcoal 
laboratory standard to follow ASTM D3803-1989 for each BVPS Unit.
    This license amendment request also: (1) Revises the minimum amount 
of output in kilowatts needed for the control room emergency 
ventilation system heaters at each BVPS Unit; (2)

[[Page 62711]]

revises BVPS-1 SLCRS surveillance testing criteria to be consistent 
with American National Standards Institute/American Society of 
Mechanical Engineers (ANSI/ASME) N510-1980, the BVPS-1 control room 
ventilation testing, and the BVPS-2 SLCRS/control room ventilation 
testing; and (3) makes minor typographical corrections and editorial 
changes.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Does the change involve a significant increase in the 
probability or consequences of an accident previously evaluated?
    The proposed changes to the surveillance requirements for the 
laboratory testing of ventilation system charcoal are consistent 
with Generic Letter 99-02. The proposed change will adopt ASTM 
D3803-1989 as the laboratory testing standard for performing the 
surveillance associated with the Control Room emergency ventilation 
and the SLCRS charcoal filters at each BVPS Unit. Thus this proposed 
change will not involve a significant increase in the probability or 
consequences of a previously evaluated accident since this standard 
provides the assurance for continuing to comply with the current 
BVPS Unit 1 and Unit 2 licensing basis as it relates to the dose 
limits of GDC 19 and 10 CFR Part 100.
    The change in the control room emergency ventilation system 
heater minimum output at both BVPS Units does not change the system 
ability to meet its design bases. The change in the BVPS Unit 1 
SLCRS testing frequency for adsorber/filter in-place testing and the 
adsorber laboratory testing does not change the SLCRS system's 
ability to meet its design bases. The change in the BVPS Unit 1 
SLCRS testing frequency for SLCRS air flow distribution testing does 
not change the SLCRS system's ability to meet its design bases. 
Therefore, these changes will not increase the probability of 
occurrence of a postulated accident or the consequences of an 
accident previously evaluated since these systems' ability to 
operate as required remains unchanged.
    2. Does the change create the possibility of a new or different 
kind of accident from any accident previously evaluated?
    The proposed license amendment to the control room emergency 
ventilation system and SLCRS at both BVPS Units does not change the 
way the system is operated. The proposed changes only involve 
changes to the surveillance testing. These testing modifications do 
not alter these systems' ability to perform their design bases. 
Therefore, these proposed changes do not create the possibility of a 
new or different kind of accident from any previously evaluated 
accident since the control room emergency ventilation system and 
SLCRS will continue to operate in accordance with their previous 
design bases.
    3. Does the change involve a significant reduction in a margin 
of safety?
    The proposed amendment does not involve revisions to any safety 
limits or safety system setting that would adversely impact plant 
safety. The proposed amendment does not affect the ability of 
system, structures or components important to the mitigation and 
control of design bases accident conditions within the facility. In 
addition, the proposed amendment does not affect the ability of 
safety systems to ensure that the facility can be maintained in a 
shutdown or refueling condition for extended periods of time.
    The proposed license amendment to the control room emergency 
ventilation system and SLCRS at both BVPS Units does not change the 
way the system is operated. The proposed changes only involve 
changes to the surveillance testing. These testing modifications do 
not alter these systems' ability to perform their design bases. The 
existing safety analyses remain bounding. Therefore, the margin of 
safety is not adversely affected.

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

Duquesne Light Company, et al., Docket No. 50-412, Beaver Valley Power 
Station, Unit 2, Shippingport, Pennsylvania

    Date of amendment request: September 22, 1999.
    Description of amendment request: The proposed amendment would 
allow a one-time only extension to the surveillance interval of 
Technical Specification Surveillance 4.7.12.d for functional testing of 
snubbers. The proposed extension would be limited to the end of the 8th 
refueling outage or November 30, 2000, whichever occurs sooner.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Does the change involve a significant increase in the 
probability or consequences of an accident previously evaluated?
    The proposed change is for a one-time extension to the 
surveillance interval for functional testing of snubbers specified 
in Technical Specification (TS) 4.7.12.d. The proposed change 
involves revising the calendar time allowed between functional tests 
and would result in a maximum surveillance interval extension of 
approximately 6.5 months.
    The proposed change continues to adequately limit plant 
operation between required snubber surveillances by ensuring the 
required surveillances are performed by November 30, 2000. 
Therefore, the proposed change continues to limit snubber wear due 
to vibration and elevated temperatures. The elevated temperatures 
and vibration experienced during plant operation are the primary 
contributors to snubber wear.
    In addition, snubber-testing experience has shown that the 
historical failure rate of snubbers is low. There have been seven 
refueling outages since Unit 2's startup in 1987. Only during the 
first refueling outage, 2R01, did the snubber functional test sample 
plan identify any inoperable snubbers. In that outage, seven 
snubbers tested inoperable. All failed due to damage sustained 
during original construction and startup activities. Since 2R01, no 
inoperable snubbers were found by sample plan functional testing 
performed during each surveillance interval. Also, the latest visual 
inspections performed on the Unit 2 snubbers (during 2R07) revealed 
no evidence of damage or potential problems with any snubber.
    Due to the low incidence of snubber functional test failures 
resulting from sample plan testing and the limited plant operating 
time between tests, the possibility of a snubber failure resulting 
from this one-time surveillance extension is low. No changes are 
being made to any accident initiator. No analyzed accident scenario 
is being changed. The initiating conditions and assumptions of 
previously analyzed accidents remain unchanged. Therefore, the 
proposed change does not involve a significant increase in the 
probability of a previously evaluated accident.
    This change does not involve a physical change to the plant and 
does not affect the acceptance criteria specified in the TS for 
snubber functional testing, nor does this change reduce the remedial 
actions required for inoperable snubbers. Therefore, the proposed 
change does not involve a significant increase in the consequences 
of an accident previously evaluated.
    2. Does the change create the possibility of a new or different 
kind of accident from any accident previously evaluated?
    The proposed amendment does not involve any physical changes to 
the plant or the modes of plant operation defined in Appendix A of 
the operating license. 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. The one-time surveillance 
interval extension proposed by this change will not reduce the 
capability of the snubbers to perform their design function.
    Therefore, operation of the facility in accordance with the 
proposed amendment will not create the possibility of a new or 
different kind of accident from any accident previously evaluated.
    3. Does the change involve a significant reduction in a margin 
of safety?
    The margin of safety depends on the maintenance of specific 
operating parameters

[[Page 62712]]

and systems within design requirements and safety analysis 
assumptions.
    The proposed amendment does not involve revisions to any safety 
limits or safety system setting that would adversely impact plant 
safety. The proposed amendment does not affect the ability of 
systems, structures or components important to the mitigation and 
control of design bases accident conditions within the facility. In 
addition, the proposed amendment does not affect the ability of 
safety systems to ensure that the facility can be maintained in a 
shutdown or refueling condition for extended periods of time, and 
sufficient instrumentation and control capability is available for 
monitoring and maintaining the unit status.
    The proposed change is for a one-time extension to the 
surveillance interval for functional testing of snubbers specified 
in Technical Specification 4.7.12.d. The proposed change continues 
to adequately limit plant operation between required snubber 
surveillances by ensuring the required surveillances are performed 
by November 30, 2000. Therefore, the proposed change continues to 
limit snubber wear due to vibration and elevated temperatures. The 
elevated temperatures and vibration experienced during plant 
operation are the primary contributors to snubber wear.
    In addition, snubber-testing experience has shown that the 
historical failure rate of snubbers is low. There have been seven 
refueling outages since Unit 2's startup in 1987. Only during the 
first refueling outage, 2R01, did the snubber functional test sample 
plan identify any inoperable snubbers. In that outage, seven 
snubbers tested inoperable. All failed due to damage sustained 
during original construction and startup activities. Since 2R01, no 
inoperable snubbers were found by sample plan functional testing 
performed during each surveillance interval. Also, the latest visual 
inspections performed on the Unit 2 snubbers (during 2R07) revealed 
no evidence of damage or potential problems with any snubber.
    This change does not involve a physical change to the plant and 
does not affect the acceptance criteria specified in the TS for 
snubber functional testing, nor does this change reduce the remedial 
actions required for inoperable snubbers. The snubbers and systems 
supported by the snubbers will continue to be available to perform 
their intended safety functions during the requested extension 
period.
    Therefore, operation of the facility in accordance with the 
proposed amendment 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 request involves no significant hazards consideration.
    Attorney for licensee: Jay E. Silberg, Esquire, Shaw, Pittman, 
Potts & Trowbridge, 2300 N Street, NW., Washington, DC 20037.
    NRC Section Chief: Sheri R. Peterson.

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: July 30, 1999.
    Description of amendment request: The request proposes changes to 
the Technical Specifications (TSs) and the operating license to extend 
operation of the station from its licensed power of 2894 megawatts 
thermal (MWt) to the uprated power level of 3039 MWt, an increase of 5 
percent. The proposed changes are to (1) extend the definition of rated 
thermal power in TS Section 1.1 and the operating license to 3039 MWt; 
(2) reduce the thermal power safety limit of TSs 1.4, 2.1.1.1, 3.2.1, 
3.2.2, 3.2.3, 3.3.1.1, 3.4.3, and 3.7.5; (3) increase the reactor steam 
dome pressure in TS Table 3.1.4-1, TS 3.4.12, and SR 3.5.3.3; (4) 
increase the control rod drive charging water header pressure in TSs 
3.1.5, 3.9.5, and 3.10.8; (5) increase the standby liquid control (SLC) 
system Boron-10 enrichment and concentration criteria in TS 3.1.7; (6) 
increase the surveillance test discharge pressure for the SLC pump in 
surveillance requirement (SR) 3.1.7.7; (7) increase the allowable value 
of the reactor vessel steam dome pressure--high scram setpoint in TS 
Table 3.3.1.1-1; (8) increase the allowable value for the anticipated 
transient without scram--reactor pressure trip reactor steam dome 
pressure--high setpoint in SR 3.3.4.2.4; (9) revise the safety, relief, 
and low low set function of the main steam safety/relief valves (SRVs) 
in SRs 3.3.6.4.3 and 3.4.4.1; (10) increase the upper and lower bounds 
on reactor pressure for the purposes of performing reactor core 
isolation cooling pump flow rate surveillance at high pressure in SR 
3.5.3.3; (11) increase the main steam line flow--high reactor isolation 
trip in TS Table 3.3.6.1-1; (12) reduce the thermal power limits for 
single loop operation in TS 3.4.1; (13) increase the upper and lower 
bounds on reactor pressure for the purposes of performing pressure 
isolation valve surveillance at high pressure in SR 3.4.6.1; and (14) 
revise the reactor coolant system pressure/temperature limits in TS 
3.4.11 (including replacing TS Figure 3.4.11-1 with figures for 14 and 
32 effective full power years of operation). Item (9) includes 
increasing the main steam SRV setpoint tolerance from +0%, -2% to [plus 
or minus] 3% in SR 3.4.4.1.
    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) Will the change involve a significant increase in the 
probability or consequences of an accident previously evaluated?
    The increase in power level discussed herein will not 
significantly increase the probability or consequences of an 
accident previously evaluated.
    The probability (frequency of occurrence) of Design Basis 
Accidents occurring is not affected by the increased power level, as 
the regulatory criteria established for plant equipment (ASME 
[American Society of Mechanical Engineers] Code, IEEE [Institute of 
Electrical and Electronic Engineers] standards, NEMA [National 
Equipment Manufacturers Association] standards, Reg[ulatory] Guide 
criteria, etc.) will still be complied with at the uprated power 
level. An evaluation of the BWR [boiling water reactor] 
probabilistic risk assessments concludes that the calculated core 
damage frequencies will not significantly change due to [the] power 
uprate. Scram setpoints (equipment settings that initiate automatic 
plant shutdowns) will be established such that there is no 
significant increase in scram frequency due to [the] uprate. No new 
challenges to safety-related equipment will result from [the] power 
uprate.
    The changes in consequences of hypothetical accidents which 
would occur from 102% of the uprated power, compared to those 
previously evaluated from [greater than or equal to] 102% of the 
original power, are in all cases insignificant, because the accident 
evaluations from [the] power uprate to 105% of original power 
([approximately] 106% of original steam) flow will not result in 
exceeding the NRC-approved acceptance [criteria] limits. The 
spectrum of hypothetical accidents and transients has been 
investigated, and are shown to meet the plant's currently licensed 
regulatory criteria. In the area of core design, for example, the 
fuel operating limits such as Maximum Average Planar Linear Heat 
Generation Rate (MAPLHGR) and Safety Limit Minimum Critical Power 
Ratio (SLMCPR) are still met at the uprated power level, and fuel 
reload analyses will show plant transients meet the criteria 
accepted by the NRC as specified in NEDO-24011, ``GESTAR II''. 
Challenges to fuel or ECCS [emergency core cooling system] 
performance are evaluated, and shown to still meet the criteria of 
10 CFR 50.46 and Appendix K [to 10 CFR 50], (Section 4.3 above, and 
Regulatory Guide 1.70 and USAR [Updated Safety Analysis Report] 
Section 6.3).
    Challenges to the containment have been evaluated, and the 
containment and its associated cooling systems will continue to meet 
10 CFR 50 Appendix A [General Design Criteria] Criterion 38, Long 
Term Cooling, and Criterion 50, Containment.
    Radiological release events (accidents) have been evaluated, and 
shown to meet the guidelines of 10 CFR 100 (Regulatory Guide 1.70 & 
USAR Chapter 15).
    (2) Will the change create the possibility of a new or different 
kind of accident from any accident previously evaluated?

[[Page 62713]]

    As summarized below, this change will not create the possibility 
of a new or different kind of accident from any accident previously 
evaluated.
    Equipment that could be affected by [the] power uprate has been 
evaluated. No new operating mode, safety-related equipment lineup, 
accident scenario, or equipment failure mode was identified. The 
full spectrum of accident considerations defined in Regulatory Guide 
1.70 have been evaluated and no new or different kind of accident 
has been identified. [The power] Uprate uses already developed 
technology, and applies it within the capabilities of already 
existing plant equipment in accordance with presently existing 
regulatory criteria to include NRC approved codes, standards, and 
methods. GE [General Electric] has designed BWRs of higher power 
levels than the uprated power of any of the currently operating BWR 
fleet and no new power dependent accidents have been identified.
    The Technical Specification changes needed to implement [the] 
power uprate require some small adjustments, but no change to the 
plant's physical configuration. All changes have been evaluated, and 
are acceptable.
    (3) Will the change involve a significant reduction in a margin 
of safety?
    The calculated loads on all affected structures, systems and 
components will remain within their design allowables for all design 
basis event categories. No NRC acceptance criteria will be exceeded. 
Only some design and operational margins are affected by [the] power 
uprate. The margins of safety originally designed into the plant are 
not affected by [the] power uprate. Because the plant configuration 
and reactions to transients and hypothetical accidents will not 
result in exceeding the presently approved NRC acceptance limits, 
[the] power uprate can 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 for the power uprate.
    Although not required for the power uprate, the licensee also 
requested a change to technical specifications to increase the main 
steam SRV setpoint tolerance from +0%, -2% to [plus or minus] 3%. 
However, the licensee's no significant hazards consideration for the 
power uprate does not expressly address the change to the SRV setpoint 
tolerance. Therefore, the NRC staff's review of this change is 
presented below:
    (1) Will the change involve a significant increase in the 
probability or consequences of an accident previously evaluated?
    The main steam SRV's safety function lift setpoints are tested in 
accordance with ASME Code requirements and the licensee's inservice 
testing program. The setpoint tolerance determines whether the SRV 
passes or fails the surveillance requirement and if additional valves 
are to be tested. Notwithstanding the results of the safety function 
lift setpoint test, if the measured value is outside a tolerance of 
[plus or minus] 1%, the valve is reset to within [plus or minus] 1% of 
the design lift setpoint. Therefore, the change to the SRV setpoint 
tolerance does not affect the performance of any structure, system, or 
component in the plant and does not affect the operation of the plant. 
Accordingly, the change will not significantly increase the probability 
or consequences of an accident previously evaluated.
    (2) Will the change create the possibility of a new or different 
kind of accident from any accident previously evaluated?
    The setpoint tolerance change does not alter the function of the 
valves' over-pressure protection features, and the release of steam/
water through the SRVs is addressed in previously evaluated accident 
analysis. Therefore, the change does not create the possibility of a 
new or different kind of accident from any accident previously 
evaluated.
    (3) Will the changes involve a significant reduction in a margin of 
safety?
    The change only affects whether a SRV passes or fails its safety 
function surveillance requirement, as well as the total number of 
valves to be tested. Regardless the outcome of these tests, all valves 
tested will be returned to within [plus or minus] 1% of the design lift 
setpoint. The 2% nominal ``as-left'' tolerance span is effectively the 
same tolerance span as specified in the current technical 
specifications. As a result, there is no significant reduction in a 
margin of safety.
    Therefore, based on its review, it appears that the three standards 
of 10 CFR 50.92(c) are satisfied, and the NRC staff proposes to 
determine that the amendment request involves no significant hazards 
consideration.
    Attorney for licensee: Mark Wetterhahn, Esq., Winston & Strawn, 
1400 L Street, NW., Washington, DC 20005.
    NRC Section Chief: Robert A. Gramm.

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

    Date of amendment request: March 3, 1999.
    Description of amendment request: Entergy Operations, Inc. 
(licensee) has proposed to revise Final Safety Analysis Report (FSAR) 
Section 9.5.4.1, ``Diesel Generator Fuel Oil Storage and Transfer 
Systems.'' The revision will change this section of the FSAR to 
explicitly list the Waterford Steam Electric Station, Unit 3 (Waterford 
3) deviations from the guidance described in American National 
Standards Institute (ANSI) N195-1976, ``Fuel Oil Storage System for 
Standby Diesel Generator.'' The licensee determined that these proposed 
changes require Nuclear Regulatory Commission staff approval prior to 
implementation.
    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 revises the Waterford 3 FSAR to match the 
current design of the Waterford 3 fuel oil storage and transfer 
system. The change effectively requests deviations from portions of 
ANSI N195-1976. None of these changes significantly increases the 
probability of an accident because the Emergency Diesel Generator 
(EDG) fuel oil system is not an initiator of any analyzed event. 
There are no accidents analyzed in the Final Safety Analysis Report 
(FSAR) that are initiated by the systems or components affected by 
these changes.
    The deviation from ANSI N195-1976, which allows less than the 
ANSI Standard recommended volume to be stored in the existing EDG 
Fuel Oil Storage Tanks (FOSTs) A and B, will not significantly 
increase the consequences of an accident. Waterford 3 contains at 
least seven days of fuel oil in each FOST. Although the Waterford 3 
FOSTs do not contain a 10% margin, there are numerous diesel fuel 
oil vendors nearby from which to obtain fuel oil. Waterford 3 also 
has the capability to transport EDG fuel oil from vendors by tanker 
truck, train, or barge. This situation ensures that Waterford 3 will 
have fuel oil readily available when there is a need for 
replenishment. Waterford 3 does not store the additional amount of 
fuel oil required for testing. A previous Technical Specification 
(TS) Amendment addressed the Waterford 3 FOSTs not containing enough 
fuel oil for testing. However, an exception to this requirement was 
previously approved in TS Amendment 92.
    The request for deviation from the ANSI N195-1976 requirement 
for the feed tank suction to be from above the bottom, will not 
increase the consequences of any accident. Previous operating 
experience at Waterford 3 has shown that since initial startup there 
have not been any water or filter blockage problems attributed to 
the bottom suction from the feed tank. The fuel oil in each feed 
tank is replenished every 31 days during the EDG monthly 
Surveillance Requirement (SR). Blockage problems are further 
minimized because testing the FOSTs for particulates is performed 
with a more conservative filter size than installed on the EDG 
engine (0.8

[[Page 62714]]

microns versus 5 microns). Also, TS Surveillances require water and 
sediment content to be verified and if water is present, for it to 
be removed.
    The request for deviation from the ANSI N195-1976 requirement 
for the feed tank overflow to discharge to the FOST will not 
increase the consequences of any accident. The feed tank is equipped 
with design features to ensure fuel oil is not depleted due to over-
filling the feed tank. The feed tank contains a high level switch 
that stops the transfer pump upon indication of high level and a 
high level alarm that alerts the Control Room of high level in the 
tank. A failure of both the feed tank high level switch and high 
level alarm occurring simultaneously is very remote. These measures 
will not prevent the loss of some fuel oil; however, two failures 
would have to occur to prevent the Control Room from being notified. 
Even if one EDG FOST were depleted because of the above failures, 
the other EDG FOST would be available to ensure seven days of fuel 
oil for one EDG.
    The request for deviation from the ANSI N195-1976 requirement to 
have one pressure indicator located in the discharge of the fuel oil 
transfer pump will not increase the consequences of any accident. A 
pressure indicator on the discharge of the transfer pump could 
indicate performance degradation of the pump; however, the Waterford 
3 transfer pumps are designed for automatic operation. If a failure 
of the transfer pump occurred, indication would appear in the 
Control Room via the alarm for low feed tank level. The alarm for 
low feed tank level is adequate to alert the Control Room of a 
transfer pump malfunction. If a transfer pump were to malfunction, 
the other transfer pump would be available to deliver fuel oil to 
operate one EDG for at least seven days. ASME Section XI testing is 
performed on the transfer pump once per quarter (temporary pressure 
instrumentation is installed on the discharge of the pump to measure 
pump differential pressure) to verify that pump performance has not 
degraded. In addition, the transfer pumps are functionally tested 
every month during routine testing of the EDGs.
    The requested deviations from ANSI N195-1976 do not affect the 
consequences of an accident because none of the requested deviations 
will prevent the EDG from having seven days of fuel oil available 
(without multiple failures). Therefore, the EDG fuel oil system will 
perform as required to provide sufficient fuel oil to the EDG to 
mitigate the consequences of design basis accidents.
    Therefore, based on all the above, the proposed changes do not 
involve a significant increase in the probability or consequences of 
an accident previously evaluated.
    2. Will operation of the facility in accordance with this 
proposed change create the possibility of a new or different type of 
accident from any accident previously evaluated?
    Response: No.
    The proposed change revises the Waterford 3 FSAR to match the 
current design of the Waterford 3 fuel oil storage and transfer 
system. This change is a change to a commitment, and has no [a]ffect 
on the current diesel fuel oil storage system or how it is operated, 
nor does it [a]ffect any other safety systems or components, or the 
way the plant is operated. The change does not affect any accident 
analysis assumptions (including a loss of offsite power) or accident 
analysis conclusions. Therefore, the proposed change will not create 
the possibility of a new or different kind of accident from any 
accident 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 revises the Waterford 3 FSAR to match the 
current design of the Waterford 3 fuel oil storage system. Although 
Waterford 3 deviates from certain ANSI N195-1976 requirements, these 
deviations do not result in any changes to the fuel oil storage 
system or accident analyses. The deviations do not affect the 
ability of any safety systems required to protect the multiple 
barriers. No accident mitigatiors are affected by the change because 
the amount of available fuel oil has not changed. As a result, the 
proposed deviations will not cause a significant decrease in the 
margin of safety or prevent Waterford 3 from safely shutting down. 
The result of using Probabilistic Safety Assessment techniques 
conclude that increasing the fuel oil storage capacity at Waterford 
3 to comply with the ANSI requirements has no risk significance. The 
specific [a]ffects of the deviations on the margin of safety are 
addressed below.
    The current TS for stored EDG fuel oil ensures there is 
sufficient fuel oil to operate one EDG for seven days assuming the 
worst case single active or passive failure. Fuel oil is readily 
available due to the number of vendors in the vicinity of Waterford 
3. Waterford 3 is also capable of replenishing EDG fuel oil via 
tanker truck, train, or barge. Therefore, this change does not 
affect the supply of EDG fuel oil being maintained at Waterford 3. 
This supply of fuel oil is sufficient to power the ESF systems 
required to mitigate design basis accidents. A previous TS Amendment 
addressed the Waterford 3 FOSTs not containing enough fuel oil for 
testing.
    The current feed tank design with the suction from the bottom 
instead of on the side as required by ANSI N195-1976 will not 
significantly decrease the margin of safety. Waterford 3 has not 
experienced particulate or water accumulation in the feed tanks. The 
fuel oil in the tank is essentially turned-over every 31 days during 
the EDG monthly SR, and TS Surveillances ensure water and sediment 
content are verified. Additionally, particulate testing is performed 
on the EDG FOSTs using a test filter with a smaller micron size than 
is on the engine. This will assure the EDG engine is not subject to 
failures due to particulate or water accumulation in the feed tanks.
    The request for deviation from the ANSI N195-1976 requirement 
for the feed tank overflow to discharge to the FOST will not 
significantly decrease the margin of safety. The feed tank is 
equipped with two safety measures that would have to fail in order 
to allow a loss of EDG fuel oil due to over-filling a feed tank. A 
failure of these safety measures (high level switch to stop the 
transfer pump and a high level alarm in the feed tank) occurring 
simultaneously is very remote.
    The request for deviation from ANSI N195-1976 to have one 
pressure indicator located at the discharge of the fuel oil transfer 
pump will not significantly decrease the margin of safety. A 
pressure indicator on the discharge of the transfer pump could 
indicate performance degradation of the pump. If a failure of the 
transfer pump occurred, indication would appear in the Control Room 
via the alarm for low feed tank level. The alarm for low feed tank 
is adequate to alert the control room of a transfer pump 
malfunction. However, if the transfer pump were to malfunction, the 
other transfer pump would be available to deliver fuel oil to 
operate one EDG for at least seven days. ASME Section XI testing is 
performed on the transfer pump once per quarter (temporary pressure 
instrumentation is installed on the discharge of the pump to measure 
pump differential pressure) to verify that pump performance has not 
degraded. In addition, the transfer pumps are functionally tested 
every month during routine testing of the EDGs.
    Therefore, based on all the above, the proposed changes 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 request involves no significant hazards consideration.
    Attorney for licensee: N.S. Reynolds, Esquire, Winston & Strawn 
1400 L Street NW., Washington, DC 20005-3502.
    NRC Section Chief: Robert A. Gramm.

PECO Energy Company, Docket Nos. 50-352 and 50-353, Limerick Generating 
Station, Units 1 and 2, Montgomery County, Pennsylvania

    Date of amendment request: September 27, 1999.
    Description of amendment request: The proposed change to the 
Technical Specifications (TSs), if approved, will clarify several 
administrative requirements, delete redundant requirements, and correct 
typographical errors. These revisions affect TS Sections 3.8.3.1, 
3.8.3.2, 6.2.2, 6.5.1.2, 6.8.2, 6.9.1.5, and 6.9.1.6.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. The proposed TS changes do not involve a significant increase 
in the probability or

[[Page 62715]]

consequences of an accident previously evaluated.
    The changes are administrative in nature and do not impact the 
operation, physical configuration, or function of plant equipment or 
systems. The changes do not impact the initiators or assumptions of 
analyzed events, nor do they impact mitigation of accidents or 
transient events. Therefore, these changes do not increase the 
probability of occurrence or consequences of an accident previously 
evaluated.
    2. The proposed TS changes do not create the possibility of a 
new or different kind of accident from any accident previously 
evaluated.
    The proposed changes are administrative in nature and do not 
alter plant configuration, require that new equipment be installed, 
alter assumptions made about accidents previously evaluated, or 
impact the operation or function of plant equipment. Therefore, 
these changes do not create the possibility of a new or different 
kind of accident than previously evaluated.
    3. The proposed TS changes do not involve a significant 
reduction in a margin of safety.
    The proposed changes are administrative in nature and do not 
involve any physical changes to plant structures, systems, or 
components (SSCs), or the manner in which these SSCs are operated, 
maintained, modified, tested, or inspected. The proposed changes do 
not involve a change to any safety limits, limiting safety system 
settings, limiting conditions of operation, or design parameters for 
any SSC. The proposed changes do not impact any safety analysis 
assumptions and do not involve a change in initial conditions, 
system response times, or other parameters affecting any accident 
analysis. Therefore, these changes do not involve any 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: J.W. Durham, Sr., Esquire, Sr. V.P. and 
General Counsel, PECO Energy Company, 2301 Market Street, Philadelphia, 
PA 19101.
    NRC Section Chief: James W. Clifford.

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

    Date of amendment request: October 1, 1999.
    Description of amendment request: The proposed amendments would 
revise the minimum fuel oil level for the diesel generator day tanks in 
Surveillance Requirement 3.8.1.3 and would change the acceptable fuel 
oil level storage band in Required Action Statement B of Limiting 
Condition for Operation 3.8.3.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Do the proposed changes involve a significant increase in the 
probability or consequences of an accident previously evaluated?
    The diesel generators are designed to supply power to the 
emergency systems needed to mitigate the consequences of design 
basis accidents such as LOCA/LOSP [loss-of-coolant accident/loss-of-
offsite power]. They (the diesel generators) do not function to 
prevent accidents. Reducing the level requirement in the day tanks 
and raising the level requirement in the fuel oil storage tanks will 
therefore not increase the probability of occurrence of a LOCA/LOSP 
event. Furthermore, this proposed change does not affect any other 
system or piece of equipment designed to prevent the occurrence of 
any other design basis accident or transient. Therefore, reducing 
the required level in the day tanks and raising the level in the 
fuel oil storage tanks will not increase the probability of 
occurrence of any previously evaluated accident or transient.
    The consequences of previously evaluated events will not be 
significantly increased because, with the 500-gallon day tank 
requirement and the increased storage tank supply, ample fuel will 
be available to supply the diesel generators for the duration of a 
LOCA/LOSP event or a station blackout event. Therefore, the 
consequences of an accident previously evaluated are not increased 
by this modification.
    2. Do the proposed changes create the possibility of a new or 
different kind of accident from any accident previously evaluated?
    Lowering TS SR 3.8.1.3 from [greater than or equal to] 900 
gallons to [greater than or equal to] 500 gallons and raising TS SR 
3.8.3.1 from [greater than or equal to] 33,000 gallons to [greater 
than or equal to] 33,320 gallons will have no impact on the normal 
or emergency operation of the diesel generator and its support 
systems. For example, diesel generator transfer pumps and supply 
tank transfer pumps will continue to perform as necessary to insure 
an adequate supply in the respective tanks for accident mitigation.
    As a result, since no new unanalyzed modes of operation are 
introduced, the possibility of a new or different type of accident, 
from any previously evaluated is not introduced.
    3. Do the proposed changes involve a significant reduction in a 
margin of safety?
    The Bases for TS SR 3.8.1.3 states that the day tank must carry 
enough fuel oil to provide for one hour of operation, plus a 10 
percent margin. This requirement is based on ANSI N195-1976 (Section 
6.1).
    The present 900-gallon requirement in the present Technical 
Specifications provides for 3.5 hours of continuous operation. 
Reducing the volume requirement to 500 gallons will continue to 
provide ample margin above the 1-hour requirement. In fact, 500 
gallons in the day tank provides for 1.89 hours of continuous 
operation.
    The Bases for TS SR 3.8.3.1 states that the fuel in the storage 
tanks (33,000 gallons) alone is sufficient to account for seven days 
of continuous operation. This is true for 33,000 gallons of usable 
fuel. However, each storage tank contains approximately 1,438 
gallons of unusable fuel. Additionally, part of the current design 
bases for the emergency diesel generators is the ability to run four 
of the five diesels continuously for seven days at a load of 3250 
kW. With 500 gallons in each of the four diesel's day tanks and 
33,320 gallons in each of the five storage tanks, the system is 
capable of running continuously for 7 days. Ample onsite fuel 
capacity remains to operate the diesels continuously for a longer 
period than required to replenish the supply from outside sources. 
For the above reasons, the margin of safety is not significantly 
reduced.

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

Southern Nuclear Operating Company, Inc., Georgia Power Company, 
Oglethorpe Power Corporation, Municipal Electric Authority of Georgia, 
City of Dalton, Georgia, Docket No. 50-321, Edwin I. Hatch Nuclear 
Plant, Unit 1, Appling County, Georgia

    Date of amendment request: October 15, 1999.
    Description of amendment request: The proposed amendment would 
change the Safety Limit Minimum Critical Power Ratios (SLMCPR) in 
Technical Specification (TS) 2.1.1.2 to reflect results of a cycle-
specific calculation performed for Unit 1 Operating Cycle 19. The 
calculation was done using the new NRC-approved methodology for 
determining SLMCPRs.
    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 Technical Specification changes do not involve a 
significant increase in the probability of an accident previously 
evaluated.
    The derivation of the revised SLMCPRs for Plant Hatch Unit 1 
Cycle 19 for incorporation

[[Page 62716]]

into the TS, and their use to determine cycle-specific thermal 
limits, have been performed using NRC-approved methods and 
procedures. The procedures incorporate cycle-specific parameters and 
reduced power distribution uncertainties in the determination of the 
lower value for SLMCPRs. These calculations do not change the method 
of operating the plant and have no effect on the probability of an 
accident initiating event or transient.
    The basis of the MCPR Safety Limit is to ensure no mechanistic 
fuel damage is calculated to occur if the limit is not violated. The 
new SLMCPRs preserve the existing margin to transition boiling and 
the probability of fuel damage is not increased. Therefore, the 
proposed changes do not involve an increase in the probability or 
consequences of an accident previously evaluated.
    2. The proposed TS change does not create the possibility of a 
new or different kind of accident from any accident previously 
evaluated.
    The proposed changes result only from a revised method of 
analysis for the Unit 1 Cycle 19 core reload. These changes do not 
involve any new method for operating the facility and do not involve 
any facility modifications. No new initiating events or transients 
result from these changes. Therefore, the proposed TS changes do not 
create the possibility of a new or different kind of accident from 
any accident previously evaluated.
    3. The proposed TS changes do not involve a significant 
reduction in a margin of safety.
    The margin of safety as defined in the TS bases will remain the 
same. The new SLMCPRs are calculated using NRC-approved methods and 
procedures which are in accordance with the current fuel design and 
licensing criteria. The SLMCPRs remain high enough to ensure that 
greater than 99.9% of all fuel rods in the core are expected to 
avoid transition boiling if the limit is not violated, thereby 
preserving the fuel cladding integrity.
    Therefore, the proposed TS changes do not involve a reduction in 
the margin of safety.

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

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

    Date of amendment request: October 15, 1999.
    Description of amendment request: The proposed amendment would 
change the Safety Limit Minimum Critical Power Ratios (SLMCPR) in 
Technical Specification (TS) 2.1.1.2 to reflect results of a cycle-
specific calculation performed for Unit 2 Operating Cycle 16. The 
calculation was performed using the new NRC-approved methodology for 
determining SLMCPRs.
    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 Technical Specification changes do not involve a 
significant increase in the probability of an accident previously 
evaluated.
    The derivation of the revised SLMCPRs for Plant Hatch Unit 2 
Cycle 16 for incorporation into the TS, and their use to determine 
cycle-specific thermal limits, have been performed using NRC-
approved methods and procedures. The procedures incorporate cycle-
specific parameters and reduced power distribution uncertainties in 
the determination of the lower value for SLMCPRs. These calculations 
do not change the method of operating the plant and have no effect 
on the probability of an accident initiating event or transient.
    The basis of the MCPR Safety Limit is to ensure no mechanistic 
fuel damage is calculated to occur if the limit is not violated. The 
new SLMCPRs preserve the existing margin to transition boiling and 
the probability of fuel damage is not increased. Therefore, the 
proposed changes do not involve an increase in the probability or 
consequences of an accident previously evaluated.
    2. The proposed TS change does not create the possibility of a 
new or different kind of accident from any accident previously 
evaluated.
    The proposed changes result only from a revised method of 
analysis for the Unit 2 Cycle 16 core reload. These changes do not 
involve any new method for operating the facility and do not involve 
any facility modifications. No new initiating events or transients 
result from these changes. Therefore, the proposed TS changes do not 
create the possibility of a new or different kind of accident from 
any accident previously evaluated.
    3. The proposed TS changes do not involve a significant 
reduction in a margin of safety.
    The margin of safety as defined in the TS bases will remain the 
same. The new SLMCPRs are calculated using NRC-approved methods and 
procedures which are in accordance with the current fuel design and 
licensing criteria. The SLMCPRs remain high enough to ensure that 
greater than 99.9% of all fuel rods in the core are expected to 
avoid transition boiling if the limit is not violated, thereby 
preserving the fuel cladding integrity.
    Therefore, the proposed TS changes do not involve a reduction in 
the margin of safety.

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

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

    Date of amendment request: October 18, 1999.
    Description of amendment request: The proposed amendment would 
revise the activated charcoal testing methodology in accordance with 
the guidance provided in NRC Generic Letter 99-02, ``Laboratory Testing 
of Nuclear Grade Activated Charcoal.''
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration which is presented below:

    1. Will the proposed changes involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    The Standby Gas Treatment (SBGT) system is used to support 
mitigation of the consequences of postulated accidents. The SBGT 
system is not considered an initiator of any analyzed accident. 
There is no change in function or operation of the system. The 
proposed change only revises the charcoal laboratory testing 
protocol to a more current standard that is more reliable, accurate 
and conservative. The change in relative humidity proposed is 
likewise in accordance with accepted guidance and reflective of the 
Vermont Yankee system configuration, which utilizes heaters to 
reduce the incoming humidity. The change in iodide removal 
efficiency is also more conservative.
    Thus, the probability or consequences of previously analyzed 
accidents is not significantly increased.
    2. Will the proposed changes create the possibility of a new or 
different kind of accident from any previously evaluated?

[[Page 62717]]

    This change does not affect the design or mode of operation of 
any plant system, structure or component. No physical alteration of 
plant structures, systems or components is involved and no new or 
different equipment will be installed. The proposed change only 
modifies the laboratory testing protocol and acceptance criteria to 
a more currently accepted standard.
    Thus, the proposed change does not create the possibility of a 
new or different [kind of] accident from those previously evaluated.
    3. The operation of Vermont Yankee Nuclear Power Station in 
accordance with the proposed amendment will not involve a 
significant reduction in a margin of safety.
    The proposed changes in laboratory test protocol do not 
adversely affect the operation of any systems, structures or 
components. In fact, adopting the newer test standard will provide 
greater assurance that the charcoal will perform its intended 
function of accident consequence mitigation.
    Thus, the proposed change does not significantly reduce a margin 
of safety.

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

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

    Date of amendment request: October 21, 1999.
    Description of amendment request: The proposed amendment makes 
editorial and administrative changes to the Technical Specifications 
(TSs) by correcting two administrative errors and changing the 
designation of a TS-referenced figure. These changes do not materially 
change the meaning or application of any TS requirement.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration which is presented below:

    1. Will the proposed changes involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    The proposed changes are administrative or editorial in nature 
and do not involve any physical changes to the plant. The 
administrative changes do not materially affect any existing 
technical requirement and do not reduce the actions that are 
currently taken to ensure operability of plant structures, systems 
or components.
    The changes correct past administrative errors and change a 
reference in the Technical Specifications and do not revise the 
methods of plant operation which could increase the probability or 
consequences of previously evaluated accidents. No new modes of 
operation are introduced by the proposed changes such that a 
previously evaluated accident is more likely to occur or more 
adverse consequences would result.
    Therefore, the proposed change does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    2. Will the proposed changes create the possibility of a new or 
different kind of accident from any accident previously evaluated?
    These changes are administrative in nature and do not affect the 
operation of any systems or components, nor do they involve any 
potential initiating events that would create any new or different 
kind of accident. There are no changes to the design assumptions, 
conditions, configuration of the facility, or the manner in which 
the plant is operated and maintained.
    The changes do not affect assumptions contained in plant safety 
analyses or the physical design and/or modes of plant operation. 
Consequently, no new failure mode is introduced due to the 
administrative changes.
    Therefore, the proposed change does not create the possibility 
of a new or different kind of accident from any accident previously 
evaluated for Vermont Yankee.
    3. Will the proposed changes involve a significant reduction in 
a margin of safety?
    There are no changes being made to the Technical Specification 
safety limits or safety system settings. The operating limits and 
functional capabilities of systems, structures, and components are 
unchanged as a result of these administrative changes. These 
proposed changes do not affect any equipment involved in potential 
initiating events or safety limits. There is no change to the basis 
for any Technical Specification that is related to the establishment 
of, or the maintenance of, a nuclear safety margin.
    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: Mr. David R. Lewis, Shaw, Pittman, Potts and 
Trowbridge, 2300 N Street, NW., Washington, DC 20037-1128.
    NRC Section Chief: James W. Clifford.

Wisconsin Electric Power Company, Docket Nos. 50-266 and 50-301, Point 
Beach Nuclear Plant (PBNP), Units 1 and 2, Town of Two Creeks, 
Manitowoc County, Wisconsin

    Date of amendment request: October 5, 1999.
    Description of amendment request: The proposed amendments would 
make changes to the Technical Specifications (TSs) that are necessary 
to eliminate inconsistencies in the TSs pertaining to decay heat 
removal requirements (TSs 15.3.1.A.3, 15.3.3.A, and 15.3.3.C). An 
additional change to the requirements in TS 15.3.1.A.4 for pressurizer 
safety valve operability is also proposed to provide appropriate 
coordination with low temperature overpressure protection requirements. 
Bases revisions are provided consistent with the proposed amendments 
and to administratively correct references related to accumulator 
operability in the Bases for TS 15.3.3.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration which is presented below:

    1. Operation of the Point Beach Nuclear Plant in accordance with 
the proposed amendments will not create a significant increase in 
the probability or consequences of an accident previously evaluated.
    Technical Specifications 15.3.1.A.3, 15.3.3.A.3 and 15.3.3.C are 
all interrelated in that they each provide direction for required 
decay heat removal capability, either directly or indirectly by 
providing requirements for both support and supported systems. TS 
15.3.1.A.3 provides requirements for the operation of the reactor 
coolant system loops, steam generators, reactor coolant pumps and 
residual heat removal loops as necessary to support decay heat 
removal from a shutdown unit. TS 15.3.3.A provides requirements for 
operation of the high head safety injection and low head residual 
heat removal system. Specifically, TS 15.3.3.A.3 provides 
requirements for inoperability of the residual heat removal system 
which accounts for the dual purpose of injection and decay heat 
removal. TS 15.3.3.C.2 provides requirements for operation of the 
Component Cooling Water System, a primary support system for both 
Residual Heat Removal System and Reactor Coolant Pump operation. The 
proposed Specifications require redundancy of decay heat removal and 
require placing the plant in a safe condition, maximizing the 
availability of decay heat removal methods when redundancy is lost. 
Appropriate allowances and actions are required to ensure uniform 
mixing of boron for reactivity control with the unit shutdown and 
provide for appropriate allowances to facilitate surveillance 
testing, and refueling operations. The time limits placed on all 
actions are consistent with safe operations, industry and NRC 
guidance. Therefore the probability of a

[[Page 62718]]

loss of shutdown cooling or loss of subcooling; or a loss of 
shutdown reactivity control is minimized.
    Amendments are also proposed to provide for coordination of 
Pressurizer Safety Valve and Pressurizer Power Operated Relief Valve 
operability requirements to ensure redundant overpressure protection 
is provided for all operating conditions. Proposed actions for 
inoperability of Pressurizer Safety Valves minimizes the time in 
that condition. Operation of the valves is not changed. Thus, the 
probability of a loss of coolant due to inadvertent opening of the 
valves is not increased. In addition, overpressure protection is 
maintained under all conditions such that the probability of an 
overpressure due to an analyzed event is not increased.
    The proposed changes do not affect potential leakage paths for 
radiation to the environment, or of key safety barriers, and ensure 
appropriate system and function redundancy is maintained. Therefore 
the consequences of an accident previously evaluated will not 
increase.
    Therefore, operation of the Point Beach Nuclear Plant in 
accordance with the proposed amendments does not result in a 
significant increase in the probability or consequences of an 
accident previously evaluated.
    2. Operation of the Point Beach Nuclear Plant in accordance with 
the proposed amendments will not create the possibility of a new or 
different kind of accident from any accident previously evaluated.
    The proposed amendments do not alter the operation or method of 
function of the Residual Heat Removal System, Component Cooling 
Water System, Pressurizer Safety Valves, or Power Operated Relief 
Valves. The amendments provide for consistency of decay heat removal 
and pressure relief requirements within the Specifications providing 
assurance these functions can be maintained during all required 
plant conditions. Operations are not altered in any way that could 
introduce a new accident initiator not previously considered in the 
PBNP Safety Analyses. Therefore, operation of the Point Beach 
Nuclear Plant in accordance with the proposed amendments cannot 
create the possibility of a new or different kind of accident than 
any previously evaluated.
    3. Operation of the Point Beach Nuclear Plant in accordance with 
the proposed amendments will not result in a significant reduction 
in a margin of safety.
    The proposed amendments ensure redundancy of the decay heat 
removal and overpressure protection over the complete range of 
operating conditions. Limitations are provided to ensure timely 
action to restore the functions to an operable condition consistent 
with their importance to safety. Appropriate allowances and actions 
are required to ensure uniform mixing of boron for reactivity 
control with the unit shutdown and provide for appropriate 
allowances to facilitate surveillance testing, and refueling 
operations consistent with overall safety. The functions or method 
of function of the systems or components affected are not being 
altered. Therefore, operation of the Point Beach Nuclear Plant in 
accordance with the proposed amendments cannot result in a 
significant reduction in a margin of safety.

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

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

    Date of amendment request: October 21, 1999.
    Description of amendment request: The request proposes to revise 
Technical Specification (TS) 3.4.10, Pressurizer Safety Valves (PSV), 
of the improved Technical Specifications issued March 31, 1999. The 
proposed revision is to reduce the safety valve set pressure in 
Limiting Condition for Operation (LCO) 3.4.10, and increase the 
setpoint tolerance in Surveillance Requirement (SR) 3.4.10.1. The PSV 
setpoint and setpoint tolerance is proposed to be changed from 2485 
psig plus or minus 1% to 2460 psig plus or minus 2% in the LCO. The 
tolerance of plus or minus 1% in the SR is for resetting the setpoint 
after testing, if this is needed. The licensee also submitted the Bases 
pages for TS 3.4.10, which show modifications to reflect the changes to 
the TSs.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. The proposed change does not involve a significant increase 
in the probability or consequences of an accident previously 
evaluated.
    Any evaluations performed on an overpressure transient 
conservatively assume the upper limit of the pressurizer safety 
valve (PSV) tolerance as the pressure to which the reactor coolant 
system (RCS) is subjected. The proposed change to the lower 
tolerance limit of the pressure set point means that an overpressure 
transient may be terminated at a pressure that is lower than assumed 
in the analysis. It has also been determined that the design 
transients are not adversely affected because the limiting 
transients are not sensitive to the pressure tolerance decrease. 
Therefore, the primary system pressure boundary is not challenged by 
the PSV lower tolerance limit change. The change in the upper limit 
of the PSV tolerance does not challenge the upper limit of the 
overpressure protection. The maximum opening set pressure is not 
changed, and therefore, does not impact analyses performed for 
overpressure transients. Although the lower PSV set point would 
result in a lower qualified valve flow rate, the slightly lower 
valve flow rate would be more than compensated for by the reduced 
valve opening pressure. The change to the PSV set point and set 
point tolerance does not change the conclusions of the existing 
thermal hydraulic analysis for the pressurizer safety and relief 
system. The design function of the valves is not being changed. 
Therefore, the proposed change does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated in the USAR [Wolf Creek Updated Safety Analysis 
Report].
    2. The proposed change does not create the possibility of a new 
or different kind of accident from any accident previously 
evaluated.
    The proposed change would allow the PSV minimum actuation 
pressure to be as low as 2411 psig. The pressurizer power-operated 
relief valve (PORV) actuation set point is 2335 psig. Therefore, the 
margin between the PORV and PSV actuation set points could be as low 
as 76 psi, which is a reduction of 49 psi from the current 125 psi 
margin. Even with the 30 psi pressure control uncertainty, the 
actuation set point margin of 76 psi is considered adequate and the 
PORVs are expected to continue to actuate before the PSVs during 
Condition 1 transients. As such, the proposed change will not have 
any adverse effect on the control systems. Except for the reduced 
lower set point, the design and operation of the PSVs are not being 
changed. The maximum opening pressure is not being changed. The only 
effect of this change would be that the PSVs could open at a lower 
pressure, but still above the PORV actuation set point. Therefore, 
the possibility of a new or different kind of accident from any 
accident previously evaluated is not created.
    3. The proposed change does not involve a significant reduction 
in a margin of safety.
    The PSVs provide, in conjunction with the reactor protection 
system, overpressure protection for the RCS. The PSVs are designed 
to prevent the system pressure from exceeding the system safety 
limit, 2735 psig, which is 110% of the design pressure. The change 
in the upper limit of the PSV tolerance from plus or minus 1% to 
plus or minus 2% with a reduction in the nominal set point from 2485 
psig to 2460 psig does not challenge the upper limit of the 
overpressure protection. The maximum opening pressure set point is 
not changed, and therefore, does not impact analyses performed for 
overpressure transients. The change to PSV set point and set point 
tolerance does not change the conclusions of the existing thermal 
hydraulic analysis for the pressurizer safety and relief system. For 
all non-LOCA [non-loss of coolant accident] events the analyses 
support the change in PSV set point and set point tolerance from 
2485 psig plus or minus 1% to 2460 psig plus or minus 2%. The change 
in the PSV set

[[Page 62719]]

point and set point tolerance also has no effect on the Reactor 
Protection or Engineered Safety Features Systems trip set points. 
Thus, the proposed change does not involve a significant reduction 
in any margin of safety.

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

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

    The following notices were previously published as separate 
individual notices. The notice content was the same as above. They were 
published as individual notices either because time did not allow the 
Commission to wait for this biweekly notice or because the action 
involved exigent circumstances. They are repeated here because the 
biweekly notice lists all amendments issued or proposed to be issued 
involving no significant hazards consideration.
    For details, see the individual notice in the Federal Register on 
the day and page cited. This notice does not extend the notice period 
of the original notice.

Indiana Michigan Power Company, Docket No. 50-315 and 50-316, Donald C. 
Cook Nuclear Plant, Units 1 and 2, Berrien County, Michigan

    Date of amendment request: September 23, 1999, as supplemented 
October 11, 1999.
    Brief description of amendment request: The proposed amendments 
involve movement of loads in excess of the design-basis seismic 
capability of the auxiliary building load handling equipment and 
structures. The proposed amendment requests approval to move the steam 
generator sections through the auxiliary building and to disengage 
crane travel interlocks, and also requests relief from performance of 
Technical Specification Surveillance Requirement 4.9.7.1.
    Date of publication of individual notice in Federal Register: 
October 26, 1999 (64 FR 57665).
    Expiration date of individual notice: November 26, 1999.

Indiana Michigan Power Company, Docket No. 50-315 and 50-316, Donald C. 
Cook Nuclear Plant, Units 1 and 2, Berrien County, Michigan

    Date of amendment request: October 1, 1999.
    Brief description of amendment request: The proposed amendments 
involve the resolution of an unreviewed safety question related to 
certain small-break loss-of-coolant accident scenarios for which there 
may not be sufficient containment recirculation sump water inventory to 
support continued operation of the emergency core cooling system and 
containment spray system pumps during and following switchover to cold 
leg recirculation. Resolution of this issue consists of a combination 
of physical plant modifications, new analyses of containment 
recirculation sump inventory, and resultant changes to the accident 
analyses to ensure sufficient water inventory in the containment 
recirculation sump. In addition, the licensee proposes to change the 
Technical Specifications dealing with the refueling water storage tank 
inventory and temperature, the required amount of ice in each ice 
basket in the containment, and the delay to start the containment air 
recirculation/ hydrogen skimmer fans.
    Date of publication of individual notice in Federal Register: 
October 29, 1999 (64 FR 58458).
    Expiration date of individual notice: November 29, 1999.

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, the Gelman Building, 2120 L 
Street, NW., Washington, DC, and electronically from the ADAMS Public 
Library component on the NRC Web site, http://www.nrc.gov (the 
Electronic Reading Room).

Consolidated Edison Company of New York, Docket No. 50-247, Indian 
Point Nuclear Generating Unit No. 2, Westchester County, New York

    Date of application for amendment: April 21, 1999, as supplemented 
October 15, 1999.
    Brief description of amendment: The amendment allows for a one-time 
extension of the reactor protection system and engineered safety 
features actuation system instruments.
    Date of issuance: October 29, 1999.
    Effective date: As of the date of issuance to be implemented within 
30 days.
    Amendment No.: 205.
    Facility Operating License No. DPR-26: Amendment revised the 
Technical Specifications.
    Public comments requested as to proposed no significant hazards 
consideration (NSHC): Yes October 14, 1999 (64 FR 55777). The October 
15, 1999, letter provided clarifying information that did not change 
the initial proposed no significant hazards consideration. The notice 
provided an opportunity to submit comments on the Commission's proposed 
NSHC determination. No comments have been received. The notice also 
provided for an opportunity to request a hearing by October 28, 1999, 
but indicated that if the Commission makes a final NSHC determination, 
any such hearing would take place after issuance of the amendment.
    The Commission's related evaluation of the amendment, finding of 
exigent circumstances, and final determination of NSHC are contained in 
a Safety Evaluation dated October 29, 1999.

[[Page 62720]]

    Attorney for licensee: Mr. Brent L. Brandenburg, Assistant General 
Counsel, Consolidated Edison Company of New York, Inc., 4 Irving 
Place--1822, New York, NY 10003.
    NRC Section Chief: Sheri Peterson.

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

    Date of application for amendments: August 4, 1999.
    Brief description of amendments: The amendments revise the TS 
(Appendix A of the Catawba operating licenses) to: (1) modify Section 
3.3.2 regarding the Nuclear Service Water System, and (2) Section 5.3.1 
regarding operating personnel qualifications.
    Date of issuance: November 2, 1999.
    Effective date: As of the date of issuance and shall be implemented 
within 45 days from the date of issuance.
    Amendment Nos.: Unit 1-181; Unit 2-173.
    Facility Operating License Nos. NPF-35 and NPF-52: Amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: September 8, 1999 (64 
FR 48861).
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated November 2, 1999.
    No significant hazards consideration comments received: No

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

    Date of application for amendment: October 22, 1997.
    Brief description of amendment: This amendment approves a proposed 
modification that changes the Perry facility as described in the 
Updated Safety Analysis Report. The change incorporates temperature 
control valves and associated bypass lines around the Emergency Closed 
Cooling system heat exchangers. These features are designed to ensure 
operability of the Control Complex Chilled Water System under post-
accident load conditions, without the need for compensatory measures.
    Date of issuance: October 29, 1999.
    Effective date: October 29, 1999.
    Amendment No.: 107.
    Facility Operating License No. NPF-58: This amendment authorizes 
the revision of the Updated Safety Analysis Report.
    Date of initial notice in Federal Register: November 5, 1997 (62 FR 
59922).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated October 29, 1999.
    No significant hazards consideration comments received: No

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

    Date of application for amendments: May 18, 1999, as supplemented 
by letter dated September 22, 1999.
    Brief description of amendments: The amendments revised 
Surveillance Requirements (SR) 3.8.1.3 and 3.8.1.13 to reduce the 
loading requirements for the emergency diesel generators (EDGs). 
Revised SR 3.8.1.3 requires the EDGs be loaded and operated for 
[greater than or equal to] 60 minutes at a load [greater than or equal 
to] 6500 kW and [less than or equal to] 7000 kW at least every 31 days. 
Revised SR 3.8.1.13 requires the EDGs to be loaded [greater than or 
equal to] 6900kW and [less than or equal to] 7700 kW and operated as 
close as practicable to 3390 kVA for 2 hours. For the remaining hours 
of the test, the EDGs would be loaded [greater than or equal to] 6500 
kW and [less than or equal to] 7000 kW and operated as close as 
practicable to 3390 kVA.
    Date of issuance: October 25, 1999.
    Effective date: As of the date of issuance and shall be implemented 
within 30 days from the date of issuance.
    Amendment Nos.: Unit 1-109; Unit 2-87.
    Facility Operating License Nos. NPF-68 and NPF-81: Amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: August 11, 1999 (64 FR 
43780) The supplemental letter dated September 22, 1999, provided 
clarifying information that did not change the scope of the May 18, 
1999, application and the initial proposed no significant hazards 
consideration determination.
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated October 25, 1999.
    No significant hazards consideration comments received: No

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

    Date of application for amendments: April 28, 1999.
    Brief Description of amendments: These amendments revise TS Section 
3.4.A.4 for Units 1 and 2. The changes relax the minimum volume 
requirement for the refueling water Chemical Addition Tank (CAT) from 
4200 gallons to 3930 gallons. A minor administrative change is also 
being made to TS Table 4.1-2B to correct an earlier printing error and 
to delete a reference which no longer applies.
    Date of issuance: November 1, 1999.
    Effective date: As of the date of issuance and shall be implemented 
within 30 days from the date of issuance.
    Amendment Nos.: 222 and 222.
    Facility Operating License Nos. DPR-32 and DPR-37: Amendments 
change the Technical Specifications.
    Date of initial notice in Federal Register: September 8, 1999 (64 
FR 48869).
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated November 1, 1999.
    No significant hazards consideration comments received: No

Notice of Issuance of Amendments to Facility Operating Licenses and 
Final Determination of No Significant Hazards Consideration and 
Opportunity for a Hearing (Exigent Public Announcement or Emergency 
Circumstances)

    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 for the 
amendment 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.
    Because of exigent or emergency circumstances associated with the 
date the amendment was needed, there was not time for the Commission to 
publish, for public comment before issuance, its usual 30-day Notice of 
Consideration of Issuance of Amendment, Proposed No Significant Hazards 
Consideration Determination, and Opportunity for a Hearing.
    For exigent circumstances, the Commission has either issued a 
Federal Register notice providing opportunity for public comment or has 
used local media to provide notice to the public in the area 
surrounding a licensee's facility of the licensee's application and of 
the Commission's proposed determination of no significant hazards 
consideration. The Commission has provided a

[[Page 62721]]

reasonable opportunity for the public to comment, using its best 
efforts to make available to the public means of communication for the 
public to respond quickly, and in the case of telephone comments, the 
comments have been recorded or transcribed as appropriate and the 
licensee has been informed of the public comments.
    In circumstances where failure to act in a timely way would have 
resulted, for example, in derating or shutdown of a nuclear power plant 
or in prevention of either resumption of operation or of increase in 
power output up to the plant's licensed power level, the Commission may 
not have had an opportunity to provide for public comment on its no 
significant hazards consideration determination. In such case, the 
license amendment has been issued without opportunity for comment. If 
there has been some time for public comment but less than 30 days, the 
Commission may provide an opportunity for public comment. If comments 
have been requested, it is so stated. In either event, the State has 
been consulted by telephone whenever possible.
    Under its regulations, the Commission may issue and make an 
amendment immediately effective, notwithstanding the pendency before it 
of a request for a hearing from any person, in advance of the holding 
and completion of any required hearing, where it has determined that no 
significant hazards consideration is involved.
    The Commission has applied the standards of 10 CFR 50.92 and has 
made a final determination that the amendment involves no significant 
hazards consideration. The basis for this determination is contained in 
the documents related to this action. Accordingly, the amendments have 
been issued and made effective 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 
application for amendment, (2) the amendment to Facility Operating 
License, 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, the Gelman Building, 2120 L Street, NW., Washington, DC, and 
electronically from the ADAMS Public Library component on the NRC Web 
site, http://www.nrc.gov (the Electronic Reading Room).
    The Commission is also offering an opportunity for a hearing with 
respect to the issuance of the amendment. By December 17, 1999, 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, the Gelman Building, 2120 L Street, 
NW., Washington, DC 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. Since the Commission has made a final determination 
that the amendment involves no significant hazards consideration, if a 
hearing is requested, it will not stay the effectiveness of the 
amendment. Any hearing held would take place while the amendment is in 
effect.
    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: Rulemakings and 
Adjudications Staff or may be delivered to the Commission's Public 
Document Room, the Gelman Building, 2120 L Street, NW., Washington, DC, 
by the above date. A copy of the petition should also be sent

[[Page 62722]]

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 the 
factors specified in 10 CFR 2.714(a)(1)(i)-(v) and 2.714(d).

Consolidated Edison Company of New York, Docket No. 50-247, Indian 
Point Nuclear Generating Unit No. 2, Westchester County, New York

    Date of application for amendment: March 26, 1999, as supplemented 
October 15, 1999.
    Brief description of amendment: The amendment allows for a one-time 
extension of system functional tests. The test intervals are extended 
for 37 months to coincide with the next refueling outage scheduled to 
commence on June 3, 2000.
    Date of issuance: October 29, 1999.
    Effective date: As of the date of issuance to be implemented upon 
receipt.
    Amendment No.: 204.
    Facility Operating License No. DPR-26: Amendment revised the 
Technical Specifications.
    Press release issued requesting comments as to proposed no 
significant hazards consideration: Yes, October 22 and 24, 1999, 
Peekskill Evening Star.
    The October 15, 1999, letter provided clarifying information that 
did not change the initial proposed no significant hazards 
consideration. The notice provided an opportunity to submit comments on 
the Commission's proposed NSHC determination. No comments have been 
received. The notice also provided for an opportunity to request a 
hearing by October 28, 1999, but indicated that if the Commission makes 
a final NSHC determination, any such hearing would take place after 
issuance of the amendment.
    The Commission's related evaluation of the amendment, finding of 
exigent circumstances, and final determination of NSHC are contained in 
a Safety Evaluation dated October 29, 1999.
    Attorney for licensee: Mr. Brent L. Brandenburg, Assistant General 
Counsel, Consolidated Edison Company of New York, Inc., 4 Irving 
Place--1822, New York, NY 10003 NRC Section Chief: Sheri Peterson.

    Dated at Rockville, Maryland, this 9th day of November 1999.

    For the Nuclear Regulatory Commission.
John A. Zwolinski,
Director, Division of Licensing Project Management Office of Nuclear 
Reactor Regulation.
[FR Doc. 99-29846 Filed 11-16-99; 8:45 am]
BILLING CODE 7590-01-P